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Koyama, Shinichi; Namekawa, Takashi
JNC TN9400 2002-060, 28 Pages, 2002/12
JNC-TN9400-2002-060.pdf:0.91MB
Solubility test was performed to evaluate the fundamental solubility characteristics of neptunium and insoluble residue after dissolution of MOX fuels irradiated in LWRs. (1)Neptunium was observed in insoluble residue of unirradiated MOX fuel as well as dissolved solution. (2)
Np content in dissolution liquid of unirradiated fuels ranged from 0.06 to 0.13% and that of irradiated fuels ranged from 0.03 to 0.1%. (3)Np content in MOX fuels decreased exponentially with increasing burn-up. (4)The solubility of neptunium in unirradiated fuels ranged from 84.1 to 98.1% and that in irradiated fuels ranged from 95.9 to 99.9%. (5)The solubility of neptunium in unirradiated and irradiated fuels slightly decreased both with initial plutonium content.
Np reaction amount by chemical analysis of Neptunium sample irradiated at experimental fast reactor "JOYO"Osaka, Masahiko; Koyama, Shinichi; Mitsugashira, Toshiaki; Morozumi, Katsufumi; Namekawa, Takashi
JNC TN9400 2001-016, 54 Pages, 2000/08
JNC-TN9400-2001-016.pdf:1.33MB
The chemical analysis technique was established to determine the nuclide generated in Neptunium (Np) sample with a high accuracy, to contribute to evaluation of transmutation characteristics of Np in the fast reactor. (1)Establishment of chemical analysis technique The chemical analysis technique containing determination technique of fission amount of Np, which was consist of Vanadium (V) of capsule material removal and Neodymium (Nd) recovery at high efficiency, was established with optimization of experimental conditions. Four Np saples irradiated in "JOYO" were analyzed using this technique. Results were as follows. (a)Np were determined with high accuracy (relative error was 2.2% at maximum). (b)Errors of fission amount monitoring nuclides 
Nd were half less than that of 
Cs. (c)Small amount of 
Pu was able to determined. (2)Evaluation of Np reaction amount The reaction amount of capture, fission and (n,2n) reactions were evaluated using analyzed values. Transmutation characteristics of Np were evaluated using reaction amount. Evaluated results were as follows. (a)The ratio of capture or fission amount to unirradiated Np amount were 6.1
25.5 at%, 0.73.6 at%, respectively. (b)The Np (n,2n) 
Np reaction amount was 7.0
10
times of Np amount at maximum. (c)The dependences of Np reaction amount to neutron energy spectrum were revealed from the fact such as linearity of fission to capture reaction amount ratio against fast neutron ratio in same fuel assembly.
Osaka, Masahiko; Koyama, Shinichi; Mitsugashira, Toshiaki; Morozumi, Katsufumi; Namekawa, Takashi
JNC TN9400 2000-058, 49 Pages, 2000/04
JNC-TN9400-2000-058.pdf:1.22MB
The analytical technique for Cm contained in a MOX FUEL was developed and analysis of Cm contained in irradiated fuel of experimental fast reactor "JOYO" was carried out, to contribute to evaluation of transmutation characteristics of MA nuclide in the fast reactor. The procedure of ion-exchange separation of Cm with nitric acid-methanol mixed media essential for the isotopic analysis in irradiated MOX fuel was adopted considering for being rapid and easy. The fundamental test to grasp separation characteristics of this procedure, such as Cm elution position and separation capacity between Cm and Am or Eu, was carried out. ln applying this procedure to the analysis of Cm contained in actual specimen, separation condition was evaluated and optimized, and the procedure consist of impurity removal and Am removal process was devised. This procedure resulted in high recovery rate of Cm and high removal rate of Am and impurity which becomes a problem in sample handling and mass-spectrometry such as Eu and Cs. The Cm separation test from irradiated MOX fuel was carried out using this technique, and Cm isotopic ratio analysis was enabled. The analytical technique for Cm contained in irradiated MOX fuel was established using the procedure of ion-exchange separation with nitric acid-methanol mixed media. The analysis of Cm contained in irradiated MOX fuel of experimental fast reactor "Joyo" was carried out. As a result, it was revealed from measured data that Cm content rate was 1.4 4.0lO
atom%, small amount of 
Cm was generated and Cm isotopic ratio was constant above burn-up 60GWd/t.
; ;
JNC TN9400 2000-029, 87 Pages, 1999/11
JNC-TN9400-2000-029.pdf:5.11MB
The second Power-To-Melt (PTM) test, PTM-2, was performed in the experimental fast reactor "JOYO". AIl of the twenty-four fuel pins of the irradiation vehicle, B5D-2, for the PTM-2 test, were provided for post-irradiation-experiment (PIE) to evaluate the PTM values. ln this study, the PIE technique for PTM test was established and the PTM results were evaluated. The findings are as follows: (1) The maximum fuel-melting ratio on the transverse section was 10.7%, and was within the limit of fuel-melting in this PTM test enough. Unexpected fuel-melting amount to a ratio of 11.8% was found at 24 mm below the peak power elevation in a test fuel pin, lt is possible that this arose from secondary fuel-melting. (2) Combination of metallographical observation with X-ray microanalysis of plutonium distribution was very effective for the identification of once-molten fuel zone. (3) The PTM evaluation suggested that dependence of the PTM on the fuel pellet density was stronger than that of previous foreign PTM tests, while the dependence on the pellet-cladding gap and the oxygen-to-metal ratio was indistinctly. The dependence on the cladding temperature and the fill gas composition was not shown as well.