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Journal Articles

Fabrication and short-term irradiation behaviour of Am-bearing MOX fuels

Kihara, Yoshiyuki; Tanaka, Kosuke; Koyama, Shinichi; Yoshimochi, Hiroshi; Seki, Takayuki; Katsuyama, Kozo

NEA/NSC/R(2017)3, p.341 - 350, 2017/11

In order to investigate the effect of the addition of americium to MOX fuels on the irradiation behaviour, the "Am-1" program is being conducted at the JAEA. The Am-1 program consists of two short-term irradiation tests of 10-min and 24-h irradiation periods, and a steady-state irradiation test. The short-term irradiation tests and their post irradiation examinations (PIEs) have been successfully completed. To date, the data for PIE of the Am-MOX fuels focused on the microstructural evolution and redistribution behaviour of Am at the initial stage of irradiation have been obtained and reported. In this paper, the results obtained from the Am-1 program are reviewed and detailed descriptions of the fabrication and inspection techniques for the Am-MOX fuels prepared for the program are provided. PIE data for the Am-MOX fuels at the initial stage of irradiation have been accumulated. In this paper, unpublished PIE data for the Am-MOX fuels are also presented.

Journal Articles

Outline of Japan Atomic Energy Agency's Okuma Analysis and Research Center, 2; Labolatory-1

Sugaya, Yuki; Sakazume, Yoshinori; Akutsu, Hideyuki; Inoue, Toshihiko; Yoshimochi, Hiroshi; Sato, Soichi; Koyama, Tomozo; Nakayama, Shinichi

Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 8 Pages, 2017/00

The Japan Atomic Energy Agency has been developing the research and development facilities, "Okuma Analysis and Research Center", in order to ascertain the properties of radioactive wastes and fuel debris towards the decommissioning of TEPCO's Fukushima Daiichi Nuclear Power Station. This paper outlines the concept of "Laboratory-1" which will analyze low and medium level samples in the Okuma Analysis and Research Center with a focus on the research plan.

Journal Articles

Outline of Japan Atomic Energy Agency's Okuma Analysis and Research Center, 3; Laboratory-2

Ito, Masayasu; Ogawa, Miho; Inoue, Toshihiko; Yoshimochi, Hiroshi; Koyama, Shinichi; Koyama, Tomozo; Nakayama, Shinichi

Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 7 Pages, 2017/00

Laboratory-2 of the Okuma Analysis and Research Center will be used for the technological development of techniques to treat and dispose fuel debris, etc. The specific analytical content and its importance has been discussed by an experts committee in FY 2016. The committee regarded fuel debris retrieval and criticality control related topics as the most important content. As a result, it will be a priority to introduce equipment to perform examination such as shape and size measurement, compositional and nuclide analysis, hardness and toughness test, and radiation dose rate measurement. In addition, since sample will have high dose rates (1 Sv/h or more) at the time of reception, hot cells with enough radiation shielding ability will be used. In the hot cell, the pre-processing will be performed, such as cutting and dissolution of samples. Processed samples will be examined in concrete cells, steel cells, glove boxes and fume hoods. Detail design of Laboratory-2 started on FY 2017.

Journal Articles

The Outline of Japan Atomic Energy Agency's Okuma Analysis and Research Center, 1; The Total progress of Labolatory-1 and Labolatory-2

Inoue, Toshihiko; Ogawa, Miho; Sakazume, Yoshinori; Yoshimochi, Hiroshi; Sato, Soichi; Koyama, Shinichi; Koyama, Tomozo; Nakayama, Shinichi

Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 7 Pages, 2017/00

Decommissioning of TEPCO's 1F is in progress according to the Roadmap. The Roadmap assigned the construction of a hot laboratory and analysis to the JAEA. The hot laboratory, Okuma Analysis and Research Center consists of the three buildings; Administrative building, the Laboratory-1 and Laboratory-2. The Laboratory-1 and Laboratory-2 are hot laboratories. Laboratory-1 is for radiometric analysis of low and medium level radioactive rubble and secondary wastes. The license of the Laboratory-1's implementation was approved by The Secretariat of the Nuclear Regulation Authority and the construction started in April 2017 and plans an operational start in 2020. Laboratory-2 provides concrete cells, steel cells for the analysis of the fuel debris and high level radioactive rubble. The Laboratory-2's major analysis items is reviewed by review meeting organized of cognoscente.

