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Journal Articles

2016 Professional Engineer (PE) test preparation course "Nuclear and Radiation Technical Disciplines"

Takahashi, Naoki; Yoshinaka, Kazuyuki; Harada, Akio; Yamanaka, Atsushi; Ueno, Takashi; Kurihara, Ryoichi; Suzuki, Soju; Takamatsu, Misao; Maeda, Shigetaka; Iseki, Atsushi; et al.

Nihon Genshiryoku Gakkai Homu Peji (Internet), 64 Pages, 2016/00

no abstracts in English

Journal Articles

A Summary of sodium-cooled fast reactor development

Aoto, Kazumi; Dufour, P.*; Hongyi, Y.*; Glats, J. P.*; Kim, Y.-I.*; Ashurko, Y.*; Hill, R.*; Uto, Nariaki

Progress in Nuclear Energy, 77, p.247 - 265, 2014/11

 Times Cited Count:99 Percentile:99.52(Nuclear Science & Technology)

Much of the basic technology for the Sodium-cooled fast Reactor (SFR) has been established through long term development experience with former fast reactor programs, and is being confirmed by the Ph$'e$nix end-of-life tests, the restart of Monju, the lifetime extension of BN-600 and the startup of CEFR. Planned startup in 2014 for BN-800 and PFBR will further enhance the confirmation of the SFR basic technology. Nowadays, the SFR development has advanced to aiming at establishment of the Generation-IV system which is dedicated to sustainable energy generation and actinide management, and several advanced SFR concepts are under development. Generation-IV International Forum is an international collaboration framework where various R&D activities are progressing for the Generation-IV SFR development, and will play a beneficial role of promoting them thorough providing an opportunity to share the past experience and the latest data of design and R&D among countries developing SFR.

Journal Articles

The Screening methodologies and/or achievement evaluation in Japanese FR cycle development program with the changing needs for evaluation

Shiotani, Hiroki; Uto, Nariaki; Kawaguchi, Koichi; Shinoda, Yoshihiko*; Ono, Kiyoshi; Namba, Takashi

Proceedings of 20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference (ICONE-20 & POWER 2012) (DVD-ROM), 10 Pages, 2012/07

This paper argues the characteristics evaluation of Fast Reactor and fuel cycle concepts in the FS "Feasibility Study on commercialized fast reactor cycle systems" and the achievement of the performance evaluation conducted in FaCT (Fast Reactor Cycle System Technology Development) project in Japan. The methodologies and way of achievement evaluation has been changed according to the evaluation needs and objectives, etc. Some decision-making methodologies are tried to be applied in the FS, FaCT phase I evaluation put emphasis on the confirmation of the direction of FR cycle development. Although some items of respective facilities showed insufficient achievements because of the challenging design requirements to achieve higher performance, a comprehensive evaluation determined that the performance criteria set by the Japan Atomic Energy Commission were achieved in FaCT phase I evaluation in general.

Journal Articles

Thermal analysis on shipping cask for JSFR fresh fuel

Kato, Atsushi; Chikazawa, Yoshitaka; Uto, Nariaki; Hirata, Shingo; Obata, Hiroyuki*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 12 Pages, 2012/00

A basic feasibility of the helium gas cask has been evaluated by thermal analyses. There have been conducted two analyses: whole cask and detail inside subassembly analyses. The detail inside subassembly analysis has shown that the temperature distribution is mainly governed by thermal conductivity and natural convection of coolant helium hardly contributes heat removal. In the case of a cask with five subassemblies with 2.2 kW decay heat per each, the maximum cladding temperature is evaluated to be 361 $$^{circ}$$C satisfying cladding temperature limit of 395 $$^{circ}$$C. Those results have shown the basic feasibility of the helium gas fresh fuel shipping cask.

Journal Articles

Development of transfer pot for JSFR ex-vessel fuel handling

Hirata, Shingo; Chikazawa, Yoshitaka; Kato, Atsushi; Uto, Nariaki; Obata, Hiroyuki*; Kotake, Shoji*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00

In Fast Reactor Cycle Technology Development (FaCT) project, Japan Sodium-cooled Fast Reactor (JSFR) is going to adopt an advanced fuel handling system. As for ex-vessel spent fuel handling, a pot which contains two fuel subassemblies simultaneously and is applicable size to compact reactor vessel, has been developing so as to shorten a refueling period leading to an improvement of plant availability. The pot is required to provide sufficient cooling capability even in case of transportation malfunction during transportation of spent fuel subassemblies with high decay heat. In the present study, experimental and analytical studies to evaluate the cooling capacity of the pot are summarized.

