In Japan, at the middle of the year 2022, 24 Nuclear Power plants have been planned to stop and to prepare towards decommissioning. At these plants and at Fukushima Dai-ichi plant, safe and suitable management of radioactive materials is one of essential issues before decommissioning. For such a purpose, accurate radiation measurement is required however, target materials structures, or equipment are generally large, and heavy. Furthermore, they are not easy to cut into pieces or separate into parts. Classifying them efficiently, to not overestimate the amount of low level waste and put as much materials through clearance while operators are under as little dose as possible is a challenge. Considering these situations, radiation measurements should be on-site, able to characterize wide targets, keeping own shape and structures of the target, time efficient and accurate as possible.
To identify radioactive nuclides, it is necessary to determine the photons (γ-ray) energies from the target material accurately. Germanium detectors have been being utilized widely relied on their very high ability of energy resolution. However, to determine accurate radioactivity, not only photon counts but also geometrical efficiency between the detector and the target is required. Normally, the latter is not easy to evaluate accurately.
Conventionally, two methods are applied to determine geometrical efficiencies. One is the standard source method, the other is the numerical calculation method.
The standard source calibration method has been applied to this issue for a long time based on high reliability of the method. On the other hand, it requires costly and time consuming complex procedures. Moreover, it is often impossible to obtain an equivalent source that is representative of the target (in terms of geometry and density).
The alternative method is numerical calculation. General-purpose Monte Carlo codes, MCNP for example, have been widely used for this purpose however, calculation preparations require expert knowledge. In addition, due to the nature of Monte Carlo simulation, considerable computing time is required, increasing easily with the complexity of the model.
To solve In situ measurement problems, the In Situ Object Counting System (ISOCSTM) technique was developed. This technique has been already widely utilized around the world, especially it was well recognized by the NRC (United States Nuclear Regulatory Commission). ISOCS easily enables users to realize an accurate efficiency calculation in a quite short time.
This paper presents the abilities and capabilities of ISOCS. Furthermore, verification examples are shown through measurements with Eu-152 standard sources. Calculation values are also compared with those of MCNP.
