Takumaro Momose

Estimation of internal exposure of the thyroid to 131I on the basis of 134Cs accumulated in the body among evacuees of the Fukushima Daiichi Nuclear Power Station accident

Namie Town was heavily contaminated by the Fukushima Daiichi Nuclear Power Station accident. The thyroid equivalent dose for residents who lived in Namie was estimated using results of whole body counting examinations which were carried out by the Japan Atomic Energy Agency a few months after the nuclear accident. Photon peaks of (131)I and (134)Cs were previously measured by the authors using a NaI(Tl) scintillation spectrometer and that information was used to estimate the (131)I/(134)Cs activity ratio of total intake in the present study. The maximum values of (131)I/(134)Cs activity ratio corresponding to thyroid uptake factors of 0.3, 0.1 and 0.03 were evaluated to be 0.9, 2.6 and 8.7, respectively. The maximum value of the (131)I/(134)Cs activity ratio was used to obtain the most conservative thyroid equivalent dose estimation. The maximum internal exposure of the thyroid to (131)I on the basis of (134)Cs accumulated in the body measured by the whole body counter was estimated to be 18mSv. This value was much smaller than 50mSv that the International Atomic Energy Agency recommends as the dose at which exposed persons should take stable iodine tablets.

Internal thyroid doses to Fukushima residents—estimation and issues remaining

Enormous quantities of radionuclides were released into the environment following the disastrous accident at the Fukushima Daiichi Nuclear Power Plant (FDNPP) in March 2011. It is of great importance to determine the exposure doses received by the populations living in the radiologically affected areas; however, there has been significant difficulty in estimating the internal thyroid dose received through the intake of short-lived radionuclides (mainly, 131I), because of the lack of early measurements on people. An estimation by the National Institute of Radiological Sciences for 1 April 2012 to 31 March 2013 was thus performed using a combination of the following three sources: thyroid measurement data (131I) for 1080 children examined in the screening campaign, whole-body counter measurement data (134Cs, 137Cs) for 3000 adults, and atmospheric transport dispersion model simulations. In this study, the residents of Futaba town, Iitate village and Iwaki city were shown to have the highest thyroid equivalent dose, and their doses were estimated to be mostly below 30 mSv. However, this result involved a lot of uncertainties and provided only representative values for the residents. The present paper outlines a more recent dose estimation and preliminary analyses of personal behavior data used in the new method.

EARLY INTAKE OF RADIOCESIUM BY RESIDENTS LIVING NEAR THE TEPCO FUKUSHIMA DAI -ICHI NUCLEAR POWER PLANT AFTER THE ACCIDENT.PART 1: INTERNAL DOSES BASED ON WHOLE-BODY MEASUREMENTS BY NIRS

AbstractThe Tokyo Electric Power Company’s Fukushima Dai-ichi Nuclear Power Plant (FDNPP) accident in 2011 resulted in a release of radionuclides into the environment (131I: 142.9 PBq, 137Cs:12.4 PBq). This study presents the results of internal doses to 174 residents living near the FDNPP at the time of the accident based on whole-body (WB) measurements performed by the National Institute of Radiological Sciences (NIRS) during the period between 27 June and 28 July 2011. The 174 subjects consisted of 125 adults (≥18-y) and 49 children (<18-y) and included 90 persons of Namie town, one of the municipalities heavily contaminated with the radionuclides. The number of subjects with significant detection of both 134Cs and 137Cs was relatively small: 28.8% for the adults and 4.1% for the children. A significant gender difference in the Cs detection rate (males > females) was observed in the adults but not the children. In this study, the committed effective dose (CED) from 134Cs and 137Cs was calculated based on individual WB contents (134Cs) corrected against body size, the observed body content ratio of 137Cs to 134Cs, and the assumed intake scenario (namely, acute inhalation of Type F compounds on 12 March 2011 when the first explosive event occurred at the site of the FDNPP). The 90th-percentile CED value for the adults was around 0.1 mSv and the maximum CED (0.63 mSv) was found in an elderly male. Comparable CED results were obtained in other WB measurements subsequently performed by the Japan Atomic Energy Agency (JAEA) in a similar manner to that of the NIRS, suggesting that the contribution of ingestion to the WB content observed would be trivial for most of the JAEA subjects. The intake ratio of 131I to 134Cs was evaluated to be 3~5 based on the 131I thyroid measurement data of Tokonami et al. Using the average intake ratio of 3.8, the resulting median and maximum thyroid-equivalent doses to the adult subjects of this study were estimated at 3.5 mSv and 84 mSv, respectively.

Measurements of 131I in the thyroids of employees involved in the Fukushima Daiichi nuclear power station accident

The Great East Japan Earthquake Disaster on 11 March 2011 caused an unprecedented accident at the Fukushima Daiichi nuclear power station operated by Tokyo Electric Power Company (TEPCO). Nuclear Fuel Cycle Engineering Laboratories of Japan Atomic Energy Agency performed internal dose measurements of 560 employees involved in the accident during the period from 20 April to 5 August in 2011 at the request of TEPCO. The present paper describes our measurements of 131I in the thyroid that is the predominant contributor to the internal dose. These measurements were carried out using an HPGe detector installed in a low-background shielded chamber made of 20-cm-thick steel and the detector was placed adjacent to the subjects neck. The typical minimum detectable activity of this technique was 10 Bq for a counting time of 10 min; however, this sensitivity made it difficult to identify a residual thyroid content of 131I corresponding to a committed effective dose of 20 mSv for late subjects. This paper discussed technical issues experienced through the measurements such as the influence of 131I in the rest of the body, the calibration phantom of use, and so on.