Journal Articles

Program of the analysis and research laboratory for Fukushima-Daiichi and advanced techniques to be applied in the laboratory

Sekio, Yoshihiro; Yoshimochi, Hiroshi; Kosaka, Ichiro; Hirano, Hiroyasu; Koyama, Tomozo; Kawamura, Hiroshi

Proceedings of 52nd Annual Meeting of Hot Laboratories and Remote Handling Working Group (HOTLAB 2015) (Internet), 8 Pages, 2015/09

Due to the Fukushima Daiichi Nuclear Power Plant accident in March 2011, the safe and secure implementations of the decommissioning for Fukushima Daiichi Nuclear Power Plant has been positioned as the urgent tasks in Japan. Japan Atomic Energy Agency has a critical mission of analysing radioactive wastes having generated by the accident for long-term storage and disposal methods. This will be performed in two hot laboratories to be constructed in Okuma Analysis and Research Center at Fukushima Daiichi Nuclear Power Plant site. In one laboratory, radioactive wastes such as rubbles and secondary wastes will be treated, whereas debris such as fuel debris and high dose structural materials will be handled in the other laboratory. The detail considerations for advanced techniques and experimental apparatus to be installed are underway.

Journal Articles

Oxidation behavior of Am-containing MOX fuel pellets in air

Tanaka, Kosuke; Yoshimochi, Hiroshi; Obayashi, Hiroshi; Koyama, Shinichi

Energy Procedia, 71, p.282 - 292, 2015/05

 Times Cited Count:2 Percentile:8.81

Americium-containing MOX (Am-MOX) fuels were subjected to heating tests using thermogravimetric and differential thermal analysis (TG-DTA) measurements in a flowing gas atmosphere of dry air to investigate the effect of Am addition on oxidation behavior of MOX fuel.

Journal Articles

Restructuring and redistribution of actinides in Am-MOX fuel during the first 24h of irradiation

Tanaka, Kosuke; Miwa, Shuhei; Sekine, Shinichi; Yoshimochi, Hiroshi; Obayashi, Hiroshi; Koyama, Shinichi

Journal of Nuclear Materials, 440(1-3), p.480 - 488, 2013/09

 Times Cited Count:3 Percentile:66.1(Materials Science, Multidisciplinary)

In order to confirm the effect of minor actinide addition on irradiation behavior of MOX fuel pellets, 3% and 5% americium-containing MOX (Am-MOX) fuels were irradiated for 10 minutes at 43 kW/m and for 24 hours at 45 kW/m in the experimental fast reactor Joyo. Two nominal values of the fuel pellet oxygen-to-metal ratio (O/M), 1.95 and 1.98, were used as a test parameter. Emphasis was placed on the behavior of restructuring and redistribution of actinides which directly affect the fuel performance and the fuel design for fast reactors. Microstructural evolutions in the fuels were observed by optical microscopy and redistribution behavior of constituent elements was determined by mapping and quantitative point analyses of EPMA.

Journal Articles

MgO-based inert matrix fuels for a minor actinides recycling in a fast reactor cycle

Miwa, Shuhei; Osaka, Masahiko; Usuki, Toshiyuki; Sato, Isamu; Tanaka, Kosuke; Hirosawa, Takashi; Yoshimochi, Hiroshi; Onose, Shoji

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 8 Pages, 2011/12

A new fast reactor (FR) cycle concept was previously proposed that incorporates MgO-based inert matrix fuels (IMFs) containing minor actinides harmonious with the existing FR cycle technologies. A basic study of MgO-based IMFs was made regarding their fabrication, characterization and reprocessing in terms of applicability to existing FR cycle technology. It was concluded from these basic investigations of MgO-based IMFs that the existing FR cycle technologies can be applied to those for MgO-based IMFs, and the basic technologies of MgO-based IMFs containing minor actinides harmonious with the existing FR cycle technologies were established.

