The present work concerns the examination of management options for VLLW drums [1] with ion exchange resin waste from the Greek Research Reactor (GRR-1). According to the general clearance criterion [2, 3], after cementation, 75 % of the VLLW could be cleared and for the remaining 25 % the summation formula result is lower than 1.6. The proposed management option for the total amount of the VLLW resin is the spread of the cemented resin, before thickening, over the ground around the interim storage facilities, inside the controlled area, for pavement construction. Additionally, a quantity of the cemented waste will be used for preparation of blocks for quality and homogeneity examination of the waste form.
In this study, a semi-empirical calibration method for NORM samples measurement by using a LaBr3(Ce) scintillator was developed based on a combination of experimental gamma spectrometry measurements and MCNP-X simulations. The aim of this work is to provide us with full energy peak efficiency calibration curves in a wide photon energy range which is of particular importance when selected photon energies of 234Th, 214Pb, 214Bi, 228Ac, 208Tl and 226Ra are to be measured with accuracy.
The objective for decommissioning planning, is to obtain a radiological understanding of the involved installation. The characterization at this stage could be carried out by means of: (a) neutron activation calculations based on reactor design and neutron flux; (b) dose rate measurements; (c) in-situ gamma spectrometry; (d) sampling for determination of the scaling factors in activated and contaminated components.Neutron activation calculations contains several uncertainties. These uncertainties are based on the input data - such as material data (composition and impurities), neutron flux and energy, nuclear data libraries - and on the methodology of the process and the simulation codes.Taking into consideration all these modeling uncertainties, this work is focused on the development of a technique for validation of the calculations. A non-destructive gamma spectrometry technique by using MCNP6 simulations is under development for interpretation of the resulting gamma-ray spectra of the radionuclides in activated components. In particular, a spectrum will be produced, based on the activities of the main radionuclides in the activated component and the results of MCNP6 simulations. This spectrum will be compared with the experimental spectrum.Furthermore, the radiological characterization of activated components, which appeared with surface contamination, is essential for the decision making process during decommissioning. The cutting techniques to be followed in order to reduce the production of secondary waste and limit the doses to personnel and the selection of decontamination techniques should be based on accurate determination of the radionuclides inside the material and/ or in the surface contamination. The proposed method could also be helpful in this case. The activities inside and on the surface of the components could be determined by comparing the experimental spectrum with that produced by MCNP6 simulations, using the arisen activities from the scaling factors and the dose rate measurements.
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