Numerical system utilising a Monte Carlo calculation method for accurate dose assessment in radiation accidents.
A system utilising radiation transport codes has been developed to derive accurate dose distributions in a human body for radiological accidents. A suitable model is quite essential for a numerical analysis. Therefore, two tools were developed to setup a 'problem-dependent' input file, defining a radiation source and an exposed person to simulate the radiation transport in an accident with the Monte Carlo calculation codes-MCNP and MCNPX. Necessary resources are defined by a dialogue method with a generally used personal computer for both the tools. The tools prepare human body and source models described in the input file format of the employed Monte Carlo codes. The tools were validated for dose assessment in comparison with a past criticality accident and a hypothesized exposure.
Takahashi
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F
Research Group for Radiation Protection, Division of Environment and Radiation Sciences, Nuclear Science and Energy Directorate, Japan Atomic Energy Agency (JAEA), Shirakata 2-4, Tokai-mura, Ibaraki 319-1195, Japan. takahashi.fumiaki@jaea.go.jp
Endo
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D003198:Computer Simulation; D004867:Equipment Design; D019544:Equipment Failure Analysis; D006801:Humans; D008954:Models, Biological; D009010:Monte Carlo Method; D009502:Neutrons; D016273:Occupational Exposure / Q000032:analysis*; D011829:Radiation Dosage; D011834:Radiation Monitoring / Q000379:methods*; D011835:Radiation Protection / Q000379:methods*; D018788:Radioactive Hazard Release; D015203:Reproducibility of Results; D018570:Risk Assessment / Q000379:methods*; D012680:Sensitivity and Specificity