Influence of short-lived radioiodines other than 131I on screening direct thyroid measurements with TCS-172 NAI(TL) survey meters.

In a nuclear emergency, one of the actions taken for the sake of public is to monitor thyroid exposure to radioiodines. Japan's Nuclear Regulation Authority recently published a report on such monitoring and proposed direct thyroid measurements with conventional NaI(Tl) survey meters (e.g. Hitachi model TCS-172) as a primary screening method. A previous study proposed screening levels (SLs) used in these simplified measurements as the net reading values of the TCS-172 device. Age-specific SLs were derived from a thyroid equivalent dose of 100 mSv due to the inhalation intake of 131I. This study addressed the possible influence of short-lived iodine isotopes other than 131I on the simplified measurements. In preparation for such measurements, the responses of the device for 132I as an ingrowth component from 132Te, 133I, 134I and 135I in the thyroid were evaluated by numerical simulations using age-specific stylized phantoms in addition to those obtained for 131I in the previous study. The radioactivity ratios of the relevant isotopes were taken from the inventory data of the Fukushima Daiichi Nuclear Power Plant. The results were used to predict the net readings of the device when 132Te-132I and 133I as well as 131I were inhaled at 24 or 72 h after the shutdown of a nuclear power plant. In these cases, the signals from 132Te-132I and 133I become undetectable a couple of days after intake, which could lead to underestimations of the thyroid dose. To estimate the thyroid dose accurately from the simplified measurements, it is necessary to identify the exact time of intake after the shutdown and the actual physiochemical property of 132Te that affects the thyroid uptake of 132I.

Biokinetic model analysis with DTPA administration for a case of accidental inhalation of actinides in Japan.

Accidental inhalation intake of plutonium isotopes and 241Am occurred at a Pu research facility in Japan in 2017, and the five workers involved in this accident were treated by the administration of Ca/Zn-diethylenetriaminepentaacetic acid (DTPA). For the worker who was most internally exposed, the therapy was continued over 1 y after the accident. Urinary samples collected before and after each administration were subject to bioassay to evaluate the efficacy of the dose reduction. This study performed numerical analyses using a biokinetic model dealing with 241Am-DTPA with reference to the European Coordinated Network on Radiation Dosimetry approach, which assumes that the complex of actinides and Ca/Zn-DTPA is generated in the designated compartments in the biokinetic model. The results of the model prediction well captured the trend of the observed urinary excretion in the long-term bioassay and would be useful to evaluate the efficacy of the Ca/Zn-DTPA administration for the worker involved in the accident.

Screening levels of TCS-172 NaI(Tl) survey meters used for direct thyroid measurements in nuclear disasters.

It is a challenging task to establish a feasible and robust method for the population monitoring of individuals' thyroid exposure following an accidental intake of radioiodines in a nuclear emergency, because of the time restriction. The authors previously proposed a method for such monitoring to obtain as many reliable human data as possible and one of the components is simplified measurements by conventional NaI(Tl) survey meters that are intended to be used for the initial triage to identify significantly exposed individuals and get an overall picture of the exposure levels in a target population in a timely manner. This study determined screening levels (SLs) for a conventional NaI(Tl) survey meter (model TCS-172, Hitachi, Japan) using the conversion factor (131I kBq in the thyroid per µSv h-1) that were obtained from experiments and simulations with age-specific phantoms. The results demonstrated that the derived SLs for 100 mSv thyroid equivalent dose were as follows: 0.2 µSv h-1 (SL1) for the age group ≤ 5-y-olds, 0.5 µSv h-1 (SL2) for the 10- and 15-y-old age groups and 1.0 µSv h-1 (SL3) for adults. These SLs would be reasonably available within 1 week after the intake of 131I on the safe side.

Reassessment of Internal Thyroid Doses to 1,080 Children Examined in a Screening Survey after the 2011 Fukushima Nuclear Disaster.

The dose reconstruction of populations potentially affected by the accident at the Fukushima Daiichi nuclear power plant in March 2011 is of great importance. However, it has been difficult to assess internal thyroid doses to Fukushima residents (mainly from their intake of I) due to the lack of direct measurements. For the residents, only about 1,300 data points related to I are available, and 1,080 of the data points were obtained from the screening campaign that was conducted by the Nuclear Emergency Response Local Headquarters at the end of March 2011 in Kawamata Town, Iwaki City, and Iitate Village. Here, we reassessed thyroid doses to 1,080 subjects aged ≤15 y old using new age-specific conversion factors to determine I thyroid contents from net signals of the devices used, with consideration for the possible uncertainty related to the measurements. The results demonstrated that thyroid equivalent doses to the subjects were <30 mSv (excluding outliers). We also demonstrate dose distributions of each age group from the above three municipalities and those of subjects from Minamisoma City and Fukushima City. One of the findings was that the I intake was similar among different age groups in each of the three municipalities. This was consistent with the assumption that ingestion was a dominant route of intake rather than inhalation. The range of thyroid doses to Iitate Village residents was similar to that to Iwaki City residents even though the I concentration in tap water was much higher in Iitate Village than Iwaki City. The range of thyroid doses to Minamisoma City residents was similar to that to Iitate Village and Iwaki City residents, and the range for Fukushima City residents was smallest among the five municipalities. Since the major route of intake has remained unclear, this paper presents the plausible upper and lower thyroid doses, between which the actual doses are thought to mostly exist, based on two intake scenarios: single inhalation and repeated ingestion. Further research is thus necessary to extract useful evidence from the individual evacuation behaviors for improving the present internal thyroid dose assessment.

