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Journal Articles

Now is the time of fast reactor

Negishi, Hitoshi; Kamide, Hideki; Maeda, Seiichiro; Nakamura, Hirofumi; Abe, Tomoyuki

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 62(8), p.438 - 441, 2020/08

Prototype Fast Breeder Reactor, Monju, was under decommission since April, 2018. It is the first time for Japan to make a sodium cooled reactor into decommission. It is significant work and will take 30 years. The Monju has provided wide spectrum and huge amount of findings and knowledge, e.g., design, R&D, manufacturing, construction, and operation up to 40% of full power over 50 years of development history. It is significant to utilize such findings and knowledge for the development and commercialization of a fast rector in Japan.

Journal Articles

Outline of the R&D plan for the fast reactor cycle system development in JAEA

Hayafune, Hiroki; Maeda, Seiichiro; Ohshima, Hiroyuki

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 61(11), p.798 - 803, 2019/11

In the "Strategic Roadmap" of Fast Reactor Development decided at the Inter-Ministerial Council for Nuclear Power in December 2018, the development works for the around next 10 years were identified, and the role of JAEA was presented. In response, JAEA has prepared a framework for R&D plans for about 5 years on the fast reactor technology and the fuel cycle technology (reprocessing, fuel manufacturing, fuel and material development). In the future, JAEA will promote independent R&D works based on these plans, and provide the obtained R&D results together with various testing functions of JAEA to the activities of the private sector, etc. Through these actions, JAEA will actively contribute to the future fast reactor development. This article outlines JAEA's policy and the R&D items (development of ARKADIA; Advanced Reactor Knowledge- and AI-Aided Design Integration Approach through the whole Plant Life Cycle, development of standards and standards system, development of safety improvement technology, research in the fuel cycle technology), the policy of international cooperation, the human resource development, and the future perspective were explained.

Journal Articles

Nuclide partitioning and transmutation technology; Transmutation using fast reactor

Yanagisawa, Tsutomu*; Usami, Shin; Maeda, Seiichiro

Genshiryoku Nenkan 2018, p.90 - 95, 2017/10

no abstracts in English

Journal Articles

Current status of the next generation fast reactor core & fuel design and related R&Ds in Japan

Maeda, Seiichiro; Oki, Shigeo; Otsuka, Satoshi; Morimoto, Kyoichi; Ozawa, Takayuki; Kamide, Hideki

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06

The next generation fast reactor is being investigated in Japan, aiming at several targets such as "safety", "reduction of environmental burden" and "economic competitiveness". As for the safety aspect, FAIDUS concept is adopted to avoid re-criticality in core destructive accidents. The uranium-plutonium mixed oxide fuel, in which minor actinide elements are included, will be applied to reduce the amount and potential radio-toxicity of radioactive wastes. The high burn-up fuel is pursued to reduce fuel cycle cost. The candidate concept of the core and fuel design, which could satisfy various design criteria by design devisals, has been established. In addition, JAEA is investigating material properties and irradiation behavior of MA-MOX fuel. JAEA is developing the fuel design code especially for the fuel pin with annular pellets of MA-bearing MOX. Furthermore, JAEA is developing oxide dispersion strengthened (ODS) ferritic steel cladding for the high burnup fuel.

Journal Articles

Core design of the next-generation sodium-cooled fast reactor in Japan

Kan, Taro*; Ogura, Masashi*; Hibi, Koki*; Oki, Shigeo; Maeda, Seiichiro; Maruyama, Shuhei; Ohgama, Kazuya

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04

Journal Articles

Development and verification of the thermal behavior analysis code for MA containing MOX fuels

Ikusawa, Yoshihisa; Ozawa, Takayuki; Hiroka, Shun; Maeda, Koji; Kato, Masato; Maeda, Seiichiro

Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 6 Pages, 2014/07

In order to develop MA contained MOX (MA-MOX) fuel design method, the analysis models to predict irradiation behavior of MA-MOX fuel have to be developed and the accuracy of irradiation behavior analysis code should be evaluated with the result of post-irradiation examinations (PIEs) for MA-MOX fuels. In this study, we developed the computer module "TRANSIT" to compute thermal properties of MA-MOX fuel. TRANSIT can give thermal conductivity, melting temperature and vapor pressures of MA-MOX. By using this module, we improved the thermal behavior analysis code "DIRAD" and developed DIRAD-TRANSIT code system to compute the irradiation behavior of MA-MOX fuel. This system was verified with the results of PIEs for the conventional MOX fuels and the MA-MOX fuels irradiated in the experimental fast reactor "JOYO". As the result of the verification, it can be mentioned that the DIRAD-TRANSIT system would precisely predict the fuel thermal behavior, i.e. fuel temperature and fuel restructuring, for oxide fuels containing several percent minor actinides.

