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Hotoku, Shinobu; Ban, Yasutoshi; Konda, Miki; Kitatsuji, Yoshihiro
JAEA-Technology 2025-009, 33 Pages, 2025/11
JAEA-Technology-2025-009.pdf:1.9MB
High-level liquid waste (HLLW) produced from reprocessing of spent nuclear fuels contains heat generating nuclides such as Sr-90, Y-90, Cs-137, Ba-137m, and Am-241. Separation and recovery of these nuclides lead to reduce the volume and toxicity of high-level waste. Furthermore, the recovered nuclides and elements could be utilized as resources after purification. In this test, Sr separation by extraction chromatography using Sr resin and Pb resin, Cs separation by co-precipitation using ammonium phosphomolybdate (AMP), and Am separation by solvent extraction using alkyl diamideamine (ADAAM) were carried out, cold tests were performed for the separation of Cs and Sr in a nitric acid solution. Based on the results, hot tests were performed using dissolution solutions of spent fuel at the Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF), and each component contained in the separated solution was analyzed. In the Sr separation by extraction chromatography, most of Sr was separated from other elements using 8 mol/L nitric acid for absorption and 0.02 mol/L nitric acid for elution. In the separation of Cs, more than 99.9% of Cs was selectively co-precipitated by adding AMP to the HLLW, in which nitric acid concentration was adjusted to 3.1 mol/L. In solvent extraction of Am by ADAAM, 81.4% of Am-241 was recovered by a single stage batch experiment. Since Sr, Cs, and Am were properly separated and recovered from HLLW, the effectiveness of the present separation method was successfully demonstrated.
Nakashio, Nobuyuki*; Osugi, Takeshi; Kurosawa, Shigenobu; Ishikawa, Joji; Hemmi, Ko; Iketani, Shotaro; Yokobori, Tomohiko
JAEA-Technology 2022-016, 47 Pages, 2022/08
JAEA-Technology-2022-016.pdf:2.23MB
The Nuclear Science Research Institute (NSRI) of the Japan Atomic Energy Agency (JAEA) started operation of the Advanced Volume Reduction Facilities (AVWF) for production of waste packages for disposal of low-level radioactive solid wastes (LLW). To clarify the operating conditions for homogenization of non-metallic LLW, preliminary tests were carried out using the plasma melting furnace of the non-metal melting unit. The fluidity of molten waste influences homogenization conditions of solidified products. It was clarified that the viscosity, which is determined by the chemical composition and the melting temperature, influence the fluidity of molten waste greatly through previous literature review and the small-scale melting tests. In the preliminary tests, the simulated waste with a cold tracer loaded in 200 L drums were melted. Using the waste chemical components (basicity, iron oxide concentration) as an experimental parameter, the homogeneity of the chemical components of the solidified product was investigated and the homogenization conditions of melting tests were examined. The retention ratio of the tracer in the molten bath was also confirmed. The viscosity of the molten wastes was measured and the correlation with homogeneity was examined. In addition, the technical requirements that should be concerned in advance for future actual operation were discussed.
Suzuki, Hideya*; Sasaki, Yuji; Sugo, Yumi; Apichaibukol, A.; Kimura, Takaumi
Radiochimica Acta, 92(8), p.463 - 466, 2004/08
Times Cited Count：95 Percentile：97.58(Chemistry, Inorganic & Nuclear)The promising extractant for the partitioning of HLLW, TODGA, was used and investigated for the extraction of Sr(II) and separation from Am(III). Both metal ions can be extracted by TODGA based on the extraction reaction accompanying the neutral HNO
as well as the counter anion, NO
. The mixture of TODGA and monoamide can reduce the distribution ratio of Sr(II), compared to the D(Sr) without monoamide, this solvent may extract only Am(III) with holding Sr(II) in the aqueous phase. After extraction of An by TODGA and monoamide, Sr(II) remaining in HLLW can be extracted by using enough high concentration of TODGA at the next step. Because of its high D value, Sr(II) can be coextracted with An by TODGA. It was observed that D(Sr) decrease with an increase of HNO from 3M to 6M HNO at the same TODGA concentration, while Am(III) has still high D values at least until 6M HNO. By using 6M HNO of aqueous phase, An and Sr(II) can be separated after coextraction.
