日本原子力学会和文論文誌
Online ISSN : 2186-2931
Print ISSN : 1347-2879
ISSN-L : 1347-2879
3 巻 , 4 号
選択された号の論文の12件中1~12を表示しています
  • 文沢 元雄, 田中 学, 趙 紅, 菱田 誠, 椎名 保顕
    2004 年 3 巻 4 号 p. 313-322
    発行日: 2004/12/25
    公開日: 2010/01/21
    ジャーナル フリー
    This paper deals with numerical analysis of helium-air counter flow in a rectangular channel with an aspect ratio of 10. The channel has a cross sectional area of 5-50mm and a length of 200mm. The inclination angle was varied from 0 to 90 degree.
    The velocity profiles and concentration profiles were analyzed with a computer program [FLUENT]. Following main features of the counter flow are discussed based on the calculated results. (1) Time required for establishing a quasi-steady state counter flow. (2) The relationship between the inclination angle and the flow patterns of the counter flow. (3) The developing process of velocity profiles and concentration profiles. (4) The relationship between the inclination angle of the channel and the velocity profiles of upward flow and the downward flow. (5) The relationship between the concentration profile and the inclination angle. (6) The relationship between the net in-flow rate and the inclination angle.
    We compared the computed velocity profile and the net in-flow rate with experimental data. A good agreement was obtained between the calculation results and the experimental results.
  • 近藤 宏一, 吉田 憲司, 大川 富雄, 片岡 勲
    2004 年 3 巻 4 号 p. 323-330
    発行日: 2004/12/25
    公開日: 2010/01/21
    ジャーナル フリー
    Experiment and numerical calculation were made on the multi-dimensional behavior of upward gas-liquid bubbly flow in a vertical pipe with an axisymmetric sudden expansion. The variation of the void fraction distribution in the sudden expansion was measured at the different axial and radial positions using a point-electrode resistivity probe for various gas and liquid flow conditions. Based on the measured void fraction distribution in the sudden expansion, the axial profiles for the phase distribution parameters were evaluated in the sudden expansion. Numerical calculations were carried out to predict the cross-sectional averaged void fractions along the flow direction by using the one-dimensional two-fluid model considering the phase distribution parameter to verify the applicability of the computations. From these results, it concretely pointed out that some multi-dimensional modeling or modifications for numerical simulation would be needed and the distribution parameter was one of the most important parameters for one-dimensional two-fluid model to evaluate accurately the void fraction distriblltions affected by the multi-dimensional channel geometry.
  • 飛田 徹, 相澤 一也, 鈴木 雅秀, 岩瀬 彰宏
    2004 年 3 巻 4 号 p. 331-339
    発行日: 2004/12/25
    公開日: 2010/01/21
    ジャーナル フリー
    In addition to fast neutron, thermal neutron and gamma-ray also contribute to the irradiation induced embrittlement of reactor pressure vessel steels. Although there is a possibility that gamma-ray induces embrittlement more efficiently than fast neutron, the irradiation effect of gamma-ray has not been fully clarified. In this study, to simulate hardening by gamma-ray irradiation, comparative experiments on irradiation hardening by 2.5 MeV electrons and by neutrons were performed for Fe-Cu model alloys. During these irradiations, temperature and dose rate were accurately controlled. We obtained the dose dependence and the temperature dependence of irradiation induced hardening. The growth of Cu clusters with increase in irradiation dose was examined by using the small angle neutron scattering measurements. Although the electron irradiation hardening was initiated slightly earlier than that of the neutron irradiation, the differences in hardening between electron and neutron irradiations were very small on a displacement-per-atom (dpa) basis. The growth of Cu clusters with increase in irradiation dose was a principal cause of hardening, and it became saturated before the doses reached 10-3 dpa. The present results suggest that, from an engineering point of view, both the gammaray induced and neutron induced hardening can be well scaled by using dpa as an index of irradiation dose.
  • 瀬古 典明, 玉田 正男, 吉井 文男
    2004 年 3 巻 4 号 p. 340-345
    発行日: 2004/12/25
    公開日: 2010/01/21
    ジャーナル フリー
    Fibrous adsorbents having chelate agent such as amidoxime (AN/MAA-ad), iminodiacetic acid (IDA-ad), and phosphoric acid (HMPA-ad) were prepared by radiation-induced graft polymerization. AN/MAA-ad was synthesized by co-grafting of acrylonitrile and methacrylic acid. Then the introduced functional group of CN was converted to the amidoxime with chemical treatment. Grafting of glycidyl methacrylate and chemically introducing the IDA group synthesized IDA-ad. HMPA-ad was directly synthesized by grafting of phosphoric acid monomer. The densities of the chelate agency were 3.5mol/kg for AN/MAA-ad, 2.0mol/kg for IDA-ad and 2.0mol/kg for HMPA-ad after the grafting time of 2, 1.5 and 8h, respectively. HMPA-ad had 200 times higher distribution coefficient for uranium than that of a commercial adsorbent (Diaion-PK216) at pH 0.5. The distribution coefficient for AN/MAA-ad became 500 times higher than that of Diaion-CR11 more than pH 8. These chelate adsorbents are promising materials for removal of uranium from acid or alkali treated waste sludge contaminated uranium.
