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Sato, Isamu*; *; ; Arima, Tatsumi*; ; Kajitani, Yukio
PNC TY9606 97-001, 117 Pages, 1997/07
no abstracts in English
Koyama, Shinichi; Osaka, Masahiko; Mitsugashira, Toshiaki; Konno, Koichi; Kajitani, Yukio
PNC TN9410 97-054, 44 Pages, 1997/04
For analysis of a small quantity of americium, it is necessary to separate from curium which has similar chemical property. As a chemical separation method for americium, and curium, the oxidation of americium with pentavalent bismuth and subsequent co-precipitation of trivalent curium with BIPO
were applied to analyze americium in irradiated MOX fuels which contained about 3Owt% plutonium and 0.9wt% 
Am before irradiation and were irradiated up to 26.2GWd/t in the experimental fast reactor Joyo. The purpose of this study is to measure isotopic ratio of americium and to evaluate the change of isotopic ratio with irradiation. Following results are obtained in this study, (1)The isotopic ratio of americium (Am, 
Am and 
Am) can be analyzed in the MOX fuels by isolating americium. The isotopic ratio of 
Am and Am increases up to 0.62at% and 0.82at% at maximum burnup, respectively. (2)The results of isotopic analysis indicates that the contents of Am decreases, whereas Am, Am increase linearly with increasing burnup.
Kajitani, Yukio; ; Abe, Kazuyuki; Osaka, Masahiko; ; Hirosawa, Takashi; Koyama, Shinichi
PNC TN9440 97-004, 186 Pages, 1997/02
The 25th anniversary meeting of the Alpha-Gamma Facility (AGF) at O-arai Engineering Center of PNC was held on February 7. The AGF started to examine irradiated materials on october 1 and fuel pins irradiated in the Dounreay Fast Reactor, DFR332/2 on December 1, 1971. The contents in this paper of the anniversary meeting are as follows. (1)25 years history and challenging plan for 2000 year. (2)Maintenance logbook of the facility, apparatus and manipulators for 25 years. (3)Recent results of melting temperature, thermal conductivity and lattice constants in irradiated MOX fuels. (4)Development on fission products release measuring apparatus and results of cold run tests. (5)Post irradiated examination results operated at the metallography cell in the Fuels Monitoring Facility (FMF). (6)Development on chemical analysis method for minor actinides (MA) in irradiated MOX fuels. (7)Refurbishment for MA containing MOX fuels, status and specifications for the fabrication and quality control apparatus.
Kakehi, Isao; ; ; ; ; Kajitani, Yukio;
PNC TN9410 97-015, 382 Pages, 1996/12
For the purpose of developing the future nuclear fuel recycle system, the design study of the advanced nuclear fuel recycle system is being conducted. This report describes intermediate accomplishments in the conceptual system study of the advanced nuclear fuel recycle system. Fundamental concepts of this system is the recycle system using molten salt which intend to break through the conventional concepts of purex and pellet fuel system. Contents of studies in this period are as follows, (1)feasibility study of the process by Cd-cathode for nitride fuel (2)application study for the molten salt of low melting point (AlCl
+organic salt)(3)research for decladding (advantage of decladding by heat treatment)(4)behavior of FPs in electrorefinning (behavior of iodine and volatile FP chlorides, FPs behavior in chlorination) (5)criticaliy analysis in electrorefiner (6)drawing of off-gas flow diagram (7)drawing of process machinery concept (cathode processor, vibration packing) (8)evaluation for the amounts of the high level radioactive wastes (9)quality of the recycle fuels (FPs contamination of recycle fuel) (10)conceptual study of in-cell handling system (11)meaning of the advanced nuclear fuelrecycle system. The conceptual system study will be completed in describing concepts of the system and discussing issues for the developments.
