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Chitose, Hiromasa*; Hotta, Akitoshi*; Onuki, Akira; Fujimura, Ken*
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 6 Pages, 2006/07
no abstracts in English
Goto, Takehiro; Tsushima, Hiroyuki; Sakurai, Naoto; Jo, Takahisa
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 13 Pages, 2006/07
MONJU is a prototype fast breeder reactor. Modification work commenced in March 2005. Since June 2004, MONJU has changed one-loop operation of the primary heat transport system with all of the secondary heat transport systems drained of sodium. Purposes of this change are to shorten the modification period and to reduce the cost incurred for circuit trace heating electrical consumption. Before changing condition, the following issues were investigated to show that this mode of operation was possible. The heat loss from the reactor vessel and the single primary loop must exceed the reactor core decay heat by an acceptable margin but the capacity of the preheater to keep the sodium within the primary vessel at about 200
C must be maintained. With regard to heat loss and core decay heat, the estimated heat loss in the primary system was in the range of 90-170kW in one-loop operation, and the calculated reactor decay heat was 21.2kW. Although the heat input of the primary pump was considered, it was clear that circuit heat loss greatly exceeded the core decay heat. As for the preheater, effective capacity was less than the heat loss. Therefore, the temperature of the reactor vessel room was raised to reduce the heat loss. One-loop operation of the primary heat transport system was able to be executed by means of these measures. The cost of electrical consumption in the power plant has been reduced by one-loop operation of the primary heat transport system. The modification period was shortened.
Shen, X.*; Mishima, Kaichiro*; Nakamura, Hideo
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 8 Pages, 2006/07
Since the transport of momentum, heat and mass tightly links with local interfacial characteristics it is essential to know the local interfacial parameters in various two-phase flows. The interfacial velocity plays a determinant role in determining the other interfacial parameters such as the interfacial area concentration and so on. It is accordingly one of the most important parameters in analyzing two-phase flow. However, it also is one of the most difficult parameters to measure up to now. Based on the application of the interfacial measurement theorem to several four-sensor probes, the present study established a theoretical foundation of the measurement method for the local instantaneous interfacial velocity in multi-dimensional two-phase flow by using three independent four-sensor probes. Since we can find three independent four-sensor probes in a multi-sensor probe, which has more than four sensors, by sharing the sensors of the first four-sensor probe with the sensors of the others, a five- or six-sensor probe including at least one set of three four-sensor combinations was recommended to measure the local instantaneous interfacial velocity, interfacial area concentration and so on in multi-dimensional two-phase flow. A six-sensor probe was developed and employed in the practical measurement in an air-water multi-dimensional two-phase flow in a pool. The six-sensor probe measurements were checked against the gas flow rate measurement using a rotameter and a manometer. The comparing results were very satisfactory.
Ito, Kazuhiro*; Ito, Taro*; Kukita, Yutaka*; Koterazawa, Hiroyuki*; Kondo, Hiroo*; Yamaoka, Nobuo*; Horiike, Hiroshi*; Ida, Mizuho; Nakamura, Hideo; Nakamura, Hiroo; et al.
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 6 Pages, 2006/07
Waves on a liquid-lithium jet flow, simulating a proposed high-energy beam target design, have been measured using an optical technique based on specular reflection of a single laser beam on the jet surface. The streamwise and spanwise fluctuations of the local free-surface slope were least-square fitted with a sinusoidal curve to makeup the signals lost due to the constriction in the optical arrangement. The waveform was estimated with an assumption that wave phase speed can be calculated using the dispersion relation for linear capillary gravity waves. The direction of propagation on the jet surface was also evaluated so that the wave amplitudes, calculated by integral of slope angle signal, agree consistently in streamwise and spanwise direction. These measurements and analyses show that the waves at the measurement location for a jet velocity of 1.2 m/s can best be represented by oblique waves with an inclination of 0.32 rad, a wavelength of 4.2 mm and a wave amplitude of about 0.06 mm.
