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Journal Articles

High temperature physicochemical properties of irradiated fuels

Ishikawa, Takashi; Onishi, Takashi; Hirosawa, Takashi; Tanaka, Kosuke; Katsuyama, Kozo

Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 10 Pages, 2017/00

Journal Articles

Thermophysical properties of americium-containing barium plutonate

Tanaka, Kosuke; Sato, Isamu; Hirosawa, Takashi; Kurosaki, Ken*; Muta, Hiroaki*; Yamanaka, Shinsuke*

Journal of Nuclear Science and Technology, 52(10), p.1285 - 1289, 2015/10

 Times Cited Count:2 Percentile:74.39(Nuclear Science & Technology)

Polycrystalline specimens of americium-containing barium plutonate have been prepared by mixing the appropriate amounts of (Pu$$_{0.91}$$Am$$_{0.09}$$)O$$_{2}$$ and BaCO$$_{3}$$ powders followed by reacting and sintering at 1600 K under the flowing gas atmosphere of dry-air. The sintered specimens had a single phase of orthorhombic perovskite structure and were crack-free. The elastic moduli were determined from the longitudinal and shear sound velocities. The Debye temperature was also determined from the sound velocities and lattice parameter measurements. The thermal conductivity was calculated from the measured density at room temperature, literature values of heat capacity, and thermal diffusivity measured by laser flash method in vacuum. The thermal conductivity of americium-containing barium plutonate was roughly independent of the temperature and was almost the same magnitude as that of BaPuO$$_{3}$$ and BaUO$$_{3}$$.

Journal Articles

Research program for the evaluation of fission product release and transport behavior focusing on FP chemistry

Sato, Isamu; Miwa, Shuhei; Tanaka, Kosuke; Nakajima, Kunihisa; Hirosawa, Takashi; Iwasaki, Maho; Onishi, Takashi; Osaka, Masahiko; Takai, Toshihide; Amaya, Masaki; et al.

Proceedings of 2014 Water Reactor Fuel Performance Meeting/ Top Fuel / LWR Fuel Performance Meeting (WRFPM 2014) (USB Flash Drive), 6 Pages, 2014/09

A new research program on severe accidents is lunched for the evaluation of FP release and transport behavior in BWR system. The purpose of the program is to improve the FP release and transport model using experimental database about FP chemistry focusing on Cs and I chemistry. In this program, effects of B including in control rod materials, B$$_{4}$$C for the Cs and I chemistry are paid attention. The experimental database used for the improvement will consist of results to obtain with newly-prepared test device under atmosphere with broad-ranging oxygen and/or steam partial pressure simulated those in BWR. The state of preparation for these experimental studies and analyses is introduced. In addition, the preliminary test was moved into action to show B chemical effect on Cs and I transport under one of the processes, which is deposited Cs compounds and B vapor and aerosol interaction. In this experiment, a "B stripping effect" to deposited CsI was observed.

Journal Articles

Effects of interaction between molten zircaloy and irradiated MOX fuel on the fission product release behavior

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

Journal of Nuclear Science and Technology, 51(7-8), p.876 - 885, 2014/07

 Times Cited Count:5 Percentile:51.92(Nuclear Science & Technology)

As a first step for obtaining experimental data on the effects of high-temperature chemical interaction on fission product (FP) release behavior, we focused on the dissolution of irradiated uranium plutonium mixed oxide (MOX) fuel by molten zircaloy (Zry), and carried out a heating test under the reducing atmosphere. Pieces of an irradiated MOX fuel pellet and cladding were subjected to the heating test at 2373 K for 5 min. The fractional release rate of cesium (specifically $$^{137}$$Cs) was monitored during the test and its release behavior was evaluated. The observation of microstructures and measurements of elemental distribution in the heated specimen were also performed. We demonstrated experimentally that the fuel dissolution by molten Zry accelerated the release of Cs from the fuel pellets.

JAEA Reports

Evaluation of fission product and actinide release behaviors focusing on their chemical forms; Phase relation and fission product release behavior resulting from interaction between molten zircaloy and irradiated MOX fuel

Tanaka, Kosuke; Miwa, Shuhei; Sato, Isamu; Hirosawa, Takashi; Sekine, Shinichi; Seki, Takayuki*; Tokoro, Daishiro*; Obayashi, Hiroshi; Koyama, Shinichi

JAEA-Research 2013-022, 62 Pages, 2014/01

JAEA-Research-2013-022.pdf:33.64MB

In order to establish the method for heating tests focused on the fission product release resulting from the high temperature chemical interaction between fuel and cladding material and to obtain the novel data on fission product release behaviors, the heating test was carried out with irradiate MOX fuel pellet and cladding.

Journal Articles

Americium and plutonium release behavior from irradiated mixed oxide fuel during heating

Sato, Isamu; Suto, Mitsuo; Miwa, Shuhei; Hirosawa, Takashi; Koyama, Shinichi

Journal of Nuclear Materials, 437(1-3), p.275 - 281, 2013/06

 Times Cited Count:5 Percentile:53.31(Materials Science, Multidisciplinary)

To obtain the source term data in severe accidents for advanced reactors, americium and plutonium release behaviors were evaluated with thermochemical consideration for release kinetics and adhere mechanism.

