Development of a new integrated IN-VIVO counting system at the QST.

A new in-vivo counting system that functions as both a whole-body counter (WBC) and a lung counter (LC) was developed at the QST to enhance its dose assessment capability. This paper presents an overview of this system and the results of its performance tests. For use of the system as a WBC, three high purity germanium (HPGe) detectors installed in a 20-cm-thick iron shielding chamber are linearly arrayed over a subject lying on the bed, whereas two of the three HPGe detectors are placed over the subject's chest from side to side when using the system as an LC. The new in-vivo system was calibrated using three de-facto phantoms owned by the QST: an adult-male BOttle Manikin ABsorption (BOMAB) phantom, a Lawrence Livermore National Laboratory (LLNL) phantom and a Japan Atomic Energy Research Institute (JAERI) phantom. Monte Carlo simulations were also performed to determine an optimum location for the three detector array in the WBC mode and revealed that the peak efficiency for the BOMAB phantom (662 keV) was little varied as long as the middle detector was placed above the thorax and abdomen parts of the phantom. The calculated peak efficiencies agreed well with the observed peak efficiencies for photons with energies over 100 keV. For lung counting, a tentative Minimum Detectable Activity of 241Am was evaluated as 9.5 Bq for a counting time of 30 minutes, and a Japanese male subject with an average chest wall thinness (2.27 cm). The developed system is now ready for use.

The new QST bioassay laboratory in Chiba, Japan.

Japan's National Institutes for Quantum Science and Technology (QST) was designated as the core radiation emergency medical support center by the country's Nuclear Regulation Authority (NRA) in 2019. One of the main missions of the QST is to maintain and improve its dose assessment capability for radiation-exposed individuals. Toward the goal of effectively fulfilling this mission, a new facility-the Dose Assessment Building for Advanced Radiation Emergency Medicine-was constructed at the Chiba base of the QST in 2020. An integrated bioassay laboratory was installed in this facility for assessing subjects' internal doses, along with a new integrated in vivo counter. The bioassay capability of the new laboratory is currently expected to screen 5-10 persons simultaneously assuming internal contamination with actinides such as Pu, Am/Cm and U, although this is dependent on the specific contamination circumstances.

Rapid measurement of actinides in urine by mass and alpha spectrometric methods.

To provide timely information for prompting decision-making in emergency radiation therapy, we developed simple and rapid mass and alpha spectrometric methods for urinary bioassays to determine ultra-trace actinide isotopes. For the mass spectrometric method, after organic matter decomposition, LaF3/CaF2 co-precipitation and chromatographic purification using 2 ml of AG MP-1 M anion exchange resin, U and Pu isotopes were measured in a 20-ml urine sample by inductively coupled plasma-mass spectrometry. In the alpha spectrometric method, after organic matter decomposition, iron hydroxide co-precipitation and chromatographic purification using 2 ml of TEVA and 2 ml of DGA resin cartridges, Pu, U and Am/Cm isotopes were measured in a 500-ml urine sample by alpha spectrometry. These alpha and mass spectrometric methods were then applied for participation in the 2020 intercomparison organized by the Association for the PROmotion of Quality COntrol in RADiotoxicological Analysis (PROCORAD), France, for method validations.

Occupational radiation dose on the hand of assisting medical staff in diagnostic CT scans.

Chronic radiation exposure increases the risk of skin damage of medical personnel engaged in radiology. However, hand dose measurements in computed tomography (CT) for diagnostic purposes have not been evaluated. The occupational radiation dose to the hands of CT assistants was herein investigated to evaluate its compliance with the equivalent dose limit for the hand (500 mSv/year). The occupational doses of nine CT assistants were measured in 89 cases (April 2017-May 2018) by installing radio-photoluminescence glass dosemeters (GD-302 M) (70-µm dose-equivalent conversion coefficient = 0.37) on the dorsal aspect of both hands. The occupational dose to the hand was the highest with head holding (right: 1.14 mSv/CT scan, left: 1.07 mSv/CT scan). Considering the results for annual work, even for head holding, the hand dose of the CT-assisting personnel was insignificant. However, CT assistants should be mindful of the possibility of locally higher doses to hands.

Estimation of inhomogeneous occupational exposure to the lens of the eyes and the extremities of radiation workers in a research accelerator facility.

We performed an experimental investigation on occupational exposure of the eye lens and the extremity of radiation workers engaged in handling of highly activated materials in a small research accelerator facility. Using a simplified physical phantom to simulate the relevant inhomogeneous radiation exposure situations, the personal dose equivalents obtained at the eye lens and the extremities of radiation workers handling heavily radioactive converters were measured together with the dose measured by personal dosemeters worn on their trunk. Results of mockup experiments and the Monte Carlo calculations suggest that the quantitative estimation of the eye lens doses can be estimated from the trunk dose, while the extremity doses vary considerably from the dose readings from the trunk, depending on the use of simple point-source or volume source geometry.

