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Journal Articles

Modeling and simulation of redistribution of oxygen-to-metal ratio in MOX

Hirooka, Shun; Kato, Masato; Watanabe, Masashi

Transactions of the American Nuclear Society, 118, p.1624 - 1626, 2018/06

This study suggested the time development of oxygen-to-metal ratio (O/M) redistribution model with oxygen-related properties in MOX. Irradiation simulation including the suggested O/M redistribution and pore migration with vaporization-condensation model which bares density redistribution was demonstrated. The simulation results showed that O/M redistribution proceeded at lower temperature than density redistribution, which indicated that oxygen diffusion got influential at lower temperature than vaporization-condensation of MOX. Another find was that O/M redistribution was very slow at the surface because temperature kept low. However, near the surface (inside from the surface) where the temperature exceeded 1000 K, O/M redistribution was rather recognizable with oxygen flown from inner region to the near-surface. The results will be evaluated by comparison with post-irradiation examination data.

Journal Articles

Effect of a raw material powder on sintered CeO$${2}$$ pellets by 28 GHz microwave irradiation

Akashi, Masatoshi; Matsumoto, Taku; Kato, Masato

Transactions of the American Nuclear Society, 118, p.1391 - 1394, 2018/06

In this study, CeO$${2}$$ pellet sintering by irradiating microwave at a frequency of 28 GHz was carried out to investigate the effect of particle diameter of raw powder on the density of sintered pellet. The highest bulk density is 94.2 %T.D. under the condition of 30 min holding at 1473 K. The bulk density decreases with increasing the particle diameter of used raw powder. On the other hand, all of the apparent density of sintered pellet is more than 93.5 %T.D.. The difference between the bulk density and the apparent density is caused by the difference of open porosity for each sample pellet. It seems that the high density sintered pellets with porous structure are obtained because sample pellet is heated internally and uniformly in microwave sintering.

Journal Articles

Towards enhancing Fukushima environmental resilience

Miyahara, Kaname

Transactions of the American Nuclear Society, 117(1), p.51 - 52, 2017/10

This presentation highlights JAEA's challenges for contributing to recovering the previous life of residents and the development of resilient communities in Fukushima Prefecture based on considering needs of local people on the environmental restoration categorized by the state of evacuation orders and the lifting of such orders.

Journal Articles

Validation study in SAS4A code in simulated mild TOP condition

Kawada, Kenichi; Suzuki, Toru

Transactions of the American Nuclear Society, 115(1), p.1597 - 1598, 2016/11

Journal Articles

The Research of MOX fuels in Japan

Kato, Masato

Transactions of the American Nuclear Society, 114, p.987 - 988, 2016/06

In Japan, uranium and plutonium mixed oxide (MOX) has been developed as fuels of sodium-cooled fast reactors. The developing MOX fuels come in variety of O/M ratio, Pu content, minor actinide (MA) content and density. We have studied a science based fuel technology to evaluate fuel behaviors in fabrication process and irradiation condition of such various fuels. The technologies which are constructed based on experimental database can apply to mechanistic evaluation of fuel behaviors. To develop the science based fuel technology, many different varieties of basic properties have been investigated, and experimental database was constructed. And a mechanistic physical property model has been studied. The models contribute to describe various behaviors in fuel fabrication process and irradiation condition.

Journal Articles

Oxygen potential measurement and point defect chemistry of UO$$_{2}$$

Watanabe, Masashi; Kato, Masato; Sunaoshi, Takeo*

Transactions of the American Nuclear Society, 114, p.1081 - 1082, 2016/06

Many studies on the oxygen potential of UO$$_{2}$$ have been carried out so far. However, the oxygen potential data for UO$$_{2}$$ near the stoichiometric composition in the high temperature region (1673-1873 K) are limited. In this work, the oxygen potential data of UO$$_{2+x}$$ were extended to high temperature range of 1673-1873 K by gas equilibrium method. The measured data were analyzed based on a defect chemistry model.

