Assessment of Human Body Surface and Internal Dose Estimations in Criticality Accidents Based on Experimental and Computational Simulations

Abstract

For a study on the applicability of a personal dosimetry method to criticality accident dosimetry, an assessment of the human body surface and internal dose estimations was performed by experimental and computational simulations. The experimental simulation was carried out in a criticality accident situation created at the TRACY facility. The neutron and gamma-ray absorbed doses in muscle tissue were separately estimated by a dosimeter set of an alanine dosimeter and a thermoluminescence dosimeter made of enriched lithium tetra borate with a phantom. The computational simulation was conducted with a Monte Carlo code taking account of dose components of neutrons, prompt gammarays and delayed gamma-rays. Prior to the assessment of the body surface and internal dose estimations, the computational simulation was ascertained to be valid by comparison between the calculated dose distributions in the phantom and the measured ones. The assessment of the body surface and internal dose estimations based on the experimental and computational simulations confirmed that the personal dosimetry using the dosimeter set provided a first estimation of the body surface and internal doses with precision enough for emergency medical care.

Retrospective Estimation of the Spatial Dose Distribution and the Number of Fissions in Criticality Accident Using Area Dosimeters

“…The uncertainty of 10-20% is allowable in radiation emergency medicine: its criterion is roughly 50% in 48 hrs and 25% in a week after the accident. 11) After the thorough investigation of the accident, the Monte Carlo simulation 10,12) under the condition at the time of the accident can remove the uncertainty from the first estimation.…”

“…The unit dose D f can be determined by a Monte Carlo simulation taking account of the dose components of the prompt neutrons, the prompt gamma-rays and the delayed gamma-rays emitted in criticality accident situations, 10,12) based on the conditions of fissile materials which have been investigated. Before determining the D f value, the following values may be used as an initial guess of D f to obtain a first estimation of the number of fissions: ð21 AE 3Þ Â 10 À17 Gy/ fission for the sum of neutron and gamma-ray doses D n measured with the alanine dosimeter, ð13 AE 2Þ Â 10 À17 Gy/ fission for the gamma-ray dose D measured with the 7 Li 2 11 B 4 O 7 TLD, and ð8 AE 1Þ Â 10 À17 Gy/fission for the neutron dose D n (the difference in these two dose values).…”

“…These values were determined as gross doses exclusively for the TRACY core where 10 wt%-enriched uranium nitrate solution was fed into a cylindrical core tank having a 10 mmthick wall. 9,10,12) The D f value can vary with the condition parameters of radiation shielding such as the thickness of tank wall. According to the Monte Carlo simulation under the varying conditions of the TRACY core, 10,12) the relative variation of the D f value with the thickness of tank wall is estimated to be in the range of AE10% for the neutron dose and AE30% for the gamma-ray dose, as shown in Fig.…”