Ionising radiation dose calculations for the release of 131I during accident conditions at the SAFARI-1 materials test reactor

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dc.contributor.author Bekker, Willem Adriaan
dc.date.accessioned 2011-03-14T09:06:26Z
dc.date.available 2011-03-14T09:06:26Z
dc.date.issued 2011-03-14
dc.identifier.uri http://hdl.handle.net/10539/9153
dc.description.abstract This study demonstrates how the Monte Carlo N-Particle (MCNP) radiation transport code can be used as a contribution to the family of Safety Analysis tools for Research Reactors. Since Research Reactors are used worldwide over a wide range of research and commercial applications it can be justified that effort must be spent to improve on safety analysis. While the advantages of these installations to society are widely recognized, it is still necessary that safety analysis provides the necessary assurance that these installations do not cause an undue risk to society. A Materials Test Reactor (MTR) is used as an example to describe the Safety Analysis steps that need to be done, limited to the reactor and its building. A radioactive inventory in the reactor core is determined. For further evaluation of the effects of the release due to a hypothetical accident in such a reactor, the 131I radioisotope is chosen to demonstrate the capabilities of the MCNP code. A 131I cloud resulting from the release is simulated together with an MCNP model of the reactor building. Personnel in the proximity of the 131I cloud for short times, either due to emergency actions or accidental entrapment, are also modelled. The external photon whole-body doses and -decay (electron) skin doses are determined. The MCNP code results are also benchmarked against another method. The findings show that the method presented in the study could be used to predetermine emergency actions that could be incorporated into emergency planning and even into design of a new research reactor. This is substantiated by the conclusion that only the external photon dose could result in unacceptable doses above 500 mSv for short exposure times. The study concludes that the MCNP code can be used effectively for Safety Analyses and leaves the opportunity open to expand the use of the code to other fields related to research reactors. en_US
dc.language.iso en en_US
dc.title Ionising radiation dose calculations for the release of 131I during accident conditions at the SAFARI-1 materials test reactor en_US
dc.type Thesis en_US


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