Rivard, M. J., Granero, D., Perez-Calatayud, J., & Ballester, F. (2010). Influence of photon energy spectra from brachytherapy sources on Monte Carlo simulations of kerma and dose rates in water and air. Med. Phys., 37(2), 869–876.
Abstract: Methods: For Ir-192, I-125, and Pd-103, the authors considered from two to five published spectra. Spherical sources approximating common brachytherapy sources were assessed. Kerma and dose results from GEANT4, MCNP5, and PENELOPE-2008 were compared for water and air. The dosimetric influence of Ir-192, I-125, and Pd-103 spectral choice was determined. Results: For the spectra considered, there were no statistically significant differences between kerma or dose results based on Monte Carlo code choice when using the same spectrum. Water-kerma differences of about 2%, 2%, and 0.7% were observed due to spectrum choice for Ir-192, I-125, and Pd-103, respectively (independent of radial distance), when accounting for photon yield per Bq. Similar differences were observed for air-kerma rate. However, their ratio (as used in the dose-rate constant) did not significantly change when the various photon spectra were selected because the differences compensated each other when dividing dose rate by air-kerma strength. Conclusions: Given the standardization of radionuclide data available from the National Nuclear Data Center (NNDC) and the rigorous infrastructure for performing and maintaining the data set evaluations, NNDC spectra are suggested for brachytherapy simulations in medical physics applications.
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Hornillos, M. B. G., Gorlychev, V., Caballero, R., Cortes, G., Poch, A., Pretel, C., et al. (2011). Monte Carlo Simulations for the Study of a Moderated Neutron Detector. J. Korean Phys. Soc., 59(2), 1573–1576.
Abstract: This work presents the Monte Carlo simulations performed with the MCNPX and GEANT4 codes for the design of a BEta deLayEd Neutron detector, BELEN-20. This detector will be used for the study of beta delayed neutron emission and consists of a block of polyethylene with dimensions 90 x 90 x 80 cm(3) and 20 cylindrical (3)He gas counters. The results of these simulations have been validated experimentally with a (252)Cf source in the laboratory at UPC, Barcelona. Also the first experiment with this detector has been carried out in November 2009 in JYFL, Finland. In this experiment the neutron emission probability after beta decay of the fission products (88)Br, (94,95)Rb, and (138)I has been measured; this data is still under analysis. Simulations with MCNPX and GEANT4 have been performed in order to obtain the efficiency of the BELEN-20 detector for each of the above nuclei using the neutron energy distribution corresponding to each nucleus.
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n_TOF Collaboration(Mendoza, E. et al), Giubrone, G., & Tain, J. L. (2011). Improved Neutron Capture Cross Section Measurements with the n_TOF Total Absorption Calorimeter. J. Korean Phys. Soc., 59(2), 1813–1816.
Abstract: The n_TOF collaboration operates a Total Absorption Calorimeter (TAC) [1] for measuring neutron capture cross-sections of low-mass and/or radioactive samples. The results obtained with the TAC have led to a substantial improvement of the capture cross sections of (237)Np and (240)Pu [2]. The experience acquired during the first measurements has allowed us to optimize the performance of the TAC and to improve the capture signal to background ratio, thus opening the way to more complex and demanding measurements on rare radioactive materials. The new design has been reached by a series of detailed Monte Carlo simulations of complete experiments and dedicated test measurements. The new capture setup will be presented and the main achievements highlighted.
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