日本原子力学会和文論文誌
Online ISSN : 2186-2931
Print ISSN : 1347-2879
ISSN-L : 1347-2879
5 巻 , 1 号
選択された号の論文の8件中1~8を表示しています
  • 杉山 均, 向井 秀明, 人見 大輔
    2006 年 5 巻 1 号 p. 1-13
    発行日: 2006/03/25
    公開日: 2009/04/21
    ジャーナル フリー
    Numerical analysis has been performed for three-dimensional developing turbulent flow in a rectangular duct with sharp 180-degree turn by using an algebraic Reynolds stress model. It is interesting point whether the present method can correctly predict the flow separation in rectangular duct, or not. In the numerical calculation, wall function method is modified by applying log law velocity used for roughened wall to separated flow region instead of usual log law velocity for smooth wall, because flow separation works as flow resistance. Calculated results are compared with the experimental data in order to examine the validity of the modified wall function method and an algebraic Reynolds stress model. As a result of this calculation, it was found that the present method with modified wall function could predict the reattachment location of flow separation more correctly than the usual wall function method. Besides, the present method could reproduce the streamwise velocity and the secondary flow pattern of the experiment except for cross section located in downstream of flow separation. Characteristic features of normal stresses are predicted qualitatively but the presented method has tendency to underestimate the experimental results. As for the modified wall function, calculated results suggest that changing log law velocity in the separation region is effective way to predict the flow separation accurately.
  • 佐藤 道雄, 師岡 慎一, 白川 健悦, 山本 泰, 渡部 和美, 新井 良一
    2006 年 5 巻 1 号 p. 14-24
    発行日: 2006/03/25
    公開日: 2010/03/08
    ジャーナル フリー
    This paper deals with behavior of liquid film on the fuel rod which is very important for the critical power prediction. In this study, the liquid film measurement device using an ultrasonic transducer has been developed and the liquid film thickness data has been obtained for a simulated BWR 4×4 rod bundle under 1MPa condition. The cooling fluid is steam-water mixture and flow direction is vertical. Also the following results were obtained. Firstly, the liquid film thickness becomes thinner with increasing quality and the liquid film thickness is about 0.2mm at 9.3% of quality. Secondary, the time change of liquid film thickness becomes smaller with increasing the quality. It was found that the change of liquid film thickness becomes more smoothly near the dryout condition.
  • 藤原 健一, 佐々木 朋三, 亀井 篤志
    2006 年 5 巻 1 号 p. 25-33
    発行日: 2006/03/25
    公開日: 2009/04/21
    ジャーナル フリー
    In this study, fluorination decontamination was investigated from the viewpoint of treating uranium bearing waste which is in a condition that is difficult to decontaminate like sludge. Since the clearance level of uranium bearing waste has not been defined yet in Japan, a radiation level of 0.3Bq/g which is recommended by IAEA TECDOC855 was chosen and a decontamination technique was developed to achieve this level. The principle of the fluorination decontamination technique is to fluorinate uranium to get gaseous UF6 and thus remove uranium from the uranium bearing waste. ClF3 was chosen as a fluorination decontamination gas because ClF3 can convert uranium of various chemical forms into UF6 and its fluorination reactivity is higher than other interhalogen compounds.
    In experiments, 1 g of uranium bearing sludge was put in a rotary kiln filled with ClF3. The experimental variables were decontamination temperature, duration of ClF3 treatment, revolution speed of the kiln, pretreatment grinding period of sludge, and weight of sludge in the kiln. The results showed the sludge could be decontaminated below the clearance level when treated at 550°C for 2 h. The residue of uranium increased proportionally to the weight of sludge. An empirical formula was derived to predict decontamination performance which includes experimental variables.
  • 石井 一弥, 伏見 篤, 日野 哲士, 丸山 博見, 井筒 定幸, 笹川 勝, 岩田 豊
    2006 年 5 巻 1 号 p. 34-44
    発行日: 2006/03/25
    公開日: 2010/01/21
    ジャーナル フリー
    For the purpose of verification of the nuclear analysis method of BWR for mixed oxide (MOX) cores, UO2 and MOX fuel critical experiments EPICURE and MISTRAL were analyzed using nuclear design codes HINES and CERES with ENDF/B nuclear data file. The critical keffs of the absorber worth experiments, the water hole worth experiments and the 2D void worth experiments agreed with those of the reference experiments within about 0.1%Δk. The root mean square differences of radial power distributions between calculation and measurement were almost less than 2.0%. The calculated reactivity worth values of the absorbers, the water hole and the 2D void agreed with the measured values within nearly experimental uncertainties. These results indicate that the nuclear analysis method of BWR in the present paper give the same accuracy for the UO2 cores and the MOX cores.
