Future R&D programs using Monju

Konomura, Mamoru; Ichimiya, Masakazu; Mukai, Kazuo

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 9 Pages, 2012/00

http://www-pub.iaea.org/mtcd/meetings/PDFplus/2009/cn176/cn176_BoeS.pdf

Japanese prototype fast breeder reactor, Monju, will soon restart. Monju has three sodium loops with steam generators and a turbin-generator, 280 MWe. Monju is expected to demonstrate the sodium technology and the sodium-water heat exchange technology in Japan. It will be operated at full power operation during around 10 years after restart in order to accumulate operation & maintenance experience and to evaluate its design technology. After the system start-up test (SST), Monju will be operated under full power. In this stage, the main object of Monju operation will be to achieve its initial targets which were fixed when its construction was decided around thirty years ago. The targets are to demonstrate a safe and reliable operation, that is, accumulation of the operation & maintenance experience and evaluation of the design technology, and to establish sodium handling technology. For example, inspection and diagnosis technologies are important for maintenance of a sodium cooled reactor. An out-of-pile sodium test facility will be constructed near Monju in order to make many tests of inspection devices and research many chemical tests. At the same time, the activities for the performance improvement, for example, a new licence, will be prepared in order to utilize Monju as a R&D facility. After the accumulation of operating experience, Monju will be enhanced the performance as a R&D facility in order to demonstrate innovative technologies, for example, irradiation of advanced fuel, longer operation cycle, higher burnup. For this purpose, Monju will be needed to get a new licence and core modification. And Monju on-site non-destructive Post Irradiated Evaluation facility will be expected at this stage. There were many R&D works in Japan with sodium out-of-pile facilities. All the experience were reflected in the design of Monju. Monju will demonstrate a handling of sodium technologies under power plant operation.

Measurement of relative biological effectiveness of protons in human cancer cells using a laser-driven quasimonoenergetic proton beamline

Yogo, Akifumi; Maeda, Takuya; Hori, Toshihiko; Sakaki, Hironao; Ogura, Koichi; Nishiuchi, Mamiko; Sagisaka, Akito; Kiriyama, Hiromitsu; Okada, Hajime; Kanazawa, Shuhei; et al.

Applied Physics Letters, 98(5), p.053701_1 - 053701_3, 2011/02

https://doi.org/10.1063/1.3551623

Superconducting Pb island nanostructures studied by scanning tunneling microscopy and spectroscopy

Nishio, Takahiro*; An, Toshu*; Nomura, Atsushi*; Miyachi, Kosuke*; Eguchi, Toyoaki*; Sakata, Hideaki*; Lin, S.*; Hayashi, Nobuhiko; Nakai, Noriyuki; Machida, Masahiko; et al.

Physical Review Letters, 101(16), p.167001_1 - 167001_4, 2008/10

https://doi.org/10.1103/PhysRevLett.101.167001

Superconductivity of nanosized Pb-island structures whose radius is 0.8 to 2.5 times their coherence length was studied under magnetic fields using low-temperature scanning tunneling microscopy and spectroscopy. Spatial profiles of superconductivity were obtained by conductance measurements at zero-bias voltage. Critical magnetic fields for vortex penetration and expulsion and for superconductivity breaking were measured for each island. The critical fields depending on the lateral size of the islands and existence of the minimum lateral size for vortex formation were observed.

The H-Invitational Database (H-InvDB); A Comprehensive annotation resource for human genes and transcripts

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

https://doi.org/10.1093/nar/gkm999

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.

A New measurement of the Li(,n)B reaction for astrophysical interest

Das, S. K.*; Fukuda, Tomokazu*; Mizoi, Yutaka*; Ishiyama, Hironobu*; Miyatake, Hiroari*; Watanabe, Yutaka*; Hirayama, Yoshikazu*; Tanaka, Masahiko*; Yoshikawa, Nobuharu*; Jeong, S.-C.*; et al.

AIP Conference Proceedings 847, p.374 - 376, 2006/07

https://doi.org/10.1063/1.2234430

no abstracts in English

Rewetting characteristics of a wire-wrapped fuel rod

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.

Heavy Liquid Metal Coole; Results 2002

Enuma, Yasuhiro; Hayafune, Hiroki; Soman, Yoshindo; Konomura, Mamoru; Mizuno, Tomoyasu; Sakai, Takaaki

JNC TN9400 2003-081, 234 Pages, 2003/09

JNC-TN9400-2003-081.pdf:13.34MB

Main issues of structural integrity, seismic integrity and operation are studied to confirm feasibilities of the medium size lead-bismuth cooled reactor as a candidate for one of the concepts of fast reactor in the feasibility studies on commercialized FBR cycle system. The results are following. (1) Integrity under loading of earthquake (a) Reactor system: Seismic integrity of the reactor is confirmed. (b) Upper core structures and fuel exchanging machine: Those structures are necessary for increasing stiffness by bonding. (2) Structural integrity for thermal loading Characteristics of both the medium size lead-bismuth cooled reactor and the large size sodium cooled reactor in the feasibility studies are arranged and structural integrity for thermal loading is evaluated about particular phenomena of thermal stratification and thermal striping. (3) Operation Feasibility of the natural circulation operation is confirmed. (a) Water-steam supply system concerned with control system (b) Control characteristics and transient characteristics of lead-bismuth reactor (c) Initial start-up operation concerned with preheating and charging of coolant (4) Safty Core integrity on SG tube rupture is evaluated to be confirmed its safety, by analyzing thermal hydraulic behavior of transient of power, flow and temperature, by calculating characteristics of core and by evaluating additional reactivity by steam coming into core region. (5) Internal core structures design and system power balance design.

Development of Advanced Austenitic Stainless Steel for Fast Reactor Core Material

Shikakura, Sakae; Ukai, Shigeharu; Sato, Yoshinori; Harada, Makoto; Koyama, Shinichi; ; Nomura, Shigeo; Shibahara, Itaru

Nihon Genshiryoku Gakkai-Shi, 36(5), p.441 - 455, 1994/00

None

None

Nomura, Kazunori; ; Sakai, Toshiyuki*; *;

PNC TN8410 91-180, 51 Pages, 1991/06

PNC-TN8410-91-180.pdf:1.06MB

None

Study on the design limit of the FBR fuel cladding at the anticipated transient event; Evaluation of the thermal transient test results

Seshimo, Ichiro; Ukai, Shigeharu; Nomura, Shigeo; Shikakura, Sakae

PNC TN9410 89-122, 47 Pages, 1989/08

PNC-TN9410-89-122.pdf:5.34MB

Fuel cladding integrity must be confirmed even at the anticipated transient event in LMFBR. In case of loss of coolant accident, cladding highest temperature is limited to 830 for Monju design from a view point of preventing the cladding creep rupture failure by increasing internal gas pressure. In this study, using the recent thermal transient test results of modified SUS316 stainless steel, Larson-Miller Parameter Life Fraction method was applied for predicting the optimum failure temperature. With this method, different thermal transient data can be evaluated systematically, and the effect of irradiation on cladding failure temperature was analyzed. Through this analysis, the Monju design limit temperature of 830 for cladding failure can be changed to 966, and alternative limiting temperature of 920 defined for preventing the coolant sodium boiling becomes a critical factor. This results shows the possibility of improvement for the design limit of this event.