Shin-ichi Koyama

Fabrication Technology for MOX Fuel Containing AmO2 by an In-cell Remote Process

An in-cell remote fabrication technique was developed for MOX fuel pellets containing 3 and 5% americium (Am-MOX fuel pellet). The fuel pellet was fabricated by means of conventional powder metallurgy. A series of fuel pellet fabrication apparatuses were systematically installed in the alpha-gamma cell (hot cell) to protect workers from a strong γ-ray exposure from 241 Am, and were remotely controlled from a panel in the operation room outside the hot cells as much as possible. From a preliminary UO2 pellet fabrication run, ball milling of powder for 4h, pressing at 4t/cm2 and sintering at 1,700°C for 2h were determined as a good fabrication, but the ball milling time was too short for the UO2 and Am-PuO2 powders of different morphologies to be uniformly mixed. Then, the 5% Am-MOX fuel pellet of density more than 93% T.D. which is proper to the irradiation in FBR was successfully fabricated by extending the ball milling time for more than 10 h. It was, furthermore, found that the complete cleanup of the powder feeder was necessary in the transition of fabrication runs to prevent the formation of uranium and plutonium spots in the pellets.

Development of a Multi-functional Reprocessing Process based on Ion-exchange Method by using Tertiary Pyridine-type Resin

A series of separation experiment was performed in order to study a multi-functional spent fuel reprocessing process based on ion-exchange technique. The tertiary pyridine-type anion-exchange resin was used in this experiment and the mixed oxide fuel highly irradiated in the experimental fast reactor “JOYO” was used as a reference spent fuel. As the result, 106Ru+125Sb, 137Cs+155Eu+144Ce, plutonium, americium and curium could be separated from the irradiated fuel by only three steps of ion-exchange. The decontamination factor of 137Cs and trivalent lanthanides (155Eu, 144Ce) in the final americium product exceeded 3.9x104 and 1.0x105, respectively. The decontamination factor for the mutual separation of 243Cm and 241Am was larger than 2.2x103 for the americium product and, moreover, the content of 137Cs, trivalent lanthanides and 243Cm included in 241Am product did not exceed 2 ppm. These results prove that the proposed simplified separation process has a reality as a candidate for future reprocessing process based on the partitioning and transmutation concept.

Study on adsorption behavior of cesium using ammonium tungstophosphate (AWP)-calcium alginate microcapsules

A functional microcapsule was prepared by encapsulating the fine crystalline ammonium tungstophosphate (AWP) in calcium alginate polymer (CaALG). The characterization of AWP-CaALG microcapsule was examined by SEM and EPMA. The adsorption behavior of Cs(I), Rb(I), Sr(II), Pd(II), Ru(III), Rh(III), La(III), Ce(III), Dy(III) and Zr(IV) was investigated by the batch method. The batch experiments were carried out by varying the shaking times, HNO3 concentration, and initial concentration of metal ions. Relatively large Kd values above 105 cm3/g for Cs(I) were obtained in the range of 0.1–5 M HNO3, resulting in a separation factor of Cs/Rb exceeding 102. In contrast, the Kd values of Sr(II), Pd(II), Ru(III), La(III), Dy(III), Ce(III) and Zr(IV) were considerably lower than 50 cm3/g. The Kd value of Cs(I) decreased in the order of the coexisting ions, H+ > Na+ ≫ NH4+, and a linear relationship with a slop of about −1 was obtained between log Kd and log [NH4+] ([NH4+] > 0.01 M). The adsorption of Cs(I) was found to be controlled by chemisorption mechanism, and followed a Langmuir-type adsorption equation. A high uptake percentage of 99.4% for Cs(I) was obtained by using the dissolved solutions of spent fuel from FBR-JOYO (JAEA).

Chemical Analysis of Americium Samples Irradiated under Fast Neutron Spectra

In order to evaluate the transmutation behavior of americium under fast neutron spectra, two irradiated 241 Am samples (No. 69, No. 70) were analyzed by radiochemical methods. About 100μg of 241 Am (Am oxide, 99.9%) samples encapsulated in a small vanadium capsule were irradiated in the experimental fast reactor JOYO for 275 effective full power day (EFPD) by the fast neutron flux of 1.08 × 1015 (No. 69, in the reflector region) and 3.25 × 1015 (No. 70, at the center core) n.cm−2 s−1 (E ≥ 0.1 MeV). After dissolution of these samples, americium, curium, and plutonium were chemically separated and the isotopic composition was determined by alpha and gamma-ray spectrometries and mass spectroscopy. 242mAm, 243Am, 238–242Pu and 242–248Cm were clearly observed. The isotopic composition of 242mAm exceeded 1.01 at.% for sample No. 69 and 1.48 at.% for No. 70. It was suggested that this difference came from the different rates of neutron capture reaction. The main detected curium isotope 243Cm was formed through 242m,gAm by the neutron capture of 241Am. The transmutation ratios of americium could be evaluated from these experimental results and were around 7.8% and 11.1% for the sample No. 69 and No. 70, respectively.