Iwatsuki, Jin; Terada, Atsuhiko; Noguchi, Hiroki; Imai, Yoshiyuki; Ijichi, Masanori; Kanagawa, Akihiro; Ota, Hiroyuki; Kubo, Shinji; Onuki, Kaoru; Hino, Ryutaro
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 8 Pages, 2006/07
JAEA has been conducting the HTTR project from the view to establishing technology base on HTGR and also on the IS process. Based on the test results and know-how obtained through the bench-scale tests, a pilot test plant that can produce hydrogen of about 30 Nm
/hr is being designed. The test plant will be fabricated with industrial materials such as glass coated steel, SiC ceramics etc, and operated under high pressure condition up to 2 MPa. The test plant will consist of a IS process plant and a helium gas (He) circulation facility (He loop). In parallel to the design study, key components of the IS process such as the sulfuric acid (H
SO
) and the sulfur trioxide (SO
) decomposers working under-high temperature corrosive environments have been designed and test-fabricated to confirm their fabricability. Also, other R&D's are under way such as corrosion, processing of HIx solutions. This paper describes present status of these activities.
Satou, Akira; Maruyama, Yu; Asaka, Hideaki; Nakamura, Hideo
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 8 Pages, 2006/07
Series of out-of-pile experiments to obtain the knowledge on the transient void behavior during reactivity initiated accidents are in progress at JAEA. In the present series of experiments, the transient void behavior in a test section of 2
2 bundle geometry under atmospheric pressure condition was measured using an impedance technique. The measuring areas and the arrangement of electrodes for the impedance technique were defined on the basis of numerical analyses and scaled model experiments. The comparison was made between the impedance and differential pressure techniques for steady boiling experiments to estimate the accuracy of the impedance technique. The impedance technique showed a good agreement with the void fraction estimated from the differential pressure. The transient void behavior in the bundle geometry was measured using the impedance technique. It was clarified that the transient void behavior depends on both the subcooling of inlet water and the heat generation rate of simulated fuel rod. Local void fraction was influenced by the ratio of flow area to heat transfer area of the simulated fuel rod.
Takeda, Tetsuaki; Yan, X.; Kunitomi, Kazuhiko
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 5 Pages, 2006/07
Japan Atomic Energy Agency (JAEA) has been developing the analytical code for the safety characteristics of the HTGR and carrying out design study of the gas turbine high temperature reactor of 300 MWe nominal-capacity for hydrogen production, the GTHTR300H (Gas Turbine High Temperature Reactor 300 for Hydrogen). The objective of this study is to clarify safety characteristics of the GTHTR300H for the pipe rupture accident. A numerical analysis of heat and mass transfer fluid flow with multi-component gas mixture has been performed to obtain the variation of the density of the gas mixture, and the onset time of natural circulation of air. From the results obtained in this analysis, it was found that the duration time of the air ingress by molecular diffusion would increase due to the existence of the recuperator in the GTHTR300H system.
Ito, Kazuhiro*; Ito, Taro*; Kukita, Yutaka*; Koterazawa, Hiroyuki*; Kondo, Hiroo*; Yamaoka, Nobuo*; Horiike, Hiroshi*; Ida, Mizuho; Nakamura, Hideo; Nakamura, Hiroo; et al.
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 6 Pages, 2006/07
no abstracts in English
Sumita, Junya; Hanawa, Satoshi; Shibata, Taiju; Tada, Tatsuya; Iyoku, Tatsuo; Sawa, Kazuhiro
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 8 Pages, 2006/07
An analytical study on micro-indentation method to integrity evaluation for graphite components was carried out. The indentation method is used as simplicity test to measure mechanical properties of materials. This method is thought to be applicable to evaluate the residual stress from the relationship between indentation load and indentation depth. In this study, in order to confirm the applicability of the micro-indentation method for lifetime evaluation of the graphite component, indentation load-depth behavior under stress/strain condition was evaluated taking account of the specified minimum ultimate strength of IG-110 graphite. Moreover, analytical investigations of indentation load-depth behavior for oxidized graphite and oxidized graphite with residual strain were also carried out. As a result, it can be said that the indentation method is potentially applicable to evaluate the integrity of graphite components.