Journal Articles

Thermophysical properties of perovskite type alkaline-earth metals and plutonium complex oxides

Tanaka, Kosuke; Sato, Isamu; Hirosawa, Takashi; Kurosaki, Ken*; Muta, Hiroaki*; Yamanaka, Shinsuke*

Journal of Nuclear Materials, 422(1-3), p.163 - 166, 2012/03

 Times Cited Count:7 Percentile:43.25(Materials Science, Multidisciplinary)

Polycrystalline specimens of strontium plutonate, SrPuO$$_{3}$$, have been prepared by mixing the appropriate amounts of PuO$$_{2}$$ and SrCO$$_{3}$$ powders followed by reacting and sintering at 1600 K under the flowing gas atmosphere of dry-air. The sintered specimens had a single phase of orthorhombic perovskite structure and were crack-free. The elastic moduli of SrPuO$$_{3}$$ were determined from the longitudinal and shear sound velocities. The Debye temperature was also determined from the sound velocities and lattice parameter measurements. The thermal conductivity of SrPuO$$_{3}$$ was calculated from the measured density at room temperature, literature values of heat capacity, and thermal diffusivity measured by laser flash method in vacuum.

Journal Articles

Preparation and characterization of the simulated burnup americium; Containing uranium-plutonium mixed oxide fuel

Tanaka, Kosuke; Osaka, Masahiko; Miwa, Shuhei; Hirosawa, Takashi; Kurosaki, Ken*; Muta, Hiroaki*; Uno, Masayoshi*; Yamanaka, Shinsuke*

Journal of Nuclear Materials, 420(1-3), p.207 - 212, 2012/01

 Times Cited Count:4 Percentile:61.15(Materials Science, Multidisciplinary)

In order to investigate the effect on fuel thermophysical properties when adding americium and selected fission products to uranium-plutonium mixed fuel, simulated low decontamination MOX fuel with high burn-ups to 250 GWd/t, has been prepared and subjected to characterization tests, elastic moduli measurements, melting temperature measurement.

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

Melting temperature evaluation for burnt fast reactor (U, Pu)O$$_{2}$$ fuels

Hirosawa, Takashi; Sato, Isamu; Tanaka, Kosuke; Miwa, Shuhei

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 7 Pages, 2011/12

The melting temperature for burnt fast reactor fuels is evaluated in order to increase the precision of fuel thermal design by accurately deciding the safety margin. Fast reactor (U,Pu)O$$_{2}$$ fuels burnt in the experimental fast reactor JOYO were measured through a new method using a Re inner capsule. The effect of FPs on the melting temperature was discussed with these obtained data, previous information on FP behavior studies and computational studies. In respond to the discussion, necessary research for melting temperature evaluation of high burnup fuels are suggested.

Journal Articles

Burnup dependence of melting temperature of FBR mixed oxide fuels irradiated to high burnup

Hirosawa, Takashi; Sato, Isamu

Journal of Nuclear Materials, 418(1-3), p.207 - 214, 2011/11

 Times Cited Count:8 Percentile:39.39(Materials Science, Multidisciplinary)

The melting temperatures of FBR MOX fuels irradiated to 112.5 MWd kg$$^{-1}$$ were measured using a rhenium inner capsule to hold the specimens. These results were compared with those obtained using a tungsten capsule. The former melting temperatures were about 30 K higher than the latter, although the slopes of the respective lines plotting melting temperature versus burnup were not so different. This was attributed to the difference in the chemical reaction between the irradiated MOX fuels and the capsules.

Journal Articles

Thermal conductivity of BaPuO$$_{3}$$ at temperatures from 300 to 1500 K

Tanaka, Kosuke; Sato, Isamu; Hirosawa, Takashi; Kurosaki, Ken*; Muta, Hiroaki*; Yamanaka, Shinsuke*

Journal of Nuclear Materials, 414(2), p.316 - 319, 2011/07

 Times Cited Count:12 Percentile:26.02(Materials Science, Multidisciplinary)

Polycrystalline specimens of barium plutonate, BaPuO$$_{3}$$, have been prepared by mixing the appropriate amounts of PuO$$_{2}$$ and BaCO$$_{3}$$ followed by reacting and sintering at 1600 K under the flowing gas atmosphere of dry-air. The thermal conductivity of BaPuO$$_{3}$$ was almost the same magnitude as that of BaUO$$_{3}$$.