  • 河村 拓己, 堺 公明
    2004 年 3 巻 4 号 p. 346-353
    発行日: 2004/12/25
    公開日: 2010/03/08
    ジャーナル フリー
    The oxide electrolysis method is one of the candidates for fuel reprocessing in the fast breeder reactor cycle. In the molten salt electrolysis process, the optimization of deposit conditions and the amount evaluation of U, Pu deposits have been important subjects for feasibility of the system. Mass transportation dominates the deposit reaction rate in the molten salt electrolysis. The amount of deposits, therefore, is estimated by the Nernst's equation and the Fick.'s law. However, diffusion layer thickness, which shows the transportation characteristic of a reactant, is necessary to be treated as an empirical parameter which depends on hydro-dynamic conditions. This research examines the applicability of the nondimensional number correlations for mass transfer in molten salt electrolysis. The nondimensional number correlation is verified by using the numerical analysis for molten salt electrolysis.
    As a result, the prediction by the nondimensional number correlation agreed well with the numerical analysis. Therefore, it can be concluded that the diffusion layer thickness in molten salt electrolysis can be evaluated by using nondimensional number correlation.
  • 亀尾 裕, 原賀 智子, 中塩 信行, 星 亜紀子, 中島 幹雄
    2004 年 3 巻 4 号 p. 354-362
    発行日: 2004/12/25
    公開日: 2010/01/21
    ジャーナル フリー
    In order to investigate chemical stability of solidified products made from Low Level Radioactive Wastes (LLW) by plasma melting, a leaching test based on the MCC (Material Characterization Center)-3S Agitated Powder Leach Test Method was performed to determine Normalized Elemental Mass Loss (NLi) of both main components (Na, Al, Si, Ca, and Fe) of the solidified product and radioactive tracers (60Co, 137Cs, and 152Eu) incorporated into it. The results of leaching test indicated that NLi value was greatly affected by basicity defined as weight ratio of CaO to SiO2 in the solidified product, while effect of FeO concentration on NLi value was small. In the case of basicity less than 0.8, logarithm of NLi linearly increased with the basicity, implying that NLi value can be estimated by chemical composition of the solidified products.
  • 服部 隆利
    2004 年 3 巻 4 号 p. 363-368
    発行日: 2004/12/25
    公開日: 2010/03/08
    ジャーナル フリー
    On the monitoring for compliance with clearance level, the concentrations of objective nuclides, such as alpha or lowenergy beta emitters, can be estimated without direct gamma measurement by assuming the existence of objective nuclides with geometric mean concentrations or using previously assessed information on nuclide spectra and measurement results for a key gamma nuclide. To determine whether clearance can be carried out, deviations in the mean concentrations and nuclide ratios to the key gamma nuclide should be appropriately considered, in addition to the measurement error. In this paper, the concept of clearance level has been reconsidered and a new approach has been proposed to establish an appropriate safety factor of the monitoring for compliance with clearance level. To apply the approach to practical use, a probability distribution calculation system has been developed and verified. A simple method has also been proposed to determine whether the safety factor is required. The other option is to adopt initially a conservative value for the mean concentrations or the nuclide ratios. It has also been clarified that the use of arithmetic means for the mean concentrations or the nuclide ratios is more conservative than the present approach.
  • 高松 邦吉, 中川 繁昭
    2004 年 3 巻 4 号 p. 369-380
    発行日: 2004/12/25
    公開日: 2010/01/21
    ジャーナル フリー
    Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to demonstrate the excellent inherent safety features of the High Temperature Gas-cooled Reactors (HTGRs), to improve the safety design and to evaluate the technologies for HTGRs. The coolant flow reduction test by tripping one or two out of three gas circulators is one of the safety demonstration tests. The reactor power safely brings to a stable level without a reactor scram and the temperature transient of the in-core is very slow. Through the safety demonstration test, the temperature transient analysis code (TAC-NC code) is improved. This paper describes the validation of the TAC-NC code using the measured value of the test by tripping of one and two out of three gas circulators at 30% (9MW). By upgrading the code, the analytical model can evaluate accurately the transient temperature distribution of the in-core within 20°C during the test. Through the temperature transient analysis by the improved code, it was confirmed that the reactor became stable safely without fuel temperature rise during all gas circulators stop test planned as near future safety demonstration test. The result of this study can be applied to development for not only the commercial HTGRs but also the Very High Temperature Reactor (VHTR) such as one of the Generation IV reactors.