  • 高松 邦吉, 中川 繁昭
    2006 年 5 巻 1 号 p. 45-56
    発行日: 2006/03/25
    公開日: 2010/01/21
    ジャーナル フリー
    The HTTR (High Temperature Engineering Test Reactor), which has thermal output of 30MW, coolant inlet temperature of 395°C and coolant outlet temperature of 850°C/950°C is the first high temperature gas-cooled reactor (HTGR) in Japan. HTGRs have high inherent safety potential to accident condition. Safety demonstration tests using the HTTR are underway in order to demonstrate such excellent inherent safety features of HTGRs. The reactivity insertion tests demonstrate rapid increase of reactor power by withdrawing the control rod without operating the reactor power control system. The experimental results show the negative reactivity feedback effect and the slow temperature transient. A one-point core dynamics approximation with one fuel channel had applied to our analysis. It was found that the analytical results couldn't simulate accurately the reactor power behavior. This report proposes an original new method with fuel channels and temperature coefficients in some regions of the core. It is crucial that the analytical results using this method can simulate experimental values and excellent inherent safety features of HTGRs under accident conditions.
  • 栃尾 大輔, 角田 淳弥, 高田 英治, 藤本 望, 中川 繁昭
    2006 年 5 巻 1 号 p. 57-67
    発行日: 2006/03/25
    公開日: 2010/01/21
    ジャーナル フリー
    High Temperature Engineering Test Reactor (HTTR) of high temperature gas-cooled reactor at Japan Atomic Energy Agency (JAEA) achieved the reactor outlet coolant temperature of 950°C for the first time in the world at Apr. 19, 2004. To ensure the thermal integrity of fuel in high temperature test operation, it is necessary that fuel temperature is designed appropriately by fuel temperature designing method, and that estimated maximum fuel temperature is lower than the thermal limit temperature. In this report, by constructing newly a realistic core-shape representing model, the current fuel temperature estimation model is improved. Moreover fuel temperature in high-temperature test operation is estimated with the newly-constructed model, and it is confirmed that estimated maximum fuel temperature in high temperature test operation is lower than the thermal limit temperature.
  • 寺田 敦彦, 大田 裕之, 野口 弘喜, 小貫 薫, 日野 竜太郎
    2006 年 5 巻 1 号 p. 68-75
    発行日: 2006/03/25
    公開日: 2010/01/21
    ジャーナル フリー
    The Japan Atomic Energy Agency has been conducting R & D on high-temperature gas-cooled reactor (HTGR) technology and also on thermo-chemical water splitting hydrogen production technology by using an iodine-sulfur cycle (IS process) in the high temperature engineering test reactor (HTTR) project. The sulfuric acid (H2SO4) decomposer is one of the key equipments in the IS process, in which concentrated sulfuric acid is evaporated and decomposed into SO3and H2O with the heat of high temperature helium gas supplied by HTGR. A concept of the decomposer consisting of multiblock-type heat exchanger made of SiC ceramics was proposed, and its feasibility was examined by preliminary analyses of thermal-hydraulic and structural strength and test-fabrication of SiC block components as well as elementary tests of seal performance in SiC blocks and metal flanges.
  • 尾方 義人, 川崎 和男
    2006 年 5 巻 1 号 p. 76-80
    発行日: 2006/03/25
    公開日: 2010/03/08
    ジャーナル フリー
    The expected nuclear powered battery is the isotope battery, which applies photoconduction or radiant rays, in other words, releases electron. Because of its lightweight and constant stability and 80 years long (=half time of radiant rays energy) running span, this battery is very effective for artificial satellites, desert island, and artificial hears. For this small battery, production method, waste disposal method, and emergency risk management need to be designed as international standards. To make that possible, design of the battery should achieve universal consensus with its form and shape. True meaning of peace could be shared worldwide, this isotope battery will be international standard.
    And, On this research as Design Development of Thermoelectric Generator by Design Engineering, we used a "Design Activities Theory" in which the activities involved in Knowledge Management and Project Management aimed at ensuring the efficiency of that Knowledge Management function as a single system.
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