Journal Articles

Status of PIE technology development in JAEA-Oarai

Tanaka, Kosuke; Kawamata, Kazuo; Yoshimochi, Hiroshi; Sozawa, Shizuo; Onose, Shoji; Niimi, Motoji; Asaka, Takeo

Proceedings of 1st Asian Symposium on Material Testing Reactors (ASMTR 2011), p.71 - 76, 2011/02

Post irradiation examination (PIE) facilities have been operated for about 40 years at the Oarai Research and Development Center of the Japan Atomic Energy Agency to investigate the performance and soundness of irradiated fuels and materials. The JMTR Hot Laboratory (JMTR-HL) was founded in 1971 mainly to examine the objects irradiated in the Japan Material Testing Reactor (JMTR). The Alpha-Gamma Facility (AGF) was constructed as the first laboratory to perform PIE of plutonium-bearing fuels for Japanese fast reactor development programs. This facility started hot operation in 1971 and has performed physical, metallurgical, and chemical examinations of irradiated fuels including uranium plutonium mixed oxide fuels. A renewal plan for the JMTR-HL and AGF is now in progress, associated with re-operation of the JMTR.

Journal Articles

Development of a fabrication method for oxide fuels containing metallic dopant materials

Ishii, Tetsuya; Yoshimochi, Hiroshi; Tanaka, Kenya

Nippon Genshiryoku Gakkai Wabun Rombunshi, 9(2), p.207 - 218, 2010/06

In order to develop an innovative fuel fabrication method for americium containing oxide fuels, a feasibility study of metallic U and Mo-doped oxide fuel concept with extruding granulated oxide material was conducted using UO$$_{2}$$. In the concept, it is expected that doped U should reduce the effective oxygen potential and doped Mo should increase the thermal conductivity of the fuel. In this study, sintering tests of U and Mo-doped UO$$_{2}$$ powder were done and thermal conductivities of the sintered material were evaluated. From the results, it can be seen that the doped U and Mo would function as a oxygen potential reducer and thermal conductivity improver, respectively. And it can be seen that the U and Mo doped oxide fuel pellets would be fabricated successfully using hot pressing. Also, from the results of a sintering test of U and Mo-doped extruding granulated UO$$_{2}$$, it can be seen that the extruding granulated substances have a preferable sintering characteristic.

Journal Articles

Microstructural evolution and Am migration behavior in Am-containing MOX fuels at the initial stage of irradiation

Tanaka, Kosuke; Miwa, Shuhei; Sato, Isamu; Osaka, Masahiko; Hirosawa, Takashi; Obayashi, Hiroshi; Koyama, Shinichi; Yoshimochi, Hiroshi; Tanaka, Kenya

Actinide and Fission Product Partitioning and Transmutation, p.179 - 187, 2010/00

In order to investigate the effect of americium addition to MOX fuels on the irradiation behavior, the "Am-1" program is being conducted in JAEA. The Am-1 program consists of two short-term irradiation tests of 10-minute and 24-hour irradiations and a steady-state irradiation test. The short-term irradiation tests were successfully completed and the post irradiation examinations (PIEs) are in progress. The PIEs for Am-containing MOX fuels focused on the microstructural evolution and redistribution behavior of Am at the initial stage of irradiation and the results to date are reported. Successful development of fabrication technology with remote handling and evaluation of thermo-chemical properties based on the out-of-pile experiments are described with an emphasis on the effects of Am addition on the MOX fuel properties.

Journal Articles

Evaluation of MA recycling concept with high Am-containing MOX (Am-MOX) fuel and development of its related fuel fabrication process

Tanaka, Kenya; Ishii, Tetsuya; Yoshimochi, Hiroshi; Asaka, Takeo

Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.2045 - 2050, 2009/09

As a part of the economic evaluation of the MA recycling system, the management cost of high level radioactive waste was estimated quantitatively. The development of an innovative fuel fabrication process has been done by using UO$$_{2}$$ powder, U metal particles and Mo powder. From comparisons of granulated material characteristics, two candidate methods, mixing granulation (MIX/G) and extruding granulation (EXT/G), were considered to have good feasibility as the fuel fabrication process. In the preliminary sintering test of granulated UO$$_{2}$$ obtained by EXT/G, a high density UO$$_{2}$$ pellet (97% of TD) with 5wt% of U and 5wt% of Mo was successfully sintered. From the results of thermal conductivity measurements, it was confirmed that the dispersion of Mo powder and U metal into the oxide matrix was an effective way to improve the characteristic.