Journal Articles

Study on the mechanism of diametral cladding strain and mixed-oxide fuel element breaching in slow-ramp extended overpower transients

Uwaba, Tomoyuki; Maeda, Seiichiro; Mizuno, Tomoyasu; Teague, M. C.*

Journal of Nuclear Materials, 429(1-3), p.149 - 158, 2012/10

 Times Cited Count:3 Percentile:28.72(Materials Science, Multidisciplinary)

Cladding strain caused by fuel/mechanical interaction (FCMI) was evaluated for mixed-oxide fuel elements subjected to 70-90% slow-ramp extended overpower transient tests in EBR-II. Calculated transient-induced cladding strains were correlated with cumulative damage fractions (CDFs) using cladding strength correlations. In a breached high-smeared density solid fuel element with low strength cladding, cladding thermal creep strain was significantly increased to approximately half the transient-induced cladding strain due to the tertiary creep when the CDF was close to the breach criterion (= 1.0). In low-smeared density annular fuel elements, FCMI load was significantly mitigated and resulted in little cladding strain. The CDFs of the annular fuel elements were lower than 0.01 at the end of the overpower transient, indicating a substantial margin to breach. A substantial margin to breach was also maintained in a high-smeared density fuel element with high strength cladding.

Journal Articles

Short-term irradiation behavior of low-density americium-doped uranium-plutonium mixed oxide fuels irradiated in a fast reactor

Maeda, Koji; Katsuyama, Kozo; Ikusawa, Yoshihisa; Maeda, Seiichiro

Journal of Nuclear Materials, 416(1-2), p.158 - 165, 2011/09

 Times Cited Count:12 Percentile:71.79(Materials Science, Multidisciplinary)

In order to evaluate the thermal behavior of low-density uranium and plutonium mixed oxide fuels containing several percent of americium (Am-MOX), fuel irradiation test (B14) was conducted using the experimental fast reactor. Pellet-cladding gap width and O/M ratio of oxide fuels were specified as experimental parameters. Four fuel pins were irradiated step-by-step in consideration of fuel restructuring during 48 hours as pre-conditioning before full power reactor operation. The irradiation history, i.e. linear power, was simulated the conventional FBR oxide fuel pins. And the linear power was rapidly increased up to 47 kW/cm for 10 minutes to simulate the transient condition. After the irradiation, ceramography samples were taken from the axial position of each fuel pins where the fuel centerline temperature reached the maximum during irradiation. The result was investigated relative to those of other irradiated fuels.

JAEA Reports

Study of ageing effect of long-term storage fuel in prototype fast breeder reactor Monju

Kato, Yuko; Umebayashi, Eiji; Okimoto, Yutaka; Okuda, Eiichi; Takayama, Koichi; Ozawa, Takayuki; Maeda, Seiichiro; Matsuzaki, Masaaki; Yoshida, Eiichi; Maeda, Koji; et al.

JAEA-Research 2007-019, 56 Pages, 2007/03

JAEA-Research-2007-019.pdf:6.79MB

In order to resume the System Startup Test (SST) of Monju, replacement fuel have to be loaded in exchange for some of initial fuel now loaded in the core to compensate core reactivity lost by decay of Pu-241 in them. The replacement fuel were being stored either in sodium in an ex-vessel storage tank or in air in a storage rack for about 10 years since their fabrication. The initial fuel were irradiated during the SST which was suspended in the end of 1995 and then stayed being loaded in the sodium-circulated core. As this long-term storage and loading may deteriorate mechanical integrity of the assemblies, a study has been made thoroughly on its thermal-hydraulic, structural and material effects on them that might be caused by irradiation in the core, sodium and mechanical environment. The study has shown that the mechanical integrity of them is well maintained even with this long-term storage and loading.

JAEA Reports

Design study on a demonstration core for a practical LMFBR in Monju, 2

Saito, Kosuke; Maeda, Seiichiro; Higuchi, Masashi*; Takano, Mitsuhiro*; Nakazawa, Hiroaki

JAEA-Technology 2006-035, 76 Pages, 2006/06

JAEA-Technology-2006-035.pdf:5.25MB

Because of the revision on the standardized strength of the ODS steel, the previous design study of MONJU demonstrative core has been obliged to be reconsidered. For economical advantages, only a 127 pins-bundle core was selected to be redesigned. For the sake of cladding endurance, the ratio of cladding thickness to outer diameter was reset incrementally followed by the determination of the basic specification of a pin. Notwithstanding some deterioration thanks to the reduction of a fuel volume fraction, the prospect in neutronics was obtained. Coolant flow distribution design which was based on power distribution was successfully carried out without overheating cladding. Average burn-up of 150 GWd/t and 380 days-long operational period per cycle are to be attained, and the designed core can thermally afford to receive test fuels. The study has necessity to be advanced extensively for the purpose of materialization according to the circumstances of MONJU in future.

