Fission cross section ratios of ²⁴⁰Pu and ²⁴²Pu relative to ²³⁵U were measured by using the 4.5 MV Dynamitron accelerator of Tohoku University. The measurement using mono-energetic neutrons was performed in the neutron energy range of 0.67 MeV with the time-of-flight method. Prior to the measurement, a fast timing back-to-back fission chamber was developed with good time resolution to reduce the backgrounds due to α-particles and spontaneous fissions. Furthermore, we took account of the effect of the nonuniformity of fission sample thickness for accurate determination of fission cross section ratio. The uncertainty was estimated by analyzing the correlation between the error sources. The correlation matrix between the measured data was given. The overall uncertainty of the present results is about 2%. For both nuclides, the present results agree well with those by Meadows and by Kuprijanov et al. The JENDL-3 evaluation generally has good agreement with the present results. However, the evaluated data are slightly higher around 1 MeV and lower above 6 MeV than the present results.

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A recurrence equation has been derived, which presents an element of the neutron scattering transformation matrix as a sum of elements of lower order and their derivatives with respect to the mass number A. This facilitates the generation of analytic expressions for elements of high orders, including those matrices with A<1, and particularly that of hydrogen. The derived equation is used here to evaluate expressions of matrix elements corresponding up to P₇ order of scattering of the neutron transport equation. Some numerical results for transformation matrices of A>1 and A<1 are shown, as well as a comparison of those of A≠1 with the matrix of A=1.

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Efficacy of the forward-adjoint Monte Carlo coupling technique, which is characterized by forward-and-adjoint-fluxes folding, has been investigated to accurately calculate shielding characteristics with less computer time in complex geometry, including a long streaming path from source to detection point. The fundamental formulas for detector response and statistical error calculations are given. Sample calculations for neutrons streaming through a mockup geometry of a sodium duct penetrating an iron shield clarify that the coupling technique is several times faster than the ordinary Monte Carlo method.

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This paper presents a method for localizing faulty components of control systems by replaceable parts such as print boards and cables, in a large scale plant like a nuclear power plant. Most of today's control systems form a distributed configuration including many digital controllers interconnected by data communication networks. Usually, to localize the faulty components in nuclear plant control systems, suspected faulty components are narrowed down by executing manual tests to examine whether the objects are normal or abnormal based on design documents and personnel know-how, besides the use of self-diagnosis functions built into the control systems. In the present method, procedures of various tests including the know-how and checking of self-diagnosis functions are provided as knowledge of tests. The test to be executed is determined by considering failure probabilities of objects, and easiness and effectiveness of testing. Then, the suspects are narrowed down sequentially based on the test result. In checking feasibility of this diagnosis method for a simulated control system, intended faults are satisfactorily localized. This method is confirmed to be practicable for diagnosis of large scale digital control systems.

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Equilibrium distributions of actinides and fission products were determined in a LiCl-KCl/Cd system at 800973 K. The redox potential of the system was controlled with the addition of reductant Li. Different distribution behaviors due to different group elements were observed. The group partitioning in this system was thus supported to be feasible. On reduction, however, the actinide and lanthanide elements were found to be less soluble in the Cd phase and to remain at the interface in the form of their solid intermetallics. A group partitioning process in which reductive extraction is combined with filtration was proposed on the basis of the present observations.

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A measurement method of tritiated moisture in gas stream has been developed by taking advantage of the high hygroscopic and deliquescent performances of phosphorus pentoxide P₂O₅. The method includes two techniques; one is of collecting moisture from gas stream using the desiccant, the other is of preparing homogeneous liquid scintillation cocktails containing the phosphide and the collected moisture. By introducing an idea of common use of standard vials as the collector cells disposable and the scintillation cocktail containers, the treatment process has been simplified and the chances of error arising in process for subsequent tritium analysis have been decreased. From results of a typical experiment, a relationship between the collec-tion efficiencies and the moisture contents of P₂O₅ has been obtained and the isotope effect on the moisture collection has been elucidated. The collecting performance and the isotope effect are discussed here considering the mass transport phenomena in collector cell.

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Experimental study on reactivity worth for absorber material in HCLWR core has been carried out in a series of experiments using the Fast Critical Assembly (FCA) in Japan Atomic Energy Research Institute (JAERI). The central reactivity worth as well as the simulated control rod worth of B₄C with different B content and of Hf was measured in FCA-HCLWR core fueled with enriched uranium. Both reactivity worths of B₄C increase with ¹⁰B content. These increasing trends do not saturate to 90% enriched B₄C. The Hf has the smaller reactivity worth than the 20% B₄C. The experimental values are compared with the calculated ones which obtained from JENDL-2 data and the SRAC system. The calculation predicts well the dependence of reactivity worth on ¹⁰B content and underestimates the reactivity worth ratios of the Hf to the 20% B₄C.

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