Journal Articles

Microstructure and elemental distribution of americium-containing uranium plutonium mixed oxide fuel under a short-term irradiation test in a fast reactor

Tanaka, Kosuke; Miwa, Shuhei; Sato, Isamu; Hirosawa, Takashi; Obayashi, Hiroshi; Koyama, Shinichi; Yoshimochi, Hiroshi; Tanaka, Kenya

Journal of Nuclear Materials, 385(2), p.407 - 412, 2009/03

 Times Cited Count:18 Percentile:19.06(Materials Science, Multidisciplinary)

In order to confirm the effect of americium addition on irradiation behavior of MOX fuel, the "Am-1" program is being conducted in Joyo. The Am-1 program consists of two short-term irradiation tests and a steady-state irradiation test. The short-term irradiation tests were successfully completed and the post irradiation examinations are in progress. This paper reports on the results of PIEs for Am-containing MOX fuel irradiated for 10 minutes. MOX fuel pellets containing 3% or 5% Am were fabricated in a shielded air-tight hot cell using a remote handling technique. The oxygen to metal ratio (O/M) of these fuel pellets was 1.98. They were irradiated at peak linear heating rate of about 43 kW/m. The ceramography results showed that structural changes such as lenticular voids and a central void occurred early, within the brief 10 minutes of irradiation. The results of EPMA revealed that Am migrated to the radial center of the fuel pellet up the temperature gradient.

Journal Articles

Microstructure and elemental distribution of americium-containing MOX fuel under the short-term irradiation tests

Tanaka, Kosuke; Hirosawa, Takashi; Obayashi, Hiroshi; Koyama, Shinichi; Yoshimochi, Hiroshi; Tanaka, Kenya

JAEA-Conf 2008-010, p.288 - 296, 2008/12

In order to investigate the effect of americium addition to MOX fuels on the irradiation behavior, the "Am-1" program is being conducted in JAEA. The Am-1 program consists of two short-term irradiation tests of 10-minute and 24-hour irradiations and a steady-state irradiation test. The short-term irradiation tests were successfully completed and the post irradiation examinations (PIEs) are in progress. The PIEs for Am-containing MOX fuels focused on the microstructural evolution and redistribution behavior of Am at the initial stage of irradiation and the results to date are reported.

Journal Articles

Phase behavior of PuO$$_{2-x}$$ with addition of 9% Am

Miwa, Shuhei; Osaka, Masahiko; Yoshimochi, Hiroshi; Tanaka, Kenya; Kurosaki, Ken*; Uno, Masayoshi*; Yamanaka, Shinsuke*

Journal of Alloys and Compounds, 444-445, p.610 - 613, 2007/10

 Times Cited Count:14 Percentile:34.94(Chemistry, Physical)

no abstracts in English

Journal Articles

Innovative oxide fuels doped with minor actinides for use in fast reactors

Osaka, Masahiko; Miwa, Shuhei; Tanaka, Kosuke; Sato, Isamu; Hirosawa, Takashi; Obayashi, Hiroshi; Mondo, Kenji; Akutsu, Yoko; Ishi, Yohei; Koyama, Shinichi; et al.

WIT Transactions on Ecology and the Environment, Vol.105, p.357 - 366, 2007/06

no abstracts in English

JAEA Reports

Behavior on O/M ratio for Am containing MOX fuel

Sato, Isamu; Seki, Takayuki*; Ishi, Yohei; Mondo, Kenji; Yoshimochi, Hiroshi; Tanaka, Kenya

JAEA-Research 2007-013, 63 Pages, 2007/03

JAEA-Research-2007-013.pdf:4.89MB

In air atmosphere, the weight of Am-MOX fuel relatively rapidly increased, which change rate strongly depended on the initial O/M ratio. The lower initial O/M ratio is, the higher the rate is. However, for the other MOX fuel containing little Am, the equivalent behavior have been observed, which indicated that not only Am but also the property of law material powder affected the behavior. The X-ray diffraction pattern change as time goes by was observed, which was a evidence that the O/M ratio change might arise from crystallographic one. The rate of O/M ratio change is a function of water vapor pressure in the atmosphere. If the water vapor pressure would is set to be quite low (ex. $$<$$ 1ppm), the O/M ratio change could effectively been avoided. On a model basis of Am(III) and U(V) existence, we could explain the O/M ratio dependence of the lattice parameter of Am-MOX fuel near O/M ratio, 2.00 better.