Hanawa, Satoshi; Sato, Tomonori; Mori, Yuichiro; Ogiyanagi, Jin; Kaji, Yoshiyuki; Uchida, Shunsuke*
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 9 Pages, 2006/07
no abstracts in English
Toya, Yuichi; Washiya, Tadahiro; Koizumi, Kenji; Morita, Shinichi
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 6 Pages, 2006/07
Japan Atomic Energy Agency (JAEA) has been leading feasibility study on commercialized fast reactor cycle systems in Japan. In this study, we have proposed a new disassembly technology by mechanical disassembly system that consists of a mechanical cutting step and a wrapper tube pulling step. In the mechanical tool system, high durability mechanical cutter cuts the wrapper tube (Slit-Cut (S/C) operation in circle direction), and then the wrapper tube is pulled out and removed from the fuel assembly. Then the fuel pins are cut (Crop-Cut (C/C) operation at entrance nozzle side) and the entrance nozzle is removed. The fuel pins are transported to the shearing machine in next process. The Fundamental tests were carried out with simulated FBR fuel pins and wrapper tube, and cutting performance and wrapper tube pulling performance has been confirmed by engineering scale. As the results, we established the disassembly procedure and the fundamental design of mechanical disassembly system.
Suda, Kazunori; Watanabe, Akira*; Ohshima, Hiroyuki
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 9 Pages, 2006/07
The purpose of this paper is to study numerical simulation of sodium-water reaction (SWR) phenomena under small leakage condition in a steam generator of a sodium cooled fast reactor. In this study, a numerical simulation was carried out regarding an experimental result of sodium water reaction test rig (SWAT-1R) located at Japan Atomic Energy Agency (JAEA). In the simulation, void fraction that could not be measured in the experiment concerning hydrogen and water vapor, development of high temperature region, maximum temperature, and flow characteristics of sodium and reaction products were estimated. At the same time, simulated temperature distributions and leakage flow rate were compared with measured data in the experiment, and it was confirmed that SERAPHIM was able to predict the maximum temperature that is important to evaluate overheating rupture of adjoining hear transfer tubes.
Soga, Kazuo*; Niikura, Hideto*; Sugiyama, Kenichiro*; Narabayashi, Tadashi*; Ohshima, Hiroyuki; Suda, Kazunori
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 6 Pages, 2006/07
A series of experiments that investigate the entrainment process of ambient liquid toward jet interior are carried out by using a laser-sheet visualization and a void meter in water pool in the present work. It was observed that the entrainment of water into Ar gas jet is constantly caused in two regions just above the nozzle and just below the single rod. In the region just above the nozzle, negative pressure causes the entrainment of water. In the region below the rod, the entrainment of water is caused because the preceding Ar gas jet is caught up by the succeeding gas jet. The basic behavior of Ar gas jet causing the entrainment of water was confirmed to be almost same over the Reynolds number range of Ar gas jet, 2.17103 to 2.17104, in the present study.
Yamaguchi, Akira*; Takata, Takashi*; Ohshima, Hiroyuki; Suda, Kazunori
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 8 Pages, 2006/07
A sodium-water reaction in a counter-flow diffusion flame is studied mostly by numerical simulations. The numerical results are compared with an experimental observation. From the comparison of the numerical simulation and experimental observation, it has been found that the performance of the present simulation is satisfactorily accurate. Also, a new observation is that the reaction region and airborne particulates production region are separated in location, which could be explained with the numerical simulation results.
Shaaban, N.*; Masuda, Fukuzo*; Nasif, H.*; Yamada, Masao*; Sawamura, Hidenori*; Morota, Hidetsugu*; Sato, Satoshi; Iida, Hiromasa; Nishitani, Takeo
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 7 Pages, 2006/07
no abstracts in English
Takata, Takashi*; Yamaguchi, Akira*; Tanaka, Masaaki; Ohshima, Hiroyuki
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 8 Pages, 2006/07
Turbulent statistics near a structural surface, such as a magnitude of temperature fluctuation and its frequency characteristic, play an important role in damage progression due to thermal stress. A Large Eddy Simulation (LES) has an advantage to obtain the turbulent statistics especially in terms of the frequency characteristic. However, it still needs a great number of computational cells near a wall. In the present paper, a two-layer approach based on boundary layer approximation is extended to an energy equation so that a low computational cost is achieved even in a large-scale LES analysis to obtain the near wall turbulent statistics. The numerical examinations are carried out based on a plane channel flow with constant heat generation. The friction Reynolds numbers of 395 and 10,000 are investigated, while the Prandtl number (Pr) is set to 0.71 in each analysis. It is demonstrated that the present method is cost-effective for a large-scale LES analysis.