Feasibility of in vivo measurement of 239Pu distribution in lungs using an imaging plate

An in vivo measurement system using an imaging plate (IP) system was developed, which displayed images reflecting (239)Pu distribution in the lung of a phantom. The detection limits of the IP system for 1-12h exposures were between 1670 and 245 Bq at a 1.6 cm chest wall thickness. The detection limit of the IP system for a 2.5h exposure was equal to that of a germanium detector for a 0.5h measurement. The IP system could be used as a new device for in vivo measurement of (239)Pu in the lung.

Development of 241Am lung monitoring system using an imaging plate

A new 241Am lung monitoring system without shielding was devised by using an imaging plate system. The Lawrence Livermore National Laboratorys realistic torso phantom containing a 241Am lung was covered by imaging plates sealed in lightproof bags. The imaging plate system displayed 241Am lung images characteristic of the lung shape of the torso phantom. The imaging plate systems lower detection limits of 14 Bq for 60 min exposure and 6 Bq for 300 min were the same levels as those of the phoswich detectors and the germanium detectors placed in shielded rooms. The imaging plate system for 60 min exposure detected about 2% of the annual limit of 740 Bq for 241Am inhalation. A lung monitoring system using imaging plates is applicable for 241Am lung monitoring. Health Phys. 93(1):28–35; 2007

INTERNAL DOSES OF THREE PERSONS STAYING 110 KM SOUTH OF THE FUKUSHIMA DAIICHI NUCLEAR POWER STATION DURING THE ARRIVAL OF RADIOACTIVE PLUMES BASED ON DIRECT MEASUREMENTS

The authors describe the results of direct measurements made on three persons who stayed in Tokai-mura, a village located ∼110 km south of the Fukushima Daiichi Nuclear Power Station (FDNPS), during the arrival of significant radioactive plumes released from the FDNPS as a consequence of the Tohoku earthquake/tsunami/FDNPS accident in March 2011. These measurements were made using a NaI(Tl) spectrometer and a whole-body counter shortly after the accident. Their thyroid equivalent doses ((131)I) were estimated to be 0.9-1.4 mSv under the assumption of acute intake via inhalation on 15 March, when the first significant release event was observed. Although greatly depending on the physicochemical form of iodine, the intake amount ratios of (131)I to (137)Cs for the three subjects were calculated as 2.7-3.7, which were much smaller than the radioactivity ratio (7.8) found in air sampling at the same site.

Radiation protection at nuclear fuel cycle facilities

Radiation protection methodologies concerning individual monitoring, workplace monitoring and environmental monitoring in nuclear fuel facilities have been developed and applied to facilities in the Nuclear Fuel Cycle Engineering Laboratories (NCL) of Japan Atomic Energy Agency (JAEA) for over 40 y. External exposure to photon, beta ray and neutron and internal exposure to alpha emitter are important issues for radiation protection at these facilities. Monitoring of airborne and surface contamination by alpha and beta/photon emitters at workplace is also essential to avoid internal exposure. A critical accident alarm system developed by JAEA has been proved through application at the facilities for a long time. A centralised area monitoring system is effective for emergency situations. Air and liquid effluents from facilities are monitored by continuous monitors or sampling methods to comply with regulations. Effluent monitoring has been carried out for 40 y to assess the radiological impacts on the public and the environment due to plant operation.

REIDAC-A Software Package for Retrospective Dose Assessment in Internal Contamination with Radionuclides

For cases of internal contamination with radionuclides, it is necessary to perform an internal dose assessment to facilitate radiation protection. For this purpose, the ICRP has supplied the dose coefficients and the retention and excretion rates for various radionuclides. However, these dosimetric quantities are calculated under typical conditions and are not necessarily detailed enough for dose assessment situations in which specific information on the incident or/and individual biokinetic characteristics could or should be taken into account retrospectively. This paper describes a newly developed PC-based software package called Retrospective Internal Dose Assessment Code (REIDAC) that meets the needs of retrospective dose assessment. REIDAC is made up of a series of calculation programs and a package of software. The former calculates the dosimetric quantities for any radionuclide being assessed and the latter provides a user with the graphical user interface (GUI) for executing the programs, editing parameter values and displaying results. The accuracy of REIDAC was verified by comparisons with dosimetric quantities given in the ICRP publications. This paper presents the basic structure of REIDAC and its calculation methods. Sensitivity analysis of the aerosol size for 239Pu compounds and provisional calculations for wound contamination with 241Am were performed as examples of the practical application of REIDAC.

AN ACCIDENT OF INTERNAL CONTAMINATION WITH PLUTONIUM AND AMERICIUM AT A NUCLEAR FACILITY IN JAPAN: A PRELIMINARY REPORT AND THE POSSIBILITY OF DTPA ADMINISTRATION ADDING TO THE DIAGNOSIS

This article introduces the first accident of internal contamination with plutonium (Pu) or americium (Am) in Japan for which treatment was carried out. An accident of internal contamination with Pu and Am occurred at a Pu research facility at Oarai-town of Ibaraki prefecture in Japan. A plastic bag containing these radionuclides ruptured when five workers were inspecting a storage container in a hood. As a consequence, these workers were internally contaminated with Pu and Am. Although contamination on the body surface was observed in all five workers, a positive nasal swab was detected in only three of them. A chelating agent, calcium diethylenetriaminepenta-acetate (CaDTPA), was administered to all of them including the two workers without a positive nasal swab. However, bioassay detected a significant amount of Pu and Am in urine after administration of DTPA in these two workers, whereas the levels of these nuclides were below minimum detectable levels in urine before the administration. Since the prevalence of adverse reactions in DTPAs is low, the present results suggest that administration of DTPA can be used for the diagnosis of internal contamination even when a nasal swab is negative or contamination around body orifices is not detected.