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JAEA Reports

Separation test of heat generating nuclides from high-level liquid waste

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.

JAEA Reports

Analysis of the radioactivity concentrations in radioactive waste generated from JRR-2 and JRR-3 and Stored at the Waste Storage Facility L as compressed packages

Kinase, Akari; Goto, Katsunori*; Aono, Ryuji; Konda, Miki; Sato, Yoshiyuki; Haraga, Tomoko; Ishimori, Kenichiro; Kameo, Yutaka

JAEA-Data/Code 2024-004, 60 Pages, 2024/07

JAEA-Data-Code-2024-004.pdf:2.05MB

Radioactive wastes generated from nuclear research facilities in Japan Atomic Energy Agency are planning to be buried in the near surface disposal field as trench and pit. Therefore, it is required to establish the method to evaluate the radioactivity concentrations of radioactive wastes until the beginning of disposal. In order to contribute to this work, we collected and analyzed the samples generated from JRR-2 and JRR-3 and stored at the waste storage facility L. In this report, we summarized the radioactivity concentrations of 20 radionuclides ($$^{3}$$H, $$^{14}$$C, $$^{36}$$Cl, $$^{60}$$Co, $$^{63}$$Ni, $$^{90}$$Sr, $$^{94}$$Nb, $$^{99}$$Tc, $$^{108m}$$Ag, $$^{129}$$I, $$^{137}$$Cs, $$^{152}$$Eu, $$^{154}$$Eu, $$^{234}$$U, $$^{238}$$U, $$^{238}$$Pu, $$^{239}$$Pu, $$^{240}$$Pu, $$^{241}$$Am, $$^{244}$$Cm) which were obtained from radiochemical analysis of the samples in fiscal year 2022.

JAEA Reports

Analysis of the radioactivity concentrations in radioactive waste generated from JRR-3, JRR-4 and JRTF facilities, 2

Tobita, Minoru*; Goto, Katsunori*; Omori, Takeshi*; Osone, Osamu*; Haraga, Tomoko; Aono, Ryuji; Konda, Miki; Tsuchida, Daiki; Mitsukai, Akina; Ishimori, Kenichiro

JAEA-Data/Code 2023-011, 32 Pages, 2023/11

JAEA-Data-Code-2023-011.pdf:0.93MB

Radioactive wastes generated from nuclear research facilities in Japan Atomic Energy Agency are planning to be buried in the near surface disposal field as trench and pit. Therefore, it is required to establish the method to evaluate the radioactivity concentrations of radioactive wastes until the beginning of disposal. In order to contribute to the study of radioactivity concentration evaluation methods for radioactive wastes generated from nuclear research facilities, we collected and analyzed concrete samples generated from JRR-3, JRR-4 and JAERI Reprocessing Test Facility. In this report, we summarized the radioactivity concentrations of 23 radionuclides ($$^{3}$$H, $$^{14}$$C, $$^{36}$$Cl, $$^{41}$$Ca, $$^{60}$$Co, $$^{63}$$Ni, $$^{90}$$Sr, $$^{94}$$Nb, $$^{rm 108m}$$Ag, $$^{137}$$Cs, $$^{133}$$Ba, $$^{152}$$Eu, $$^{154}$$Eu, $$^{rm 166m}$$Ho, $$^{234}$$U, $$^{235}$$U, $$^{238}$$U, $$^{238}$$Pu, $$^{239}$$Pu, $$^{240}$$Pu, $$^{241}$$Am, $$^{243}$$Am, $$^{244}$$Cm) which were obtained from radiochemical analysis of the samples in fiscal years 2021-2022.

