Fuel Unloading Operations -2019- in the decommissioning of prototype fast breeder reactor "Monju"

Yabe, Takanori; Murakami, Makio; Shiota, Yuki; Isobe, Yuta; Shiohama, Yasutaka; Hamano, Tomoharu; Takagi, Tsuyohiko; Nagaoki, Yoshihiro

JAEA-Technology 2022-002, 66 Pages, 2022/07

JAEA-Technology-2022-002.pdf:10.45MB

In the first stage of "Monju" decommissioning project, "Fuel Unloading Operations" have been carrying out. The operations consists of two processes. The first process is "Fuel Treatment and Storage" is that the fuel assemblies unloaded from the Ex-Vessel fuel Storage Tank are canned after sodium cleaning, and then transferred to the storage pool. The second process is "Fuel Unloading" that the fuel assemblies in the reactor core are replaced with dummy fuel assemblies and stored in the Ex-Vessel fuel Storage Tank. "Fuel Treatment and Storage" and "Fuel Unloading" are performed alternately until 370 fuel assemblies in the core and 160 fuel assemblies in the Ex-Vessel fuel Storage Tank are all transferred to the storage pool. In fiscal 2018, as "Fuel Treatment and Storage", 86 fuel assemblies were transferred to the storage pool. As "Fuel Unloading", 76 dummy fuel assemblies were stored in the Ex-Vessel fuel Storage Tank. In fiscal 2019, as "Fuel Unloading", 60 fuel assemblies and 40 blanket fuel assemblies were unloaded from the core. These assemblies were stored in the Ex-Vessel fuel Storage Tank, and dummy fuel assemblies were loaded into the core instead. During these operations, a total of 38 cases of alarming or equipment malfunctions classified into 24 types occurred. However, no significant events that menaces to safety have occurred. The operations were continued safely by removing the direct factors for the malfunctions in the equipment operation and performance.

Evolution of the reaction and alteration of granite with Ordinary Portland cement leachates; Sequential flow experiments and reactive transport modelling

Bateman, K.*; Murayama, Shota*; Hanamachi, Yuji*; Wilson, J.*; Seta, Takamasa*; Amano, Yuki; Kubota, Mitsuru*; Ouchi, Yuji*; Tachi, Yukio

Minerals (Internet), 12(7), p.883_1 - 883_20, 2022/07

https://doi.org/10.3390/min12070883

Development of methodology to evaluate mechanical consequences of vapor expansion in SFR severe accident transients; Lessons learned from previous France-Japan collaboration and future objectives and milestones

Bachrata, A.*; Gentet, D.*; Bertrand, F.*; Marie, N.*; Kubota, Ryuzaburo*; Sogabe, Joji; Sasaki, Keisuke; Kamiyama, Kenji; Yamano, Hidemasa; Kubo, Shigenobu

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04

In the frame of France-Japan collaboration, one of the objectives is to define and assess the calculation methodologies, and to investigate the phenomenology and the consequences of severe accident scenarios in sodium fast reactors (SFRs). A methodology whose purpose is to assess the loadings of the structures induced by a Fuel Coolant Interaction (FCI) taking place in the sodium plenum of SFR has been defined in the frame of the collaboration between France and Japan during 2014-2019. The work progress will be spread over the period 2020-2024 and the main objectives and milestones will be introduced in the paper. The objective of studies is to comprehensively address the margin between the limit of integrity of the main vessel structures and the loadings resulting from severe accidents. For this purpose, the SIMMER mechanistic calculation code simulates core disruptive accident sequences in SFRs. A fluid structure dynamics tool evaluates this interaction i.e. EUROPLEXUS is used in CEA studies and AUTODYN tool is used in JAEA studies. In the paper, a benchmark study is described in order to illustrate the evaluation of vapour expansion phase in the hot plenum. To do that, joint input data are used on the basis of an ASTRID 1500 MWth core degraded state after the power excursion which leads to vapour expansion. The most penalizing case was evidenced in this study by suppressing the action of transfer tube in-core mitigation devices in SIMMER input deck and thus privileging the upward molten core ejection. Even if the most penalizing case was evidenced in this paper, no significant RV deformation was observed in both EUROPLEXUS and AUTODYN calculation results. The assumed mechanical energy was small for the core expansion phase.

Patterns of gene content and co-occurrence constrain the evolutionary path toward animal association in Candidate Phyla Radiation bacteria

Jaffe, A. L.*; Thomas, A. D.*; He, C.*; Keren, R.*; Valentin-Alvarado, L. E.*; Munk, P.*; Bouma-Gregson, K.*; Farag, I. F.*; Amano, Yuki; Sachdeva, R.*; et al.

mBio, 12(4), p.e00521-21_1 - e00521-21_21, 2021/08

https://doi.org/10.1128/mBio.00521-21

Clades of huge phages from across Earth's ecosystems

Al-Shayeb, B.*; Sachdeva, R.*; Chen, L.-X.*; Ward, F.*; Munk, P.*; Devoto, A.*; Castelle, C. J.*; Olm, M. R.*; Bouma-Gregson, K.*; Amano, Yuki; et al.