Numerical Simulation Based on Individual Voxel Phantoms for a Sophisticated Evaluation of Internal Doses Mainly From 131I in Highly Exposed Workers Involved in the TEPCO Fukushima Daiichi NPP Accident.

As a response to the Tokyo Electric Power Company's Fukushima Daiichi nuclear power plant accident in 2011, seven TEPCO workers whose exposure doses were expected to be >250 mSv (a tentative dose limit stipulated by the Japanese central authority) attended Japan's National Institute for Radiological Sciences for additional internal dose measurements. The National Institute for Radiological Sciences examination revealed that these workers' internal doses came mainly from their intake of the radionuclide I during emergency operations. In this study, we performed numerical simulations based on individual volume-pixel (voxel) phantoms of six of the seven workers for a more sophisticated evaluation of their internal doses, taking into account the individual thyroid size and other specific parameters. The voxel phantoms were created from magnetic resonance imaging scan images. As a result, the individual thyroid volumes ranged from 6.5 to 28.2 cm and were considerably smaller than the reference value (~20 cm) adopted in the International Commission on Radiation Protection's dosimetric model for four of the six subjects. Compared to the original estimates of the thyroid absorbed dose, our preliminary evaluation revealed values that were increased by approximately 3-fold or decreased by 30% at maximum. A wide difference in the individual thyroid size would be one of the significant modifiers in the current dose estimation of subjects of the ongoing epidemiological study project. The present simulations also provided evidence that the direct thyroid measurements by the National Institute for Radiological Sciences to determine the workers' I thyroid contents were sufficiently accurate.

A revised model for the deposition and clearance of inhaled particles in human extra-thoracic airways.

The International Commission on Radiological Protection (ICRP) Task Group that developed the Human Respiratory Tract Model for Radiological Protection (HRTM) identified a lack of published information on aspects of the clearance of inhaled particles deposited in the human nasal passage. Using the results of a recent human volunteer study on the clearance of inhaled particles from the nose, a revised model of clearance from the extra-thoracic (ET) airways has been developed that addresses important issues for which simplifying assumptions had to be made in the ICRP Publication 66 HRTM ET model. This ET clearance model has been adopted by ICRP for inclusion in the revised HRTM. The derivation of the model and parameter values from the experimental data are explained.

Radiological protection issues arising during and after the Fukushima nuclear reactor accident.

Following the Fukushima accident, the International Commission on Radiological Protection (ICRP) convened a task group to compile lessons learned from the nuclear reactor accident at the Fukushima Daiichi nuclear power plant in Japan, with respect to the ICRP system of radiological protection. In this memorandum the members of the task group express their personal views on issues arising during and after the accident, without explicit endorsement of or approval by the ICRP. While the affected people were largely protected against radiation exposure and no one incurred a lethal dose of radiation (or a dose sufficiently large to cause radiation sickness), many radiological protection questions were raised. The following issues were identified: inferring radiation risks (and the misunderstanding of nominal risk coefficients); attributing radiation effects from low dose exposures; quantifying radiation exposure; assessing the importance of internal exposures; managing emergency crises; protecting rescuers and volunteers; responding with medical aid; justifying necessary but disruptive protective actions; transiting from an emergency to an existing situation; rehabilitating evacuated areas; restricting individual doses of members of the public; caring for infants and children; categorising public exposures due to an accident; considering pregnant women and their foetuses and embryos; monitoring public protection; dealing with 'contamination' of territories, rubble and residues and consumer products; recognising the importance of psychological consequences; and fostering the sharing of information. Relevant ICRP Recommendations were scrutinised, lessons were collected and suggestions were compiled. It was concluded that the radiological protection community has an ethical duty to learn from the lessons of Fukushima and resolve any identified challenges. Before another large accident occurs, it should be ensured that inter alia: radiation risk coefficients of potential health effects are properly interpreted; the limitations of epidemiological studies for attributing radiation effects following low exposures are understood; any confusion on protection quantities and units is resolved; the potential hazard from the intake of radionuclides into the body is elucidated; rescuers and volunteers are protected with an ad hoc system; clear recommendations on crisis management and medical care and on recovery and rehabilitation are available; recommendations on public protection levels (including infant, children and pregnant women and their expected offspring) and associated issues are consistent and understandable; updated recommendations on public monitoring policy are available; acceptable (or tolerable) 'contamination' levels are clearly stated and defined; strategies for mitigating the serious psychological consequences arising from radiological accidents are sought; and, last but not least, failures in fostering information sharing on radiological protection policy after an accident need to be addressed with recommendations to minimise such lapses in communication.