JAEA Reports

Analysis of the radioactivity concentrations in radioactive waste generated from JRR-2, JRR-3 and Hot laboratory

Aono, Ryuji; Mitsukai, Akina; Tsuchida, Daiki; Konda, Miki; Haraga, Tomoko; Ishimori, Kenichiro; Kameo, Yutaka

JAEA-Data/Code 2023-002, 81 Pages, 2023/05

JAEA-Data-Code-2023-002.pdf:3.0MB

Radioactive wastes generated from nuclear research facilities in Japan Atomic Energy Agency are planning to be buried in the near surface disposal field as trench and pit. Therefore, it is required to establish the method to evaluate the radioactivity concentrations of radioactive wastes until the beginning of disposal. In order to contribute to this work, we collected and analyzed the samples generated from JRR-2, JRR-3 and Hot laboratory facilities. In this report, we summarized the radioactivity concentrations of 20 radionuclides ($$^{3}$$H, $$^{14}$$C, $$^{36}$$Cl, $$^{60}$$Co, $$^{63}$$Ni, $$^{90}$$Sr, $$^{94}$$Nb, $$^{99}$$Tc, $$^{rm 108m}$$Ag, $$^{129}$$I, $$^{137}$$Cs, $$^{152}$$Eu, $$^{154}$$Eu, $$^{234}$$U, $$^{238}$$U, $$^{238}$$Pu, $$^{239}$$Pu, $$^{240}$$Pu, $$^{241}$$Am, $$^{244}$$Cm) which were obtained from radiochemical analysis of the samples in fiscal year 2020.

JAEA Reports

Analysis of the radioactivity concentrations in radioactive waste generated from JRR-3, JRR-4 and JRTF facilities

Tobita, Minoru*; Konda, Miki; Omori, Takeshi*; Nabatame, Tsutomu*; Onizawa, Takashi*; Kurosawa, Katsuaki*; Haraga, Tomoko; Aono, Ryuji; Mitsukai, Akina; Tsuchida, Daiki; et al.

JAEA-Data/Code 2022-007, 40 Pages, 2022/11

JAEA-Data-Code-2022-007.pdf:1.99MB

Radioactive wastes generated from nuclear research facilities in Japan Atomic Energy Agency are planning to be buried in the near surface disposal field. Therefore, it is required to establish the method to evaluate the radioactivity concentrations of radioactive wastes until the beginning of disposal. In order to contribute to this work, we collected and analyzed concrete, ash, ceramic and brick samples generated from JRR-3, JRR4 and JRTF facilities. In this report, we summarized the radioactivity concentrations of 24 radionuclides ($$^{3}$$H, $$^{14}$$C, $$^{36}$$Cl, $$^{41}$$Ca, $$^{60}$$Co, $$^{63}$$Ni, $$^{90}$$Sr, $$^{94}$$Nb, $$^{99}$$Tc, $$^{rm 108m}$$Ag, $$^{129}$$I, $$^{137}$$Cs, $$^{133}$$Ba, $$^{152}$$Eu, $$^{154}$$Eu, $$^{rm 166m}$$Ho, $$^{234}$$U, $$^{238}$$U, $$^{238}$$Pu, $$^{239}$$Pu, $$^{240}$$Pu, $$^{241}$$Am, $$^{243}$$Am, $$^{244}$$Cm) which were obtained from radiochemical analysis of the samples in fiscal years 2020-2021.

Journal Articles

Improvement of adsorption performances of Sr adsorption fiber and investigation for realizing simple $$^{90}$$Sr analysis

Horita, Takuma; Asai, Shiho*; Konda, Miki; Matsueda, Makoto; Hanzawa, Yukiko; Kitatsuji, Yoshihiro

Bunseki Kagaku, 69(10-11), p.619 - 626, 2020/10

https://doi.org/10.2116/bunsekikagaku.69.619
 Times Cited Count：0 Percentile：0.00(Chemistry, Analytical)

We have developed a Sr adsorption fiber for rapid analysis of $$^{90}$$Sr. The prepared Sr adsorption fiber has a Sr-extraction layer that densely retains a Sr-selective extractant, an 18-crown-6 ether derivative, on the fiber surface. Hydrophobic group-containing polymer chains embedded onto the surface of the fiber allow to form a hydrophobic phase, incorporating Sr-selective extractants. This unique surface structure provides high adsorption capacity, leading to rapid and highly efficient adsorption of Sr$$^{2+}$$. The adsorption capacity of the Sr adsorption fiber was 3 times higher than commercially available 18-crown-6 ether derivative-impregnated resin (Sr Resin). The equilibrium adsorption capacity of the Sr adsorption fiber was comparable to the Sr Resin. The retained $$^{90}$$Sr was finally determined by a GM counter. The total analysis time including the Sr adsorption and measurement was about 1 hour.

