ESTIMATION OF THE GENERATION OF 13C AND 14C IN THE REACTOR GRAPHITE USING MCNP6 MODELLING, ISOTOPE RATIO MASS SPECTROMETRY AND 14C MEASUREMENT TECHNIQUE

Characterization of irradiated graphite in terms of 14C activity is crucial for the optimization of treatment technology: geological disposal, landfill storage, recycling, etc. The main contributor to 14C generation in the RBMK reactor graphite is 14N(n, p)14C reaction. The generation of carbon isotopes 13C and 14C in the virgin RBMK graphite samples irradiated at the LVR-15 research reactor (Research Centre Řež, Ltd.) were investigated in order to obtain the impurity concentration level of 14N. Afterwards the modeling of graphite activation in the RBMK-1500 reactor was performed by computer code MCNP6 using obtained 14N impurity concentrations and new nuclear data libraries. The irradiation parameters – neutron fluence have been checked by method based on coupling of stable isotope ratio mass spectrometry and computer modelling. The activity of 14C in the different constructions of irradiated graphite of the RBMK-1500 reactor has been measured by the β spectrometry technique (LSC) and has been compared with the simulated one. Obtained results have indicated the importance of 14C production from 14N in the RBMK-1500 reactor and in the LVR-15 neutron spectrum. Measured 14C specific activity values in the samples varied from 130-700 kBq/g in the RBMK-1500 irradiated samples and from 3-12.5 Bq/g in the LVR-15 irradiated graphite samples. This corresponds to 15±4 - 80±10 ppm impurity of 14N in various graphite samples of RBMK reactor.

EVALUATION OF THE DOSE LEVELS IN THE IRRADIATION CHANNELS OF PU-BE NEUTRON SOURCES STORAGE DEVICE BY USING NEUTRON ACTIVATION ANALYSIS AND MCNP6 MODELLING

PuBeneutrons sources of Educational Neutron Laboratory in the Center for Physical Sciences and Technology are used in several activities including neutron activation for educational purposes and laboratory irradiation experiments. In this work we address the dose rate estimation in the irradiation channels of reconstructed PuBe neutron sources storage device in the particular positions dedicated for neutron irradiation experiments using neutron activation analysis (NAA) and MCNP6 modelling.The results on neutron activation of V2O5, MnO2, Al and Na2CO3were analyzed and compared with MCNP6 model prediction results. The conclusions on the actual neutron flux energy distribution and model corrections are drawn. The absolute neutron intensity at dedicated irradiation points as well as neutron and gamma dose rates are obtained.

Rapid analysis method for the determination of 14C specific activity in irradiated graphite

14C is one of the limiting radionuclides used in the categorization of radioactive graphite waste; this categorization is crucial in selecting the appropriate graphite treatment/disposal method. We propose a rapid analysis method for 14C specific activity determination in small graphite samples in the 1-100 μg range. The method applies an oxidation procedure to the sample, which extracts 14C from the different carbonaceous matrices in a controlled manner. Because this method enables fast online measurement and 14C specific activity evaluation, it can be especially useful for characterizing 14C in irradiated graphite when dismantling graphite moderator and reflector parts, or when sorting radioactive graphite waste from decommissioned nuclear power plants. The proposed rapid method is based on graphite combustion and the subsequent measurement of both CO2 and 14C, using a commercial elemental analyser and the semiconductor detector, respectively. The method was verified using the liquid scintillation counting (LSC) technique. The uncertainty of this rapid method is within the acceptable range for radioactive waste characterization purposes. The 14C specific activity determination procedure proposed in this study takes approximately ten minutes, comparing favorably to the more complicated and time consuming LSC method. This method can be potentially used to radiologically characterize radioactive waste or used in biomedical applications when dealing with the specific activity determination of 14C in the sample.

Uncertainty of determination of 158Tb in the RBMK nuclear reactor waste

The activity of ¹⁵⁸Tb was measured in waste samples from the Ignalina NPP Unit I RBMK-1500 reactor using gamma-ray spectrometry. The origin of ¹⁵⁸Tb and the other observed gamma-ray emitters has been studied by using SCALE 6.1 modeling and comparing radionuclide ratios in the RBMK-1500 radioactive waste. The results of the calculation of the massic activity of gamma-ray emitters were used for interpretation of the total gamma-ray spectrum and the determination of ¹⁵⁸Tb massic activity uncertainty in the waste of RBMK-1500.