Journal Articles

Development and demonstration of ATR-MOX fuel

Abe, Tomoyuki; Maeda, Seiichiro; Nakazawa, Hiroaki

Proceedings of 13th International Conference on Nuclear Engineering (ICONE-13), 0 Pages, 2005/05

Japan Nuclear Cycle Development Institute (JNC) developed plutonium and uranium mixed oxide (MOX) fuels for an advanced thermal reactor (ATR) for a flexible utilization of plutonium. JNC made endeavors to obtain well-homogenized MOX pellets by a ball mill mixing method with a variety of raw powders, including MOX powder by a microwave-heating denitration process. A total of 772 MOX fuel assemblies were utilized in the ATR prototype reactor

Journal Articles

Status of studies on HLW glass performance for confirming its validity in assessment

Inagaki, Yaohiro*; Mitsui, Seiichiro*; Makino, Hitoshi*; Ishiguro, Katsuhiko*; Kamei, Gento*; Kawamura, Kazuhiro*; Maeda, Toshikatsu; Ueno, Kenichi*; Bamba, Tsunetaka*; Yui, Mikazu*

Genshiryoku Bakkuendo Kenkyu, 10(1-2), p.69 - 84, 2004/03

no abstracts in English

JAEA Reports

Status of Studies on HLW Glass Performance for Increasing Its Validity in Assessment

Inagaki, Yaohiro*; Mitsui, Seiichiro; Makino, Hitoshi; Ishiguro, Katsuhiko; Kamei, Gento; Kawamura, Kazuhiro; Maeda, Toshikatsu*

JNC TN8400 2003-036, 53 Pages, 2003/12

JNC-TN8400-2003-036.pdf:0.51MB

Obtain of sufficient data for the performance of high-level radioactive waste(HLW) glass and verification of a model for the radionuclide release from the HLW glass in the disposal condition are required in order to show the objective reliability. In this paper, some reports about performance assessment of HLW glass are reviewed and we clarify the problems to raise reliability comparing these reports.

JAEA Reports

Design Study on a Demonstration Core for a Practical LMFBR in Monju

Maeda, Seiichiro; Togashi, Nobuhito*; Higuchi, Masashi*; Takano, Mitsuhiro*; Abe, Tomoyuki

JNC TN8400 2003-028, 135 Pages, 2003/12

JNC-TN8400-2003-028.pdf:85.68MB

The Monju advanced core concept to demonstrate a practical LMFBR core with 150GWd/t (average discharged burnup) was embodied in this design study. A high performance fuel with annular pellets of a large diameter filled in ODS (oxide dispersion strengthen ferritic steel) claddings was applied in the advanced core. This enables improvement of an internal conversion ratio in combination with increase of effective fuel volume fraction, achievement of high burnup up to 150GWd/t and a long operation period beyond 1 year in Monju. The core in which the practical high burnup lessens a burden for a fuel cycle system including fuel fabrication and reprocessing can be demonstrated. In the first step, constraints in the existing plant and requirements to demonstrate the practical LMFBR were clarified. The core and fuel specifications were surveyed with parameters of a number of fuel pins in an assembly and so on. Two types of cores with 127-pin-bundle and 91-pin-bundle were selected as candidates. In the second step, performances of these core options were specified in this design study. It was shown that major parameters in neutronic design, hydraulic design and fuel design would meet criteria. The application of the high performance fuel significantly contributes the enhancement of economical efficiency of Monju itself. The net operation cost will be greatly reduced by increase of the annual electricity generated caused by a boost of the plant operating rates and by saving of the annual discharged fuel assemblies up to 1/2 or 1/3. The deliberate margin for thermal limits ensures the irradiation field to develop new type fuels and core materials and to demonstrate a low decontaminated fuel with miner actinides as a candidate of advanced fuel cycle. The results in this study may become a technically credible guideline to make the future management plan of Monju.

JAEA Reports

Design Study on In-core Breeding Concept Using Annular Thick Fuel Pins

Maeda, Seiichiro; Takashita, Hirofumi; Okawa, Tsuyoshi; Higuchi, Masashi*; Abe, Tomoyuki

