Blind Benchmark Exercise for Spent Nuclear Fuel Decay Heat

The decay heat rate of five spent nuclear fuel assemblies of the pressurized water reactor type were measured by calorimetry at the interim storage for spent nuclear fuel in Sweden. Calculations of the decay heat rate of the five assemblies were performed by 20 organizations using different codes and nuclear data libraries resulting in 31 results for each assembly, spanning most of the current state-of-the-art practice. The calculations were based on a selected subset of information, such as reactor operating history and fuel assembly properties. The relative difference between the measured and average calculated decay heat rate ranged from 0.6% to 3.3% for the five assemblies. The standard deviation of these relative differences ranged from 1.9% to 2.4%

Disappearing and already gone villages of Jonava district

The aim of this research is to analyze the disappearing and already gone villages of Jonava district and to determine the causes of their disappearance. In the course of the research, literature sources, statistical data, legal acts and other documents were analyzed. For the fulfilment of the research Rural Plans M 1:10 000 of the Jonava district for the end of the 20th century were used. Analytical, statistical grouping and comparison methods were used as well. This topic is relevant in many countries. Since the middle of the 20th century in both Lithuania and other countries, the processes of urbanization have intensified. For economic and social reasons, many rural residents went to cities. These processes also touched Jonava district situated near Kaunas. There are 8 rural subdistricts in this district. At present, there are 259 villages. From 1979 to 2017 the number of villages has decreased by 25%. In 1979, there were 344 place names of villages in Jonava district, in 1989 – 274, in 2001 – 269, and in 2017 – 259Vytauto Didžiojo universitetasŽemės ūkio akademij

Modeling of neutron and photon transport in iron and concrete radiation shields by using Monte Carlo method

Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 (AVF cyclotron of Research Center of Nuclear Physics, Osaka University, Japan) – γ-ray beams (1-10 MeV), HIMAC (heavy-ion synchrotron of the National Institute of Radiological Sciences in Chiba, Japan) and ISIS-800 (ISIS intensive spallation neutron source facility of the Rutherford Appleton laboratory, UK) – high energy neutron (20-800 MeV) transport in iron and concrete. The calculation results were then compared with experimental data.compared with experimental data

ESTIMATION OF THE GENERATION OF

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