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Osaka, Masahiko; Miwa, Shuhei; Tanaka, Kosuke; Akutsu, Yoko; Ikeda, Kaoru*; Mimura, Hitoshi*; Suzuki, Tatsuya*; Usuki, Toshiyuki; Yano, Toyohiko*
Annals of Nuclear Energy, 38(12), p.2661 - 2666, 2011/10
Times Cited Count：2 Percentile：18.18(Nuclear Science & Technology)Novel concepts for effective utilization of molybdenum (Mo) from nuclear waste and magnesium silicates from hazardous asbestos wastes are proposed. A fast reactor cycle scheme that incorporates each material is described in the present paper. Basic studies on some fundamental technologies for the present cycle are given. Basic separation aspects for Mo by using LIX63 micro capsules and tertiary pyridine resin were investigated. A simple chemical synthesis route for Mo precursor powder from Mo containing HNO
solution was tested. Effects of impurities in recovered Mo on sintering behavior were experimentally investigated.
Miwa, Shuhei; Osaka, Masahiko; Usuki, Toshiyuki; Yano, Toyohiko*
Progress in Nuclear Energy, 53(7), p.1045 - 1049, 2011/09
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)We proposed a new concept for densification of minor actinides-containing inert matrix fuels by using asbestos waste-derived materials for the effective utilization of resources and protection of public safety. In this concept, magnesium silicates, which are mainly generated by the decomposition of asbestos in low temperature heat-treatment, are used as a sintering additive for the achievement of high density of magnesia-based inert matrix fuels at relatively low sintering temperature. In this study, preliminary fabrication tests of magnesia-based inert matrix fuels with magnesium silicates were carried out by using cerium oxides as a representative of minor actinides oxides.
Usuki, Toshiyuki; Yoshida, Katsumi*; Yano, Toyohiko*; Miwa, Shuhei; Osaka, Masahiko
Progress in Nuclear Energy, 53(7), p.1078 - 1081, 2011/09
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)The effects of sintering additives of magnesium silicates, i.e. enstatite (MgSiO), steatite (MgSiO) and forsterite (Mg
SiO
), on the fabrication properties and characteristics of the silicon nitride ceramics based inert matrix fuels were experimentally investigated. CeO was selected as simulating element of AmO. Sintered pellets were characterized in term of their densities, thermal conductivities and solubility to nitric acid. The densifications of sintered bodies were enhanced by using additives of magnesium silicates at relative low sintering temperature. The relative density of silicon nitride ceramics based inert matrix fuels with forsterite were achieved above 90% at 1723 K. The thermal conductivities of silicon nitride ceramics based inert matrix fuels varied according to sintering temperature, and those sintered at 1923 K were above 34 W/m K. The grain boundary phases in Silicon nitride ceramics based inert matrix fuels found to be dissolved into HNO.
N Inert MatrixYano, Toyohiko*; Osaka, Masahiko; Namekawa, Takashi
JNC TY9400 2004-002, 84 Pages, 2004/03
None
Akiyoshi, Masafumi; Akasaka, Naoaki; Tachi, Yoshiaki; Yano, Toyohiko*
Abstract p303,22-P-02, 303 Pages, 2003/00
Focusing on the cover layer materials (as the Radon Barrier Materials), which could have the effect to restrain the radon from scattering into the air and the effect of the radiation shielding, we produced the radon barrier materials with crude bentonite on an experimental basis, using the rotary type comprehensive unit for grinding and mixing, through which we carried out the evaluation of the characteristics thereof.
; ; ; Onose, Shoji
JNC TY9400 2002-011, 87 Pages, 2002/08
JNC-TY9400-2002-011.pdf:7.54MB
no abstracts in English
Sekimoto, Hiroshi*; Igashira, Masayuki*; Yano, Toyohiko*; Obara, Toru*; Osaki, Toshiro*
JAERI-Tech 2002-008, 58 Pages, 2002/03
no abstracts in English
Ishino, Shiori*; Terai, Takayuki*; Oku, Tatsuo*; Arai, Taketoshi; Hayashi, Kimio; Ito, Hisayoshi; Yano, Toyohiko*; Motohashi, Yoshinobu*; ; ; et al.
