To obtain fundamental data for the research of the transmutation of nuclear waste, the thermal neutron cross section of the reaction ⁹⁰Sr(n, γ)⁹¹Sr has been measured by means of a radiochemical method.
Strontium chloride targets containing about 2 MBq of ⁹⁰Sr were irradiated for 10min with reactor neutrons. Activation detectors of Co/Al and Au/A1 alloy wires were irradiated simultaneously to monitor the neutron flux and Westcott's epithermal index r(T/T₀)^1/2.The irradiated Sr samples were purified chemically and their γ-ray spectra were measured with a high purity Ge detector. Since the targets included ⁸⁵Sr, a γ-ray emitter and a tracer for ⁹⁰Sr, and a ratio of the amount of ⁸⁵Sr to that of ⁹⁰Sr was measured in advance, an effective cross section of the ⁹⁰Sr(n, γ)⁹¹Sr reaction for a reactor neutron spectrum was determined from the ratio of radioactivities of ⁹¹Sr and ⁸⁵Sr, and neutron flux data. The resonance integral was also measured by a modified Cd-ratio measurement.
Considering an upper limit of the resonance integral to be 0.16b, the thermal neutron cross section (2, 200 m/s neutron cross section) of ⁹⁰Sr(n, γ)⁹¹Sr reaction was found to be 0.0153±^0.0013_0.0042 b.
The value obtained is only 1/50 of the value reported by Zeisel (1966), and is in good agreement with the figure reported by McVey et al. (1983).

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A simple semi-time-optimal controller is proposed for a linear single-input single-output system. The controller is presented as a state feedback with time-varying gains. Stability analysis of the control system shows that the system is stable if the closed loop coefficient matrix is within a limited range.A method of eigenvalue assignment for the closed loop is provided and coordinate transformation is applied in the present paper to design the controller. Simulations of the control system for BWR indicate that the present controller realizes the time-optimal control better than the conventional PID controller.

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Temperature noise, measured by thermocouples mounted at each core fuel subassembly, is considered to be the most useful signal for detecting and locating local cooling anomalies in an LMFBR core. However, the core outlet temperature noise contains background noise due to fluctuations in the operating parameters including reactor power. It is therefore necessary to reduce this background noise for highly sensitive anomaly detection by subtracting predictable components from the measured signal. In the present study, both a physical model and an autoregressive model were applied to noise data measured in the experimental fast reactor JOYO. The results indicate that the autoregressive model has a higher precision than the physical model in background noise prediction. Based on these results, an "autoregressive model modification method" is proposed, in which a temporary autoregressive model is generated by interpolation or extrapolation of reference models identified under a small number of different operating conditions. The generated autoregressive model has shown sufficient precision over a wide range of reactor power in applications to artificial noise data produced by an LMFBR noise simulator even when the coolant flow rate was changed to keep a constant power-to-flow ratio.

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For the establishment of liquid metal cooled fast breeder reactors (LMFBRs) as commercial reactors, one of the most important tasks is to develop highly reliable heat transfer tubes for steam generators (SGs) used in LMFBRs because their failure could lead to a potentially explosive sodium/water reaction. The authors proposed a new concept of a wire-packed duplex tube in order to improve stress concentration after its bending, thermal stress and so on against the contact type duplex tubes and knitted wire-packed duplex tubes which had been proposed.
In this paper, two-phase flow in wire-packed beds, which simulate the wire-packed layer of the proposed duplex tubes, is experimentally investigated using air and water as test fluids in order to obtain a fundamental data base for a water-leakage detection method using absolute pressure and differential pressure measurements. The measured pressure drops are analyzed by using the following analytical models; (1) homogeneous flow model, (2) channeling model and (3) annular flow model.
It is made clear that the hydrodynamics of two-phase flow in wire-packed beds can be evaluated well by the annular flow model.

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Liquid film thickness measurements and flow pattern observations near a spacer have been carried out using 4×4 and 9×9 mock-up fuel bundles with air-water co-current flow. The spacer had two different effects on the liquid film flow in the downstream region: (a) liquid droplet entrainment and deposition induced by turbulence enhancement in gas core flow ; and (b) liquid droplet entrainment and deposition induced by a direct interaction of liquid film flow and spacer elements. The former effect was taken into account for development of a new spacer model to estimate dependence of critical power on spacer pitch. The liquid film model, including the spacer model, was confirmed to predict the critical power with high accuracy.

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A simple model based on the equation of energy with constant radial velocity and thermal conductivity being input was proposed to predict the change in hot-wire temperature caused by flow circulation near the top and bottom plates of a thermal diffusion column. A resultant differential equation on radial temperature distribution was solved approximately by a collocation method to give a formula for the change in hot-wire temperature explicitly in terms of pressure, constant radial velocity and physical properties of gas and geometry of the column. The formula explains qualitatively well the change in hot-wire temperature, which was found in the previous work on the rigorous calculation of temperature : (1) The magnitude of the change in hot-wire temperature is almost proportional to the square of the pressure, and (2) the magnitude of the change in hot-wire temperature at the bottom is larger than that near the top plate.

