Measurement of the neutron capture reaction rate of 232Th has been made (1) in the energy range from 1 to 408 keV, and (2) with 55 and 146 keV neutrons by thick silicon filtered neutron beams, using a linac time-of-flight technique. The capture reaction rate of 232Th has been measured relative to that for the ¹⁰B(n, αγ) reaction, whose cross section values have been taken from ENDF/B-V File. The results have been normalized to the 24 keV value previously measured by the present authors. Capture r-rays were detected by a pair of C₆D₆ scintillators located at 12 m on the neutron flight path. Thorium samples were chemically purified in order to reduce the γ-ray background from the daughter nuclides in the thorium decay chain.
The point data measured with the 55 and 146 keV neutrons have been in good agree-ment with the energy dependent cross section for the ²³²Th (n, γ) reaction. It has been found that the results support recent measurements and the ENDF/B-V values of the cap-ture cross section, while the discrepancies with the older measurements and the ENDF/B-IV values above 50 keV are found to be about 10 to 30%. The experimental uncertainties have been analyzed by using covariance information which combines correlation between the data.

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This paper is concerned with a method to evaluate signal transmission paths (STP) in dynamic systems. The STP analysis in the present study intends to clarify through fluctua-tion analysis the cause-effect relationship between measured signals in both qualitative and quantitative manner. In order to facilitate the analysis, we define a new function called signal power contribution ratio which represents relative contributions to the output from the input. The analysis is executed through comparative evaluation of the coherence, noise power contribution and signal power contribution functions calculated based upon autoregres-sive modeling. It is shown that comparison of these three functions provides information useful to evaluate the signal transmission paths.
The present method has been applied to noise signals collected at Japan Power Demon-stration Reactor (JPDR)-II. The result of the analysis suggested its effectiveness for in-vestigation of at-power reactor noise, especially in identifying possible noise sources and their propagation mechanisms.

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Heat transfer coefficients of steam condensation on the containment vessel wall at a LOCA are studied. As to the steady state heat transfer coefficients, though Sagawa's data are a little smaller than Uchida's data, they are very close to the analytical solutions by Mori-Hijikata. Transient heat transfer coefficients are represented by the steady state heat transfer coefficients multiplied by a factor. The factor expresses the agitation effect weaken-ing with time during blowdown. Values of parameters in the factor are determined so as the heat transfer coefficients to fit Sagawa's data. These heat transfer coefficients are ap-plied to the analyses of the experiment with the simulation apparatus of an integrated type marine water reactor. Values of the parameters are also determined so that the tempera-ture transients on the containment vessel wall by analyses fit them of experiment. The differences of the values of parameters by the analyses and by Sagawa's data are discussed.

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This paper reports the adsorption and heat transfer characteristics of Kr, N₂ and mo-lecular sieve type 13X systems for the realization of a ⁸⁵Kr storage system. Measurements were carried out on the adsorbed amount and the effective thermal conductivity using a high pressure experimental apparatus. Adsorption isotherms of Kr, N₂ and the mixed gas on molecular sieve type 13X were obtained in the temperature and pressure range of 293 573 K and 02 MPa respectively, which obey the Henry's law. Effective thermal con-ductivity of the Kr-molecular sieve type 13X system was also obtained. Adsorption and heat transfer analysis of a storage cylinder was performed by means of the finite differential method and the obtained data. Consequently, molecular sieve type 13X proved to be one of the advantageous adsorbents for ⁸⁵Kr storage system.

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This paper describes the experimental results and discussions on the changes in dimen-sion, volume and thermal expansion coefficient of five nuclear grade graphites which were made from petroleum coke or gilsonite coke and irradiated in the JMTR at 670-1, 260°C up to the maximum neutron fluence 3.3 × 10²¹ n/cm² (E>0.18 MeV).
Dimensional or volume shrinkages for all samples were observed after irradiation and their changes depended on the kind of graphite studied. The changes in thermal expansion coefficient were also observed for three of five samples.
The relations between the changes in dimension or volume and thermal expansion co-efficient were obtained, and it was clarified that both the rates of irradiation shrinkage in dimension and volume of the sample having larger thermal expansion coefficient were smaller than for the sample having smaller thermal expansion coefficient.

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An analysis is made on the trapping efficiency of a sodium mist trap brought by change in the mist particle size distribution. The trapping efficiency is represented in terms of the rate of mist adhesion onto the mesh wires. A log-normal distribution function is assumed for the particle size distribution in the mist at trap inlet. The analytical results yield the following conclusions :
(1) The trapping efficiency is greatly influenced by d_p, the mean particle diameter of the mist, and reduces to a minimum in the region of d_p=110 μm, for a trap packed with mesh of diameter currently practiced (0.1 mm). Compared with d_p, change in the standard deviation of particle size distribution exerts a much smaller influence on the trapping efficiency.
(2) The efficiency also depends on the flow rate of the mist through trap, represented by the Reynolds number Re expressed in reference to the wire diameter. With the trap configuration and other factors adopted in the analysis, lowering Re to 0.01, improves almost to unity the trapping efficiency, which is governed in this condition principally by the rate of mist diffusion.
(3) The calculated trapping efficiency agrees well with the corresponding values deter-mined from the measured sodium concentrations in the inlet and the outlet gases.

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The computer simulation method has been applied to the analysis of the NO oxidation and decomposition in the dry and moist NO-O₂-N₂ mixtures by electron beam irradiation. The calculated results were in good agreement with the experimental results for both mix-tures. The NO in the dry NO-O₂-N₂ mixture is decomposed by the reaction with N(⁴S) formed by the radiolysis of the O₂-N₂ mixture and is oxidized by the reactions with 0 and with O₃ formed through the reaction of O₂ with O. The NO in the moist NO-O₂-N₂ mix-ture is oxidized not only by the reaction with OH formed by the radiolysis, but also by the reaction with O₃ formed through the reaction of O₂ with O at low dose. The G values of the formations of the active species obtained by the computer simulation were G(N(⁴S))=1.1, G(N(²D))=0.9 and G(O)=1.7 for the dry NO-O₂-N₂ mixture and G(O)=2.0 and G(HO₂)=0.17 for the moist NO-O₂-N₂ mixture, Also, the analysis by the computer simulation con-firmed the value of G(OH)=3.2 obtained in the moist O₂-N₂ mixture by the competition method.

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By acid leaching of several kinds of irradiated coated particle fuels, uranium- and ¹³⁷Cs-leaching fractions were measured. These fractions were compared with the surface failure fraction of the coated particles by a visual inspection in the post-irradiation examination. From the measurements it was found that ¹³⁷Cs was unsuitable as the objective nuclide for acid leaching, because this nuclide escaped from failed coated particles during irradiation or trapped in the graphite grains of the fuel-compact matrix, and therefore, it was difficult to relate the leaching fraction of ¹³⁷Cs to the failure fraction of the coated particles. Cesium-137 leaching fraction was less than 10% of the uranium fraction. The uranium leaching fraction agreed fairly well with the surface failure fraction. Therefore, uranium is regarded as the suitable objective nuclide for acid leaching. Also, combination of the acid leaching and the surface inspection is an useful method for detection of failed coated particles.

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