JNC TN8400 2003-019, 185 Pages, 2003/08

JNC-TN8400-2003-019.pdf:11.78MB

We are studying on an in-core breeding concept as a candidate for a practical FBR fuel cycle system attainable in an early stage on the premise that sodium coolant and mixed oxide fuel should be adopted, since the technical issues with these combination are most advanced and common with the fuel cycle for a LWR-MOX system. An enhancement of fuel volume fraction using thick fuel pins enables the in-core breeding. The fuel material flow can be greatly lessened by minimizing amount of the blanket with the in-core breeding core. The low material flow leads to significant reduction of the fuel cycle cost. We investigated a 3500MWth large-scale core adjusting several conditions presented in JNC's feasibility study program for a commercialized FBR system in this study. These were shown in this study that a discharged burnup averaged over the core and the blanket could reach approximately 130GWd/t (core averaged about 150GWd/t) within the maximum fast neutron fluence about 5x1023/cm2, that the small reactivity loss with burnup easily enabled long operation and that stable power distribution during operation significantly improved hydraulic property in this type core. We investigated measures to reduce sodium void reactivity, because core height enlargement to enhance neutron efficiency caused the increase of sodium void reactivity.We also investigated feasibility of a high breeding type core with low burnup considering a variety of FBR introducing scenarios and a trade-off correlation between breeding performance and burnup extension. The performance in this core design at core disruption accidents is not revealed enough. Further investigation should be made in detail to confirm that the in-core breeding concept could be accepted in a safety aspect.

JAEA Reports

Proposal of a Nuclear Cycle Research and Development Plan in Tokai Works -The Roadmap from LWR Cycle to FBR Cycle-

Nakamura, Hirofumi; Abe, Tomiyuki; Kashimura, Takao; Nagai, Toshihisa; Maeda, Seiichiro; Yamaguchi, T.; Kuroki, Ryoichiro

JNC TN8440 2003-016, 39 Pages, 2003/07

JNC-TN8440-2003-016.pdf:0.79MB

The Generation-II Project Task Force Team has investigated a research and development plan on a future nuclear fuel cycle in Tokai works for about three months from December 19,2002. First we have discussed about the present condition of Japanese nuclear fuel cycle and have recognized it as the following. *The relation of the technology between the LWR-cycle and the FBR-cycle is not clear. *MOX Fuel Use in Light Water Reactors is important to establish technology of the FBR fuel cycle. *Radioactive waste disposal issue is urgent. Next we have proposed the three basic policies on R&D plan of nuclear fuel cycle in consideration of the F.S. on FBR-cycle. *Establishment and advancement of "the tough nuclear fuel cycle". *Early establishment of the FBR cycle technology to be able to supply energy stably for long-term. *Establishment of the radioactive waste treatment and disposal technology, and optimization of nuclear fuel cycle technology from the viewpoint of radioactive waste.And we have proposed the Japanese technical holder system to integrate all LWR and FBR cycle technology.

Journal Articles

ATR-MOX Fuel Design and Development

Maeda, Seiichiro; Abe, Tomoyuki; Nakazawa, Hiroaki

GENES4/ANP2003, CD-ROM, Paper1150, 8p., 8 Pages, 2003/00

None

Oral presentation

High linear power irradiation tests with Am-MOX fuels, 1; Outline of the irradiation test program

Maeda, Seiichiro; Mizuno, Tomoyasu; Aoyama, Takafumi; Kitamura, Ryoichi; Takeuchi, Norihiko; Abe, Tomoyuki

no journal, , 

To confirm thermal performance of FBR-MOX fuels containing minor actinides such as americium, the irradiation experiment was performed with low-density MOX fuels containing americium at a high linear power in the fast experimental reactor "Joyo". The post-irradiation-experiment is successively proceeding at hot cell facilities. An overall plan and an outline of the irradiation program are presented in this report.

Oral presentation

High linear power irradiation test with Am-MOX fuels, 7; Evaluation of the thermal design aspect for FBR MOX fuels

Ikusawa, Yoshihisa; Kikuchi, Keiichi; Ozawa, Takayuki; Maeda, Seiichiro; Nakajima, Hiroshi*; Koike, Naoto

no journal, , 

To study the thermal performance of Am-MOX fuels, the high linear power irradiation test with Am-MOX fuels "B14 irradiation test" was carried out in the fast experimental reactor "JOYO". From the result of the irradiation test, the thermal performance for Am-MOX fuel was confirmed and the fuel irradiation behavior code was verified.

Oral presentation

Burn-up effect on MOX fuel thermal conductivity

Ikusawa, Yoshihisa; Morimoto, Kyoichi; Maeda, Seiichiro; Ogasawara, Masahiro*

no journal, , 

Thermal conductivity of oxide fuel is important for fuel design and performance analyses. The thermal conductivity of oxide fuel would decrease with burn-up increase and irradiated UO$$_{2}$$ fuels have been measured in various laboratories. In this study, burn-up effect on MOX fuel thermal conductivity was discussed. The influence of burn-up on the thermal conductivity of irradiated and un-irradiated MOX fuels with plutonium content of 2.5$$sim$$4.5 wt% were evaluated by means of the thermal diffusivity measurement. The evaluated thermal conductivities showed the tendency of decrease with burn-up. Such decrease of thermal conductivity with burn-up can be explained in terms of phonon conduction model. In addition, the decrease of MOX fuel thermal conductivity would be smaller than that of UO$$_{2}$$.

33 (Records 1-20 displayed on this page)