JAERI-Review 99-019, 238 Pages, 1999/08
JAERI-Review-99-019.pdf:14.88MB
no abstracts in English
JNC TJ9400 99-012, 96 Pages, 1999/03
Succeeding to the report on the post-irradiation experiments conducted in the previous year, this is a summary report on the post-irrdiation experiments of physical, thermal and microstructural properties of neutron-irradiated various ceramics, which are expected to be applied to the in-corc materials of an Advanced Fast Breeder Reactor in near future. Four candidate ceramics, AlO, AlN, SiC and SiN were fast-neutron-irradiated up to a fluence of 3.9
10
n/m
, different irradiation conditions from the previous report specimens, in the CMIR-4 rig in the JOYO expelimental fast reactor in JNC. The following observations were performed: (1)Microstructural observation by means of transmission electron microscopy, (2)Measurement of swelling, (3)Measurement of thermal diffusivity by a laser-flush method, (4)Recovery of swelling by isochronal annealing, and (5)Recovery of thermal diffusivity by isochronal annearling. 0btained main results are summarized as follows. Macroscopic length changes by neutron irradiation of AlO and AIN were measured to be 1.8-2.0% and these of SiC and SiN to be 0.2-0.4%, respectively. Thermal diffusivities of all irradiated materials degraded to 0.03-0.05 cm/s, irrespective of materials which had large difference before irradiation. Microstructural observation of irradiated materials by TEM revealed that AlO contained high-density loops, microvoids in grains, and microcracking along grain boundaries, AIN contained high-density loops and microcracking along grain boundalies, SiC contained high-density loops, and SiN contained loops lying on the planes parallel to the c-axis, respectively. Macroscopic length of AlO and AIN started to recover at around 800
or 1100C, respectively, irrespective of irradiation temperature, and reduced quickly. Macroscopic length of SiC recovered gradually from near the irradiation temperature.
PNC TJ9607 98-002, 78 Pages, 1998/03
This is a summary report on the post-irradiation experiments of physical, thermal and microstructural properties of neutron-irradiated various ceramics, which are expected to be applied to the in-core materials of an Advanced Fast Breeder Reactor in near future. Four candidate ceramics, AlO, AlN, SiC and SiN were fast neutron irradiated up to 4.210n/m in the CMIR-4 rig in the JOYO reactor. The following observations were performed and the obtained results are mentioned. (1)Microstructural observation by means of transmission electron microscopy (2)Measurement of swelling (3)Measurement of thermal diffusivity by a laser-flush method (4)Recovery of swelling by isochronal annealing (5)Recovery of thermal diffusivity by isochronal annealing
; Maruyama, Tadashi*;
PNC TJ9607 89-001, 100 Pages, 1989/06
The purpose of this study is to improve the accuracy of SiC temperature monitors by clarifing the effects of (a)heavy neutron irradiation, (b)flux and fluence of neutron, and (c)variation of temperature during shut-down sequence on recovery behavior of SiC monitors. Several types of uninstrumented in-core temperature monitors were reviewed. Especially, from the literature survey of the SiC temperature monitor, some problems which should be solved were presented. Recovery curves of eighteen SiC temperature monitors irradiated in the fast breeder reactors, JOYO, Rapsodie and Phenix to fast neutron fluences from 1.510
to 1.710
n/cm (E
0.1MeV) were measured by means of the following three methods; (1)Macroscopic length measurement using a micrometer, (2)Lattice parameter measurement by X-ray diffractometry and (3)Macroscopic length measurement at high temperature by a step-heating dilatometry. Merit and demerit, and requirements for specimen of the three methods were pointed out. The method of step-heating dilatometry was recommended as more convenient and suitable method to obtain annealing curves than the others. The following results were obtained from the recovery curve measurement. (1)For heavily irradiated specimens up to 1.710n/cm, recovery curve is obtainable by means of the macroscopic length measurement. (2)Macroscopic length or lattice parameter change shows a step-like decrease with increase in annealing temperature for the specimens which were finally irradiated in the slowly shut-downed reactor cycle. (3)Different neutron-spectrum irradiation affects on the microstructure development, but slightly on recovery behavior of SiC temperature monitor. Relation between annealing intersection temperature and irradiation temperature was not assessed because insufficient specimens were available. (4)Macroscopic length or lattice parameter did not always decrease constantly with increasing annealing ...
