Norio Tsujimura

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.

Development of a hand-held fast neutron survey meter

A neutron survey meter with a ZnS(Ag) scintillator to measure recoil protons was built. The detection probe weighs ~2 kg, therefore providing us with true portability. Performance tests exhibited satisfactory neutron dosimetry characteristics in unmoderated or lightly moderated fission neutron fields and in particular work environments at a mixed oxide fuel facility. This new survey meter will augment a routine of neutron monitoring that is inconveniently being carried out by moderator-based neutron survey meters.

Energy and Angular Responses of the Criticality Accident Detector using a Plastic Scintillator

The Japan Atomic Energy Agency (JAEA), Nuclear Fuel Cycle Engineering Laboratories, operates a spent fuel reprocessing plant and MOX (Plutonium-Uranium Mixed Oxide) fuel fabrication plants. Criticality accident detectors have been installed in these facilities. The detector, the Toshiba RD120, is composed of a plastic scintillator coupled to a photomultiplier tube, and an operational amplifier. The alarm triggering point is set to 1.0–3.6 mGy.h−1 in photon dose rate to detect the minimum accident of concern. However, a plastic scintillator is principally sensitive not only to primary photons but also to neutrons by secondary photons and heavy charged particles produced in the detector itself. The authors calculated energy and angular responses of the RD120 criticality accident detector to photons and neutrons using Monte Carlo computer codes. The response to primary photons was evaluated with the MCNP-4B and EGS4 calculations, and photon and X-ray irradiation experiments. The response to neutrons that produce secondary photons and heavy charged particles from neutron interactions was computed using the MCNP-4B and SCINFUL, respectively. As a result, reliable response functions were obtained. These results will be a great help in reassessing the coverage area and in determining the appropriate triggering dose rate level in criticality accidents.

Installation Places of Criticality Accident Detectors in the Plutonium Conversion Development Facility

At the Plutonium Conversion Development Facility (PCDF) in the Nuclear Fuel Cycle Engineering Laboratories, the co-conversion technologies to purify the mixed plutonium and uranium nitrate solution discharged from a reprocessing plant have been developed. The probability of a criticality accident in PCDF is extremely low. However, the criticality accident alarm system (CAAS) has been in place since 1982 to reduce the radiation dose to workers in case of such a rare criticality accident. The CAAS contains criticality accident detector units (CADs), one unit consisting of three plastic scintillation detectors, and using the 2 out of 3 voting system for the purpose of high reliability. Currently, eight CADs are installed in PCDF evaluating the dose using a simple equation allowing for a safety margin. The purpose of this study is to show the determination procedures for the adequate relocation of the CADs which adequately ensures safety in PCDF.

Current situations and discussions in Japan in relation to the new occupational equivalent dose limit for the lens of the eye

Since the International Commission on Radiological Protection recommended reducing the occupational equivalent dose limit for the lens of the eye in 2011, there have been extensive discussions in various countries. This paper reviews the current situation in radiation protection of the ocular lens and the discussions on the potential impact of the new lens dose limit in Japan. Topics include historical changes to the lens dose limit, the current situation with occupational lens exposures (e.g., in medical workers, nuclear workers, and Fukushima nuclear power plant workers) and measurements, and the current status of biological studies and epidemiological studies on radiation cataracts. Our focus is on the situation in Japan, but we believe such information sharing will be useful in many other countries.

Prediction analysis of dose equivalent responses of neutron dosemeters used at a MOX fuel facility

To predict how accurately neutron dosemeters can measure the neutron dose equivalent (rate) in MOX fuel fabrication facility work environments, the dose equivalent responses of neutron dosemeters were calculated by the spectral folding method. The dosemeters selected included two types of personal dosemeter, namely a thermoluminescent albedo neutron dosemeter and an electronic neutron dosemeter, three moderator-based neutron survey meters, and one special instrument called an H(p)(10) monitor. The calculations revealed the energy dependences of the responses expected within the entire range of neutron spectral variations observed in neutron fields at workplaces.

