日本原子力学会和文論文誌
Online ISSN : 2186-2931
Print ISSN : 1347-2879
14 巻 , 4 号
選択された号の論文の5件中1~5を表示しています
総説
  • 吉田 一雄, 石川 淳, 阿部 仁
    2015 年 14 巻 4 号 p. 213-226
    発行日: 2015年
    公開日: 2015/11/15
    [早期公開] 公開日: 2015/10/07
    ジャーナル フリー
     An accident of evaporation to dryness caused by boiling of high-level liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, some amount of fission products (FPs) will be transferred to the vapor phase in the tank, and could be released to the environment. Therefore, the quantitative estimation of the transport and release behavior of FPs is one of the key issues in the assessment of the accident consequence. To resolve this issue, a systematic analysis method with computer codes has been developed on the basis of the phenomenological behavior in the accident of evaporation to dryness caused by boiling of HLLW. A simulation study demonstrated that the behavior of liquid waste temperature and the entrainment of mists were in good agreement with the experimental results during the early boiling stage, and that some issues to be resolved were pointed out for the estimation of the amount of transferred Ru to the vapor phase at the late boiling stage.
論文
  • 田代 信介, 天野 祐希, 吉田 一雄, 山根 祐一, 内山 軍蔵, 阿部 仁
    2015 年 14 巻 4 号 p. 227-234
    発行日: 2015年
    公開日: 2015/11/15
    [早期公開] 公開日: 2015/08/20
    ジャーナル フリー
     The release characteristics of Ru from highly active liquid waste (HALW) have been investigated under the condition of accidental evaporation to dryness by boiling of HALW. Using a laboratory-scale apparatus, the simulated HALW (s-HALW) was heated with an external heater to dryness to observe the release characteristics of Ru and gaseous nitrogen oxides. As a result, Ru was significantly released between 120 and 300℃ of the s-HALW. The cumulative release ratio of Ru was 0.088. It was also found that the partially released amount of Ru against the temperature of the s-HALW had two peaks at about 140℃ and about 240℃. Referring to the results of the release rate of gaseous nitrogen oxides and the volume of condensate, which was a collection of the mixed vapors of steam and nitric acid released from the s-HALW, we discussed the causes of Ru release around these peaks.
  • 栗原 成計, 梅田 良太, 菊地 晋, 下山 一仁, 大島 宏之
    2015 年 14 巻 4 号 p. 235-248
    発行日: 2015年
    公開日: 2015/11/15
    [早期公開] 公開日: 2015/09/15
    ジャーナル フリー
     A sodium-water reaction (SWR) would take place in the case of a breach of the heat transfer tube in the steam generator (SG) of a sodium-cooled fast reactor (SFR), and the reaction jet may cause wear to the neighboring tubes by thermal and chemical effects, which is so-called target wastage. Accordingly, failure propagation caused by target wastage may potentially reduce the secondary cooling system integrity. In a previous study, a great number of experiments on target wastage were carried out for candidate materials under practical SG operation conditions. The target wastage rate for cited materials was derived from macroscopic boundary factors of the reaction jet. However, this mock-up approach is not versatile and is not suitable for large-scale SG design. Therefore, target wastage should be focused on for the safety assessment of various SG designs. In this study, experimental apparatus and a technique for producing composite oxidation-type corrosion with flow (COCF), which is an integral part of target wastage, were developed to determine the separation effect of local wastage factors under the high-temperature sodium hydroxide (NaOH) and sodium monoxide (Na2O) environment mainly generated by the SWR. The authors quantitatively evaluated the effect of the material temperature and fluid velocity on the COCF rate. It was revealed that COCF produced sodium-iron composite oxidation-type corrosion from metallographic observation and element assay.
  • 稲垣 健太
    2015 年 14 巻 4 号 p. 249-260
    発行日: 2015年
    公開日: 2015/11/15
    [早期公開] 公開日: 2015/09/18
    ジャーナル フリー
     The development of analysis methods for severe accidents in nuclear reactors is a key issue for nuclear safety. It is difficult to estimate the behaviors of several phenomena in reactor accidents, such as the melting and relocation of structural materials, the spreading of corium on the ground, and the molten core concrete interaction (MCCI), because they involve large changes of the geometry. In the present study, a new method was developed to simulate these phenomena by using a moving particle semi-implicit (MPS) method with models for surface tension, rigid bodies, melting and freezing, heat conduction, interfacial heat transfer, and heat radiation. As benchmarks, the melting of a metal cylinder on a hot plate and the freezing of a molten metallic drop in a coolant are simulated. The characteristic behaviors in each experiment agreed well with the simulation results, which indicates that the developed method is applicable for simulation to evaluate the behavior of corium in severe reactor accidents.
技術資料
  • 酒谷 圭一, 中谷 隆良, 船橋 英之
    2015 年 14 巻 4 号 p. 261-267
    発行日: 2015年
    公開日: 2015/11/15
    [早期公開] 公開日: 2015/08/20
    ジャーナル フリー
     Corrosion rate data for activated metal wastes are necessary for the prediction of the radiological impact of radioactive waste disposal. However, there are no such data available in the consideration of Zr-2.5 wt%Nb alloy, which is used in pressure tubes of the Fugen Nuclear Power Plant. Since the pressure tubes are destined for sub-surface disposal, it is necessary to obtain the corrosion rate of Zr-2.5 wt%Nb alloy under the disposal conditions. In this study, corrosion tests were conducted in high alkalinity and deoxidized water at 30℃ by a gas-accumulating-type corrosion test. The corrosion rate decreased to 3.3-3.9 nm/y and the corrosion thickness increased in proportion to the cubic root of corrosion time until 24 months after the commencement of the test. The results indicate that the corrosion rate would decrease in proportion to the minus cubic root of corrosion time squared if the empirical corrosion characteristic that had been obtained in material research for light-water reactors is applicable.
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