日本原子力学会誌
Online ISSN : 2186-5256
Print ISSN : 0004-7120
ISSN-L : 0004-7120
29 巻, 1 号
選択された号の論文の11件中1~11を表示しています
  • 西川 雅弘, 渡辺 健二
    1987 年 29 巻 1 号 p. 2-10
    発行日: 1987/01/30
    公開日: 2010/01/08
    ジャーナル フリー
  • 佐藤 一男
    1987 年 29 巻 1 号 p. 11-14
    発行日: 1987/01/30
    公開日: 2009/04/21
    ジャーナル フリー
    The worst accident in the nuclear energy history happened at the Unit 4 of the Chernobyl Nuclear Power Station in the USSR on 26th April, 1986. The reactor and its building were severely damaged and a large amount of radioactive materials was released to the environment.
    In August, a post accident review meeting was held at the IAEA, Vienna, where the USSR presented the main feature of the reactor (RBMK-1000), accident cause, process and consequence. According to the USSR report, the direct cause of the accident was multiple serious violations of the operational rules by the operators which, combined with rather unique characteristics and design of the reactor, resulted in a very severe power excursion. In the present report, is outlined the accident and its consequence based on the USSR report at the above mentioned meeting.
  • 市川 龍資
    1987 年 29 巻 1 号 p. 15-17
    発行日: 1987/01/30
    公開日: 2009/04/21
    ジャーナル フリー
    The USSR experts have reported the results of their works on the effects of the Chernobyl plant's accident at IAEA meeting in August. Plant staffs and fire men with acute radiation syndrome were hospitalized and treated with special care including bone marrow transplantation. Whole population (135, 000) within the area of 30km radius from the plant evacuated during a few days after the accident. Collective dose to this population was estimated as 1.6×106 person rem.
  • 田坂 完二, 小泉 安郎
    1987 年 29 巻 1 号 p. 18-30
    発行日: 1987/01/30
    公開日: 2009/04/21
    ジャーナル フリー
    The Large Scale Test Facility (LSTF) of the Rig of Safety Assessment (ROSA)-IV Program is an integral test facility to investigate thermal-hydraulic response of a pressurized water reactor (PWR) system during small break loss-of-coolant accidents (LOCAs) and operational transients. The LSTF is volumetrically scaled at 1/48 and has the same height as the PWR. The construction of the LSTF was completed in May 1985 and several small break LOCAs, TMI-2 accident type simulation and natural circulation tests have been successfully conducted. Presented in the present report are outline of the test program at the LSTF and some test results including transient core liquid level depression during cold leg small-break LOCAs, break orientation effect in cold leg small-break LOCAs on the system transient and thermal-hydraulic phenomena during the TMI-2 type multifailure accident simulation test.
  • 徳永 弘倫
    1987 年 29 巻 1 号 p. 31-33
    発行日: 1987/01/30
    公開日: 2009/04/21
    ジャーナル フリー
  • 奈良 義彦, 阿部 康宏, 後藤 義則, 毛呂 達, 山口 友久
    1987 年 29 巻 1 号 p. 48-57
    発行日: 1987/01/30
    公開日: 2009/04/21
    ジャーナル フリー
    In the course of the development of the commercial Liquid Metal Fast Breeder Reactor (LMFBR), the mitigation of seismic load on building and components should be performed. This is required not only from the standpoint of compatibility of seismic design and high-temperature structural design, but also effective for cost reduction of the plant.
    We studied following two schemes on the reactor building of 1, 000MWe loop type LMFBR:
    (1) Application of seismic isolation
    (2) Estimation of effects of building embedment.
    The results showed that above two schemes were effective to mitigate the seismic load of reactor building and component.
  • 米沢 仲四郎, 星 三千男, アブドラ モハメッド, 比佐 勇, 山本 克宗
    1987 年 29 巻 1 号 p. 58-63
    発行日: 1987/01/30
    公開日: 2009/04/21
    ジャーナル フリー
    A method has been developed for the determination of ppb levels of U contents in reactor materials by neutron activation analysis. The method was applied to the determination of U in various materials, most of which are used in the JMTR (Japan Materials Testing Reactor). The sample specimens were irradiated in a nuclear reactor, together with U standard solutions. For Zircaloy, stainless steel, Al and demineralized water, analyses were carried out by measuring the γ-ray spectra of 239Np after being separated by TTA-xylene liquid-liquid extraction method. However the analyses of graphite and Be samples were carried out without the chemical separation. The analytical results were as follows, Zircaloy: 40-200ppb, stainless steel: less than 40ppb, graphite: less than 0.4ppb, demineralized water for JMTR: less than 0.5ppb, Al: 200-300ppb, Be: 10-40ppm. Agreements were satisfactory between the present results and the certified values in reference materials of Zircaloy.
  • 落合 政昭
    1987 年 29 巻 1 号 p. 64-71
    発行日: 1987/01/30
    公開日: 2009/04/21
    ジャーナル フリー
    Failure behavior of a waterlogged fuel rod is analyzed with WTRLGD, a computer code for the analyses of pressure behavior of the waterlogged fuel rod, in order to evaluate its failure threshold energy under a RIA condition of a power reactor. Parameters considered here are pulse width of a transient power, fuel stack length, volumetric fraction of water, total energy deposition and an axial power profile.
    The numerical results show that water in the gap region changes nearly iso-volumetrically When the fuel stack length is as long as that of a power reactor, that is, they show that the process of change is like that of a fully waterlogged fuel rod and that of a partially waterlogged one with the end peaks in the NSRR experiments. Additionally, the calculated failure threshold energies agree with those obtained in these experiments. It is also revealed that the calculated failure threshold energy of the waterlogged fuel rod with a long fuel stack is hardly influenced by the analytical parameters.
    Consequently, the failure threshold energy of the waterlogged fuel rod under a RIA condition of a power reactor is evaluated by the results of the NSRR experiments described above.
  • ノズルプレートとシーブプレート
    池田 秀松, 鈴木 篤之, 清瀬 量平
    1987 年 29 巻 1 号 p. 72-80
    発行日: 1987/01/30
    公開日: 2009/04/21
    ジャーナル フリー
    A pulsed perforated-plate column is controlled by varying the operating conditions as the pulsing conditions, the initial loading conditions, the A/O ratios and the flow rates of inlet and outlet streams. Changing these variables causes the holdup to vary with position in the column. As a result, it is necessary to know the axial holdup in order to obtain complete dynamic mass transfer equation.
    In the present paper, the axial holdup of dispersed phase in a pulsed perforated-plate cdumn, pulser feeder type, is studied. The column is made of Pyrex glass, 5cm I.D. and 100cm in actual length. Both nozzle plates and sieve plates with same cartridge geometries are used in the column. The liquid system is TBP in kerosene-water. The axial holdup data are obtained in the dispersed aqueous and the dispersed organic modes. Experimental results show that the plate-type affect the axial holdup distribution. The tendency of the axial holdup to the nonuniformity is counteracted by the nozzle plate.
  • 1987 年 29 巻 1 号 p. 84a
    発行日: 1987年
    公開日: 2010/01/08
    ジャーナル フリー
  • 1987 年 29 巻 1 号 p. 84b
    発行日: 1987年
    公開日: 2010/01/08
    ジャーナル フリー
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