Journal Articles

Measurement of void fraction distribution in steam-water two-phase flow in a 4$$times$$4 bundle at 2 MPa

Liu, W.; Nagatake, Taku; Shibata, Mitsuhiko; Takase, Kazuyuki; Yoshida, Hiroyuki

Transactions of the American Nuclear Society, 114, p.875 - 878, 2016/06

To contribute to the clarification of the Fukushima Daiichi Accident, JAEA is working on getting instantaneous void fraction distribution data in steam water two - phase flow in rod bundle geometry under high pressure, high temperature condition, with using Wire Mesh Sensor (WMS) developed at JAEA for high pressure, high temperature condition, focusing on the low flow rate condition after the reactor scram. This paper reports the experimental results for the measured void fraction distribution in steam vapor two-phase flow in a 4 $$times$$ 4 bundle under 1.6 MPa (202 $$^{circ}$$C), 2.1 MPa (215 $$^{circ}$$C) and 2.6 MPa (226 $$^{circ}$$C) conditions. The data is expected to be used in the validation of the detailed two-phase flow codes TPFIT and ACE3D developed at JAEA. The time and space averaged void fraction data is also expected being used in the validation of the drift flux models implemented in the two fluids codes, such as TRACE code.

Journal Articles

Fuel restructuring behavior analysis of MA-bearing MOX fuels irradiated in a fast reactor

Ozawa, Takayuki; Ikusawa, Yoshihisa; Kato, Masato

Transactions of the American Nuclear Society, 113(1), p.622 - 624, 2015/10

A recycle system for minor actinides (MAs), in which MAs are recycled by reprocessing and irradiating them in a fast reactor, is studied to reduce the degree of hazard and the amount of high-level radioactive wastes. MAs would be used as mixed oxide (MOX) fuels with plutonium and uranium in fast reactors. Since MA content of MA-bearing MOX (MA-MOX) to be used in fast reactors is assumed to reach $$sim$$5 wt%HM, the effects on not only fuel properties but also fuel behaviors have to be estimated to use MA-MOX as fast reactor fuels. As the MOX fuels to be used will be irradiated at a comparably high linear power and the fuel center temperature would be assumed to be over 2,273 K during irradiation in the fast reactors, fuel restructuring would take place due to void migration towards the fuel center under the radial temperature gradient, and a central void would be formed. Since the fuel center temperature would be decreased by the effect of formation of the central void, the fuel restructuring is one of the most important behaviors for fast reactor fuels. In this study, the effect of MA content on fuel restructuring behavior was estimated from the results of irradiation experiments such as B11 and B14 performed in Joyo to study the irradiation behaviors of MA-MOX and the calculation results using a fuel restructuring model which can take into account MA-MOX dependence on vapor pressure.

Journal Articles

Physical property model for advanced oxide fuels

Kato, Masato; McClellan, K.*

Transactions of the American Nuclear Society, 113(1), p.613 - 614, 2015/10

A joint study on advanced oxide fuels is being carried out under the Civil Nuclear Energy Working Group (CNWG) bilateral collaboration between the U.S. Department of Energy and the Japan Atomic Energy Agency. The main goal of this study is to support development and validation of a science-based fuel analysis code for minor actinide (MA) bearing MOX fuel. In analysis and evaluation of fuel performance, it is essential to understand the physical properties of the advanced oxide fuels. Therefore, we are investigating physical properties of (U,Pu)O$$_{2}$$, (U,Ce,)O$$_{2}$$, PuO$$_{2}$$, CeO$$_{2}$$ and other related compounds to prepare a physical property database and to construct an integrated mechanistic physical property model. In this paper, we describe the derivation of a model to represent heat capacity and thermal conductivity of (U,Pu)O$$_{2-x}$$ that is based on the experimental database.

Journal Articles

Early-in-life fuel restructuring behavior of Am-bearing MOX fuels

Tanaka, Kosuke; Sasaki, Shinji; Katsuyama, Kozo; Koyama, Shinichi

Transactions of the American Nuclear Society, 113(1), p.619 - 621, 2015/10

In order to evaluate the microstructural change behavior of Am-MOX fuels at the initial stage of irradiation, detailed investigations using image analysis were performed on X-ray Computed Tomography (X-ray CT) images and on ceramographs from fuels irradiated in both B11 and B14.

Journal Articles

Sintering behavior of (U,Ce)O$$_{2}$$ and (U,Pu)O$$_{2}$$

Nakamichi, Shinya; Hirooka, Shun; Sunaoshi, Takeo*; Kato, Masato; Nelson, A.*; McClellan, K.*

Transactions of the American Nuclear Society, 113(1), p.617 - 618, 2015/10

Cerium dioxide has been used as a surrogate material for plutonium dioxide. Dorr et al reported the use of hyper-stoichiometric conditions causes the start of shrinkage of (U,Ce)O$$_{2}$$ at low temperature compared with the sintering in reducing atmosphere. However, the precise stoichiometry of the samples investigated was not controlled or otherwise monitored, preventing any quantitative conclusions regarding the similarities or differences between (U,Ce)O$$_{2}$$ and (U,Pu)O$$_{2}$$. The motivation for the present work is therefore to compare the sintering behavior of MOX and the (U,Ce)O$$_{2}$$ MOX surrogates under controlled atmospheres to assess the role of oxygen defects on densification in both systems.

