Arai, Kenji*; Umezawa, Shigemitsu*; Oikawa, Hirohide*; Onuki, Akira*; Nakamura, Hideo; Nishi, Yoshihisa*; Fujii, Tadashi*
Nihon Genshiryoku Gakkai-Shi ATOMO, 58(3), p.161 - 166, 2016/03
no abstracts in English
Nakamura, Hideo; Arai, Kenji*; Oikawa, Hirohide*; Fujii, Tadashi*; Umezawa, Shigemitsu*; Abe, Yutaka*; Sugimoto, Jun*; Koshizuka, Seiichi*; Yamaguchi, Akira*
Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.5353 - 5366, 2015/08
Kawamura, Hideki*; Ando, Kenichi*; Noda, Masaru*; Tanaka, Tatsuya*; Matsuda, Takeshi*; Fujii, Haruhiko*; Hashimoto, Shuji*; Ueda, Tadashi*; Matsui, Hiroya; Takeuchi, Shinji; et al.
JAEA-Technology 2009-081, 182 Pages, 2010/03
Grouting has practical importance for the reduction of groundwater inflow into excavations during construction of underground facilities. Considering the performance assessment of a radioactive waste repository, the performance of the engineered barrier system could be adversely affected by a high pH plume generated from grout. Therefore, a quantitative estimation of the effectiveness of grouting and grout material is essential. This study has been performed in the Mizunami URL being excavated in crystalline rock as a part of the Project for Grouting Technology Development for the Radioactive Waste Repository funded by METI, Japan. The aims were to evaluate the applicability of existing grouting technology and to develop methodology to determine the distribution of grout and change in hydraulic properties of the grouted rock volume. The target rock is the volume of rock around a planned refuge niche where the pre-excavation grouting was performed at 200-m depth from ground surface. After excavation of the refuge niche, ten boreholes were drilled and different kinds of investigations were carried out during and after drilling. The results were integrated and groundwater flow analysis of pre and post excavation grouting conditions were carried out to estimate quantitatively the effect of pre-excavation grouting. The results suggest that current pre-excavation grouting technology is effective for reduction of groundwater inflow into excavations and that hydraulic conductivity of the surrounding rock can be reduced by more than one order of magnitude.
Yamasaki, Chisato*; Murakami, Katsuhiko*; Fujii, Yasuyuki*; Sato, Yoshiharu*; Harada, Erimi*; Takeda, Junichi*; Taniya, Takayuki*; Sakate, Ryuichi*; Kikugawa, Shingo*; Shimada, Makoto*; et al.
Nucleic Acids Research, 36(Database), p.D793 - D799, 2008/01
Here we report the new features and improvements in our latest release of the H-Invitational Database, a comprehensive annotation resource for human genes and transcripts. H-InvDB, originally developed as an integrated database of the human transcriptome based on extensive annotation of large sets of fulllength cDNA (FLcDNA) clones, now provides annotation for 120 558 human mRNAs extracted from the International Nucleotide Sequence Databases (INSD), in addition to 54 978 human FLcDNAs, in the latest release H-InvDB. We mapped those human transcripts onto the human genome sequences (NCBI build 36.1) and determined 34 699 human gene clusters, which could define 34 057 protein-coding and 642 non-protein-coding loci; 858 transcribed loci overlapped with predicted pseudogenes.
Chikazawa, Yoshitaka; Kisohara, Naoyuki; Hishida, Masahiko; Fujii, Tadashi; Konomura, Mamoru; Ara, Kuniaki; Hori, Toru*; Uchida, Akihito*; Nishiguchi, Yohei*; Nibe, Nobuaki*
JAEA-Research 2006-049, 75 Pages, 2006/07
In the feasibility study on commercialized fast breeder cycle system, a medium scale sodium cooled reactor with 750MW electricity has been designed. In this study, EMPs are applied to the secondary sodium main pump. The EMPs type is selected to be an annular linear induction pump (ALIP) type with double stators which is used in the 160m/min EMP demonstration test. The inner structure and electromagnetic features are decided reviewing the 160m/min EMP. Two dimensional electromagnetic fluid analyses by EAGLE code show that Rms (magnetic Reynolds number times slip) is evaluated to be 1.08 which is less than the stability limit 1.4 confirmed by the 160m/min EMP test, and the instability of the pump head is evaluated to be 3% of the normal operating pump head. Since the EMP stators are cooled by contacting coolant sodium duct, reliability of the inner structures are confirmed by temperature distribution and stator-duct contact pressure analyses. Besides, a power supply system, maintenance and repair feature and R&D plan of EMP are reported.
