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Preparatory tasks for decommissioning of nuclear power plant start with radiological characterization. Residual radioactivity inventory evaluation is a main part of the characterization. Reliable information on the inventory is important for specification for decommissioning plan. Japan Atomic Power Company (JAPC) has already started these tasks for Tsuruga Nuclear Power Plant Unit 1 (TS-1). We can optimize decommissioning plan using the information. To obtain the reliable information, we improved an evaluation procedure. The procedure is divided into two main steps. First step is neutron flux distribution calculation and second one is radioactivity distribution calculation. Radioactivity distribution is calculated using neutron flux distribution. In this work, we improved the evaluation procedure to obtain the reliable information on the inventory. Because of the limitation of computer resource, two-dimension (2D) approximation model was applied to radioactivity distribution around Reactor Pressure Vessel (RPV). We can calculate reliable 2D neutron flux distribution by having better understanding of neutron transport phenomena. Neutron flux was measured at 30 locations in TS-1 Primary Containment Vessel (PCV) using activation foils. And in order to understand the neutron transport phenomenon inside the PCV, we also calculated neutron flux distribution with the three-dimensional (3D) discrete ordinates method calculation (Sn) code. By consideration about the result of the measurement and 3D calculation, we could understand the characteristics of the neutron flux distribution inside the PCV. To simulate the neutron flux distribution well with 2D Sn code, neutron flux behaviors inside the PCV had been investigated with referencing the measurement values and with observing calculated 3D neutron flux distribution. 2D calculation model had been modified repeatedly until reliable calculation result was provided. After several model modifications, the reliable 2D calculation was accomplished and important neutron transport phenomena that are necessary to simulate the neutron flux distribution well was understood. Network-parallel-computing technique was applied to radioactivity distribution calculation. Using this technique, we could calculate radioactivity at all space mesh points that were used with 2D Sn code and we obtained the radioactivity distribution. By using this distribution, we can estimate a quantity of radioactivity around RPV more accurately and optimize dismantling designs.

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Biological shielding wall (BSW) concrete of commercial LWR is activated by neutrons emitted from the reactor core during its operation. Most of wasted BSW, having a large volume, is classified as very low radioactive level. It has been suggested that deterministic calculation methods give inaccurate estimation of activation in nuclear facilities. By using Monte-Carlo calculation methods, induced radioactivity in concrete could be estimated more accurately.The discrepancy between the calculated and measured activation in concrete is said that it attributed to water content of concrete and reinforcing bars. Concrete samples with different water to cement weight ratio without reinforcing bars and one containing reinforcing bars were exposed to neutrons in the fast column of the research reactor "YAYOI" and induced radioactivity inside were measured with gold foils by the foil activation method. Radiation transport was simulated by DORT, MCNP-5 and compared with experimental results.The build-up of 197 Au (n,) reaction rate depth profile in the sample was not found. Regardless of the water to cement ratio, the distribution of activation in the concrete samples agreed within the margin of error. The reinforced concrete was weakened reaction rate about 40 percent. Reinforcing bars did not affect epi-thermal neutron flux significantly.