Protected Plutonium Production by Transmutation of Minor Actinides for Peace and Sustainable Prosperity—Irradiation Tests of Np and Np-U Samples in the Experimental Fast Reactor JOYO (JAEA) and the Advanced Test Reactor at INL—

A project on Protected Plutonium Production (P3) was proposed by Tokyo Institute of Technology as part of a nonproliferation research program for plutonium (Pu) utilization in nuclear reactors. The project is aimed at the production of inherently protected Pu by the addition of 237Np to uranium (U) fuel. In order to validate this P3 concept, two irradiation tests were performed. In the first, a determination of Pu isotopes in 237Np samples irradiated in the experimental fast reactor JOYO was done to evaluate 238Pu production from 237Np under fast neutron spectra. The fast reactor can undertake P3, which can be better performed in the reflector region. In the test, the percentage of the total amount of 238Pu atoms transmuted from 237Np atoms by irradiation was around 90%. In the second test, 2, 5, and 10% Np-containing U samples were irradiated in the Advanced Test Reactor at the Idaho National Laboratory (INL) to evaluate 238Pu production in the thermal neutron region. The fuel specimens were removed from the core at 100, 200, and 300 effective full power days (EFPDs), and then a postirradiation examination was completed at an analytical laboratory in the Materials & Fuels Complex (MFC) at the INL. For the samples after irradiation for 300 EFPDs, Np depletions were about 60% for 2% neptunium (Np)-U samples and about 50% for 5 and 10% Np-U samples. The 238Pu-to-Pu ratios were about 20, 30, and 45% for 2, 5, and 10% Np-U samples, respectively.

Analysis of Curium in Mixed Oxide Fuel Irradiated in the Experimental Fast Reactor JOYO for the Evaluation of Its Transmutation Behavior

Curium isotopes generated in the MOX fuel irradiated in the experimental fast reactor JOYO were analyzed by applying a sophisticated radiochemical technique. Curium was isolated from the irradiated MOX fuel by anion-ex- change chromatography using a mixed medium of nitric acid and methanol. The isotopic ratio of curium and its content were determined by thermal ionization mass spectroscopy and alpha-spectrometry, respectively. The curium content was less than 0.004 at% even at high burnup of 120GWd/t, which is much smaller than that of PWR-MOX at 60 GWd/t. On the basis of present analytical results, the transmutation behavior of curium isotopes in a fast reactor was discussed from various viewpoints. Transmutation rates of curium isotopes were estimated; the rate for 246Cm, which is known to be a key nuclide in the transmutation of curium, was larger than the previously reported value. It was concluded from these evaluations that the fast reactor was suitable for the incineration of curium.

Analysis of curium isotopes in mixed oxide fuel irradiated in fast reactor

Curium is one of the key elements in recycling and transmutation of minor actinides (MA) because of its high ra- diotoxicity and difficulty of transmutation. In order to make isotopic analysis of curium in heavily irradiated fuel, the isolation technique of curium was developed by adopting anion exchange chromatography in nitric acid-methanol mixed media. The technique was successfully applied to the analysis of curium in mixed oxide (MOX) fuel irradiated in the experimental fast reactor “JOYO”. The transmutation behaviors of curium in fast reactor are discussed on the basis of observed isotopic ratio of curium.

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

As a first step for obtaining experimental data on the effects of high-temperature chemical interaction on fission product 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 five minutes. The fractional release rate of cesium (specifically 137Cs) 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.

Investigation of the cause of peculiar irradiation behavior of 9Cr-ODS steel in BOR-60 irradiation tests

Four experimental fuel assemblies (EFAs) containing 9Cr-ODS steel cladding fuel pins were previously irradiated in the BOR-60 to demonstrate the in-reactor performance of 9Cr-ODS steel for use as fuel cladding tubes. One of the EFAs achieved the best data, a peak burn-up of 11.9at% and a neutron dose of 51 dpa, without any microstructure instability or any fuel pin rupture. On the other hand, in another EFA (peak burn-up, 10.5at%; peak neutron dose, 44 dpa), peculiar irradiation behaviors, such as microstructure instability and fuel pin rupture, occurred. Investigations of the cause of these peculiar irradiation behaviors were carried out. The detection sensitivity in an ultrasonic inspection test was shown to be low for the metallic Cr and metallic Fe inclusions. The peculiar microstructure change reappeared with high-temperature thermal-aging of the 9Cr-ODS steel containing metallic Cr inclusions. The strength and ductility of the defective part containing metallic Cr inclusions were appreciably lower than those of a standard part without the inclusions. The combined effects of matrix Cr heterogeneity (presence of metallic Cr inclusions) and high-temperature irradiation were concluded to be the main cause of the peculiar microstructure change in 9Cr-ODS steel cladding tubes in the BOR-60 irradiation tests. They contributed to the fuel pin rupture.

Numerical analysis of irradiated Am samples in experimental fast reactor Joyo

Americium is a key element to design the FBR based nuclear fuel cycle, because of its long-term high radiological toxicity as well as a resource of even-mass-number plutonium by its transmutation in reactors, which contributes the enhancement of proliferation resistance. The present paper deals with the numerical analysis of the Am sample irradiation in Joyo to examine the transmutation performance of pure isotope in fast neutron environment during the irradiation, and deals with the comparison with the experimental result to evaluate the accuracy of current available numerical tool. In 241Am pure isotope sample, the burn-up calculation of Am transmutation ratio and principal nuclides accumulation are agreed with the measured data within 1-σ uncertainty caused of cross-section covariance. Isomeric ratio of 242Am in total 241Am capture reaction were calculated as 0.852±0.016 in the core and 0.85±0.025 in the axial and radial reactors. The current data and recently reported data by Koyama et. al 2008 support the latest version of nuclear data sets in ENDFB-VII and JENDL/AC-2008. From the view point of proliferation resistance, it was confirmed 241Amp reduces un-attractive Pu to abuse from the beginning to the end of irradiation, and it would have important role to denature Pu in future FBR based nuclear fuel cycle.