Journal Articles

JSFR design study and R&D progress in the FaCT project

Aoto, Kazumi; Uto, Nariaki; Sakamoto, Yoshihiko; Ito, Takaya*; Toda, Mikio*; Kotake, Shoji*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 11 Pages, 2012/00

In the FaCT project, SFR with 1,500 MWe is a target for the commercialization. R&D on innovative technologies to achieve the economic competitiveness and enhance the reliability and safety is carried out. A compact RV without wall-cooling layer is pursued in consideration of seismic reliability. For a two-loop cooling system with shortened high chromium steel piping, studies on the hydraulics in the pipe elbow and the fabrication capability of the pipes are being performed. A double-walled straight tube SG is investigated to enhance the reliability against sodium/water reaction, and developmental works are progressing including the thermal-hydraulic design and trial manufacturing for components. SASS is being developed with safety analysis of the applicability for JSFR and experimental demonstration in JOYO. An advanced fuel handling system is also pursued. Discussion on whether the innovative technologies can be adopted for JSFR is in progress to be finalized in 2010.

Journal Articles

Design study and R&D progress on Japan sodium-cooled fast reactor

Aoto, Kazumi; Uto, Nariaki; Sakamoto, Yoshihiko; Ito, Takaya*; Toda, Mikio*; Kotake, Shoji*

Journal of Nuclear Science and Technology, 48(4), p.463 - 471, 2011/04

In the FaCT project, SFR with 1,500 MWe is a target for the commercialization. R&D on innovative technologies to achieve the economic competitiveness and enhance the reliability and safety is carried out. A compact RV without wall-cooling layer is pursued in consideration of seismic reliability. For a two-loop cooling system with shortened high chromium steel piping, studies on the hydraulics in the pipe elbow and the fabrication capability of the pipes are being performed. A double-walled straight tube SG is investigated to enhance the reliability against sodium/water reaction, and developmental works are progressing including the thermal-hydraulic design and trial manufacturing for components. SASS is being developed with safety analysis of the applicability for JSFR and experimental demonstration in JOYO. An advanced fuel handling system is also pursued. Discussion on whether the innovative technologies can be adopted for JSFR is in progress to be finalized in 2010.

Journal Articles

Development of the JSFR fuel handling system and mockup experiments of fuel handling machine in abnormal conditions

Kato, Atsushi; Hirata, Shingo; Chikazawa, Yoshitaka; Uto, Nariaki; Obata, Hiroyuki*; Kotake, Shoji*; Uzawa, Masayuki*

Proceedings of 2010 International Congress on Advances in Nuclear Power Plants (ICAPP '10) (CD-ROM), p.692 - 699, 2010/06

In the JSFR design, a single rotating plug and an upper inner structure (UIS) with a vertically penetrating slit are proposed, so that the fuel handling machine (FHM) can access any subassembly by horizontal movement of the FHM arm in the slit space. As a result of a full-scale mockup test, excellent performance in normal operation has been shown. In this study, from the viewpoint of achieving reliability of the pantograph type FHM, behavior of the FHM mockup have been investigated under abnormal conditions.

Journal Articles

Development of advanced loop-type fast reactor in Japan

Kotake, Shoji; Sakamoto, Yoshihiko; Mihara, Takatsugu; Kubo, Shigenobu*; Uto, Nariaki; Kamishima, Yoshio*; Aoto, Kazumi; Toda, Mikio*

Nuclear Technology, 170(1), p.133 - 147, 2010/04

 Times Cited Count:35 Percentile:91.18(Nuclear Science & Technology)

Japan Atomic Energy Agency (JAEA) is now executing "Fast Reactor Cycle Technology Development (FaCT)" project in cooperation with the Japanese electric utilities. In the FaCT project, both the conceptual design study for Japan Sodium-cooled Fast Reactor (JSFR) and the developments of innovative technologies to be adopted to JSFR are now implemented with paying attention to the consistency between the design and the innovative technologies. The current target is that the development will be accomplished around 2015, after that a licensing procedure for the demonstration JSFR will be launched. This paper describes design requirements, design characteristics of JSFR and evaluation on the performances for economic competitiveness. Furthermore, the current status of the key technology development for JSFR is briefly introduced.