Journal Articles

Rapid separation of zirconium using microvolume anion-exchange cartridge for $$^{93}$$Zr determination with isotope dilution ICP-MS

Asai, Shiho; Hanzawa, Yukiko; Konda, Miki; Suzuki, Daisuke; Magara, Masaaki; Kimura, Takaumi; Ishihara, Ryo*; Saito, Kyoichi*; Yamada, Shinsuke*; Hirota, Hideyuki*

Talanta, 185, p.98 - 105, 2018/08

https://doi.org/10.1016/j.talanta.2018.03.026
 Times Cited Count：9 Percentile：29.37(Chemistry, Analytical)

Estimating the risks associated with radiation from long-lived fission products (LLFP) in radioactive waste is essential to ensure the long-term safety of potential disposal sites. In this study, the amount of $$^{93}$$Zr, a LLFP, was determined by ICP-MS after separating Zr from a spent nuclear fuel solution using a microvolume anion-exchange cartridge (TEDA cartridge). The TEDA cartridge achieved highly selective separation of Zr regardless of its small bed volume of 0.08 cm$$^{3}$$. The time taken to complete the Zr separation was 1.2 min with a flow rate of 1.5 mL/min, which was 10 times faster than that for a conventional anion-exchange resin column. Almost all the other elements were removed, leading to accurate measurement of $$^{93}$$Zr. The result connects experimental value to theoretical prediction provided by ORIGEN2, which requires verification. With the measured value, we demonstrated that the theoretical value is reliable enough to estimate radiation risks.

Journal Articles

Preparation of Sr adsorptive fiber by impregnating with crown ether derivative for $$^{90}$$Sr measurement

Horita, Takuma; Asai, Shiho; Konda, Miki; Hanzawa, Yukiko; Saito, Kyoichi*; Fujiwara, Kunio*; Sugo, Takanobu*; Kitatsuji, Yoshihiro

Bunseki Kagaku, 66(3), p.189 - 193, 2017/03

https://doi.org/10.2116/bunsekikagaku.66.189
 Times Cited Count：1 Percentile：2.97(Chemistry, Analytical)

A Sr-selective adsorption fiber was prepared for rapid analysis of $$^{90}$$Sr content by using radiation-induced emulsion graft polymerization and subsequent chemical modification. A polyethylene fiber with a diameter of 13 $$mu$$m was first immersed in a methanol solution of an epoxy-group-containing vinyl monomer, glycidyl methacrylate (GMA), and polyoxyethylene sorbitol ester (Tween20) as a surfactant for graft-polymerization of GMA. Octadecylamine was then bound to a polymer chain extending from the fiber surface providing hydrophobicity to the polymer chain. Dicyclohexano-18-crown-6 (DCH18C6) was finally impregnated onto the polymer chain via a hydrophobic interaction between the octadecyl moiety of the polymer chain and the cyclohexyl moiety of DCH18C6. The fiber surface structure, characterized by DCH18C6 molecules loosely entangled with polymer chains, afforded realizes the rapid and selective adsorption of Sr ions with an adsorption rate approximately 100 times higher than that of a commercially available Sr-selective resin (Sr Resin).

Journal Articles

Radiochemical analysis of rubble collected from around and inside reactor buildings at Units 1 to 4 in Fukushima Daiichi Nuclear Power Station

Sato, Yoshiyuki; Aono, Ryuji; Konda, Miki; Tanaka, Kiwamu; Ueno, Takashi; Ishimori, Kenichiro; Kameo, Yutaka

Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 13 Pages, 2017/00

http://hotlab.sckcen.be/proceedings

no abstracts in English

JAEA Reports

Preparation of Zr-91 stable isotope standard solution for determination of Zr-93 by mass spectrometry