; Takeda, Tsuneo; Muraoka, Susumu; ;
Journal of Nuclear Science and Technology, 33(3), p.268 - 270, 1996/03
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)no abstracts in English
H.K.Lee*; Ishiyama, Shintaro; Eto, Motokuni; ;
Proc. of the Int. Symp. on Material Chemistry in Nuclear Environment, p.525 - 537, 1992/00
no abstracts in English
; Kawamura, Kazuhiro; Takahashi, Takeshi
PNC TN8410 91-247, 29 Pages, 1991/12
The effects of noble metal elements such as ruthenium, rhodium and palladium on the viscosity and electrical resistivity of simulated nuclear waste glass were studied. The glass enriched with noble metals showed the viscosity of a non-Newtonian fluid. The viscosity of the waste g1ass with 10 wt% RuO
was 3 to 7 times higher than that of glass without noble metals. The RuO was mainly responsible for the increase in viscosity for the glass enriched with noble metals. Electrical resistivity of the glass with 15 wt% RuO, was one seventh to two orders of magnitude lower than that of glass without noble metals. The three noble metals contributed to the decrease in resistivity. The quantitative effects of noble metals on these properties were obtained.
Igarashi, Hiroshi; Kato, Ko; Takahashi, Takeshi
PNC TN8410 91-274, 18 Pages, 1991/11
None
; ; Takeda, Tsuneo; Muraoka, Susumu;
Hoshasei Haikibutsu Kenkyu Renrakukai Rombunshu, VI, p.64 - 75, 1991/00
no abstracts in English
Kato, Shohei; ; ; Kokubun, Morinobu
Hoken Butsuri, 22, p.453 - 465, 1987/00
no abstracts in English
Morita, Sakura; Hanaki, Shotaro; Maruyama, Shumpei; Hojo, Yuki; Nagahama, Aoi; Kudo, Isamu; Soejima, Goro; Nakamura, Yasuyuki; Kuroki, Ryoichiro
no journal, ,
The concentrated liquid waste of Fugen (ATR) is planned to be mixed and solidified using a cement mixing and solidification unit. In this facility, the concentrated liquid waste and cement as the solidification material, are mixed in the mixer and then filled into drums as the disposal containers, to produce a homogeneous and uniform waste package. The facility is being installed onsite with plans to begin service in 2026. This presentation reports on the overview of the solidification facilities and plans for putting it into service.
Hanaki, Shotaro; Morita, Sakura; Maruyama, Shumpei; Hojo, Yuki; Nagahama, Aoi; Kudo, Isamu; Sato, Junya; Tabata, Koichi; Soejima, Goro; Nakamura, Yasuyuki
no journal, ,
Waste packages are needed to satisfy the waste acceptance criteria (WAC) set by the landfill site. To produce sound waste packages that meet such criteria as uniaxial compressive strength and porosity, it is necessary to establish the range of parameters such as mixing time, rotation speed of mixing blade, and amount of cement. This presentation reports on the Beaker scale cold test conducted to determine the cement-to-waste ratio (C/W) which affects the number of waste packages produced and uniaxial compressive strength.
Suzuki, Hideya*; Ban, Yasutoshi; Tsubata, Yasuhiro; Hotoku, Shinobu; Tsutsui, Nao; Kurosawa, Tatsuya*; Shibata, Mitsunobu*; Kawasaki, Tomohiro*; Matsumura, Tatsuro
no journal, ,
A highly practical hybrid-type (soft 
-donor and hard 
-donor) extractant, which is an alkyldiamideamine (ADAAM), was investigated for the minor actinides (MA) separation. The new process aims at recovering americium (Am) alone from high-level waste liquid (HLLW) using an ADAAM. The principle of the process is based on the extraction of Am together with light lanthanides (La, Ce, Pr and Nd) and Mo having close values of distribution ratio, while curium, other lanthanides, and other fission products remain in the aqueous phase. The Am was subsequently selectively stripped from the light lanthanides and Mo using mixed solution (DTPA, malonic acid and ammonium nitrate). As a result, Am was directly separated from the simulated HLLW with high yield (95%).
Amamoto, Ippei; Kobayashi, Hidekazu; Ayame, Yasuo; Onishi, Takashi; Inaba, Yusuke*; Utsumi, Kazuo*; Takeshita, Kenji*; Onoe, Jun*; Koshizaka, Akiko*; Hasegawa, Yoshio*
no journal, ,
The development on synthesis of sorbent was carried out to separate the platinum group species such as Ru, Rh and Pd in high-level radioactive liquid waste. After sorption test using developed sorbents, the spherical porous silica impregnated aluminium ferrocyanide was effective to sorb the species of Ru and Pd in the nitric solution. This sorption effect remained stable when tested on actual high-level radioactive liquid waste. It was found that more treated waste could be loaded into the borosilicate glass than untreated waste.
-P for Am/Cm Recovery from genuine HLLWSato, Daisuke; Watanabe, So; Arai, Yoichi; Nakamura, Masahiro; Arai, Tsuyoshi*; Sano, Yuichi; Shibata, Atsuhiro; Takeuchi, Masayuki
no journal, ,