DOSE MEASUREMENT PRECISION OF AN RPLD-BASED EYE LENS DOSEMETER APPLICABLE TO THE MEDICAL SECTOR.

We demonstrate a practical calibration method and its applicability for a commercially available radiophotoluminescence dosemeter (RPLD), i.e. the GD-352M (AGC Techno Glass, Shizuoka, Japan) to eye lens dose monitoring, by performing the calibration according to the ISO recommendations. The calibration was then verified through a series of experiments. For verification of the derived calibration factor (1.21 ± 0.04, k = 1) of the RPLD, we performed standard irradiations in the ISO narrow series X-ray reference fields and the simulation measurements in the actual radiation fields in a hospital. The TLD-based commercially available dosemeters, DOSIRIS™ was also put on the ISO cylinder phantom and the RANDO phantom together with the GD-352M in the verification experiments. The personal dose equivalents Hp(3) obtained from the GD-352M and those obtained from the DOSIRIS™ were in good agreement with each other. Our results demonstrate the proper calibration of a commercially available RPLD that is applicable to the additional monitoring of the lens of the eyes for medical staff.

Estimation of the Thyroid Equivalent Doses to Residents in Areas Affected by the 2011 Fukushima Nuclear Disaster Due to Inhalation of 131I Based on Their Behavioral Data and the Latest Atmospheric Transport and Dispersion Model Simulation.

ABSTRACT: It has been challenging to obtain reliable estimates of thyroid equivalent doses (TEDs) to residents involved in the 11 March 2011 Fukushima Daiichi Nuclear Power Plant (FDNPP) accident because of the shortage of direct human measurements associated with 131I, the largest contributor to the dose of concern. The present study attempted to perform the estimation of the TEDs by inhalation of 131I to residents from Namie-town, one of the most radiologically-affected municipalities, by means of the latest atmospheric transport and dispersion model (ATDM) simulations with the Worldwide version of System for Prediction of Environmental Emergency Dose Information (WSPEEDI) ver. 2, coupled with personal behavioral data containing the history of the whereabouts of individuals shortly after the accident. We analyzed 1,637 residents who underwent direct measurements with whole-body counters several months after the accident and provided their personal behavioral data. We divided the subjects into two groups based on whether the distances between their locations and the FDNPP were >20 km as of 15:00 on 12 March in relation to the timepoint of the hydrogen explosion at the Unit 1 Reactor building. As a result, the 90th percentile TEDs of the 1,249 prompt evacuees and 388 late evacuees were 3.9 mSv (adult)-6.8 mSv (10-y-old) and 24.1 mSv (adult)-35.6 mSv (5-y-old), respectively, excluding 16 persons whose TEDs exceeded 50 mSv. The 90th percentile (median) TEDs to 1-y-old children (not included in the subjects) for the prompt and late evacuation groups were 8.1 (1.0) mSv and 36.3 (19.7) mSv, respectively. Additionally, this study provided the evidence to support the view that the explosive event at the Unit 1 Reactor building on the afternoon of 12 March 2011 could have caused the critical group among Namie-town's residents, whereas the largest release event on 15 March gave relatively small doses to the residents because their exposure took place mostly at sites that were distant from the FDNPP. However, the present dose estimation has potentially large uncertainty at the individual level; further validations are thus necessary.

PROBABILISTIC ANALYSES OF 131I THYROID ACTIVITY IN PERSONS WITH SURFACE CONTAMINATION IN DIRECT MEASUREMENT WITH A STANDING-TYPE WHOLE-BODY COUNTER AS AN EMERGENCY RESPONSE IN A RADIOLOGICAL INCIDENT.

The authors' previous study suggested that a simple standing-type whole-body counter called FASTSCAN (Canberra, Meriden, CT, USA), widely installed throughout Japan after the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident in March 2011, could be used for thyroid measurements in internal dose assessment after a future radiological incident accompanied by the release of 131I into the environment. The present study performed Monte-Carlo simulations using a computational human phantom with several patterns of body surface contamination and used the results to formulate a method for probabilistic analyses of 131I thyroid activity in persons with surface contamination. The major advantage of this method is that the upper limit of the thyroid activity can be determined from the relative frequency distribution without identifying where body surface contamination remains. Therefore, this method is especially effective for use in the early phase of a radiological incident when time and resources are limited, making it difficult to physically remove all body surface contamination for the purpose of gaining more accurate population-wide thyroid screening measurements. As a case study of the FDNPP accident, the proposed method was applied to the results of in vivo measurements for a male subject with body surface contamination. In comparing the probabilistic analyses before and after removal of the subject's contaminated work clothes, the uncertainty of the relative frequency distribution of 131I thyroid activity was reduced by their removal. Therefore, in terms of accurate estimates and avoidance of radiation exposure from their contaminated clothes, efforts to remove body surface contamination still should be made as much as possible in the chaotic situation of the early phase of a radiological incident.