Journal Articles

Revival of criticality safety research in Japan Atomic Energy Agency

Tonoike, Kotaro; Izawa, Kazuhiko; Sono, Hiroki; Umeda, Miki; Yamane, Yuichi

Transactions of the American Nuclear Society, 110(1), p.282 - 285, 2014/06

no abstracts in English

Journal Articles

Measurement of radioactive fragment production excitation functions of lead by 400 MeV/u carbon ions

Ogawa, Tatsuhiko; Morev, M. N.*; Kosako, Toshiso*

Transactions of the American Nuclear Society, 109(2, Part2), p.1253 - 1255, 2013/11

Depth distributions of radioactive fragments in a thick lead target exposed to 400 MeV/u carbon ions were measured to obtain isotopic production cross-sections of $$^{Nat}$$Pb($$^{12}$$C,x)X (X=$$^{46}$$Sc, $$^{48}$$V, $$^{54}$$Mn, $$^{56}$$Co, $$^{58}$$Co, $$^{59}$$Fe, $$^{75}$$Se, $$^{83}$$Rb, $$^{85}$$Sr, $$^{113}$$Sn, $$^{121}$$Te, $$^{127}$$Xe, $$^{133}$$Ba, $$^{139}$$Ce, $$^{143}$$Pm, $$^{144}$$Pm, $$^{146}$$Gd, $$^{148}$$Eu, $$^{149}$$Gd, $$^{172}$$Hf and $$^{175}$$Hf) reactions as excitation functions. The obtained fragment distributions were converted to excitation functions of fragmentation cross-sections by the modified stacked-foil method. This conversion procedure was validated by comparing the obtained data with the available thin target experimental data. The obtained cross-sections were in good agreement with the previously published results. Comparison of the obtained cross-sections and the simulation by PHITS showed that PHITS underestimates fragments lighter than 90 amu by factor of about 10 whereas the fragments heavier than 110 amu were predicted within a factor of 3. Energy and mass dependences of the obtained cross-sections give insight into the reaction mechanism and will be useful for radiation transport code benchmarking. Furthermore, this study clarified that excitation functions of fragmentation reactions induced by heavy ions can be obtained by applying the method adopted in this study.

Journal Articles

Calculation system for the estimation of decontamination effect

Satoh, Daiki; Kojima, Kensuke; Oizumi, Akito; Matsuda, Norihiro; Iwamoto, Hiroki; Kugo, Teruhiko; Sakamoto, Yukio*; Endo, Akira; Okajima, Shigeaki

Transactions of the American Nuclear Society, 109(1), p.1261 - 1263, 2013/11

A computer software, named CDE (Calculation system for Decontamination Effect), has been developed to support planning the decontamination. CDE is programed with VBA (Visual Basic for Applications), and runs on Microsoft Excel with a user friendly graphical interface. It calculates dose rate distributions in a target area before and after the decontamination from a radioactivity distribution and DF (Decontamination Factor), which is a ratio of original radioactivity to remaining one after the decontamination. DRRF (Dose Rate Reduction Factor) is also derived to express the decontamination effect. All the calculation results are visualized on an image of the target area with color map. Owing to its quick calculation speed, CDE is able to investigate the decontamination effect in various cases for a short period. This is very useful to establish a rational decontamination plan before an action.

Journal Articles

Simulating long-term $$^{137}$$Cs distribution on territory of Fukushima

Kitamura, Akihiro; Yamaguchi, Masaaki; Oda, Yoshihiro; Kurikami, Hiroshi; Onishi, Yasuo*

Transactions of the American Nuclear Society, 109(1), p.153 - 155, 2013/11

Long term $$^{137}$$Cs transport and its future distribution on the territory of Fukushima were predicted based on the USLE and the GIS. By modeling the soil erosion, transport, and deposition, we simulated the future distributions of air dose rates of $$^{137}$$Cs in mSv/h for 2, 6 and 21 years after the accident. The predictions made by METI were compared with the present results. The predictions of relatively high air dose rate areas were consistently matched between the two models over time. However, our model seemed to predict the decreasing rate of the $$^{137}$$Cs concentration with time to be slightly less than that of METI prediction. Some portions of the results obtained in the present study were used to provide influxes of sediments and $$^{137}$$Cs as boundary conditions and lateral inflows for the hydraulic river model.