Journal Articles

Conceptual design for Japan sodium-cooled fast reactor, 3; Development of advanced fuel handling system for JSFR

Kato, Atsushi; Hirata, Shingo; Chikazawa, Yoshitaka; Uto, Nariaki; Obata, Hiroyuki; Kotake, Shoji

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9281_1 - 9281_6, 2009/05

One of the most important challenges to commercialize a Fast Reactor is to increase economic competitiveness. For that purpose, Japan Sodium cooled Fast Reactor (hereafter JSFR) aims to simplify the plant system and reduce the raw and processed material by adopting innovative technologies. In the JSFR design, a single rotating plug and a reactor upper inner structure (hereafter UIS) with a vertically penetrating slit are proposed, so that the fuel handling machine (hereafter FHM) can access any subassembly by horizontal movement of the FHM arm in the slit space. The feature of this FHM enables no need for the UIS removal when the rotational plug moves round above the core, which can achieve a compact reactor vessel to enhance the economic competitiveness. We fabricated the full scale FHM test equipment to perform comprehensive tests in the air for demonstrating the feasibility of the key characteristics of this FHM concept.

Journal Articles

Conceptual design for Japan Sodium-cooled Fast Reactor, 1; Current status of system design for JSFR

Uto, Nariaki; Sakai, Takaaki; Mihara, Takatsugu; Toda, Mikio*; Kotake, Shoji; Aoto, Kazumi

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9298_1 - 9298_11, 2009/05

A conceptual design for JSFR and developments of innovative technologies are implemented. A compact RV has been designed to enhance the economy. The regarding development results have been reflected to the RV design. An innovative CV design has been implemented with elemental tests to reduce the construction cost. SASS and the NC DHRS have been designed to enhance the safety, with the irradiation data acquired in Joyo and the development of a 3-dimensional thermal-hydraulic evaluation method. An approach for ISI/R has been provided to be applicable for FR characteristics, and the developmental studies on innovative inspection technologies have been progressed. Other technologies including double-walled pipes with short elbows, a pump-integrated IHX are also being developed. These results, together with a preliminary conceptual design study on a demonstrative reactor for JSFR, will be utilized as resources in 2010 to determine which innovative technologies should be adopted.

JAEA Reports

Design studies on small fast reactor cores, 5; Research results in JFY2005

Uto, Nariaki; Okano, Yasushi; Naganuma, Masayuki; Mizuno, Tomoyasu; Hayashi, Hideyuki

JAEA-Research 2006-060, 68 Pages, 2006/09

JAEA-Research-2006-060.pdf:3.98MB

A design study on "Long-life Type Concept" of a 50MWe sodium-cooled metal-fueled reactor core was performed with more emphasis on irradiation results regarding fuel smear density. The concept aims at no refueling in a core life time, and achieving higher core outlet temperature such as 550$$^{circ}$$C which is advantageous to hydrogen production. The restriction of upper fuel smear density limit to 75% along with adjustments of fuel specifications showed feasibility of attaining core life time of 30 years and core outlet temperature of 550$$^{circ}$$C. No indication of occurrence of absorber-cladding mechanical interaction (ACMI) was found in the evaluation of ACMI for a control rod element. A shielding with Zr-H was selected in view of enhancement of shielding performance, and the feasibility was shown to satisfy the target allowance level of the ratio of hydrogen to zirconium, more than 1.53, with PNC316 used as the cladding material.