Estimation of the Early Cs-137 Intake of Evacuees from Areas Affected by the 2011 Fukushima Daiichi Nuclear Power Plant Accident Based on Personal Behavioral Data and the Latest Atmospheric Transport and Dispersion Model Simulation.

ABSTRACT: More than 9 y have passed since the 2011 nuclear disaster in Fukushima Prefecture, Japan. During this period, much effort has been spent on the dose reconstruction for Fukushima residents; however, the estimation of the internal dose due to the potential intake of the short-lived radionuclides (mainly iodine-131) has been challenging because of the lack of direct human measurements at the early phase of the accident. Our previous study revealed that the residual cesium body contents observed in delayed whole-body counter (WBC) measurements of residents from Namie-town, one of the most affected municipalities, varied greatly with the timepoint of their evacuations on 12 March 2011 when the first explosive event occurred at the accident site; i.e., the late evacuees had much higher residual cesium body contents compared to the prompt evacuees. The present study thus aimed to clarify this finding by reproducing the exposure situation based on the evacuees' personal behavioral data in combination with the latest atmospheric transport and dispersion model (ATDM) simulation for 356 selected subjects in adult and 15-y (13-17 y) age groups. The results demonstrated that the ATDM simulation-based method could reasonably reproduce the subjects' exposure situation, supporting the previous finding. However, the residual cesium-137 body contents calculated by this method were only 10%-20% of those in the subjects' WBC measurements. This large discrepancy was considered to be caused by both the present method's underestimation and the overestimation of the subjects' early intake in the WBC measurements due to a conservative intake scenario not assuming potential additional intake. Additional studies are needed to further clarify the reasons for the discrepancy and to evaluate the magnitude of the inhalation dose in the accident.

Rapid analysis of 237Np and Pu isotopes in small volume urine by SF-ICP-MS and ICP-MS/MS.

Internal contamination with alpha-particle emitting actinides, such as 237Np, 239Pu, 240Pu, is likely to bring a large amount of dose to the tissues of persons even if the intake amount is small. To provide timely information for prompt decision-making in radiation emergency therapy, we developed a simple and rapid method for urinary bioassay to determine ultra-trace 237Np and Pu isotopes using SF-ICP-MS and ICP-MS/MS. To avoid polyatomic interferences and tailing effects from U, 237Np and Pu isotopes were collected after removing U effectively using a simple single chromatographic column packed with 2 mL AG MP-1M anion exchange resin, exhibiting a high decontamination factor of 108 for 238U. The overall chemical fractionation between 237Np and 242Pu for the whole analytical procedure was 0.974 ± 0.064 (k = 2), allowing us to measure 237Np and Pu isotopes using 242Pu as a yield tracer with yields of 76 ± 5%. Using ICP-MS/MS with low background provided the method detection limits for 237Np, 239Pu, 240Pu, and 241Pu of 0.025, 0.025, 0.015, and 0.020 fg mL-1, respectively, for 20 mL of urine sample. Those were comparable to detection limits of SF-ICP-MS with high sensitivity. Subsequently, three urine reference materials with Pu spike, provided by the Association for the PROmotion of Quality COntrol in RADiotoxicological Analysis (PROCORAD), France, were analyzed by the developed method and the conventional alpha spectrometry technique for validation. Finally, the developed method was successfully employed to measure the contamination level of 237Np, 239Pu, 240Pu, and 241Pu in urine samples collected during decorporation therapy using DTPA, after a Pu inhalation exposure accident in Japan. The high throughput (9 h for 12 samples), simplicity, low cost, and high sensitivity of the method will allow greater numbers of related laboratories to be involved in screening activities for unexpected actinide exposure, such as in the case of a large scale radiological disaster.

EXPERIMENTAL DETERMINATION OF ANISOTROPIC EMISSION OF NEUTRONS FROM 252CF NEUTRON SOURCE WITH THE SPHERICAL PROTECTION CASE.

The anisotropic emission of neutrons from a cylindrical X1 252Cf source with the spherical external casing was experimentally determined. The influence of metal materials and shapes of the external casing to the anisotropy factor, FI(θ), was assessed by the Monte Carlo calculation, before performing the measurement. The results of the calculation implied that light- and spherical-shaped external casing decreases the anisotropic emission of neutrons from a cylindrical source and the nature of the material does not affect the anisotropic emission to a large extent. The experimental results obtained when a spherical-shaped aluminum protection case was employed also revealed that the anisotropy factor was close to 1.0 with a wide zenith angle range. Considering the source handling and measures against mechanical impact to the source, we designed an SUS304-made spherical protection case for a renovated source delivering apparatus. With the SUS304-made spherical protection case, the measured anisotropy factor FI(90) was determined to be 1.002 ± 0.002 (k = 1). Results from the experiments also indicated that the measured anisotropy factor has a flat distribution from 55 to 125° with zenith angle.

