日本原子力学会誌
Online ISSN : 2186-5256
Print ISSN : 0004-7120
ISSN-L : 0004-7120
5 巻, 7 号
選択された号の論文の13件中1~13を表示しています
  • 石原 健彦, 平野 見明
    1963 年 5 巻 7 号 p. 549-554
    発行日: 1963/07/30
    公開日: 2009/03/26
    ジャーナル フリー
    Using the difference in vapor pressure between the chlorides of uranium and those of fission products, a chlorination-distillation process of irradiated uranium dioxide fuel was investigated. The experiments were carried out with the chlorination agent of CCl4 vapor flowing through a quartz reaction tube assembly at various reaction and condensation temperatures. On a single distillation, about 80% of the uranium is recovered and about 90% of the γ-emitting fission products is separated from the uranium. Main nuclides condensed together with the uranium are 141Ce and 144Ce. For the purpose of improving the decontamination of uranium, natural CeO2 is mixed with the irradiated UO2 to the extent of 1% before chlorination. About 97% of fission products is separated from the uranium, and 103Ru, 106Ru, 137Cs and 147Nd remain as the residue.
  • 新藤 満夫
    1963 年 5 巻 7 号 p. 555-560
    発行日: 1963/07/30
    公開日: 2009/03/26
    ジャーナル フリー
    In the analytical calculation of criticality by 3 group method, the equation which determines buckling may have, in general, complex solutions. For the positive real buckling, the flux has the same form as is given in the usual bare reactors, and has an oscillatory character, and for the negative real buckling, the flux has an exponential character. For the complex buckling, however, the flux contains both an oscillating and an exponential part. These forms have been deduced explicitly for some typical reactor geometries. No other forms of functions as shown here does appear if the number of groups becomes greater than three.
  • 塩化物揮発法
    岩本 多実, 小林 紀昭, 下川 純一
    1963 年 5 巻 7 号 p. 560-566
    発行日: 1963/07/30
    公開日: 2009/03/26
    ジャーナル フリー
    The chloride-volatility process was applied to the processing of irradiated UO2-graphite fuel elements, in which most of fission products were introduced into the graphite matrix by fission recoil.
    In proper conditions, 99% of UO2 in the sample was chlorinated with CCl4 vapor. However, the major objective in the present study was to investigate the behavior of the fission products during the chlorination. It was elucidated that besides the postirradiation heating itself the existence of the chlorinating agent was additionally effective in the migration of fission products in the graphite matrix.
    The possibility of low temperature chlorination with chlorine gas was also studied.
  • 石森 富太郎, 木村 幹, 中村 永子, 鄭 文彬, 小野 麗子
    1963 年 5 巻 7 号 p. 566-571
    発行日: 1963/07/30
    公開日: 2009/03/26
    ジャーナル フリー
    The solvent extraction behavior of about sixty chemical elements was surveyed in the system of 10% (v/v) Primene JM-T xylene solution vs. hydrochloric acid, the acidity being varied 112 N. The distribution ratio, Kd, of a particular nuclide was determined by the use of radioisotope tracer.
    The Kd values of this Primene JM-T-HCl system were usually below one. Iron, gallium, antimony and protactinium were extractable above 6 N-hydrochloric acid, while gold.alone could be extracted all over the hydrochloric acidity tested.
    The present results were compared with those for secondary or tertiary amine extraction.
  • ウラン押出棒およびウラン合金の熱サイクルならびに腐食試験
    丸谷 和夫, 斎藤 利男, 前川 立夫
    1963 年 5 巻 7 号 p. 572-580
    発行日: 1963/07/30
    公開日: 2009/03/26
    ジャーナル フリー
    In the previous study, some grain-refined uranium alloys were developed by giving some additives, heat treatments and hot rollings.
    In this experiment some uranium alloys were fabricated by coreduction process, and thermal cycling tests were carried out.
    Properties of γ extruded uranium rods were tested and their merits were compared with those of alloying specimens.
    Corrosion tests of uranium and uranium alloys were undertaken in a temperature range of 300°600°C.
    Limitation of corrosion resistivity of uranium alloys were discussed. Main results were summarized as follows:
    (1) On γ extruded uranium rods, β oil quenched specimen shows the smallest growth, warp and wrinkle after the thermal cycling tests.
    After oil quenching treatment, it is not necessary to be followed by α annealing procedure.
    (2) Requirements to reduce thermal cycling growth by heat treatments are different in each specimen, so the appropriate heat treatments should be considered.
    (3) Thermal cycling growth is also large in extremely grain-refined specimens.
    (4) Weight gain of uranium alloys in oxidation by CO2 is larger than that of metallic uranium in the temperature range above 550°C.
    (5) In standpoints of grain refinement, workability, thermal cycling growth, corrosion resistivity and neutron economics, U-Nb-Mo, U-Cr-Mo and U-Zr-Mo ternary alloys are recommended as the minor additived uranium fuels.