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An experimental study using a bench-scale equipment of 1 kg-simulated spent solvents per hour has been conducted in order to evaluate the applicability of a submerged combustion technique to the treatment of spent solvents contaminated with TRU elements.
This report describes the experimental results on the combustion characteristics of the simulated spent solvents of tri-n-butyl phosphate and/or n-dodecane, and on the distribution behaviors of combustion products such as phosphoric acid, Ru, I, Zr and lanthanides as TRU simulants in the submerged combustion process. Also the experimental results of TRU separation from phosphoric acid solution by co-precipitation using bismuth phosphate are reported.
It was shown that the submerged combustion technique was applicable to the treatment of spent solvents including the distillation residues of the solvent. Based on the experimental data, a new treatment process of spent solvent was proposed which consisted of submerged combustion, co-precipitation using bismuth phosphate, ceramic membrane filtration, cementation of TRU lean phosphate, and vitrification of TRU rich waste.

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A Multi-Purpose Incinerating Melter System (MIMS) has been developed as a volume reduction technique for a wide variety of radwastes including flame retardants such as spent resin, and non-combustible materials such as concrete, glass and steel. In the MIMS, these wastes are incinerated and/or melted at temperatures between 1, 000 and 1, 500°C generated by fossil fueled burner to produce obsidian-like ingots with high integrity.
A demonstration test program was carried out from 1989 until 1991 using an engineering-scale demonstration unit. In the test program, various simulated wastes with traces of ⁶⁰Co, ⁵⁴Mn, ⁵⁹Fe, ¹³⁷Cs, ²²Na and ¹⁰⁶Ru were treated to obtain decontamination factor (DF) data and leach-resistance data of the products.
The summarized results drawn from the 13 runs of demonstrative operations are the following:
(1) Most involatile radionuclides are transferred into solified products.
(2) Global DF of the system excluding a HEPA filter ranged 1×10⁴ thru 1×10⁵ for ⁶⁰Co, 2×10² thru 2×10³ for ¹³⁷Cs and 2×10² thru 1×10⁴ for ¹⁰⁶Ru.
(3) Leaching resistance of the solidified product is a match for that of a typical borosillicate glass waste form.

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For a long time, coloration in alkali halide melts containing the alkali metals (such as Li, Na) has received much attention in the study of the properties of excess electrons in ionic solutions⁽¹⁾. Today, there are evidences that the coloration is due to the formation of the solvated electron e_s^- in the melts, whose existence is well known as a primary product of radiolysis of polar liquids such as water⁽²⁾. Of more recent attention⁽³⁾, alkali halide melts are a prospective stable material in molten salt nuclear reactor engineering in which melts are subjected to heavy irradiation by high energy nuclear radiation of different LET.
Recently, a pulse radiolysis method has been applied to study the radiation-induced decomposition of the melts of alkali halides and their mixtures⁽⁴⁾⁽⁶⁾. The proposed mechanism of decomposition has shown that ionization of halogen ion (X-) by radiation initiates the radiolysis of the melt, which is followed by the formation of a solvated electron as well as a molecular halogen ion in the melt:
Among alkali halide melts, a pulse radiolytic study of fluoride systems has not been carried out yet, though they are important melts in the nuclear engineering because a fluorine atom has a small neutron absorption cross section. In our research, we attempt to observe the absorption spectrum of primary products formed in irradiated melts of fluoride systems, especially, LiF-KF and LiF-NaF-KF eutectic mixtures.

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Radioruthenium produced in relatively high fission yield shows complex behavior in nuclear fuel reprocessing, radioactive waste treatment and environmental samples. It is known that the chemical characteristics of ruthenium is quite complicated and determination of this element is very difficult, because ruthenium possesses many oxidation states, from +8 to zero, and related compoundsm⁽¹⁾⁽³⁾.
In radiochemical analysis of ruthenium, coprecipitation with insoluble metal salts such as ferric hydroxide, cobalt sulfide is often used as the first step for concentrating or as an effective means for scavenging. The most widely used method for the separation of radioruthenium from other fission products is to distill as gaseous tetroxide (RuO₄) at high temperature. However, experimental techniques of coprecipitation and distillation are usually rather troublesome and time-consuming⁽⁴⁾.
In addition, quantitative coprecipitation and perfect distillation are difficult in some kinds of chemical species.It is known that ruthenium metal or insoluble oxides are very hard to dissolve by strong acidic solutions, such as hydrochloric acid, nitric acid and sulfuric acid. However these insoluble ruthenium species can be dissolved in a sodium hydroxidesolution by forming high oxidation state anions such as ruthenate or perruthenate with oxidizing agent, sodium hypochlorite⁽⁴⁾.
In this study the authors have made experiments on the adsorption behavior of high oxidation state anions of ruthenium on filter paper from sodium hydroxide solution and the possibility of application to simple, rapid and selective separation of radioruthenium from other important radionuclides.

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