In situ gamma spectrometry using portable HPGe detector. Radiological characterization and environmental surveillance around an operating nuclear power plant. Possibilities and limits

In situ technique for measuring radionuclides in the soil using a portable Ge detector is a highly versatile tool for both the radiological characterization and for the monitoring of operating nuclear power plants. The main disadvantage of this technique is related to the lack of knowledge of the geometry of the source whose activity concentration is to be determined. However, its greatest advantage is the high spatial representability of the samples and the lower time and resource consumption than gamma spectrometry lab measurements. In this study, the possibilities and limits offered by in situ gamma spectrometry with a high resolution gamma portable detector in two common uses are shown: First, the radiological background characterization and its relationship with the geology of an area of 2700 km2 are assessed; Secondly, its potential for monitoring man-made activity concentration in soils located around an operating nuclear power plant in Spain for surveillance purposes is evaluated. Finally, high accuracy radiation maps have been prepared from the measurements carried out. These radiation maps are essential tools to know the radioactive background of an area, especially useful to assess artificial radioactive deposits produced after a nuclear accident or incident.

Radiological characterization of a German pressurized water reactor based on a highly resolved method for activity analysis and dose rate calculation

The amendment to the atomic act in 2011 results to phase out nuclear energy in Germany until the end of 2022. Subsequently, the licensee of the nuclear power plant is responsible for decommissioning and dismantling. During operation, activation of structures near the core of the reactor occur which govern the amount of radioactive waste, the dose rate distribution and dismantling strategies. Thus, a detailed radiological characterization of in-core and out-core structures is required to optimize decommissioning processes regarding the quantification and minimization of radioactive waste, radiation protection and reducing radiation exposure. These objectives are achieved using an innovative and efficient method developed and applied at the Chair of Repository Safety (Lehrstuhl für Endlagersicherheit, ELS) RWTH Aachen University. Within the framework of the joint project „Development of a methodology for activity analysis and dose rate estimation“, funded by the Federal ministry of Education and Research, approaches the objective to develop a standardized and highly resolved method to calculate time-dependent activity of components and structures near the reactor core based on operating history of the nuclear power plant and neutron fluence distribution. The approach requires the development of a detailed model for Monte-Carlo simulations which provides the basis to neutron fluence, neutron spectra and radiation transport simulations. To calculate the nuclide specific 3-Dimensional (3D) activity distribution of the entire facility, a facility-dependent activation cross section library is produced which focuses on recent nuclear databases (ENDF/B-VIII.0). A highly resolved and space-dependent 3D activity distribution of the entire facility is obtained using a modular program package, developed at ELS, including the activation code ORIGEN2. The results are produced in the form of detailed 3D activity maps. The source terms are generated on the basis of the space-dependent 3D activity distribution using an additional module of the program package. The combination of recent nuclear databases focusing on ENDF/B-VII.1 and complemented by JEFF-3.3 ensures a comprehensive characterisation of source terms. Subsequently, source terms are prepared for 3D radiation transport simulation using the Monte-Carlo method and the computer code MCNP. The simulations are conducted separately for each individual component obtaining the partial contribution of all in-core and out-core structures as well as the dose rate distribution of the entire facility. Similar to the activity calculation, the simulation results are used to generate 3D gamma flux and dose rate maps using the graphic module of the whole program system. On the basis of the radiological characterisation and in view of a high-level radiation protection these maps allow the optimum planning and realisation of the decommissioning and dismantling process of the nuclear power plant.

Microstructural, Mechanical and Radiological Characterization of Mortars Made with Granite Sand

The study reported the effect of granite sand on strength and microstructural developments in mortars prepare from OPC with a high coal fly ash (FA) content or from hybrid alkaline cements. The radiological behaviour of the resulting mortars was compared to materials prepared with siliceous sand (with particles sizes of <2 mm) and the relationship between such radiological findings and mortar microstructure and strength was explored. A new method for determining natural radionuclides and their activity concentration Index (ACI) on cement mortars (specifically to solid 5-cm cubic specimens) was applied and validated. The microstructural changes associated in mortars have no effect on mortar radiological content measurements. The mortars with granite sand exhibited very high ACI > 0.96, which would ultimately limit their use. A conclusion of interest is that where information is at hand on the starting materials (OPC, FA, sand, admixtures), their proportions in the mortar and the mixing liquid content (water or alkaline activators) their radiological content is accurately predicted. The inference is that a mortar’s radiological content and ACI can be known prior to mixing, providing a criterion for determining its viability. That in turn lowers environmental risks and the health hazards for people in contact with such materials.

Green Tuff: Gamma Index in a Typical Building Material of Ciglio Area (Ischia Island)

Radiological characterization of tuff of Ciglio area in Ischia Island was performed to assess the potential radiological hazard associated with its use. For this purpose, high resolution gamma-ray spectroscopy was used. In five green tuff samples was measured the activity concentration of natural gamma-ray emitting radionuclides 226Ra, 232Th and 40K so, the mean gamma index was calculated. The mean value of the gamma index resulted lower than the reference level and it meets the dose criterion for the safety use of green tuff as building material.