Dose evaluation based on 24Na activity in the human body at the JCO criticality accident in Tokai-mura

An optimized method for neutron dose evaluation based on24Na activity distribution in the human body

The neutron dose resulting from external irradiation can be evaluated by measuring the counts of characteristicγrays produced by²⁴Na in the human body. The detection geometry with the highest detection efficiency for measuring the whole-body²⁴Na activity has not been studied. In this work, the MCNP code is used to calculate the spatial distribution of²⁴Na in the human body irradiated by neutrons with different energies in different irradiation geometries. The fluence distribution of²⁴Na characteristicγrays on the body surface is calculated. The counts of²⁴Na characteristicγrays induced by monochromatic neutron irradiation are simulated to fit the scenarios of neutron irradiation by a continuous energy spectrum neutron. When the spontaneous neutrons from²⁵²Cf with 1Gy dose irradiate the human body, (3.63-4.35) × 10^{10 24}Na atoms are produced. The lower detection limit for the neutron absorption dose is reduced from ∼100 to less than 1 mGy when the radiation detector is placed over the back of the human body close to the liver. The relative error between the measured counts of²⁴Na characteristic γ rays caused by²⁵²Cf neutron irradiation and the counts fitted by monochromatic neutron irradiation data is less than 5.7%. The neutron dose received from a continuous energy spectrum neutron can be acquired quickly and accurately by weighted summing of the data for monochromatic neutron irradiations calculated in this paper, which is more convenient and practical than the previous method.

Retrospective physical dosimetry in the Czech Republic: an overview of already established methods and recent research

PURPOSE

The threat of serious radiation exposures to members of the public from radiological incidents and nuclear events has led to intensive study of a number of emergency dosimetry techniques for purposes of triage. As such, a national laboratory of retrospective dosimetry was established in our institute. The purpose of this work is to provide a summary of the well-established and already implemented retrospective physical dosimetry techniques based on thermoluminescence (TL), optically stimulated luminescence (OSL) and neutron activation including their specifics. Moreover, we present some new results of the experimental work, in which we compared dosimetry potential of various dental repair materials and human teeth.

MATERIALS AND METHODS

At first, an overview of already established retrospective physical retrospective methods including their main features was compiled. As regards recent research, an experimental comparative study was performed under defined conditions. The materials used were aliquots prepared from both pure and repaired teeth and aliquots of unused dental ceramics of known type. Following irradiation, we compared TL and OSL curves of the materials. We also compared dosimetry characteristics of OSL signal as reproducibility, dose dependence and fading.

RESULTS

After irradiation, the teeth aliquots of dental enamel and dentin exhibited very low OSL and TL signals compared with aliquots containing some dental repair materials or aliquots of pure dental ceramics. With a few exceptions, the OSL signal of dental enamel and dentin aliquots irradiated to 2 Gy was hardly distinguishable from OSL signal corresponding to unirradiated aliquots. In contrast, aliquots of teeth containing some dental repair material and aliquots of pure dental ceramics provided a well reproducible OSL signal exhibiting linear dose response. All the materials tested exhibited a significant fading of the OSL signal. The loss of OSL signal during the first 24 hours after irradiation was from 20 to 99% of its original value obtained immediately after the irradiation.

CONCLUSIONS

The already established physical methods of retrospective dosimetry use a spectrum of verified materials and techniques for dose assessment in the aftermath of serious radiological incidents and nuclear events. In the comparative study, we found that the dosimetry potential of teeth in natural state is much worse compared to teeth repaired with dental ceramics or dental cement fillings. Teeth restored with dental repair materials exhibited relatively favorable dosimetry characteristics. However, they can be usable for a dose reconstruction only on condition that the main practical problems connected with fading and optical bleaching were solved.

