The neutron cross sections of 241Pu were evaluated in the energy range between 10^-5 eV and 15 MeV, and are stored in the Japanese Evaluated Nuclear Data Library Version (JENDL-1). In the energy range below 100 eV, the evaluated data contained in NDE/B-IV and the resonance parameters recommended in BNL-325 were tentatively adopted. The unresolved resonance parameters were determined between 100 eV and 21.5 keV so as to reproduce the experimental data of the fission and capture cross sections. Above 21.5 keV, the fission cross section was evaluated on the basis of the experimental data, most of which were reported as the ratio to the fission cross section ²³⁵U of and then were normalized by the fission cross section of ²³⁵U adopted in JENDL-1. The capture cross section was obtained from the experimental data of α in the energy range up to 250 keV. The capture cross section above 250 keV and the elastic and inelastic scattering, (n, 2n) and (n, 3n) reaction cross sections above 21.5 keV were obtained on the basis of the theoretical calculations. The calculated cross sections are connected smoothly with those obtained from the unresolved resonance parameters at 21.5 keV. This suggests the self-consistency of the present evaluation.

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Short-lived molybdenum isotopes produced by the ²³⁶U(n, f) reaction were isolated by rapid paper electrophoretic technique. Gamma-ray spectra of the isolated molybdenum fraction were measured with high resolution Ge(Li) detector. The half-lives of ¹⁰³Mo (70±1sec), ¹⁰⁴Mo(66±2sec) and ¹⁰⁵Mo(30±3 and 40±5sec) were determined by analyzing the genetic relationships of the known photopeaks of their technetium daughters. Decay analysis of observed photopeaks free from interference resulted in the detection of many new short-lived photopeaks with half-lives of 1070 sec, and which were not relevant to technetium isotopes. These new photopeaks were ascribed to ¹⁰³¹⁰⁶Mo based on examination of their decay behavior.

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A method of solution of a monoenergetic neutron transport equation in P_L approxi-mation is presented for x-y and x-y-z geometries using the finite Fourier transformation. A reactor system is assumed to consist of multiregions in each of which the nuclear cross sections are spatially constant. Since the unknown functions of this method are the spherical harmonics components of the neutron angular flux at the material boundaries alone, the three- and two-dimensional equations are reduced to two- and one-dimensional equations, respectively. The present approach therefore gives fewer unknowns than in the usual series expansion method or in the finite difference method. Some numerical examples are shown for the criticality problem.

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The strong dependence on molecular weight shown by the compression ratio of a turbo-molecular pump for high vacuum suggests its possible utility for isotope separation. The separation factor obtainable with a multistage turbo-machine was estimated by means of a simplified calculation assuming molecular flow, and zero net throughput. The sepa-ration factor depends on the rotating speed of the pump, and attains maximum value when the speed of the blades is close to the most probable speed of the molecules.
To study this effect experimentally, the separation factors were measured on Ar isotopes as a function of pressure and speed of rotation, using a turbo-molecular pump for high vacuum. The Ar gas was fed into the higher pressure side of the pump, to flow against the impeller action through the pump stages, and emerge on the lower-pressure side with its light component enriched. The abundance ratio between ₃₆AT and ₄₀Ar was measured to determine the separation factor. The measured values of separa-tion factor were 1.40, 1.65, 1.80 and 1.95 at 4, 500, 6, 000, 9, 000 and 12, 000 r.p.m., respectively, which were fairly close to the calculated data at low working pressures, but fell appre-ciably short of prediction as the pressure increased.

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A method of numerically solving the convection-diffusion equation is presented for a binary isotopic gas mixture in the axisymmetric velocity field. A modified Newton's method is employed to perform the numerical integration without the assumptions that the pressure can be estimated from the rigid rotation model and the temperature of gas is uniform. A suitable form of the finite difference equation gives a computationally stable integration with reasonable representation of the molar concentration distribution of isotopic molecules in a rotating cylinder. The method includes a Gaussian elimination procedure which consists of the transformation of the Jacobian matrix to a triangular matrix followed by the backward elimination.
Computations are made on UF, gas in various centrifuges which have the openings for feed, product and waste on the end plates. Discussions are also presented on gas flows and separative efficiencies for the centrifuges which have baffle plates, skirting plates and bellows.

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In order to assess the feasibility of utilizing plutonium in thermal reactors, build-up and decay of actinide nuclides have been studied for BWR, PWR, HWR, HTGR and LMFBR, which are uranium-oxide fueled or mixed-oxide fueled, and which produce electric power of 1, 000 MW. The following items were examined ;
(1) quantities of actinide nuclides build-up in the reactor,
(2) build-up and decay of activities of actinides in the spent fuel,
(3) build-up and decay of activities of actinides after reprocessing, and
(4) variation of isotopic composition of plutonium with high burn-up.
It is concluded from the calculated results that precautions should be taken against high activities of resultant actinides if plutonium is utilized as a fissile material for thermal reactors. To make reprocessing and high-level waste management easy and practical, it is recommended that a thermal reactor should be fueled with uranium, the plutonium produced in a thermal reactor should be used in a fast reactor, and plutonium produced in the blanket of a fast reactor is more appropriate as fast reactor fuel than that from a thermal reactor.

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