APPLICABILITY OF A COMMERCIALLY AVAILABLE ACTIVE EXTREMITY DOSE-RATE METER TO EYE LENS DOSE MONITORING.

A commercially available active extremity dosemeter is a promising candidate for medical staff aid individual monitoring of the eye lens. We investigated the applicability of the newly developed active extremity dosemeter, which uses a low-energy photon probe, to eye lens dose monitoring by performing a complete characterization of the dosemeters. Performance tests revealed that the active extremity dosemeter would overestimate personal dose equivalent, Hp(3), when the probe is worn close to the lens of the eye of a medial worker without any improvement in the response. Introducing an appropriate filter into the probe for low-energy photon has proven to improve the response. The dosemeter then satisfies the criteria of the personal dosemeter for eye lens dosimetry and can be applicable to individual monitoring of eye lens dose. This article also discusses the applicability of the dosemeter to area monitoring for decision making regarding additional monitoring of the eye lens and the extremities.

ESTABLISHMENT OF A LOW DOSE RATE GAMMA RAY CALIBRATION FIELD FOR ENVIRONMENTAL RADIATION MONITORING DEVICES.

For routine calibration of dosemeters used for environmental radiation monitoring, a low dose rate 137Cs gamma ray calibration field that fully satisfies the requirement of the ISO 4037 series was established in the Facility of Radiation Standards in Japan Atomic Energy Agency. Two different methods were employed to determine the reference air kerma rate, namely a conventional ionisation chamber and a G(E) function method used a newly developed scintillation spectrometer. To fulfil the requirement of the ISO 4037 and suppress scattering of Cs gamma ray within the room as far as possible, a suitable lead collimator was introduced to limit the irradiation area at test points and placed at the middle height in an irradiation room with a grating floor. From measured results of de-convoluted photon fluence spectrum and the variation of evaluated reference air kerma rates between 1.0 m and 3.0 m from the centre of the source, gamma ray scattering from the room structures was found to be negligible. The reference air kerma rate at distances between1.0 m and 3.0 m could be then interpolated by simply considering the inverse square law of the distance and air attenuation. The resulting Cs gamma ray calibration field could provide ambient dose equivalent rates of 0.7-7.2 µSv h-1 for use with environmental radiation monitoring devices. Finally, we attempted to calibrate a widely used NaI(Tl) scintillation survey metre, obtaining a quite satisfactory calibration factor. These results also imply that such survey metres can be employed to monitor affected areas and assess the progress of decontamination, as they can provide appropriate measurements of the ambient dose equivalent rate.

Establishment of 6- to 7-MeV high-energy gamma-ray calibration fields produced using the 4-MV Van de Graaff accelerator at the Facility of Radiation Standards, Japan Atomic Energy Agency.

A 6- to 7-MeV high-energy gamma-ray field, produced by the nuclear reaction of (19)F(p, αγ)(16)O, has been established at the Facility of Radiation Standards (FRS) in Japan Atomic Energy Agency for calibration purposes. Basic dosimetric quantities (i.e. averaged gamma-ray energy, air-kerma-to-dose equivalent conversion coefficients and air kerma rates at the point of test) have been precisely determined through a series of measurements using the NaI(Tl) spectrometer and an ionisation chamber coupled with an appropriate build-up material. The measurements obtained comply with values recommended by the International Organization for Standardization for an 'R-F field'. The neutron contamination component for the field has also been measured by means of a conventional neutron dose equivalent meter (the so-called neutron rem-counter) and determined to be ∼ 0.5 % of the total dose equivalent.

Measurement of air kerma rates for 6- to 7-MeV high-energy gamma-ray field by ionisation chamber and build-up plate.

The 6- to 7-MeV high-energy gamma-ray calibration field by the (19)F(p, αγ)(16)O reaction is to be served at the Japan Atomic Energy Agency. For the determination of air kerma rates using an ionisation chamber in the 6- to 7-MeV high-energy gamma-ray field, the establishment of the charged particle equilibrium must be achieved during measurement. In addition to measurement of air kerma rates by the ionisation chamber with a thick build-up cap, measurement using the ionisation chamber and a build-up plate (BUP) was attempted, in order to directly determine air kerma rates under the condition of regular calibration for ordinary survey meters and personal dosemeters. Before measurements, Monte Carlo calculations were made to find the optimum arrangement of BUP in front of the ionisation chamber so that the charged particle equilibrium could be well established. Measured results imply that air kerma rates for the 6- to 7-MeV high-energy gamma-ray field could be directly determined under the appropriate condition using an ionisation chamber coupled with build-up materials.