PAULO DE TARSO DALLEDONE SIQUEIRA

PAULO DE TARSO DALLEDONE SIQUEIRA

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Design, fabrication and modeling of an AmBe neutron irradiator for TLD screening for neutron dose measurement in mixed radiation fields
2019 - CAVALIERI, TASSIO A.; SIQUEIRA, PAULO T.D.; SHORTO, JULIAN M.B.; GENEZINI, FREDERICO A.; YORIYAZ, HELIO
TLDs dosimeters are frequently presented as a viable choice for dosimetric studies when dealing with mixed neutron-gamma radiation fields. However, this choice is not without some drawbacks, because not only TLD response is highly dependent on particle type but also on neutron energy spectrum. Therefore, a correct screening and calibration of the dosimeter are required, and a simple shift from gamma screening methodology for mixed field is not suitable. This paper presents the design, fabrication and tests of an irradiator for TLD screening for neutron dose measurement using an AmBe source and polyethylene as moderator material. The design of the irradiator was conducted through Monte Carlo simulations using the MCNP5 code. The experimental validation and tests were performed using Indium activation foils and TLD 600 dosimeters. The manufactured irradiator demonstrated to be suitable for TLD screening under neutron source radiation field, offering very good homogeneity conditions in the radiation field so to guarantee same radiation dose delivered to the TLDs.
Study of pixel damages in CCD cameras irradiated at the neutron tomography facility of IPEN-CNEN/SP
2015 - PUGLIESI, R.; ANDRADE, M.L.G.; DIAS, M.S.; SIQUEIRA, P.T.D.; PEREIRA, M.A.S.
Recalibration of U-doped standard glasses through uranium thin film for neutron-fluence measurements
2014 - SOARES, C.J.; GUEDES, S.; CURVO, E.A.C.; HADLER, J.C.; JONCKHEERE, R.; TELLO, C.A.; LIXANDRAO FILHO, A.L.; SIQUEIRA, P.T.D.; MADI FILHO, TUFIC
Validation of neutronic models and calculation systems by means of experimental results in the IPEN/MB-01 reactor
2013 - SANTOS, ADIMIR dos; GRANT, CARLOS R.; SIQUEIRA, PAULO de T.D.; TARAZAGA, ARIEL E.; ANDRADE e SILVA, GRACIETE S. de; BARBERIS, CLÁUDIA M.
Boron thin films and CR-39 detectors in BNCT: A method to measure the sup(10)B(n, 'alfa')sup(7)Li reaction rate
2013 - SMILGYS, B.; GUEDES, S.; MORALES, M.; ALVAREZ, F.; HADLER, J.C.; COELHO, P.R.P.; SIQUEIRA, P.T.D.; ALENCAR, I.; SOARES, C.J.; CURVO, E.A.C.
Study of neutron-DNA interaction at the IPEN BNCT research facility
2004 - QUAL, M.R.; RODRIGUEZ, O.; GUZMAN, F.; DEPPMAN, A.; ARRUDA NETO, J.D.T.; LIKHACHEV, V.P.; COELHO, P.R.P.; SIQUEIRA, P.T.D.
Monte Carlo simulation as an auxiliary tool for electron beam quality specification for intra-operative radiotherapy
2005 - CASTRO NETO, ALUISIO J. de; HADDAD, CECILIA M.K.; PELOSI, EDILSON L.; ZEVALLOS CHAVEZ, JUAN Y.; YORIYAZ, HELIO; SIQUEIRA, PAULO de T.D.
Intra-operative radiotherapy (IORT) using electron beams has demonstrated to be a good alternative as part of the breast-conserving surgery. Besides, as the computer processing capacity has been increasing along the years, it has become a potential auxiliary tool in radiotherapy treatment planning. In this work these streams are merged together: simulations were performed by different Monte Carlo radiation transport codes (EGS4 and MCNP in its releases 4C and 5) in an attempt to not only examine the efficacy of Al and Pb discs used in IORT procedures to protect critical regions but also to compare the performance of the forementioned codes.
Numerical analysis of an incremented statistical sampling procedure in MCNP
2008 - SIQUEIRA, PAULO de T.D.; YORIYAZ, HELIO; SANTOS, ADIMIR dos; PASCHOLATI, PAULO R.
MCNP has stood so far as one of the main Monte Carlo radiation transport codes. Its use, as any other Monte Carlo based code, has increased as computers perform calculations faster and become more affordable along time. However, the use of Monte Carlo method to tally events in volumes which represent a small fraction of the whole system may turn to be unfeasible, if a straight analogue transport procedure (no use of variance reduction techniques) is employed and precise results are demanded. Calculations of reaction rates in activation foils placed in critical systems turn to be one of the mentioned cases. The present work takes advantage of the fixed source representation from MCNP to perform the above mentioned task in a more effective sampling way (characterizing neutron population in the vicinity of the tallying region and using it in a geometric reduced coupled simulation). An extended analysis of source dependent parameters is studied in order to understand their influence on simulation performance and on validity of results. Although discrepant results have been observed for small enveloping regions, the procedure presents itself as very efficient, giving adequate and precise results in shorter times than the standard analogue procedure.