JOPSS - Search Results

Fission products (FPs) release tests from irradiated MOX fuel were performed in Alpha Gamma Facility (AGF) from viewpoint of "source term" evaluation. "Source term"means the species and the quantity of FPs and so on discharged to environment during nuclear reactor accidents. The tests are carried out two times, which names are "FP-1" and "FP-2", respectively. ln "FP-1" test, a fuel sample was heated for 30 minutes at 2000 $^{circ}$ C, and then for 30 minutes at 3000 $^{circ}$ C. In "FP-2", it did for 30 minutes at 1500 $^{circ}$ C, and then for 30 minutes at 2500 $^{circ}$ C. The heating rate is 15 K/sec for any heating processes. The samples (about 10g of weight) for the tests were irradiated to approximately 65 GWd/t in the experimental fast reactor, "Joyo", which cladding materials were removed. The Pu content and initial O/M ratio for each sample is 29wt% and 1.99, respectively. During and after the test, FP release behavior was observed with -ray spectrometry, gas mass analysis and gas chromatograph. (1)The -ray spectrometry mainly revealed the release and adhesion on sampling parts of Cs. The results of these experiments are as follows; (a)Cs suddenly was released from the sample as soon as the temperature was elevated. The peaks of Cs-134 and Cs-137 release rate are not identical. This probably results from difference of distribution between Cs-134 and Cs-137. (b)The adhesion profile of Cs on the sampling tubes is a function of the sampling tube temperature, and the adhesion quantity is large at regions less than 600 $^{circ}$ C. (c)About 70% of Cs releasing from the sample stuck on sintered metal filters for both FP-1 and FP-2 test. Cs quantities sticking on all of the sampling parts are identical for FP-1 and FP-2. (d)Rh(Ru)-106 and Eu-154 were scarcely released from the samples. (2)The gas analyses mainly revealed the release behavior of Kr and Xe. The results are summarized as follows; (a)Kr suddenly was released from the sample as ...

JAEA Reports

Evaluation of $^{237}$ 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 $^{237}$ Np in the fast reactor. (1)Establishment of chemical analysis technique The chemical analysis technique containing determination technique of fission amount of $^{237}$ 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) $^{237}$ Np were determined with high accuracy (relative error was 2.2% at maximum). (b)Errors of fission amount monitoring nuclides $^{148}$ Nd were half less than that of $^{137}$ Cs. (c)Small amount of $^{236}$ Pu was able to determined. (2)Evaluation of $^{237}$ Np reaction amount The reaction amount of capture, fission and (n,2n) reactions were evaluated using analyzed values. Transmutation characteristics of $^{237}$ Np were evaluated using reaction amount. Evaluated results were as follows. (a)The ratio of capture or fission amount to unirradiated $^{237}$ Np amount were 6.125.5 at%, 0.73.6 at%, respectively. (b)The $^{237}$ Np (n,2n) $^{236m}$ Np reaction amount was 7.010 $^{-6}$ times of $^{237}$ Np amount at maximum. (c)The dependences of $^{237}$ 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.

JAEA Reports

Analysis of Cm contained in irradiated fuel of experimental fast reactor "JOYO"; Development of the analytical technique and measurement of Cm

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 $^{-3}$ atom%, small amount of $^{247}$ Cm was generated and Cm isotopic ratio was constant above burn-up 60GWd/t.