Induced radioactivity in the blood of cancer patients following Boron Neutron Capture Therapy

Since 1990, Boron Neutron Capture Therapy (BNCT) has been used for over 400 cancer patients at the Kyoto University Research Reactor Institute (KURRI). After BNCT, the patients are radioactive and their (24)Na and (38)Cl levels can be detected via a Na-I scintillation counter. This activity is predominantly due to (24)Na, which has a half-life of 14.96 h and thus remains in the body for extended time periods. Radioactive (24)Na is mainly generated from (23)Na in the target tissue that is exposed to the neutron beam in BNCT. The purpose of this study is to evaluate the relationship between the radioactivity of blood (24)Na following BNCT and the absorbed gamma ray dose in the irradiated field. To assess blood (24)Na, 1 ml of peripheral blood was collected from 30 patients immediately after the exposure, and the radioactivity of blood (24)Na was determined using a germanium counter. The activity of (24)Na in the blood correlated with the absorbed gamma ray doses in the irradiated field. For the same absorbed gamma ray dose in the irradiated field, the activity of blood (24)Na was higher in patients with neck or lung tumors than in patients with brain or skin tumors. The reasons for these findings are not readily apparent, but the difference in the blood volume and the ratio of bone to soft tissue in the irradiated field, as well as the dose that leaked through the clinical collimator, may be responsible.

Assessment of radiological effects on the regional environment due to the operation of the Tokai Reprocessing Plant

The regional environmental radiological effects were assessed for the past 25-year operation of the Tokai Reprocessing Plant (TRP). The assessment was basically performed with the environmental radiological monitoring data around the TRP. For the environmental monitoring, various kinds of terrestrial and marine samples including air dust, surface soil, polished rice grain, leafy vegetable, milk, seawater, seabed sediments, fish, shellfish and seaweed were collected in the surrounding environment of the TRP. Radionuclides such as (3)H, (14)C, (90)Sr, (137)Cs and (239,240)Pu in the environmental samples were determined by radiochemical methods. However, they showed no significant short-term increase or long-term accumulation of radionuclides discharged from the TRP. Therefore, the public dose was evaluated using the mathematical models and the discharge data of radionuclides. The estimated annual effective dose for the public was about 0.1% of the annual effective dose limit recommended by the ICRP. The assessment showed that there were no significant radiological effects on the environment and the public due to the 25-year operation of the TRP.

Criticality accident dosimetry systems: an international intercomparison at the SILENE reactor in 2002

In criticality accident dosimetry and more generally for high dose measurements, special techniques are used to measure separately the gamma ray and neutron components of the dose. To improve these techniques and to check their dosimetry systems (physical and/or biological), a total of 60 laboratories from 29 countries (America, Europe, Asia) participated in an international intercomparaison, which took place in France from 9 to 21 June 2002, at the SILENE reactor in Valduc and at a pure gamma source in Fontenay-aux-Roses. This intercomparison was jointly organised by the IRSN and the CEA with the help of the NEA/OCDE and was partly supported by the European Communities. This paper describes the aim of this intercomparison, the techniques used by the participants and the two radiation sources and their characteristics. The experimental arrangements of the dosemeters for the irradiations in free air or on phantoms are given. Then the dosimetric quantities measured and reported by the participants are summarised, analysed and compared with the reference values. The present paper concerns only the physical dosimetry and essentially experiments performed on the SILENE facility. The results obtained with the biological dosimetry are published in two other papers of this issue.

External doses to 350 m zone residents due to anisotropic radiation from the JCO criticality accident in Tokai-mura

The validity of a method for individual dose reconstruction based on the anisotropic radiation distribution was confirmed by a comparison with the dose from Na-24 whole-body counting for seven persons at a neighboring company located on the western side of the JCO campus. The successful coincidence between Di (Na-24) and Di (present) also supports the validity of the second version of D(r) reported by the Head Office of Countermeasure. The present dose reconstruction for the 350 m zone at the western side showed an average value of 0.7 mSv and a maximal value of 3.1 mSv, as indoor-dose under the assumption of an effective transmittance of 0.4 for all of the houses. If all of the residents in 350 m zone were indoors during the accident, 83% of them might have received external doses of less than 1 mSv. The radiation exposure to the nearest residential area in the southern-west direction was significantly reduced with phi(theta(i)) between 0.4 and 0.2 by several buildings on the JCO campus. The present study on the public dose confirms that the official report on the public dose (the maximal value of 21 mSv for individual doses) from Head Office of Countermeasure is significantly overestimated due to their isotropic treatment of radiation from the source.