Calculation of absorbed dose around a facility for disposing of low activity natural radioactive waste (C3-dump).
A C3-dump is a facility for disposing of low activity natural radioactive waste containing the uranium series 238U, the thorium series 232Th and 40K. Only the external radiation owing to gamma rays, X-rays and annihilation photons is considered in this study. For two situations--the semi-infinite slab and the tourist geometry--the conversion coefficients from specific activity to air kerma rate at 1 m above the relevant level are calculated. In the first situation the waste material is in contact with the air but in the tourist geometry it is covered with a 1.35 m thick layer. For the calculations, the Monte Carlo radiation transport code MCNP is used. The yield and photon energy for each radionuclide are according to the database of Oak Ridge National Laboratory. For the tourist situation, the depth-dose distribution through the covering layer is calculated and extrapolated to determine the exit dose.
Jansen
J T M
JT
Medical Physics, Radiation Technology Group, Interfaculty Reactor Institute, Delft University of Technology, Mekelweg 15, NL-2629 JB Delft, The Netherlands. JTJansen@iri.tudelft.nl
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D003198:Computer Simulation; D004867:Equipment Design; D019544:Equipment Failure Analysis; D015233:Models, Statistical; D009010:Monte Carlo Method; D009426:Netherlands; D011829:Radiation Dosage; D011834:Radiation Monitoring / Q000379:methods* / Q000592:standards; D011835:Radiation Protection / Q000295:instrumentation* / Q000379:methods; D011850:Radioactive Waste / Q000032:analysis*; D012542:Scattering, Radiation; D014501:Uranium / Q000032:analysis*; D018505:Waste Management / Q000295:instrumentation* / Q000379:methods
D011850:Radioactive Waste; D014501:Uranium