Fujii, Tadashi; Chikazawa, Yoshitaka; Konomura, Mamoru; Kamide, Hideki; Kimura, Nobuyuki; Nakayama, Okatsu; Ohshima, Hiroyuki; Narita, Hitoshi*; Fujimata, Kazuhiro*; Itooka, Satoshi*
JAEA-Research 2006-017, 113 Pages, 2006/03
A conceptual design study of the sodium-cooled fast reactor is in progress in the Feasibility Study on Commercialized Fast Reactor Cycle Systems. Reduced scale water experiments are being performed in order to clarify the flow pattern in the upper plenum of the reactor which has higher velocity condition than the past design. In this report, the hydraulic analyses of the water experiments using the general-purpose thermal hydraulic analysis program were executed; and the applicability to evaluation of flow pattern and vortex cavitations for the designed reactor was examined. (1) Steady-state analyses under the Froude number similar condition were carried out for the 1/10th reduced scale plenum experiments. Analyses results reproduced the characteristic flow patterns in the upper plenum, such as gushed flow from the inside of the upper internal structure to reactor vessel wall and the jet flow from the slit of the upper internal structure. Further, it was confirmed that the calculated flow pattern of a designed reactor system agreed with that of the water experiment qualitatively. Moreover, the influence which setting of numerical solution and boundary condition etc. in analyzing causes to flow pattern in the plenum became clear. (2) The distribution of the vortices under the dipped plate region in the 1/10th plenum model was evaluated using the prediction method of a submerged vortex which is based on the stretching vortex theory. In case of the same velocity condition as the reactor, it identified the two vortices which were sucked into the hot leg piping from the cold leg piping wall as the submerged vortex cavitations. From this analysis result, it confirmed that the submerged vortex cavitations, which may occur in the reactor upper plenum steadily, could be identified using this prediction method.
Hishida, Masahiko; Murakami, Tsutomu*; Kisohara, Naoyuki; Fujii, Tadashi; Uchita, Masato*; Hayafune, Hiroki; Chikazawa, Yoshitaka; Usui, Shinichi; Ikeda, Hirotsugu; Uno, Osamu; et al.
JAEA-Research 2006-006, 125 Pages, 2006/03
In Phase I of the "Feasibility Studies on Commercialized Fast Reactor Cycle Systems (F/S)", an advanced loop type reactor has been selected as a promising concept of sodium-cooled reactor, which has a possibility to fulfill the design requirements of the F/S. In Phase II, design improvement for further cost reduction and the establishment of the plant concept has been performed. In this study, reactor core design and large-scale plant design have been performed by adopting the modified fuel assembly with inner duct structure and double-wall straight tube steam generator (SG), which concepts were chosen at the interim review of FY 2003. For this SG, safety logics have been studied and the structural concept has been established. And the plant designs improving the in-service inspection (ISI) and repair capability have been performed. Furthermore, elaborate confirmation of the design has been performed reflecting the development of elemental technology, back-up concepts have been proposed. Besides, cost reduction measures have been studied by reducing reactor grade materials, introducing autonomous standardizations, simplifying the design due to deregulation and adopting systemized standards for BOP and NSSS. From now on, reflecting the results of elemental experiments, in-depth design studies and examination of critical issues will be carried out and the plant concept will accomplish in preparation for the final evaluation in Phase II.
Fujii, Tadashi; Konomura, Mamoru; Kamide, Hideki; Yamaguchi, Akira; Toda, Mikio*
Proceedings of 13th International Conference on Nuclear Engineering (ICONE-13), 0 Pages, 2005/05
An innovative sodium-cooled fast reactor (JSFR) has been investigated on the Feasibility Study on Commercialized Fast Reactor (FR) Cycle Systems. In order to reduce plant construction cost, JSFR adopts compacted reactor vessel and reduction of loop number. According to adoption of compacted cooling system, sodium flow velocities in the reactor upper plenum and the pipings of the cooling system exceed to those of conventional design, therefore, flow optimization in the reactor upper plenum and structural integrity of the piping system to flow-induced vibration (FIV) have been actualized as thermal-hydraulic issues. To solve above issues, some water experiments have been performed. The thermal-hydraulic design of the primary cooling system including the reactor vessel has been advanced reflecting these experimental results.