Nature, 578(7795), p.425 - 431, 2020/02

https://doi.org/10.1038/s41586-020-2007-4

Development of the pump-integrated intermediate heat exchanger in advanced loop-type sodium-cooled fast reactor for demonstration

Amano, Katsunori; Enuma, Yasuhiro; Futagami, Satoshi; Inoue, Tomoyuki*; Watanabe, Sota*

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 7 Pages, 2016/06

http://doi.org/10.1115/ICONE24-60064

In the framework of GIF, SDC and SDG for the generation IV SFRs have been developed in the circumstance of worldwide deployment of SFRs. JAEA and MFBR have been investigating design study of an advanced loop-type SFR to satisfy SDC in the feasibility study of SDG for SFR. In this study, the ability of the pump/IHX in the advanced loop-type SFR for the safety measures has been evaluated. In addition to the safety measures, maintainability and reparability are taken into account in the advanced loop-type SFR design study. The pump/IHX has been modified to satisfy these requirements. This paper describes the modifications for the ability to withstand a severe earthquake, the reliability of the guard vessel in the primary coolant leak, and the reliability of expansion joints in a sodium-water reaction. The evaluations of thermal transient, structural vibration with pump rotation and wear-out of IHX tubes, that has been adversely effected by the modifications, were described as well.

A New view of the tree of life

Hug, L. A.*; Baker, B. J.*; Anantharaman, K.*; Brown, C. T.*; Probst, A. J.*; Castelle, C. J.*; Butterfield, C. N.*; Hernsdorf, A. W.*; Amano, Yuki; Ise, Kotaro; et al.

Nature Microbiology (Internet), 1(5), p.16048_1 - 16048_6, 2016/05

https://doi.org/10.1038/nmicrobiol.2016.48

The tree of life is one of the most important organizing principles in biology. Gene surveys suggest the existence of an enormous number of branches, but even an approximation of the full scale of The Tree has remained elusive. Here, we use newly available information from genomes of uncultivated organisms, along with other published sequences, to present a new version of the Tree of life, with Bacteria, Archaea and Eukaryotes included. The depiction is both a global overview and a snapshot of the diversity within each major lineage. The results imply the predominance of bacterial diversification and underline the importance of organisms lacking isolated representatives, with substantial evolution concentrated in a major radiation of such organisms.

Effect of deflected inflow on flows in a strongly-curved 90 degree elbow

Iwamoto, Yukiharu*; Kusuzaki, Ryo*; Sogo, Motosuke*; Yasuda, Kazunori*; Yamano, Hidemasa; Tanaka, Masaaki

Proceedings of ASME-JSME-KSME Joint Fluids Engineering Conference 2015 (AJK 2015-FED) (USB Flash Drive), p.1767 - 1773, 2015/07

Wall pressure measurements and flow visualization were conducted for a 90 degree elbow of which the axis curvature radius became the same as its inner diameter (125 mm). A deflected inflow, having an almost constant velocity slope, was introduced. Ensemble averaged pressure distributions showed no normalized pressure difference in cases of Reynolds numbers of 320,000 and 500,000. Comparisons with uniform inlet flow case proved; low-pressure region at the intrados of the elbow was weakened whereas a high pressure region outside strengthened in the deflected inflow case. Pressure distribution downstream of the elbow increased at the inside until two diameters downstream from the elbow exit. Flow visualization showed the pressure increase from collision of a strengthened secondary flow of extrados convection. The unsteady pressure distribution showed a circumferential extent of a strongly fluctuating region in and downstream the elbow decreased, comparing with the uniform inlet flow case.

Neutron angular distribution in (, n) reactions with linearly polarized -ray beam generated by laser Compton scattering

Horikawa, Ken*; Miyamoto, Shuji*; Mochizuki, Takayasu*; Amano, So*; Li, D.*; Imasaki, Kazuo*; Izawa, Yasukazu*; Ogata, Kazuyuki*; Chiba, Satoshi*; Hayakawa, Takehito

Physics Letters B, 737, p.109 - 113, 2014/10

https://doi.org/10.1016/j.physletb.2014.08.024

It was predicted in 1950's, the neutron angular distribution in (, n) reactions with a 100% linearly polarized -ray beam should be anisotropic and described by a simple function of a + b sin at 90 on the beam axis but it has not been experimentally confirmed for middle-heavy nuclides over than half a century. We have verified experimentally this angular distribution on Au, I, and natural Cu using linearly polarized laser Compton scattering -rays at NewSUBARU.

Evaluation of Earthquake and Tsunami on JSFR

Chikazawa, Yoshitaka; Enuma, Yasuhiro; Kisohara, Naoyuki; Yamano, Hidemasa; Kubo, Shigenobu; Hayafune, Hiroki; Sagawa, Hiroshi*; Okamura, Shigeki*; Shimakawa, Yoshio*

Proceedings of 2012 International Congress on Advances in Nuclear Power Plants (ICAPP '12) (CD-ROM), p.677 - 686, 2012/06

Evaluation of Earthquake and Tsunami on JSFR has been analyzed. For seismic design, safety components are confirmed to maintain their functions even against recent strong Earthquakes. As for Tsunami, some parts of reactor building might be submerged including component cooling water system whose final heat sink is sea water. However, in the JSFR design, safety grade components are independent from component cooling water system (CCWS). The JSFR emergency power supply adopts a gas turbine system with air cooling, since JSFR does not basically require quick start-up of the emergency power supply thanks to the natural convection DHRS. Even in case of long station blackout, the DHRS could be activated by emergency batteries or manually and be operated continuously by natural convection.