  • 西原 哲夫, 榊 明裕, 稲垣 嘉之, 高見 和男
    2004 年 3 巻 4 号 p. 381-387
    発行日: 2004/12/25
    公開日: 2010/01/21
    ジャーナル フリー
    Japan Atomic Energy Research Institute (JAERI) has been carried out research and development on the high temperature engineering test reactor (HTTR) hydrogen production system. One of the key components in the HTTR hydrogen production system is a high temperature isolation valve (HTIV) installed on the hot helium gas piping penetrating the reactor containment vessel. Angle valve with inner thermal insulator was selected for HTIV and conceptual design was performed. The structural integrity of HTIV was clarified by the stress analyses. Allowable helium leak rate of HTIV was discussed. Helium leak tests using small-scaled valve seat models were performed to decide the seat surface shape and valve closing force. The test results show that the leak rate of wedge shape seat increased in proportion to the number of simulated temperature and stress cycles loaded on the seat models before helium leak test, however that of flat seat did not depend on the number of cycles. So flat seat is adopted for HTIV. It was found that the seat closing force of 30MPa is reasonable to meet the allowable helium leak rate.
  • 坂場 成昭, 中川 繁昭, 古澤 孝之, 江森 恒一, 橘 幸男
    2004 年 3 巻 4 号 p. 388-395
    発行日: 2004/12/25
    公開日: 2010/01/21
    ジャーナル フリー
    Chemistry control is important for the helium coolant of High Temperature Gas-cooled Reactors (HTGRs) because impurities cause oxidation of the graphite used in the core and corrosion of high temperature materials used in the heat exchanger. In the High Temperature Engineering Test Reactor (HTTR) which is the first HTGR in Japan, the chemical impurity concentration is restricted and its behaviour is monitored during reactor operations. The impurity is reduced by the helium purification system and the concentration is measured by the helium sampling system installed to the primary and secondary helium system, continuously. This paper describes the impurity behaviour during the rise-to-power test which is the initial power-up of the HTTR. Also, the amount of the emitted impurity to the primary circuit from the graphite component and insulator used at the concentric hot gas duct are evaluated. During the power up, any abnormal impurity increases were not obtained and the chemical composition of the primary circuit was sufficiently in the stability area to avoid carbon deposition.
  • 米国原子力規制委員会による「前兆事象評価」結果に基づく分析
    渡辺 憲夫
    2004 年 3 巻 4 号 p. 396-406
    発行日: 2004/12/25
    公開日: 2010/01/21
    ジャーナル フリー
    The United States Nuclear Regulatory Commission (U.S.NRC) has been carrying out the Accident Sequence Precursor (ASP) Program to identify and categorize precursors to potential severe core damage accident sequences using the probabilistic safety assessment (PSA) technique. The ASP Program has identified a lot of risk significant events as precursors that occurred at U.S. nuclear power plants. Although the results from the ASP Program include valuable information that could be useful for obtaining and characterizing risk significant insights and for monitoring risk trends in nuclear power industry, there are only a few attempts to determine and develop the trends using the ASP results. The present study examines and discusses quantitative risk trends for the industry level, using two indicators, that is, the occurrence frequency of precursors and the annual core damage probability, deriving from the results of the ASP analyses. It is shown that the core damage risk at U.S. nuclear power plants has been lowered and the likelihood of risk significant events has been remarkably decreasing. As well, the present study demonstrates that two risk indicators used here can provide quantitative information useful for examining and monitoring the risk trends and/or risk characteristics in nuclear power industry.
  • 小澤 敏宏, 脇元 広海, 山岡 功, 青柳 春樹, 武内 豊, 荒川 秋雄, 中村 健二, 柳川 幸喜, 瀧澤 洋二, 中久木 功
    2004 年 3 巻 4 号 p. 407-416
    発行日: 2004/12/25
    公開日: 2010/01/21
    ジャーナル フリー
    The operator training simulator has been started to develop for the purpose of safety and stable operation in the Rokkasho reprocessing plant, the first commercial spent nuclear fuel reprocessing plant in Japan. At present, the safety training simulator, which had been developed for the purpose of the specified abnormal events training, has already been in operation. The simulator was designed to contribute to the acquisition of the knowledge and skills indispensable for the plant operators and to the promotion of good team-work among operators. It was also designed in consideration of the simple and consistent extensibility to the full-scope training simulator which is now planned to develop using the same hard-ware. The model building and maintenance tools were applied as trial for the model tuning and modification. They must be indispensable for the reflection of the plant operation data and for the planning full-scope training simulator development. The usefulness and the validity of the simulator were ascertained through the training and its reports as the simulator development successfulness.
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