The amount of Cs released from irradiated UO² powder in a range of temperature between 820°and 980°C was measured by sweeping with H²-1%HCL mixture.This mixed gas proved effectively to carry off the released Cs from the UO² surface, as verified by observation of the characteristics of CsCl adsorption on UO² and desorption therefrom. In addition, the amount of Cs released by knock-out during irradiation was measured. The chemical form of Cs adsorbed on the UO² surface after release by knock-out, is considered to be elemental. The activation energy for Cs migration in UO² is discussed, with account taken of the difference in behavior between single-crystals and poly-crystals.

View full abstract

The reactivity balance method commonly used for measuring inserted anomalous reactivities was applied to the HTR (Hitachi Training Reactor).
The reactivities taken into consideration included the kinetic reactivity, as well as those associated with the control rod, power feedback, coolant inlet temperature and Xe buildup.
The reactivity balance data were obtained to cover start-up operation, as well as at various power levels including full rated power. These results obtained show that the errors associated with power feedback and with Xe buildup reactivity are both below approximately 1¢, these errors being due to the one point approximation and to inaccuracies in the constants used in the equations.
The fluctuations of residual reactivity are attributed to power fluctuations, to uncertainties in the measurement of the coolant inlet temperature and in the determination of control rod position. The scattering about the zero reactivity line caused by these fluctuations was evaluated to be within 0.8¢.
Anomalous reactivity was inserted artificially by means of shim rod movement, and was detected. The discrepancy between the residual reactivity calculated by this method and the actual worth of the inserted anomalous reactivity, which was a measure of the sensitivity of this technique, was within 1¢.

View full abstract

A simple procedure has been developed for separating U from F. P. and neutron-irradiated Th. The separation is performed with tri-n-butyl phosphate in a system of dodecane-mixture of sulfuric acid and aluminium nitrate.
Uranium dioxide was irradiated with 20MeV bremsstrahlung, which produced both ²³⁷U and F. P. The target was dissolved in dilute nitric acid and U was extracted into the organic phase of the above mentioned system. Finally, U in the organic phase was back-extracted into an aqueous phase. The γ-ray spectrum and decay curve of the separated U fraction show no radioactive nuclides other than U isotopes and its decay products.
This method can also be applied to the preliminary separation of ²³³U from neutronirradiated Th.
The distribution ratios (Kd) for U, Th and some other elements in the extraction system are also given.

View full abstract

Calculations have been performed on the asymptotic angular neutron flux, critical thickness and extrapolation distance in the axial direction for a cylindrical system of finite length. Neutrons are assumed to be monoenergetic. Applying the finite Laplace transformation in the axial direction and buckling approximation in the radial, we find a solution of the transport equation that satisfies the boundary conditions of no incoming neutron and symmetry. This method is an extension of Kobayashi's Laplace transform method for slab problem, and may easily be applied to other geometries. For a rectangular prism system, numerical results based on our method are compared with results from P₃-approximation.

View full abstract

A numerical approach to the steady-state, space-, angle- and energy-dependent neutron transport equation is presented for neutron shielding calculations. The scattering integral, with anisotropic treatment of elastic scattering and isotropic treatment of inelastic scattering, is evaluated by the use of Gaussian and straightforward quadratures. A system of coupled one-group integral equations for all the energy meshes of interest, converted from the energy-dependent integral transport equation, is calculated by performing a line integration along the neutron path in the direction of motion. For this purpose the direction of neutron motion is represented by discrete-ordinate directions Ω_pq on the unit sphere.
The final presentation of the integral transport equation is derived in a difference form convenient for machine computation. A computation program PALLAS has been written in Fortran IV for IBM 360-75 computer to perform neutron transport calculations based on this approach.
Comparisons are given of the numerical solutions with analytical solutions for unscattered fluxes in various geometries such as plane, spherical and two-dimensional cylindrical, for volume sources with self absorption, and with experimental spectra for angular neutron fluxes in graphite-, polyethylene- and water-shield. Excellent agreement is obtained between the present calculations and analytical or experimental results.

View full abstract

Elementary solutions to the energy-dependent Boltzmann equation with a one-term degenerate scattering kernel are derived in plane geometry, and the weight function W(z) is obtained which makes these solutions mutually orthogonal over the half-range interval of the continuum. The weight function greatly facilitates determination of the expansion coefficients in general solutions and is applied to the problems infinite half space.
The diffusion length (discrete space eigenvalue) υ₀ is exactly expressed by using the halfrange characteristic function X(z). In a 1/ν-absorbing medium, as the absorption concentration q increases from zero to a critical value q^*, the diffusion length decreases from infinity to the end of the continuum. 1/Σmin. For q≥q^*, υ₀ vanishes and the neutron density can be represented by the transient term alone, whose exact expression is obtained.

View full abstract

Plastics, such as polycarbonate, cellulose nitrate and cellulose acetate were irradiated with fast neutrons (2.5, 14MeV) and a count was made of the number of etch-pits on the surfaces of the irradiated material after etching with suitable chemical reagents. The experiments proved that the number of etch-pits increases with the etching time, and that at least during the early period of etching, a close correlation seems to exist between ΔP the increment in the number of etch-pits per unit area and Δl the increment of thickness of layer removed for a common increment of etching time Δt. In other words, an approximate relation ΔP/Δl=ρ was observed to hold during this early etching period, ρ being a constant that seems to correspond to the number of etchable damages produced by fast neutron irradiation per unit volume. The value of ρ was found to correspond roughly to the estimated number of recoil carbon and oxygen atoms in a unit volume in the case of polycarbonate which consists of carbon, oxygen and hydrogen. This indicated that the etch-pits observed in polycarbonate might have developed at the sites of radiation damage caused by the recoil carbons and oxygens due to fast neutron irradiation.

View full abstract

The pressure drop characteristics of a JEFR type fuel pin bundle were obtained from hydraulic tests. The coefficient of drag attributable to the spiral wire spacer wound round each fuel pin, as defined by de Stordeur, was found to be approximately 0.30 for the hexagonal lattice arrangement adopted. The coefficient is independent of spiral wire pitch, which ranged from 90 to 260mm.
The pressure drop to be expected in a fuel pin bundle with spiral wire spacer, such as used in current fast breeder applications, can be satisfactorily estimated by using the coefficients reported.

View full abstract