JAEA Reports

Design Study on BN-600 Hybrid Core (II) -Evaluation of Fuel Integrity and Core Neutronic Characteristics by Japanese Analysis Methods-

Sugino, Kazuteru; Uto, Nariaki; Naganuma, Masayuki; Mizuno, Tomoyasu

JNC TN9400 2004-042, 55 Pages, 2004/08

JNC-TN9400-2004-042.pdf:2.08MB

A program of disposal of Russian surplus weapon-grade plutonium by containing the plutonium in vibropacked MOX fuel subassemblies and burning them in the BN-600 hybrid reactor core has been progressed. The relevant design works on the BN-600 hybrid core have been carried out under the contract between Japan Nuclear Cycle Development Institute (JNC) and OKB Mechanical Engineering (OKBM), Russian public enterprise. JNC obtained a series of design technical reports. Japanese analysis methods were adopted to evaluate fuel integrity in the design basis transients and neutronic characteristics of the BN-600 hybrid core, based on the design technical data described in the obtained reports. The evaluation results of the key performances, such as maximum cladding and fuel temperatures, coolant (sodium) void reactivity, reactivity coefficient, were found to satisfy the design criteria and/or target provided by Russia, and meet the Russian rule. The results of this study showed that the core and fuel specifications determined by Russia can be considered reasonable and proper from the viewpoint of safety and neutronic designs, and that the Japanese analysis methods are expected to contribute to increasing reliability of the Russian design works.

JAEA Reports

Design Studies on Small Fast Reactor Cores(III)

Sanda, Toshio; Okano, Yasushi; Takaki, Naoyuki; Naganuma, Masayuki; Uto, Nariaki; Mizuno, Tomoyasu

JNC TN9400 2004-031, 154 Pages, 2004/06

JNC-TN9400-2004-031.pdf:10.75MB

Some concepts of small fast reactors have been studied as part of the "Feasibility Studies on Commercialized Fast Reactor Cycle System (FS)", and the core design study has been performed at two main features of "long-life core " and "enhanced passive safety" in the FS phase II. Based on the previous study, 165MWe forced circulation sodium cooled reactor with control rods was studied as the promising concept from a viewpoint of economical efficiency in JFY 2003. In the present study, the fuel reloading interval of 20 years and outlet temperature of 550 deg-C are targeted under following condition as thicker metal fuel pin diameter (less than or equal) 15mm, lower pressure drop (less than or equal) 0.75kg/cm2, and smller core diameter (less than or equal) 3m by sodium void reactivity restriction relief into design conditions avoiding core melt without SASS at ATWS. The prospect of achievement of the fuel reloading interval of 20 years and outlet temperature of 550 deg-C was acquired for "Higher Temperature Core" and "Higher Temperature and Smller Core" without blanket fuels by using a sodium-cooled metal-fueled core with single Pu enrichment fuel which has high potential of small change of space distribution of power density and high breeding ratio. These cores have core height / diameter of 127/293cm and 164/260cm, fuel burnup of 77 and 80 GWd/t, burnup reactivity of 1.2 and 1.5% (delta)k/kk', breeding ratio of 1.06 and 1.07 and coolanat void reactivity of 6 and 8${$}$, respectively. Control rod reactivity balance, fuel soundness and shielding performance were checked that these were satisfied. Moreover, since the reactivity change due to burnup was small, the possibility of long-term operation which does not require a control rod movement was also examined. In addition, the "Higher Temperature Core" was recommended for a promising core of phase-II middle time since core melt would be avoided without SASS at ATWS. Furthermore, the applicability of the Feher heat cycl

JAEA Reports

Design Studies on Small Fast Reactor Cores (II)

Takaki, Naoyuki; Uto, Nariaki; Mizuno, Tomoyasu

JNC TN9400 2003-066, 110 Pages, 2003/07

JNC-TN9400-2003-066.pdf:5.15MB

Core design studies have been performed to make a comparative evaluation on the effects of different reactivity control mechanisms and different coolant circulation methods of small fast reactors as follows, (a) 150MWe reflector-controlled forced-circulation core, (b) 150MWe forced-circulation core with control rods, (c) 150Me natural-circulation core with control rods and (d) 50MWe forced-circulation core with control rods. In the present study, the fuel reloading interval of 10 years is targeted under following conditions as fuel pin diameter $$<$$ 8.5mm, pressure drop for forced-circulation core $$<$$ 0.75kg/cm2 and sodium void reactivity for forced-circulation core $$<$$ 2$.The reactivity control mechanism of the reflector-controlled core is composed of movable radial reflector and a few additional control rods. This core attains 10 years long life and burnup of about 48GWd/t. The ATWS analyses indicate the possible passive safety feature which does not rely on the effects of radial expansion of core support plate and self-actuated shutdown system.The core with control rods shows slightly superior criticality than the reflector-controlled core due to the closer arrangement of radial reflectors. The shorter core column length and reduced fuel inventory result in improvements on the sodium void reactivity and burnup of fuels. This core also shows possible passive safety features in case of ATWS events. Ten years of control rod life time is calculated to be achievable from a viewpoint of absorber cladding mechanical interactions. Large core diameter of about 3.3m and small pin gap less than 1mm are common problems of the reflector-controlled core and core with control rods. Those are ascribed to the sodium void reactivity limitation and further works considering the economy and fabrication feasibility are necessary. The natural-circulation core with low pressure drop has large core diameter of about 3.8m and low fuel burnup of around 36GWd/t. A prominent characteristic