Osaka, Masahiko; Miwa, Shuhei; Yano, Toyohiko*
no journal, ,
Novel concept for effective utilization of magnesium silicates recovered from hazardous asbestos-derived wastes has been proposed. Basic studies on sintering behavior of the inert matrix-type fuel (IMF) including the asbestos waste-derived materials as sintering additives were carried out. Three types of IMFs were tested; MgO, SiN and Mo-based IMFs. Effects of the additive on both the inert matrices and actinide oxides were experimentally investigated. Quantitative effects of the additives on the sintering densities of the IMFs were elucidated. Effects of additives on the sintering behavior were discussed in terms of high temperature chemistry. It was found that the asbestos-derived material, magnesium silicates, can control the sintered density of several types of nuclear fuel pellets. This concept potentially interconnects a hazardous social problem of asbestos to the nuclear fuel cycle issues.
, UO or PuO as a simulant for minor actinidesYamane, Junichi*; Imai, Masamitsu*; Furuta, Katsumi*; Yano, Toyohiko*; Miwa, Shuhei; Osaka, Masahiko
no journal, ,
Silicon nitride is a promising candidate to hold those nuclei for transmutation. Silicon nitride ceramics with CeO, UO or PuO as a simulant for minor actinides were fabricated through a simple powder metallurgy process.
Miwa, Shuhei; Osaka, Masahiko; Usuki, Toshiyuki*; Yano, Toyohiko*
no journal, ,
no abstracts in English
Usuki, Toshiyuki*; Yoshida, Katsumi*; Imai, Masamitsu*; Yano, Toyohiko*; Miwa, Shuhei; Osaka, Masahiko
no journal, ,
no abstracts in English
Osaka, Masahiko; Miwa, Shuhei; Kurosaki, Ken*; Yamanaka, Shinsuke*; Uno, Masayoshi*; Yamane, Yoshihiro*; Mimura, Hitoshi*; Yano, Toyohiko*
no journal, ,
Inert matrix fuel concept that is adaptable to the fast reactor cycle system was proposed. The concept has unique characteristics of a flexible management of minor actinides and effective utilization of beneficial materials in the nuclear cycle.
N-based fuelsYano, Toyohiko*; Yoshida, Katsumi*; Imai, Masamitsu*; Miwa, Shuhei; Osaka, Masahiko
no journal, ,
no abstracts in English
Usuki, Toshiyuki; Yoshida, Katsumi*; Yano, Toyohiko*; Miwa, Shuhei; Osaka, Masahiko
no journal, ,
Minor actinides (MA) have lasting radio-toxicity. One of the possible ways to reduce radio-toxicity is transmutation of MA using nuclear reactors. Inert matrix (IM) is applied to host material for MA in transmutation. Silicon nitride (SiN) ceramics considered to be a candidate material of IM since it has a high thermal conductivity and shows good resistance to neutron irradiation. In this study, We proposed fabrication of silicon nitride ceramics based IMFs with Ce as simulating element and with magnesium silicates, i.e. enstatite (MgSiO), steatite (MgSiO) and forsterite (MgSiO), as additives. The sintered densities of SiN-based IMFs with forsterite was above 96 % theoretical 1723 K. Furthermore, The thermal conductivities of IMFs sintered at 1923 K were above 34 W/mK.
