Dosimetry assessment during the sipping test in the IPR-R1 TRIGA reactor using MCNPX

Abstract

A sipping test device recently constructed at Centro de Desenvolvimento da Tecnologia Nuclear e CDTN (Nuclear Technology Development Center - CDTN), Belo Horizonte, Brazil, will be used to inspect irradiated fuel elements cladding in the IPR-R1 TRIGA reactor. The device is important to check the integrity of the irradiated fuel elements cladding of this reactor, which can have been affected by corrosion over long periods of time. This paper describes the methodology used in the characterization of gamma doses received by the personnel during the manipulation of the sipping test device. Results are presented for dose rates calculated in the reactor pool surface using the Monte Carlo code MCNPX. Validation of MCNPX calculations of gamma doses were carried out thought comparison with experimental measurements. The difference between the measured and calculated values is lower than 13%. The results provided confirmatory evidence that during the sipping test the manual handling is safety.