Konda, Miki; Asai, Shiho; Hanzawa, Yukiko; Magara, Masaaki

JAEA-Technology 2015-054, 22 Pages, 2016/03

JAEA-Technology-2015-054.pdf:3.44MB

Isotope dilution mass spectrometry (IDMS) with ICP-MS is reliable method for determination of Zr-93, which is one of the long-lived fission products found in spent nuclear fuel and high-level radioactive wastes. In order to use an isotope standard solution of zirconium as the spike for IDMS, dissolving a commercially available solid isotope standard is indispensable. Prior to the dissolution of the Zr-91 isotope standard, solubility of metal zirconium in a mixture of HNO$$_3$$ and HF was evaluated using zirconium metal chips. Then, 2 mg of the Zr-91 isotope standard was dissolved with 0.2 mL of 1 M HNO$$_3$$-3 v/v% HF mixed solution, followed by adjusting the concentration of Zr-91 to approximately 1,000 $$mu$$g/g. IDMS, in which a natural isotopic abundance standard solution of zirconium was used as the spike, was employed for the determination of the concentration of Zr-91 in the prepared Zr-91 isotope standard solution. The concentration of Zr-91 in the prepared Zr-91 isotope standard solution was (9.6$$pm$$1.0) $$times$$ 10$$^2$$ $$mu$$g/g, which is in good agreement with the predicted concentration. This indicates that the Zr-91 metal isotope standard was completely dissolved with sufficient chemical stability. Additionally, no impurities were detected in the prepared Zr-91 isotope standard solution. These positive results denote that the Zr-91 isotope standard solution with the preferable quality for IDMS of Zr-93 can be obtained by the proposed dissolution procedures.

Journal Articles

Preparation of microvolume anion-exchange cartridge for inductively coupled plasma mass spectrometry-based determination of $$^{237}$$Np content in spent nuclear fuel

Asai, Shiho; Hanzawa, Yukiko; Konda, Miki; Suzuki, Daisuke; Magara, Masaaki; Kimura, Takaumi; Ishihara, Ryo*; Saito, Kyoichi*; Yamada, Shinsuke*; Hirota, Hideyuki*

Analytical Chemistry, 88(6), p.3149 - 3155, 2016/03

https://doi.org/10.1021/acs.analchem.5b04330
 Times Cited Count：8 Percentile：25.84(Chemistry, Analytical)

Neptunium-237 ($$^{237}$$Np) is one of the major long-lived radionuclides found in spent nuclear fuel. To evaluate the long-term safety of a HLW repository, the $$^{237}$$Np content in spent nuclear fuel must be determined. In this study, micro-volume anion-exchange porous polymer disk-packed cartridges were prepared for Am-Np separation, which is required prior to the measurement of $$^{237}$$ Np with ICP-MS. Disks with a volume of 0.08 cm$$^{3}$$ were cut out from porous sheets having triethylenediamine (TEDA)-containing polymer chains densely attached on the pore surface. The resulting TEDA-introduced disk cartridge was applied to a spent nuclear fuel sample. The chemical yield of Np was 90.4%, which is sufficiently high for ICP-MS measurement of $$^{237}$$Np. Compared with the conventional separation technique using commercially available anion-exchange resin columns, the time required to adsorb, wash and elute Np using the TEDA-introduced disk cartridge was reduced by 75%.

Oral presentation

Development of in-situ analytical method of $$^{90}$$Sr in contaminated water using Sr adsorptive fiber

Konda, Miki; Horita, Takuma; Asai, Shiho; Matsueda, Makoto; Hanzawa, Yukiko; Saito, Kyoichi*; Fujiwara, Kunio*; Sugo, Takanobu*; Kameo, Yutaka

no journal, , 

no abstracts in English

Oral presentation

Determination of $$^{237}$$Np in spent nuclear fuel using anion-exchange porous polymeric filter

Asai, Shiho; Hanzawa, Yukiko; Konda, Miki; Suzuki, Daisuke; Magara, Masaaki; Kimura, Takaumi; Ishihara, Ryo*; Saito, Kyoichi*; Yamada, Shinsuke*; Hirota, Hideyuki*

no journal, , 

no abstracts in English

Oral presentation

Radiochemical analysis of rubble collected from reactor buildings at Fukushima Daiichi Nuclear Power Station