Journal Articles

Preliminary calculation of sediment and $$^{137}$$Cs transport in the Ukedo River of Fukushima

Kurikami, Hiroshi; Kitamura, Akihiro; Yamaguchi, Masaaki; Onishi, Yasuo*

Transactions of the American Nuclear Society, 109(1), p.149 - 152, 2013/11

We applied the TOMAM model to the Ukedo River as a preliminary analysis to roughly understand what was important for cesium migration. The main lessons were as follows: Cesium migrates mainly in high river discharge conditions. Migration in a dissolved form is important in low river discharge conditions, while suspended sediments, especially silt and clay, are main carriers of cesium in high discharge conditions. Bed contamination is mainly reflected by sediment erosion and deposition instead of direct sorption in the riverbed.

Journal Articles

Computational modeling of radioactive contaminants in the Fukushima environment

Kitamura, Akihiro; Machida, Masahiko; Sato, Haruo; Nakayama, Shinichi; Yui, Mikazu

Transactions of the American Nuclear Society, 109(1), p.156 - 157, 2013/11

Computational modeling and simulating team of Fukushima Environmental Safety Center, Japan Atomic Energy Agency has been started to develop a number of mathematical models of radioactive contaminants on the land and rivers, lakes, and estuaries in Fukushima, as well as the basic studies of adsorption/absorption mechanism of Cs and soils. These predictions will be utilized for the dose assessment from the environmental contamination and the proposal of countermeasures to dispersion of contaminant. In this presentation we describe the outline of our current activities.

Journal Articles

Change of corrosion characteristics of SUS304L and Zircaloy-4 by an immersion test under the pre-heat treatment and constant potential

Yamashita, Shinichiro; Osaka, Masahiko

Transactions of the American Nuclear Society, 109(1), p.140 - 142, 2013/11

The decommissioning work of Fukushima Dai-ichi Nuclear Power Plant (1F-NPP) is intensively being promoted according to the Mid-and-long Term Roadmap towards the Decommissioning of 1F-NPP units 1-4. Many research and development works are underway in order to ensure the progress of the decommissioning work. One of the R&D programs relates to the removal of fuel assemblies (FAs) from the spent fuel pool (SFP), in which the seawater was injected as an emergency counter-measure for fuel cooling during the accident. For the anticipation of the long term integrity of irradiated FA structural materials that experienced a diluted seawater exposure part of a fundamental research plan was described to support the estimation of the long-term possible influences of corrosive factors on the component materials of FAs. In this study, whether or not acceleration of corrosion occurred was investigated by the immersion test under constant potential for SUS304L and Zircaloy-4.

Journal Articles

Development of prompt $$gamma$$-ray analysis using spallation neutrons at J-PARC

Toh, Yosuke; Ebihara, Mitsuru*; Harada, Hideo

Transactions of the American Nuclear Society, 109(1), p.116 - 117, 2013/11

no abstracts in English

Journal Articles

Sensitivity analysis of low-volatile FPs and Cm-244 inventory in irradiated nuclear fuel for special nuclear material accountancy in fuel debris

Sagara, Hiroshi; Tomikawa, Hirofumi; Watahiki, Masaru; Kuno, Yusuke

Transactions of the American Nuclear Society, 107(1), p.803 - 804, 2012/11

Fission products(FPs) such as Eu, Ce, Ru have low release ratio in case of core melting event in severe accident to co-exist inside fuel debris in oxide or metallic phase, based on TMI-2 experience and source term experiments. Passive $$gamma$$ spectroscopy of Ce-144/Pr-144 was utilized in quantifying special nuclear material in fuel debris of TMI-2 historically, and the same methodology might be applied to high energy $$gamma$$ emitter, Eu-154 and Ru-106. Passive neutron measurement has been also utilized for quantifying nuclear material practically. Different from conventional spent fuel, however, fuel debris would be lost in information of irradiated profile, release ratio of volatile FPs, and so on. In the present paper, sensitivity analysis of low-volatile FPs and Cm-244 inventory in irradiated nuclear fuel in light water reactor was performed numerically to clarify the uncertainty of low-volatile FP and Cm inventory regarding fuel irradiation parameters and calculation methodology, and to derive applicable index to quantify nuclear material in fuel debris.

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