Journal Articles

An Experimental investigation of the effects of americium addtion to (U,Pu)O$$_{2-x}$$ on phase relation

Osaka, Masahiko; Miwa, Shuhei; Yoshimochi, Hiroshi; Tanaka, Kenya; Kurosaki, Ken*; Yamanaka, Shinsuke*

Recent Advances in Actinide Science, p.406 - 408, 2006/06

Phase relation of (U,Pu,Am)O$$_{2-x}$$ solid solution was experimentally investigated by using XRD, ceramograhy and DTA. It is concluded from the present investigation that Am is likely to exist as fixed valences of Am$$^{3+}$$ and corresponding amount of U$$^{4+}$$ is substituted by the U$$^{5+}$$ for the preservation of electrical neutrarity.

Journal Articles

Fabrication technology for MOX fuel containing AmO$$_{2}$$ by an in-cell remote process

Yoshimochi, Hiroshi; Ishi, Yohei; Seki, Takayuki*; Mondo, Kenji; Sekine, Shinichi*; Koyama, Shinichi

Saikuru Kiko Giho, (28), p.9 - 20, 2005/09

An in-cell remote fabrication technique has been developed for MOX fuel pellets containing 3 and 5% americium (Am-MOX) at the Alpha-gamma Facility (AGF) in O-arai Engineering Center. A series of fuel pellet and the pin fabrication apparatuses were systematically installed in hot cell to make fabrication flow easier. After cold run and some modifications, they were remotely controlled from a panel in the operation room outside the hot cell as much as possible. From a preliminary UO2 pellet test and consequently plutonium pellet fabrication run, actual range of ball milling time, pressing and sintering condition were focused for Am-MOX pellet fabrication. As the next step, moisturized atmosphere was found out to remove the heterogeneity structure of 5% Am-MOX pellet. Finally, we established an optimized fabrication condition of 5% Am-MOX pellet sintered at 1700$$^{circ}$$C for 3h in an atmosphere of 5% H2-95% Ar with total moisture of 2000 ppm. Moreover it is important that the atmosphere has to be changed to dry gas at 800$$^{circ}$$C during cool down.

JAEA Reports

Effect of Oxygen Potential on the Sintering behavior of MOX fuel containing Am

Miwa, Shuhei; Osaka, Masahiko; Yoshimochi, Hiroshi; Tanaka, Kenya; Seki, Takayuki*; Sekine, Shinichi*

JNC-TN9400 2005-023, 43 Pages, 2005/04

JNC-TN9400-2005-023.pdf:3.56MB

The effect of oxygen potential on the sintering behavior of MOX fuel containing Am (Am-MOX) was investigated. Green pellets of Am-MOX were prepared by a conventional powder metallurgical technique. For Am-MOX fuel pellets sintered at various oxygen potential conditions, density measurement, microstructural observation and elements analyses by EPMA were performed High density pellets having good structure were obtained due to oxygen potential change of sintering atmosphere from high oxygen potential to low oxygen potential at 800$$^{circ}$$C in the cooling process.For the pellets sintered at -520 kJ/mol, -390 kJ/mol and -340 kJ/mol, the sintered density increases with increase of oxygen potential up to -390 kJ/mol (threshold oxygen potential), then decreases above the threshold oxygen potential. This tendency is similar to that observed in the (U,Gd)O$$_{2 }$$ system. The differences of sintering behavior for Am-MOX pellets which were observed by changing the oxygen potential were attributable to the difference of pore structure, which was supposed to be caused by the valence state of Am in the oxides. On the other hands, the grain size of Am-MOX pellet sintered at -520 kJ/mol was almost the same as that at -390 kJ/mol. Homogeneous distribution of U, Pu and Am was obtained at pellets sintered both -520 and -390 kJ/mol in these sintering conditions. For the pellets sintered at 1500$$^{circ}$$C , 1600$$^{circ}$$C , 1700$$^{circ}$$C , the high dense pellets are obtained, therefore This results shows the the possibility of fabrication of good fuel pellets at lower temperature than 1700$$^{circ}$$C

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