Verification analysis of thermoluminescent albedo neutron dosimetry at MOX fuel facilities

Radiation workers engaging in the fabrication of MOX fuels at the Japan Atomic Energy Agency-Nuclear Fuel Cycle Engineering Laboratories are exposed to neutrons. Accordingly, thermoluminescent albedo dosemeters (TLADs) are used for individual neutron dosimetry. Because dose estimation using TLADs is susceptible to variation of the neutron energy spectrum, the authors have provided TLADs incorporating solid-state nuclear tracks detectors (SSNTDs) to selected workers who are routinely exposed to neutrons and have continued analysis of the relationship between the SSNTD and the TLAD (T/R(f)) over the past 6 y from 2004 to 2009. Consequently, the T/R(f) value in each year was less than the data during 1991-1993, although the neutron spectra had not changed since then. This decrease of the T/R(f) implies that the ratio of operation time nearby gloveboxes and the total work time has decreased.

Recalibration of indium foil for personnel screening in criticality accidents

At the Nuclear Fuel Cycle Engineering Laboratories of the Japan Atomic Energy Agency (JAEA), small pieces of indium foil incorporated into personal dosemeters have been used for personnel screening in criticality accidents. Irradiation tests of the badges were performed using the SILENE reactor to verify the calibration of the indium activation that had been made in the 1980s and to recalibrate them for simulated criticalities that would be the most likely to occur in the solution process line. In addition, Monte Carlo calculations of the indium activation using the badge model were also made to complement the spectral dependence. The results lead to a screening level of 15 kcpm being determined that corresponds to a total dose of 0.25 Gy, which is also applicable in posterior-anterior exposure. The recalibration based on the latest study will provide a sounder basis for the screening procedure in the event of a criticality accident.

Influence of the PMMA Slab and ISO Water Phantom in Calibrating Personal Dosimeter in the Energy Range of 36-662 keV

Personnel dosimeter calibration for the determination of operational quantities used in individual monitoring requires placement on a phantom that provides a reasonable approximation to the backscatter properties of the part of the body on which it is worn. The personal dose equivalent Hp(d) is defined in the human body which is not a measurable quantity. The reference International Commission on Radiation Units and Measurements (ICRU) tissue is not readily available; hence a phantom of alternative must be used for calibration. The well recognized polymethyl-methacrylate (PMMA) slab phantom of size 30 × 30 × 15 cm3 are still being used in calibrating personal dosimeters. The International Organization for Standardization (ISO)-4037(3) has proposed another phantom of the same size as ICRU, which is named as ISO water phantom. In the present study, calibrated X-ray fields are characterized for National Institute of Advanced Industrial Science and Technology (AIST) middle beam spectrum series by which the influence of these proposed phantoms has been studied by the thermoluminescence dosimetry (TLD) technique. Discrepancies of responses within 8% have been observed at lower energies up to 42 keV. There is a very good agreement in responses is observed for ISO water slab and PMMA slab phantom for photon energies above 42keV. Experimentally obtained correction factors for AIST radiation qualities of M50 and M60 for PMMA slab to ISO water phantom is proposed. Measured backscatter factor results from PMMA to ISO water-filled phantoms over the photon energy range 36 to 662 keV are reported. This shows that the new ISO water phantom is a better substitute of ICRU tissue phantom than the PMMA slab.

Anisotropic Fluence Distribution in 252Cf Neutron Calibration Field Due to Scattering from Support Structure

中性子測定器の校正に使用される252Cf標準中性子線源は, ステンレス鋼の二重構造からなるカプセルに252Cfが封じられている。252Cfそのものから放出される中性子のフルエンス分布は等方であるが, 実際には, 線源カプセルのほか, 線源ハンドリング用の金属製ホルダ等との散乱によって, その分布に歪みが生じる。そこで, 核燃料サイクル開発機構東海事業所計測機器校正施設の252Cf中性子校正場について, 線源カプセル (X1カプセル) の構造, 線源ホルダ及び照射装置等の線源周辺構造を模擬したモンテカルロ計算によって, この影響を評価し, 検証実験により確認した。その結果, 中性子測定器の設置方向における非等方性を1.11±2% (k=2) と算出した。また, 散乱による252Cf自発核分裂中性子スペクトルの変化によるフルエンス-周辺線量当量換算係数への影響も評価し, 国際規格ISO8529-3の勧告値との違いについて議論した。