Fujii, Tadashi; Chikazawa, Yoshitaka; Konomura, Mamoru; Yamaguchi, Akira
Saikuru Kiko Giho, (26), 0 Pages, 2005/00
The conceptual design study of sodium cooled reactor is in progress in the Feasibility Study on Commercialized Fast Reactor Cycle Systems. The cooling system is composed of two loops in order to reduce plant construction cost. With loop number reduction, large diameter pipings are adopted and mean velocity in the piping also increases in comparison with former design. As for these piping systems, knowledge concerning hydraulic behaviors around the elbow and vibration phenomenon which is caused by the turbulence of fluid was insufficient. Therefore, flow-induced vibration tests have been started using water test facility, which simulates a hot leg piping of primary cooling system of large-scale reactor at the 1/3 reduced scale. Until now, flow visualization was conducted using acryl model and hydraulic behaviors such as velocity distributions in the piping were clarified. Further, as for the pressure fluctuations of the fluid which become vibration sources to the piping, it confirmed that the pressure fluctuations in the piping could be divided into four sections according to the degree of the turbulence caused at flow separation in the elbow.
Fujii, Tadashi; Chikazawa, Yoshitaka; Konomura, Mamoru
JNC TN9400 2004-046, 175 Pages, 2004/08
A conceptual design study of the sodium-cooled fast reactor (JSFR) is in progress in the Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S). The cooling system of the reactor is composed of two loops in order to reduce construction cost. According to reduction of loop number, large diameter pipings are adopted in the primary cooling system (for example, outer diameter of a hot leg piping is 1.27m). And the mean velocity in the piping increases to 9 m/s level. Then, flow visualization and flow-induced vibration tests using 1/3 scale water test facility, which simulates a hot leg piping, have been started in order to confirm the realization of the piping design for primary cooling system. As a first step of the test series, flow visualization test was conducted using acryl model piping, following results are obtained. (1)Flow pattern and velocity distributions in the model piping are independent of Reynolds number. The maximum velocity in the elbow is about 1.5 times of the mean velocity. (2)Total pressure loss coefficients of the elbow under high Reynolds number conditions (up to 3.3106), which are exceeded from past test condition, are about 1.3. According to this, the turbulence energy given from flow separation of the elbow is small. (3)As for pressure fluctuation behaviors, pressure fluctuations of the region of flow separation are dominance in the model piping. An eminent peak in the power spectrum densities of pressure fluctuations at the border of region of flow separation is revealed at 0.45 of Strouhal number. (4)According to extent of the turbulence caused by flow separation of the elbow, the power spectrum densities of pressure fluctuations and the correlation lengths, which are applied in random vibration response evaluation for actual plant pipings, are classified into some sections.
Murakami, Tsutomu; Hishida, Masahiko; Kisohara, Naoyuki; Hayafune, Hiroki; Hori, Toru; Fujii, Tadashi; Uchita, Masato; Chikazawa, Yoshitaka; Uno, Osamu; Saigusa, Toshiie; et al.
JNC TY9400 2004-014, 78 Pages, 2004/07
This report summarizes the results of the design study on the sodium-cooled large-scale reactor performed in JFY2003, which is the third year of Phase 2. In the JFY2003 design study, critical subjects related to safety, structural integrity and thermal hydraulics which found in the last fiscal year has been examined and the plant concept has been modified. Furthermore, fundamental specifications of main systems and components have been set and economy has been evaluated. In addition, as the interim evaluation of the candidate concept of the FBR fuel cycle is to be conducted, cost effectiveness and achievability for the development goal were evaluated and the data of the three large-scale reactor candidate concepts were prepared.
Hishida, Masahiko; Murakami, Tsutomu; Kisohara, Naoyuki; Fujii, Tadashi; Uchita, Masato; Hayafune, Hiroki; Chikazawa, Yoshitaka; Hori, Toru; Saigusa, Toshiie; Uno, Osamu; et al.
