14C In Graphite Research Articles

14C release from steels under aerobic conditions

ABSTRACTRadiocarbon (14C) is a key radionuclide in the assessment of the safety of underground geological disposal facilities for radioactive wastes, and the understanding of the 14C behavior in stainless steel may lead to a re-evaluation of the near-surface repository for the disposal of wastes containing this radionuclide in high concentrations. To achieve this objective, leaching experiments were planned for two different scenarios. The first is where the leaching solution, NaOH solution of pH ca. 12 in aerobic conditions, simulates the expected conditions in a cement-based near-surface repository over long time periods. The other one uses an acid solution of 1M H3PO4, which has been proved as a high efficiency chemical removal agent of 14C in graphite. The development of both analytical methods and protocols to measure the release of 14C from the activated steel samples and the speciation in the aqueous and gaseous phase has been undertaken as part of the EC CAST (CArbon-14 Source Term) project. Analytical methods, suitable for identifying and quantifying low molecular weight organic molecules, comprise ion chromatography (IC) and gas chromatography coupled to mass spectrometry (GC-MS); they are described for aqueous and gaseous samples, respectively. In this paper the preparation of leaching experiments to measure the release of 14C and the results obtained are described.

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

The objective of the study has been to verify the calculated residual activity in the decommissioning waste of the TRIGA Mark II–type research reactor FiR 1 in Finland. Knowledge of the radioactive inventory of irradiated materials is important in the planning of decommissioning activities and is essential for predicting the radiological impact to personnel and the 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 14C 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 modeling the irradiation history. Decommissioning waste consists mainly of ordinary concrete, aluminum, 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.