Journal of Nuclear Energy最新文献

Asymptotic aspects of space- and time-dependent slowing down of a pulsed neutron is analytically studied, based on the time-dependent diffusion equation. The calculations are mainly concerned with the flux distribution ψ(r, u, t) and the spatially dependent most probable slowing down time t_max(r, u). The analytic solution for ψ(r, u, t) indicates that its spatial dependence is determined only by the time-dependent neutron age τ(u, t), while t_max(r, u) is shown to be dominantly influenced by the stationary neutron age τ_s(u).

Through these analyses, the interesting relations are obtained: (a) When $\text{t = t}{}_{\mathrm{s}}\text{≡ «t}\text{a}\text{̊}{}_0\text{{1 − ξ/(ξ + 4)(ξ + 2)}}$, τ(u, t), coincides with τ_s,(u), which implies that the spatial dependence of ψ(r, u, t) at that time is analogous to that of the stationary neutron distribution, where «tå₀ is the first time moment. (b) The most probable slowing down time t_max(r, u) at $\text{r = √«r}{}^2\text{a}\text{̊}\text{)}$ (standard notation) becomes equal to the most probable slowing down time t_m of the spatially independent distribution ψ(u, t).

The cross sections employed here are assumed to be constant, but different cross-sections for the source neutrons are allowed.

Gamma rays resulting from the interaction of 14·2 MeV neutrons with ²³Na, ²⁴Mg, ²⁸Si, ³²S, ⁴⁸Ti, ⁵²Cr and ⁵⁶Fe were investigated. The incident neutrons were produced by means of the T(d, n)⁴He reaction. The associated-particle time-of-flight technique was used to discriminate between the neutrons and the gamma rays resulting from the nonelastic scattering processes. The differential gamma-ray production cross-sections were measured for gamma rays with energies of 0·44, 0·64, 1·27, 1·63 MeV for ²³Na; for a gamma ray with energy of 1·37 MeV for ²⁴Mg; for a gamma ray with energy of 1·78 MeV for ²⁸Si; for gamma rays with energies of 1·27, 2·06, 2·24 MeV for ³²S; for gamma rays with energies of 0·99 and 1·31 MeV for ⁴⁸Ti; for gamma rays with energies of 0·93, 1·33, 1·43 MeV for ⁵²Cr; for gamma rays with energies of 0·85 and 124 MeV for ⁵⁶Fe. The integrated gamma ray production cross-sections were also deduced.

A method is presented which provides an analytical solution to the steady state, energy dependent neutron transport equation for the Hydrogen mixture case with isotropic elastic and inelastic scattering. The main characteristics of the method are the use of Green's function and diffusion approximation. A difference form of the Boltzman equation is derived as the final expression for machine computation.

Neutron slowing down spectrum is calculated from ³T(d, n)⁴He source put at the centre of a cubical block of water. A comparison is made with the slowing down spectrum calculated from ²⁴¹Am-⁹Be source. The effects of using different models for inelastic scattering by oxygen atom in water molecule on the slowing down spectrum are also investigated. Some thermal group cross-sections are developed and thermal spectrum in water is calculated for stationary nuclei and is compared with experiments; with good agreement obtained between them within the approximations made in the calculations and the difficulties involved in measuring the thermal spectrum in water.

A method is presented for correcting integral parameters to account for the effect of higher-order scattering anisotropy when the neutron flux used in evaluating the parameters is based upon a transport calculation with a low-order scattering approximation. The method is based upon a variational principle for the parameter of interest.

The reactor of a nuclear power station was treated as a flow calorimeter to evaluate the energy released in fission from a given loading of UO₂. The changes in the inventories of U and Pu isotopes were obtained from reprocessing of the spent fuel and the corresponding total number of fissions was determined. The resulting energy released per average fission in a fuel irradiated to 11·5GWD/MTU was found to be 203·2 ± 6·0 MeV from batch isotopic method, 204·6 ± 6·3 MeV from batch composite method and 204·8 ± 7·2 MeV from ¹⁴⁸Nd-analyses. For burnups both higher and lower than this value, the results indicate the presence of a burnup dependence somewhat greater than that expected on the basis of Pu buildup alone. The accuracy of the plant data was found to be limited by the ability to measure the steam generator feed water flow rates.

The activation cross-sections for some fast neutron reactions on nickel have been determined. At 14·7 ± 0·2 MeV, values are: ⁵⁸Ni(n, 2n)⁵⁷ Ni − 32·6 ± 2·7 mb; ⁵⁸Ni(n, np + n, d + n, pn)⁵⁷Co − 619 ± 49 mb; ⁵⁸Ni (n, p)^58m.gCo − 349 ± 30 mb; ⁶⁰Ni(n, p)^60m,gCo − 129 ± 16 mb. A comparison with previous values is given.