JNC TY9400 2004-012, 97 Pages, 2004/07
Based on the concept of a plant consisting of four modules with a capacity of 750 MWe each, which has been established by the end of FY2002, a concept of the entire plant was proposed, reflecting the modifications related to the high internal conversion type core, the double-wall straight tube steam generator (SG), and the fuel storage system. Concept studies were also performed to overcome the drawbacks of the sodium and to achieve in-service inspection and repair as easily as in light water reactor. Furthermore, feasibility studies were carried out to confirm the design, which included safety, thermal-hydraulics and the structures of the primary reactor auxiliary cooling system and the double-wall straight tube SG. A prospect for realization of this plant concept has been obtained through the evaluation results. In addition, as the interim evaluation of the candidate concepts of the FBR fuel cycle is to be conducted, cost effectiveness and achievability for the development goal were evaluated and the data of the three medium-scale reactor candidate concepts were prepared.
Sakai, Takaaki; Fujii, Tadashi; Hori, Toru; Konomura, Mamoru
paper No.550, 2004, 0 Pages, 2004/07
Rewetting velocity was measured for a wire-wrapped fuel rod to evaluate cooling ability for spent fuel direct transportation to a storage water pool in a Fast Breeder Reactor (FBR) future plant. In case of bottom flooding, the rewetting velocity of the wire-wrapped rod depended to the pressure loss of the flow system. The velocity was able to be correlated by experimental parameters. In case of top flooding, the rewetting velocity was reduced with increase its outlet flow restriction. The maximum time to rewet the heated length was 1500 sec. However, the maximum rod temperature did not increase during the rewetting time. In conclusion, it is prospective for cooling ability of the spent fuel direct transportation system in the FBR plant.
Konomura, Mamoru; Ogawa, Takashi; Okano, Yasushi; Yamaguchi, Hiroyuki; Murakami, Tsutomu; Takaki, Naoyuki; Nishiguchi, Youhei; Sugino, Kazuteru; Naganuma, Masayuki; Hishida, Masahiko; et al.
JNC TN9400 2004-035, 2071 Pages, 2004/06
The attractive concepts for Sodium-, lead-bismuth-, helium- and water-cooled FBRs have been created through using typical plant features and employing advanced technologies. Efforts on evaluating technological prospects of feasibility have been paid for these concepts. Also, it was comfirmed if these concepts satisfy design requierments of capability and performance presumed in the feasibilty study on commertialization of Fast Breeder Reactor Systems. As results, it was concluded that the selection of sodium-cooled reactor was most rational for practical use of FBR technologies in 2015.
Kisohara, Naoyuki; Hishida, Masahiko; Nibe, Nobuaki; Hori, Toru; Fujii, Tadashi; Uchita, Masato; Chikazawa, Yoshitaka; Saigusa, Toshiie; Uno, Osamu; Soman, Yoshindo; et al.
JNC TY9400 2003-015, 103 Pages, 2003/09
In Phase I of the "Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)", an advanced loop type reactor has been selected as a promising concept of sodium-cooled middle-scale modular reactor, which has a possibility to fulfill the design requirements of the F/S. This report summarizes the results of the design study on the sodium-cooled middle-scale modular reactor performed in JFY2002, which is the second year of Phase 2. The construction cost of the sodium-cooled middle-scale modular reactor, which has been constructed in JFY2002, was almost achieved the economical goal. But its achievability was not sufficient to accept the concept. In order to reduce the construction cost, the plant concept has been re-constructed based on the 50 MWe plant studied in JFY2002. After that, fundamental specifications of main systems and components for the new concept have been set, and critical subjects have been examined and evaluated. In addition, in order to achieve the further cost reduction, the plant with simplified secondary system, the plant with electric magnetic pump in secondary system, and the fuel handling system are examined and evaluated. As a result of this study, the plant concept of the sodium-cooled middle-scale modular reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000 yens/kWe, etc.) and has a prospect to solve the critical subjects. From now on, reflecting the results of elemental experiments, the preliminary conceptual design of this plant will be preceded toward the selection for narrowingdown candidate concepts at the end of Phase 2.
Kisohara, Naoyuki; Hishida, Masahiko; Nibe, Nobuaki; Hori, Toru; Fujii, Tadashi; Uchita, Masato; Chikazawa, Yoshitaka; Saigusa, Toshiie; Uno, Osamu; Soman, Yoshindo; et al.
