Preliminary Waste Characterization Measurements in FiR 1 TRIGA Research Reactor Decommissioning Project

Abstract

The objective of the study has been to verify the calculated residual activity in the decommissioning waste of TRIGA Mark II type research reactor FiR 1 in Finland. The knowledge of radioactive inventory of irradiated materials is important in the planning of the decommissioning activities and is essential for predicting the radiological impact to personnel and environment. Measurements are performed for low active material samples from outer parts of the reactor. Methods include gamma spectrometric measurements, composition measurements with mass spectrometry, oxidation measurements of especially C-14 in graphite and full combustion measurements of lithium enriched shielding materials. Results are compared to estimates calculated with a combined Monte Carlo model of the reactor and a point-depletion code modelling the irradiation history. Decommissioning waste consists mainly of ordinary concrete, aluminium, steel and graphite parts. Only preliminary measurements of low active samples are reported so far, but the same methods will be used later for characterizing and classifying dismantling waste. Some discussion of characterization requirements and future sampling is also included.

Intercomparison exercise on difficult to measure radionuclides in activated concrete—statistical analysis and comparison with activation calculations

“…The reactor was shut down permanently in 2015 and the dismantling is expected to begin in 2022. Characterisation of the FiR1 activated components has been carried out using both modelling and experimental studies [7][8][9]. Experimental characterisation of the biological shield began in 2014 with coring of inactive concrete cores to which, for example, testing of mechanical properties and chemical composition analyses were carried out.…”

“…Although the neutron flux was estimated conservatively, the chemical composition of the concrete used in the calculation model was determined from three inactive cores that had been drilled earlier from the inactive outer parts of the reactor structure as described earlier [7]. The samples were Fig.…”

Development of 3H, 14C, 41Ca, 55Fe, 63Ni radiochemical analysis methods in activated concrete samples

“…The on-going characterisation efforts of FiR1 TRIGA Mark II type research reactor have been published previously [15][16][17]. The characterisation of the biological shield concrete began with drilling of core samples from the nonactivated part of the biological shield concrete in order to determine the chemical composition of major activating elements for activation calculations.…”