  • 第4元素を添加したときの機械的性質,炭酸ガスに対する耐食性
    青木 重夫, 太郎良 績, 橋口 隆吉, 三島 良績
    1963 年 5 巻 7 号 p. 581-586
    発行日: 1963/07/30
    公開日: 2009/03/26
    ジャーナル フリー
    The effects of the fourth elements on the alloy have been investigated in order to improve the mechanical properties and the corrosion resistance of Zr-Cu-Mo alloys at high temperature and high pressure of CO2.
    Results obtained are summarized as follow: (1) Nb and Bi are excellent as the fourth element of alloying. (2) The desirable good properties are found in these alloys of Zr-0.5% Cu-1.0% Mo-1.0% Nb and Zr-0.5% Cu-1.5% Mo-1.0% Bi. (3) These alloys have better mechanical properties than ternary alloys and also these properties are much more improved by the heat-treatment. (4) These alloys have the good corrosion resistance in high temperature and pressure of CO2 in both cases of as annealed and after heat-treated state. No break away phenomenon is occurred after 2, 000 hr reaction with CO2 under pressure of 20 Kg/cm2 at 450°C.
  • X線回折による格子定数の測定
    鈴木 国雄, 丸谷 和夫, 久保田 正
    1963 年 5 巻 7 号 p. 587-592
    発行日: 1963/07/30
    公開日: 2009/03/26
    ジャーナル フリー
    Lattice-constant measurements by X-ray of the various UO2 pellets, dealt with in the 1st and 2nd reports, lead to the results:
    (1) The lattice constant of UO2+x is a=(5.4701-0.05x)±0.0002Å, where x ranges 00.09.
    (2) By being sintered at 1, 500°C for 2hr, UO2 pellets containing excess oxygen are reduced to UO2.00 in either dry or wet hydrogen, whereas in vacuo a minute amount of excess oxygen remains in the sintered pellets.
    (3) In UO2, TiO2 (or lower oxide) is soluble, BeO practically insoluble, Al2O3 completely insoluble, and MgO slightly soluble. With the last additive, the formation of solid solution depends largely upon the amount of excess oxygen of UO2.
  • 池亀 亮
    1963 年 5 巻 7 号 p. 593-600
    発行日: 1963/07/30
    公開日: 2009/03/26
    ジャーナル フリー
    The basic problems involved in the safety evaluation of power reactors were reviewed. The present days evaluating practice in Japan is considered satisfactory for the basic requirements in solving the sociological sides of reactor hazards. However, when we will proceed from subjective judgements on case-by-case basis, and give a direction to subjectivity of judgements, and further, when we intend to get a quantitative expression as site evaluation criteria, the most important problem involved will be the way how to express the empirical probability concept.
    In this paper, the safety evaluation procedures taken in over-seas countries were reviewed from this standpoint, and taking the peculiar conditions in Japan into consideration, it is attempted to find a direction in which we should proceed. The ground rules or the path to be followed in safety evaluation is proposed, to introduce a consideration of probability of more serious accidents into the present practice in Japan. But, research and development efforts in various areas are to be made before we could go further from the groud rules to the quantitative expression, i.e. the site evaluation criteria.
  • 向坊 隆, 内藤 奎爾
    1963 年 5 巻 7 号 p. 601-608
    発行日: 1963/07/30
    公開日: 2009/03/26
    ジャーナル フリー
    Although uranium carbides are becoming of considerable interest as nuclear fuel materials, few thermochemical data for these compounds are available and especially lack of specific heat data have made the assessment of these data unreliable. However, recently the authors made the measurement of specific heats on UC and UC2 above room temperature, and the assessment of experimental data bacame possible in some accuracy.
    In this paper, the specific heat functions for UC and UC2 were derived as a function of temperature, from which the heats of formation, free energies of formation and entropies of formation of these carbides were also calculated as a function of temperature respectively. Then the assessment of these thermochemical data was carried out on the values so for reported.
    In general, it was shown that the measurements at high temperatures by several authors gave a fairly good accordance with some scattering of data, which are not unreasonable in difficult experimental conditions. However, in order to make a more accurate estimation, especially for UC2, it is considered necessary to obtain further data precisely.
  • 1963 年 5 巻 7 号 p. 609-611
    発行日: 1963/07/30
    公開日: 2009/03/26
    ジャーナル フリー
  • 1963 年 5 巻 7 号 p. 611a
    発行日: 1963年
    公開日: 2009/03/26
    ジャーナル フリー
  • 1963 年 5 巻 7 号 p. 611b
    発行日: 1963年
    公開日: 2009/03/26
    ジャーナル フリー
  • Raymond L. MURRAY
    1963 年 5 巻 7 号 p. 612-616
    発行日: 1963/07/30
    公開日: 2009/03/26
    ジャーナル フリー
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