JAEA Reports

Design study on BN-600 hybrid core (I); evaluation of core neutronic and thermalhydraulic characteristics by Japanese analysis methods

Uto, Nariaki; Uto, Nariaki

JNC TN9400 2003-040, 67 Pages, 2003/06

JNC-TN9400-2003-040.pdf:3.02MB

A program of disposition of Russian weapon-grade plutonium by containing the plutonium in vibropacked MOX fuel subassemblies and burning them in the BN-600

JAEA Reports

Development of the next generation code system as an engineering modeling language (II); Study with prototyping

Yokoyama, Kenji; Hosogai, Hiromi*; Uto, Nariaki; Kasahara, Naoto; Ishikawa, Makoto

JNC TN9400 2003-021, 205 Pages, 2003/04

JNC-TN9400-2003-021.pdf:8.86MB

In the fast reactor development, numerical simulation using analytical codes plays an important role for complementing theory and experiment. It is necessary that the engineering models and analysis methods can be flexibly changed, because the phenomina to be investigated become more complicated due to the diversity of the needs for research. And, there are large problems in combining physical propaties and engineering models in many different fields. Aming to the realization of the next generation code system which can solve those problems, the authors adopted three methods, (1)Multi-language (SoftWIRE.NET, Visual Basic .NET and Fortran) (2)Fortran90 and (3)Python to make a prototype of the next generation code system. As this result, the followings were comfirmed. (1)It is possible to reuse a function of the existing codes written in Fortran as an object of the next generation code system by using visual Basic .NET. (2)The maintenanability of the existing code written by Fortran77 can be improved by using the new features of Fortran90. (3)The toolbox-type code system can be built by using Python.

Journal Articles

R&D of the object-integrated code system for fast reactors, 1

Yokoyama, Kenji; Uto, Nariaki; kasahara, Naoto; ; Ishikawa, Makoto

Nihon Genshiryoku Gakkai 2003-Nen Aki No Taikai, 2(E64), 343 Pages, 2003/00

None

Journal Articles

LMFBR Design and its Evolution, 2; Core Design of LMFBR

Mizuno, Tomoyasu; Uto, Nariaki; Takaki, Naoyuki

Proceedings of International Conference on Global Environment and Advanced Nuclear Power Plants (GENES4/ANP 2003) (Internet), 0 Pages, 2003/00

Sodium-cooled core design studies are performed. MOX fuel core with axial blanket partial elimination subassembly due to safety consideration is studied. This type of core with high internal conversion ratio possesses capability of achieving 26 months of

JAEA Reports

Development of the next generation code system as an engineering modeling language, I

Yokoyama, Kenji; Hosogai, Hiromi*; Uto, Nariaki; Kasahara, Naoto; Nagura, Fuminori; *; *; Ishikawa, Makoto

JNC TN9420 2002-004, 309 Pages, 2002/11

JNC-TN9420-2002-004.pdf:11.4MB

In the fast reactor development, numerical simulation using analytical codes plays an important role for complementing theory and experiment. It is necessary that the engineering models and analysis methods can be flexibly changed, because the phenomina to be investigated become more complicated due to the diversity of the needs for research. And, there are large problems in combining physical propaties and engineering models in many different fields. In this study, the goal is to develop a flexible and general-purposive analysis system, in which the phisical propaties and engineering models are replesented as a programming languare or a diagams that are easily understandable for humans and executable for computers. The authors named this concept the Engineering Modeling Language(EML). This report describes the result of the investigation for latest computer technologies and software development techniques which seem to be usable for a realization of the analysis code system for nuclear engineering as an EML.

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