JNC TY9400 2003-014, 52 Pages, 2003/09
In Phase 2 of the F/S, it is planed to precede a preliminary conceptual design of a sodium-cooled large-scale reactor based on the design of the advanced loop type reactor. Through the design study, it is intended to construct such a plant concept that can show its attraction and competitiveness as a commercialized reactor. This report summarizes the results of the design study on the sodium-cooled large-scale reactor performed in JFY2002, which is the second year of Phase 2. In the JFY2002 design study, critical subjects related to safety, structural integrity and thermal hydraulics which found in the last fiscal year has been examined and the plant concept has been modified. Furthermore, fundamental specifications of main systems and components have been set and economy has been evaluated.As a result of this study, the plant concept of the sodium-cooled large-scale reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000yens/kWe, etc.) and has a prospect to solve the critical subjects.
Fujii, Tadashi; Nishiguchi, Yohei; Konomura, Mamoru
JNC TN9400 2003-063, 73 Pages, 2003/08
The conceptual design study of the large-scale sodium-cooled reactor is in progress in the "Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)".The cooling system of a large-scale sodium-cooled reactor is composed of two loops in order to reduce construction cost. According to reduction of the loop number, the large diameter pipings are adopted in the primary cooling system (for example, inner diameter of a hot leg piping is 1.27m). And the average velocity in the piping increases to 9 m/s level, which is well over to a conventional plant design, therefore, Reynolds number reaches 10order levels. The hydraulic behaviors of the piping elbow, such as pressure fluctuation characteristics and formation range of flow separated layer near an elbow, under high velocity and high Reynolds number conditions are expected to be different to those under lower Reynolds number which is the highest existing number. Further, there would be the possibility of vibration of the piping by flow instability under high velocity condition. However, information of flow-induced vibration behaviors for large diameter piping is limited.Then, flow visualization and flow-induced vibration tests using 1/3 scale water test facility, which simulates hot leg piping, have been planned in order to confirm the realization of the piping design for primary cooling system. Flow pattern in the piping and pressure fluctuation near the elbow, which is a main cause of flow-induced vibration, will be measured in the flow visualization tests using the acrylic elbow model. Moreover, vibration modes and vibration response characteristics of the piping system will be measured in the flow-induced vibration tests using the stainless steel elbow model.The design of the test facility and production of the elbow models and the loop pipings were finished by 2002. In 2003, fabrication and installation of the test facility will be finished and a part of flow visualization tests will be started.
Matsuda, Makoto; Fujii, Yoshio*; Takeuchi, Suehiro; Yoshida, Tadashi
Dai-15-Kai Tandemu Kasokuki Oyobi Sono Shuhen Gijutsu No Kenkyukai Hokokushu, p.62 - 64, 2003/03
no abstracts in English
Matsuda, Makoto; Takeuchi, Suehiro; Yoshida, Tadashi; Hanashima, Susumu; Fujii, Yoshio*
Dai-14-Kai Kasokuki Kagaku Kenkyu Happyokai Hokokushu, p.170 - 172, 2003/00
no abstracts in English
Fujii, Tadashi; ; ; Sakai, Takaaki; ; Oki, Yoshihisa;
JNC TN9400 2002-049, 78 Pages, 2002/09
The conceptual design study of the large-scale sodium-cooled reactor is in progress in the "Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)". The direct water pool storage system is being examined as a candidate concept to simplify the fuel handling facility for the sodium-cooled reactor. In this concept, the decay heat of a fuel subassembly is relatively higher (18kW which is about 4.5 times of the Ex-vessel Storage Tank system in the demonstration FBR). Therefore, the information about the cooling characteristics of the fuel subassembly are lacking in cases of submergence process at a normal operation and cooling water injection process from upper part of the subassembly at a transfer accident. Accordingly, the understanding of the cooling characteristics of the fuel subassembly in higher decay heat condition was pointed out as one of the thermal hydraulic problems which influence the realization of the plant concept. Using the single heater pin equipment, fundamental tests were conducted with the parameters of the thermal conditions of a fuel pin, the outlet shapes of it, the submergence speeds and so on. Then, following basic data were acquired to be reflected in the actual plant design. (1)Cooling modes of the normally submergence tests and water injection tests were identified by visualization of the boiling behavior in the test section and the temperature change of the heater pin. (2)The initial temperature of the heater surface and the blockage size of the outlet of test section were dominating factors to the cooling completion time. (3)Maximum temperature rise of the heater surface was about 4K in normally submergence tests and 6K in water injection tests, respectively. Therefore, the heater was well cooled without significant temperature rise.(4)In the normally submergence tests, the pressure of the upper part of the test section did not exceed the lower part pressure and a water level rise in the test section was not obstructed ...