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Report No.
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Study on the evaluation method to determine the radioactivity concentration in radioactive waste generated from the dismantling of research reactors

Murakami, Masashi; Hoshino, Yuzuru; Nakatani, Takayoshi; Sugaya, Toshikatsu; Fukumura, Nobuo*; Sanda, Toshio*; Sakai, Akihiro

Toward the establishment of a common approach to determine the radioactivity concentrations in dismantling wastes arising from research reactors, radionuclide concentrations in the reactor structure materials of aluminum, carbon steel, shield concrete, and graphite of TRIGA Mark II reactor at Rikkyo University, Japan, were evaluated with both radiochemical analysis and theoretical calculation. The measured nuclides by the radiochemical analysis were $$^{3}$$H, $$^{60}$$Co, and $$^{63}$$Ni in aluminum, $$^{3}$$H, $$^{60}$$Co, $$^{63}$$Ni, and $$^{152}$$Eu in carbon steel, $$^{3}$$H, $$^{60}$$Co, and $$^{152}$$Eu in shield concrete, and $$^{3}$$H, $$^{14}$$C, $$^{60}$$Co, $$^{63}$$Ni, and $$^{152}$$Eu in graphite. Neutron-flux distributions and neutron-induced activities were computed with DORT and ORIGEN-ARP codes, respectively. Using the results of material composition analysis, radioactivity concentrations were conservatively predicted with good accuracy except for graphite material.

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