Osaka, Masahiko; Koyama, Shinichi; Otsuka, Yuko; Mitsugashira, Toshiaki; Konno, Koichi; Kajitani, Yukio
PNC TN9410 96-297, 79 Pages, 1996/11
As a part of evaluation of irradiation behavior and burnup characteristics of MA nuclides such as Np, Am and Cm in MA containing MOX fuel, we are studying the quantitative analysis techniques for MA nuclides in irradiated fuel. In this study, we studied the mutual separation method for Am and Cm to establish the analysis method for Am and Cm following Np analysis by alpha spectrometry. The measurements of Am and Cm are difficult to analyze quantitatively because the amounts of some nuclides are too small and the number of nuclides are large, whose energies of the alpha radioactive rays are almost same. Therefore we selected to analyze the trace amount of Am and Cm isotopes using mass spectrometry, and have studied the techniques for mutual separation of Am and Cm using oxidation method of Am by NaBiO for standard samples. We have also evaluated the availability of this method for irradiated fuel. Results are as follows. (1)Through the mutual separation tests, we have found the most suitable conditions for separation of both Am and Cm from each other element. The method obtaining Am which contains no Cm is used water for precipitation washing solution, containing Cm is achicved that the remaining ratio of Am (ratio of radioactivity of Am/
Cm against before separation) were reduced less than 1/10 for Cm. (2)Applying of this method to irradiated fuel, the coordinate remaining ratio and the chemical yield of Am and Cm were almost same as them in the separation tests. This method to apply various irradiated MOX Fuel is therefore possible. (3)The isotope ratio Am, Am and Am measured by mass spectrometry, which could not be analyzed by radioactive ray spectrometry causing less than detection limit, were 98.55% : 0.62% : 0.83%. We also measured zero of the mass number of 240 and 244 on the specimens and then certified no contamination of Cm to Am.
Am content and evaluation of burnup dependence in Am containing MOX fuel pinKoyama, Shinichi; Osaka, Masahiko; Otsuka, Yuko; Konno, Koichi; Kajitani, Yukio; Mitsugashira, Toshiaki
PNC TN9410 96-301, 61 Pages, 1996/10
We are studying quantitative analysis of Minor Actinides (Np, Am, Cm) in irradiated fuels as a part of the PNC research project for advanced nuclear fuel recycle system. In alpha-gamma section, irradiation behavior of MOX fuel and burning characteristic evaluation research of the MA nuclide which contain the minor actinide species are carrying out. We measured Am content of the MOX fuel pin which contained Am of about 0.9wt% before irradiation and were irradiated up to 26.2GWd/t in the JOYO reactor. The results are as follows. Burn-up dependence of the Am content in this samples was not observed. The Am content showed the fixed value of about 1% in the range from 0 to 28GWd/t. This reason is assumed that Am produced by 
-decay of Pu for cooling times between each cycles valances it in disappearance under irradiated in JOYO based on the calculated value by ORIGEN-II code.
Sato, Isamu; Yamamoto, Kazuya; Kajitani, Yukio
PNC TN9410 96-251, 82 Pages, 1996/06
The O/M ratio of fuel is related with most of fuel properties, especially it is important to evaluate radial O/M ratio distribution of fuel irradiated to high burnup in order to predict change of the properties. In this work, the radial O/M ratio distribution of irradiated FBR fuels was measured and evaluated. The fuels are irradiated in "JOYO", which were irradiated to the highest burnup(ca. 13at%). In this study, an indirect method, REDOX of Mo was used to obtain radial O/M ratio distribution, in which oxygen potential in fuel was determined by measuring oxidation and reduction states of Mo existing as a fission product (FP) in fuel. Oxygen potential distribution in fuel was determined from temperature profile and measured Mo distribution in fuel. O/M ratio distribution in fuel was evaluated from the oxygen potential, based on Catlow theory. The obtained O/M ratio distribution in fuel was compared with one calculated using Aitken model, which explains oxygen migration in fuel. Consequently, it was shown that the oxygen migration mechanism in high burnup fuel might differ from one suggested by Aitken and it might be necessary to take into consideration the effect of burnup on heat oftransport in oxygen thermal diffusion.