Aono, Ryuji; Sato, Yoshiyuki; Konda, Miki; Tanaka, Kiwamu; Ueno, Takashi; Ishimori, Kenichiro; Kameo, Yutaka

no journal, , 

A large amount of contaminated rubble was generated by the accident at the Fukushima Daiichi Nuclear Power Station (F1NPS). For safe decommissioning of F1NPS, it is important to evaluate the composition and concentration of radionuclides in the rubble. To characterize the rubble collected at F1NPS, radiochemical analysis was conducted. From the rubble collected from reactor buildings, $$^3$$H, $$^{14}$$C, $$^{60}$$Co, $$^{63}$$Ni, $$^{79}$$Se, $$^{90}$$Sr, $$^{99}$$Tc, $$^{129}$$I, $$^{137}$$Cs, $$^{154}$$Eu, $$^{238, 239+240}$$Pu, $$^{241}$$Am and $$^{244}$$Cm were detected. The radioactivity concentrations of $$^{60}$$Co, $$^{90}$$Sr and $$^{238}$$Pu are correlated that of $$^{137}$$Cs. The radioactive ratio of $$^{60}$$Co/$$^{137}$$Cs, $$^{90}$$Sr/$$^{137}$$Cs and $$^{238}$$Pu/$$^{137}$$Cs were similar between the rubble collected from 1st floor and 5th floor of unit 1 reactor building. This result implied that regardless of sampling location in reactor building, the radioactive ratios of $$^{60}$$Co/$$^{137}$$Cs, $$^{90}$$Sr/$$^{137}$$Cs and $$^{238}$$Pu/$$^{137}$$Cs were consistent.

Oral presentation

Preparation of $$^{91}$$Zr isotope standard solution for isotope dilution mass spectrometry

Konda, Miki; Asai, Shiho; Hanzawa, Yukiko; Magara, Masaaki

no journal, , 

It is important to measure the inventory of radionuclides in high-level radioactive waste for safe and reasonable geological disposal of the waste. For verification of the reliability of the inventory assessment by measured value, we planned to quantify the long-lived nuclides $$^{93}$$Zr, that is important for safety evaluation of radioactive waste disposal, using isotope dilution inductively coupled plasma mass spectrometry (ID-ICP-MS) we developed. Spike (isotope standard solution) is required to apply the IDMS. We prepare $$^{91}$$Zr isotope standard solution by dissolving the metallic $$^{91}$$Zr isotope standard. The metallic Zr dissolves in HF, but HF dissolve glass and has high toxicity. Therefore it is necessary to reduce an amount of HF and to simplify a dissolution method. At first, we examined the dissolution method of the metallic Zr. The isotope standard of $$^{91}$$Zr was dissolved under optimum condition that obtained by the previous examination. The prepared $$^{91}$$Zr isotope standard solution was measured using the IDMS to determine concentration, which gave us a high-accuracy value.

Oral presentation

Determination of Zr isotopes in spent nuclear fuel with ICP-MS

Asai, Shiho; Hanzawa, Yukiko; Konda, Miki; Suzuki, Daisuke; Magara, Masaaki; Kimura, Takaumi

no journal, , 

Various Zr isotopes generated by U fission are found in spent fuel and HLW. Among them, $$^{93}$$Zr which has a long half-life of 1.5$$times$$10$$^{6}$$y has a potential to contribute to radiation dose over an extended period of time after the implementation of HLW disposal. The inventory estimation of $$^{93}$$Zr in HLW confirmed by measured data is the key to realize a safe and cost-efficient disposal. In this study, a simple and robust analytical technique for the determination of $$^{93}$$Zr based on ion-exchange chromatography combined with ICP-MS was developed. Interference-free measurement was achieved by a single anion-exchange step, removing Sr, Nb, and Mo which would cause spectral interferences. Additionally, major component U and radioactive components, such as Cs, Ba, and Pu, were also removed concurrently. Concentration of $$^{93}$$Zr was readily calculated with measured isotope ratios of $$^{93}$$Zr/$$^{91}$$Zr in the sample and natural Zr-spiked sample with sufficient accuracy.