Shinohara, Kazunori; Okuda, Hiroshi*; Ito, Satoshi*; Nakajima, Norihiro; Ida, Masato
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 10 Pages, 2006/07
To decrease the fluid drag force on the surface of a specified object subjected to an unsteady flow, under a constant volume condition, the adjoint variable method is formulated by using FEM. Based on the Lagrange multiplier method (a conditional variational principle), this method consists of the state equation, the adjoint equation and the sensitivity equation. To solve the equations effectively using the steepest descent method, a parallel algorithm that finds the Armijo's line-search step size is constructed. The shape optimization code for solving a large scale 3D problem using a parallel algorithm was implemented on ITBL using the HPC-MW library. Results show that, by using shape optimization, the fluid drag force on the object can be reduced.
Washiya, Tadahiro; Kikuchi, Toshiaki*; Shibata, Atsuhiro; Chikazawa, Takahiro*; Homma, Shunji*
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 7 Pages, 2006/07
The crystallization is an advanced and remarkable technology in the future reprocessing process, which requires safety and cost advantages. Japan Atomic Energy Agency (JAEA), Mitsubishi Materials Corporation and Saitama University have been developing an annular-type continuous crystallizer. This paper mainly discussed about this crystallizer design and its development. JAEA has considered following two application processes of the crystallization technology. One is a uranium crystallization process, which applied before the solvent extraction process to recover excessive uranium from dissolver solution and reducing the throughput in the later extraction process. In this process, highly concentrated dissolver solution (about 500g-HM/L) is fed to this crystallizer, and only uranium is crystallized. Another is a plutonium co-crystallization process, which consists of two crystallization steps and excludes extraction process, and thus it's expected to reduce the waste generation and to improve operation safety. In this process, plutonium is co-crystallized with uranium in the first step and separated from residual solution, then the crystals are dissolved into nitric acid solution and excessive uranium is crystallized in the second step. This residual solution is recycled to fuel dissolution process, thus it contributes to reduce nitric acid quantity consumption. For both crystallization processes, same crystallizer design can be applied; we have developed a continuous crystallization system to establish high process throughput and optimizing of the crystallization processes. In the design study of the crystallizer, an annular-type was selected as the most promising design. The fundamental data was obtained by scale-down test device with uranium conditions, and an engineering scale crystallizer was fabricated to confirm the system performance in engineering scale.
Nishi, Yoshihisa*; Ueda, Nobuyuki*; Kinoshita, Izumi*; Miyakawa, Akira; Kato, Mitsuya*
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 10 Pages, 2006/07
CERES is plant system analysis code for LMRs developed by the Central Research Institute of Electric Power Industry (CRIEPI). CERES has a function of calculating multidimensional flow in the plena of a coolant in addition to that in one-dimensional plant network calculation. To verify the CERES code, analyses were performed by using the result of the plant trip test from the partial power operation of the prototype FBR "MONJU" that had been executed in December, 1995. The verification work was performed as a joint research of CRIEPI and JAEA. (1)Analysis concerning the primary/secondary/auxiliary cooling system (the plenum in the reactor vessel (R/V) is modeled in R-Z 2-dimension). (2)Analysis concerning the flow pattern in the plenum of R/V (the plenum is modeled in 3-dimension). (3)Analysis concerning the flow pattern inside the IHX plenum (the plenum in the IHX is modeled in 3-dimension). Analytical results by the CERES code showed good agreement with the results of the test of the "MONJU". Fundamental abilities of the CERES as a plant dynamics calculation code had been verified through these analyses. Additionally, some characteristic flows in plena of "MONJU" became clear by these analyses.
Nakamura, Yasuyuki; Kikuchi, Koichi; Morishita, Yoshitsugu; Usui, Tatsuo*; Ogane, Daisuke*
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 9 Pages, 2006/07
It is necessary to clarify the dismantlement method of 224 double tubes arranging both pressure and calandria tubes concentrically in the reactor as a peculiar problem of Fugen, in the case of phased dismantlement of the reactor. The machine type cutting is desirable, considering the influence on the atmospheres because the double tubes consist of the zirconium alloy and zircalloy material radio activated highly. Besides, Cutting method has long standoff to cut the double tubes at a time for to be short the term of dismantlement. is desirable. Therefore, it was examined to confirm the applicability to the double tubes cutting by abrasive water jet (hereinafter referred to as AWJ) as the machine type cutting method that can take the standoff comparatively longer. As a result, We confirmed for possibility of cutting the double tubes at a time from inside and outside tube, and cutting thick slab by abrasive water jet. Besides, We confirmed for relationship of abrasive supply and cutting velocity, properties of secondly waste.