Effective dose conversion coefficients for radionuclides exponentially distributed in the ground.

In order to provide fundamental data required for dose evaluation due to environmental exposures, effective dose conversion coefficients, that is, the effective dose rate per unit activity per unit area, were calculated for a number of potentially important radionuclides, assuming an exponential distribution in ground, over a wide range of relaxation depths. The conversion coefficients were calculated for adults and a new-born baby on the basis of dosimetric methods that the authors and related researchers have previously developed, using Monte Carlo simulations and anthropomorphic computational phantoms. The differences in effective dose conversion coefficients due to body size between the adult and baby phantoms were found to lie within 50 %, for most cases; however, for some low energies, differences could amount to a factor of 3. The effective dose per unit source intensity per area was found to decrease by a factor of 2-5, for increasing relaxation depths from 0 to 5 g/cm(2), above a source energy of 50 keV. It is also shown that implementation of the calculated coefficients into the computation of the tissue weighting factors and the adult reference computational phantoms of ICRP Publication 103 does not significantly influence the effective dose conversion coefficients of the environment. Consequently, the coefficients shown in this paper could be applied for the evaluation of effective doses, as defined according to both recommendations of ICRP Publications 103 and 60.

Distribution of thermal neutron flux around a PET cyclotron.

The number of positron emission tomography (PET) examinations has greatly increased world-wide. Since positron emission nuclides for the PET examinations have short half-lives, they are mainly produced using on-site cyclotrons. During the production of the nuclides, significant quantities of neutrons are generated from the cyclotrons. Neutrons have potential to activate the materials around the cyclotrons and cause exposure to the staff. To investigate quantities and distribution of the thermal neutrons, thermal neutron fluxes were measured around a PET cyclotron in a laboratory associating with a hospital. The cyclotron accelerates protons up to 18 MeV, and the mean particle current is 20 µA. The neutron fluxes were measured during both 18F production and C production. Gold foils and thermoluminescent dosimeter (TLD) badges were used to measure the neutron fluxes. The neutron fluxes in the target box averaged 9.3 × 10(6) cm(-2) s(-1) and 1.7 × 10(6) cm(-2) s(-1) during 18F and 11C production, respectively. Those in the cyclotron room averaged 4.1 × 10(5) cm(-2) s(-1) and 1.2 × 10(5) cm(-2) s(-1), respectively. Those outside the concrete wall shielding were estimated as being equal to or less than ∼3 cm s, which corresponded to 0.1 µSv h(-1) in effective dose. The neutron fluxes outside the concrete shielding were confirmed to be quite low compared to the legal limit.

Implementation of the NCRP wound model for interpretation of bioassay data for intake of radionuclides through contaminated wounds.

Emergency response preparedness for radiological accidents involving wound contamination has become more important, considering the current extending tendency in the nuclear industry related to the nuclear fuel cycle. The US National Council on Radiation Protection and Measurements (NCRP) proposed a biokinetic and dosimetric model for the intake of radionuclides through contaminated wounds in 2007. The present paper describes the implementation of this NCRP wound model for the prediction of systemic behaviour of some important radioactive elements encountered in workplaces related to the nuclear industry. The NCRP wound model was linked to the current ICRP systemic model at each blood compartment and simultaneous differential equations for the content of radioactivity in each compartment and excreta were solved with the Runge-Kutta method. The results of the calculation of wound, whole-body or specific organ retention and daily urinary or faecal excretion rate of some selected elements will be useful for the interpretation of bioassay data and dose assessment for cases of wound contamination.

Reconstruction of two victims' posturing based on the induced radioactivities in their bones in the criticality accident in Tokai-Mura, Japan.

On 30 September 1999, three workers were severely exposed to neutrons and gamma rays in a criticality accident that occurred at a uranium conversion facility in Tokai-mura, Ibaraki Prefecture, Japan. Radiochemical analyses of 32P and 45Ca induced by neutrons in bone matrix were carried out after the deaths of two of the victims. It was found that more than several million becquerels of both nuclides had been produced in their body skeletons. Results showed non-homogeneous distributions of neutron fluence in the bodies, from which it could be deduced how both workers were positioned relative to the fission source during exposure, i.e., at the moment of the first nuclear excursion. For the victim who died first, the activities in the central part of his body were more than those of his extremities. Also, in the central part of his body, the right side showed more activities than the left side. As for the second man, the activities indicated rather uniform exposure to neutrons to the whole body although the geometrical distribution of the activity varied enough to assume his orientation. Such information on the geometrical distribution of neutron-induced radioactivities in the skeleton can be used to reconstruct the posturing of the victims, which is necessary to estimate their apparent absorbed doses.