Oral presentation

Preparation of 18-crown-6-ether derivative-impregnated fiber based on the radiation-induced graft polymerization

Horita, Takuma; Asai, Shiho; Konda, Miki; Hanzawa, Yukiko; Saito, Kyoichi*; Fujiwara, Kunio*; Sugo, Takanobu*; Kitatsuji, Yoshihiro

no journal, , 

High activity concentrations of $$^{90}$$Sr, which greatly exceed the regulatory limit (30 Bq/L), are detected in contaminated waters sampled in Fukushima Daiichi NPP. However, analytical method for $$^{90}$$Sr associated with time-consuming procedure causes delay in understanding the current status of $$^{90}$$Sr contamination. In this study, we have prepared a Sr adsorption fiber (Sr Fiber) based on radiation-induced graft polymerization technique to achieve a rapid adsorption of Sr ions. Dicyclohexano-18-crown-6 ether that has high affinity for Sr was impregnated in the hydrophobic interfacial phase provided by hydrophobic polymer chains attached on the fiber surface through graft polymerization. The time required to reach the Sr adsorption equilibrium for the Sr Fiber is approximately 180 times shorter than that for a commercially available Sr adsorbent (Sr Resin), showing that the Sr Fiber has a potential to efficiently reduce the analytical time of $$^{90}$$Sr.

Oral presentation

Direct beta-ray measurement of $$^{90}$$Sr adsorbed on fiber surface; Preparation of 18-crown 6-ether derivative-impregnated Sr adsorption fiber based on the radiation-induced graft polymerization

Horita, Takuma; Asai, Shiho; Konda, Miki; Hanzawa, Yukiko; Saito, Kyoichi*; Fujiwara, Kunio*; Kitatsuji, Yoshihiro

no journal, , 

There has been an increasing importance of the development of rapid separation techniques for $$^{90}$$Sr analysis, responding to needs in Fukushima Daiichi NPP. However, conventional $$^{90}$$Sr analytical methods require two different separation steps for Sr and Y, respectively, resulting in a long processing time of about one month. In this study, we prepared a Sr adsorptive fiber (Sr fiber) that has a high density Sr adsorption phase on its surface, allowing to highly efficient $$beta$$-ray counting by minimizing the self-attenuation effects. The adsorption capacity of the prepared Sr fiber was about 14 g/mol, which is equivalent to that of a commercially available Sr adsorptive resin (Sr Resin). The selectivity of the Sr fiber was nearly the same as that of the Sr resin. Considering that the Sr fiber has a specific surface area 1000 times smaller than that of the Sr Resin, the Sr ions can be concentrated to 1000 times on its surface, capable of achieving highly-efficient $$beta$$-ray counting. From these result, we confirmed that Sr fiber has adsorption capacity and selectivity necessary for highly efficient $$beta$$-ray counting of $$^{90}$$Sr.

Oral presentation

Development of crown-ether-derivative impregnated fiber for in-situ analytical method of $$^{90}$$Sr; Study of safe Sr adsorption operation without using nitric acid

Konda, Miki; Horita, Takuma; Asai, Shiho*; Hanzawa, Yukiko; Saito, Kyoichi*; Fujiwara, Kunio*; Sugo, Takanobu*; Ishimori, Kenichiro; Kameo, Yutaka

no journal, , 

no abstracts in English

Oral presentation

Anion-exchange porous polymer sheet for rapid analysis of $$^{237}$$Np in spent nuclear fuel

Asai, Shiho; Hanzawa, Yukiko; Konda, Miki; Suzuki, Daisuke; Magara, Masaaki; Kimura, Takaumi; Ishihara, Ryo*; Saito, Kyoichi*; Yamada, Shinsuke*; Hirota, Hideyuki*

no journal, , 

no abstracts in English

24 (Records 1-20 displayed on this page)