Journal Articles

Melting behavior of MgO-based inert matrix fuels containing (Pu,Am)O$$_{2-x}$$

Miwa, Shuhei; Sato, Isamu; Tanaka, Kosuke; Hirosawa, Takashi; Osaka, Masahiko

Journal of Nuclear Materials, 400(1), p.32 - 36, 2010/05

 Times Cited Count:6 Percentile:54.14(Materials Science, Multidisciplinary)

The melting behavior of MgO-based inert matrix fuels containing (Pu,Am)O$$_{2-x}$$ ((Pu,Am)O$$_{2-x}$$-MgO fuels) was experimentally investigated. Heat-treatment tests were carried out at 2173 K, 2373 K and 2573 K. The fuel melted at about 2573 K in the eutectic reaction of the Pu-Am-Mg-O system. The (Pu,Am)O$$_{2-x}$$ grains, MgO grains and pores grew with increasing temperature. In addition, Am-rich oxide phases were formed in the (Pu,Am)O$$_{2-x}$$ phase by heat-treatment at high temperatures. The melting behavior was compared with behaviors of PuO$$_{2-x}$$-MgO and AmO$$_{2-x}$$-MgO fuels.

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

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:19 Percentile:20.17(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

High temperature behavior of irradiated mixed nitride fuel during heating tests

Sato, Isamu; Tanaka, Kosuke; Hirosawa, Takashi; Miwa, Shuhei; Tanaka, Kenya

Journal of Alloys and Compounds, 444-445, p.580 - 583, 2007/10

 Times Cited Count:1 Percentile:85.75(Chemistry, Physical)

JAEA performs the study on the applicability of the mixed nitride fuel for the fast reactor by the irradiation test in the experimental fast reactor, "JOYO." In this work, heating tests for the mixed nitride fuel irradiated in JOYO have been performed and the fission gas release and the decomposition behaviours are evaluated for the safety study of Pb-Bi cooled fast reactor. As to the gas release aspects, the behaviour is independent on decomposition of the Nitride fuel. The density change was observed in the fuel after the heating test by slow heating rate. Over the temperature of 2400 degree C, the nitride fuel had decomposed to make some Pu-Rh alloy inclusions.

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

Thermal stability test and analyses for irradiated nitride fuel; Nifti-1 results and analyses

Sato, Isamu; Hirosawa, Takashi; Ishida, Takashi*; Miwa, Shuhei; Tanaka, Kenya

JNC-TN9400 2005-007, 62 Pages, 2005/03

JNC-TN9400-2005-007.pdf:6.37MB

Thermal stability test for irradiated mixed nitride fuel by the low heating rate was performed with the FP release behavior device. The decomposition behavior and the release behavior of fission gas and radioactive FPs of the nitride fuel were evaluated for the thermal stability and the source term, based on the behavior of nitride decomposition and FP release observed in the test. The findings for these behavior are as follows; (1)The practical decomposition of the irradiated nitride fuel occurred over 1900$$^{circ}$$C. In the region of lower temperature than 1000$$^{circ}$$C, nitrogen could be released along with the decomposition of FP nitrides such as Ba$$_{2}$$N$$_{3}$$. (2)Fission gas release started from 1400$$^{circ}$$C, and then was terminated. After the temperature reached to 1700$$^{circ}$$C, the burst-like release of fission gas occurred again. The former was slightly related to the decomposition, but the latter was hardly dependent on it as well as the release of Cs. (3)The swelling of fuel by the heating test occurred and then the porosity reached to approximately 40%. The porosity is equal to the maximum value obtained in the evaluation of post irradiation examination, which indicated that the porosity is saturated by the heating. (4)Most of Cs and Sb were released from the fuel after the heating. The chemical form of released Cs could be mainly elemental Cs and somewhat CsI. The form of released Sb could be elemental. (5)Ru could be hardly released, nevertheless most of Eu was released from the fuel. (6)On the basis of these results and discussions, we proposed the test plan of Nifti-2 (Thermal stability test for irradiated nitride fuel under transient thermal condition).

Journal Articles

Experimental investigation of fission products release from irradiated FBR MOX fuel

Ohno, Shuji; Sato, Isamu; Nakagiri, Toshio; Hirosawa, Takashi; Miyahara, Shinya; Namekawa, Takashi

JAERI-Review-2004-021, p.199 - 208, 2004/10

Out-of-pile experiments on the release of fission products (FPs) under transient heating conditions were carried out for mixed oxide fuels. The fuels used in the experiments, the plutonium content of which was 30wt%, were irradiated up to 65 GWd/t in the experimental fast reactor JOYO. The experiments consisted of two runs, FP-1 and FP-2. In FP-1, the fuel sample was first heated to 2,000$$^{circ}$$C and then up to 3,000$$^{circ}$$C. The holding time was 30 min at each temperature. In FP-2, the terminal temperatures were 1,500$$^{circ}$$C and 2,500$$^{circ}$$C, and the holding time was 30 min in the same manner. The release of Cs, a volatile FP, was detected as soon as the fuel sample was heated up. The release rate, after peaking in several minutes, decreased gradually via a diffusion process in the fuel matrix. The "gross" diffusion coefficient agreed well with the data reported in other experiments using LWR fuels. The release fractions were identical in both experiments, namely Cs 100%, Sb~100%, Ru 10%, Ce 0% and Eu 0%.

69 (Records 1-20 displayed on this page)