Analysis of Minor Actinides in Mixed Oxide Fuel Irradiated in Fast Reactor, (I): Determination of Neptunium-237
Analysis of Minor Actinides in Mixed Oxide Fuel Irradiated in Fast Reactor, (I): Determination of Neptunium-237
复制标题
快堆辐照混合氧化物燃料中次锕系元素的分析（一）：Neptunium-237的测定
DOI：
10.1080/18811248.1998.9733882
复制
发表时间：
1998
影响因子：
1.2
通讯作者：
T. Mitsugashira
中科院分区：
文献类型：
--
作者：
S. Koyama;Y. Otsuka;M. Osaka;Katsufumi Morozumi;K. Konno;M. Kajitani;T. Mitsugashira
The MOX (mixed oxide) fuels, 29.97wt% (analytical value is 28.08wt%) plutonium in 8.3 or 12.1 wt% enriched uranium were irradiated up to 120 GWd/t in the MARK-II core of JOYO. The irradiated MOX fuels were dissolved in 8 M nitric acid solution. Anion exchange chromatographic procedures were developed for the isolation of neptunium contained in these MOX solutions. Neptunium-237 was separated from the other actinides by using the developed chromatographic technique and the content of 237 Np was determined by a spectrometry. It was found that the content of 237 Np slightly increased with increasing burnup. The content did not exceed 0.08wt% in the 30wt% MOX fuels irradiated up to about 120 GWd/t. Model calculation by ORIGEN-2 reproduces well the observed content of 237 Np.
