Shielding Calculations on Waste Packages – The Limits and Possibilities of different Calculation Methods by the example of homogeneous and inhomogeneous Waste Packages

For nuclear waste packages the expected dose rates and nuclide inventory are beforehand calculated. Depending on the package of the nuclear waste deterministic programs like MicroShield® provide a range of results for each type of packaging. Stochastic programs like “Monte-Carlo N-Particle Transport Code System” (MCNP®) on the other hand provide reliable results for complex geometries. However this type of program requires a fully trained operator and calculations are time consuming. The problem here is to choose an appropriate program for a specific geometry. Therefore we compared the results of deterministic programs like MicroShield® and stochastic programs like MCNP®. These comparisons enable us to make a statement about the applicability of the various programs for chosen types of containers. As a conclusion we found that for thin-walled geometries deterministic programs like MicroShield® are well suited to calculate the dose rate. For cylindrical containers with inner shielding however, deterministic programs hit their limits. Furthermore we investigate the effect of an inhomogeneous material and activity distribution on the results. The calculations are still ongoing. Results will be presented in the final abstract

VARIAN CLINAC 6 MeV Photon Spectra Unfolding using a Monte Carlo Meshed Model

[EN] Energy spectrum is the best descriptive function to determine photon beam quality of a Medical Linear Accelerator (LinAc). The use of realistic photon spectra in Monte Carlo simulations has a great importance to obtain precise dose calculations in Radiotherapy Treatment Planning (RTP). Reconstruction of photon spectra emitted by medical accelerators from measured depth dose distributions in a water cube is an important tool for commissioning a Monte Carlo treatment planning system. Regarding this, the reconstruction problem is an inverse radiation transport function which is ill conditioned and its solution may become unstable due to small perturbations in the input data. This paper presents a more stable spectral reconstruction method which can be used to provide an independent confirmation of source models for a given machine without any prior knowledge of the spectral distribution. Monte Carlo models used in this work are built with unstructured meshes to simulate with realism the linear accelerator head geometry.Ha recibido financiación de DESARROLLO E IMPLEMENTACION DE UN SISTEMA DE PLANIFICACION RADIOTERAPÉUTICA (TELETERAPIA Y BRAQUITERAPIA) BASADO EN EL MÉTODO MONTECARLO PARA EL HUPLAFE (UNIVERSIDAD POLITECNICA DE VALENCIA), DESARROLLO Y VALIDACION EXPERIMENTAL DE UN SISTEMA DE PLANIFICACION DE TRATAMIENTOS RADIOTERAPEUTICOS UTILIZANDO TECNICAS DE MONTE CARLO (UNIVERSIDAD POLITECNICA DE VALENCIA), Dosimetría in vivo en radioterapia sin filtro aplanador (FFF) (UNIVERSIDAD POLITECNICA DE VALENCIA), MONTE CARLO TREATMENT PLANNING SYSTEM: SOFTWARE PARA EL CALCULO DOSIMETRICO DE ALTA PRECISION EN RADIOTERAPIA. (UNIVERSIDAD POLITECNICA DE VALENCIA) y SISTEMA DE PLANIFICACION DE TRATAMIENTOS EN RADIOTERAPIA BASADO EN EL METODO DE MOTE CARLO Y EL PROCESADO 3D DE IMAGENES MEDICAS (GENERALITAT VALENCIANA)Morató-Rafet, S.; Juste Vidal, BJ.; Miró Herrero, R.; Verdú Martín, GJ. (2017). VARIAN CLINAC 6 MeV Photon Spectra Unfolding using a Monte Carlo Meshed Model. EPJ Web of Conferences. 153(04012):1-7. https://doi.org/10.1051/epjconf/201715304012S171530401

Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing