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放射線防護人材の不足が国内の政府機関でも議論されるようになって久しい。2018 年に行われた放射線防護に関連する学会の合同調査では、10 － 20 年前に比べ、放射線防護の専門家数は2 割減となっており、現リーダーと次世代リーダーの層が薄くなっていることが明らかとなった。 そこで、日本放射線安全管理学会、日本放射線影響学会、日本放射線事故・災害医学会、日本保健物理学会では、放射線防護関連人材の効果的な確保方策を検討するため、2019 年9 月1 日から10 月31 日の間、会員の専門分野や業務内容、バックグラウンドや就職の経緯等に関するWeb アンケートを合同で実施した。複数の学会に所属する会員は1 回限りの回答とし、371 名の回答が得られた。 調査の結果、放射線生物学（実験系）、放射線防護学・放射線安全管理学、放射線計測・線量評価を専門とする会員が多かった。原子力・放射線災害医療を専門とする会員は15％程度で、医師の比率が高い傾向が見られた。彼らの5 割は放射線研究歴がなく、典型的なキャリアとしては比較的多数の職場を経験した後、学生時代の研究とはほとんど関係のない教育や診療を主業務としていることが明らかになった。放射線防護や線量評価の専門家は多いものの、原子力・放射線災害医療も専門と回答した割合は1 割程度にとどまった。 調査の回答率から換算すると、緊急被ばく医療を専門とする医師および診療放射線技師有資格者は、放射線防護関連の学会内にそれぞれ60-70 名および20-30 名程度存在すると推定された。40 歳未満の会員の3 割が診療放射線技師有資格者であったことから、今後、災害時の放射線防護や線量評価を担う診療放射線技師が増えることが期待される。一方、緊急被ばく医療を専門とする医師の確保は、放射線防護関連の学術コミュニティ内だけでは難しく、放射線科や救急・災害医療関連の診療科からの参入が必要と思われる。

A high-efficiency proton recoil telescope was developed to determine neutron fluences in neutron fields using the 3H(d,n)4He reaction. A 2-mm thick plastic scintillation detector was employed as a radiator to increase the detection efficiency and compensate for the energy loss of the recoil proton within. Two silicon detectors were employed as the ΔE and E detectors. The distance between the radiator and the E detector was varied between 50 and 150 mm. The telescope had detection efficiencies of 3.5 × 10-3 and 7.1 × 10-4 cm2 for distances of 50 and 100 mm, respectively, which were high enough to determine the neutron fluence in 14.8-MeV neutron fields, with a few thousand cm-2 s-1 fluence rate, within a few hours.

In this study, we developed a 45 MeV neutron fluence rate standard of Japan. Quasi-monoenergetic neutrons with a peak energy of 45 MeV in the neutron standard field were produced by the 7Li(p,n)7Be...

Background: The moderated neutron calibration fields using 241Am-Be sources and a graphite moderator have been constructed at the Facility of Radiation Standard (FRS) in the Japan Atomic Energy Agency (JAEA). Materials and Methods: The neutron spectra of the fields were evaluated by the Monte-Carlo calculations and measurements using the Bonner Multi-sphere Spectrometer. Results and Discussion: The fields have continuous neutron spectra from several MeV to thermal neutron energy, with fluence-averaged energies of 0.84 MeV and 0.60 MeV. Reference values of fluence rates and ambient/personal dose equivalent rates were determined from neutron spectra by measurements. Conclusion: Currently, the fields are available for calibration or performance test of neutron measuring instruments.

A quasi-monoenergetic neutron field generated in the ${^7{\rm Li}({\rm p}, {\rm n})}$ reaction consists of a high-energy monoenergetic peak and a continuum to the low-energy region. In this study, the spectral fluence of the continuum was measured with the time-of-flight (TOF) method using a ${^6{\rm Li}}$ -glass scintillation detector and an organic liquid scintillation detector for the keV and MeV region, respectively. The neutron spectral fluence was determined down to the keV region by implementing a new beam chopping system and the results showed that the neutrons that came directly from the target had a lower energy limit about 100 keV. Discussions were made also on the effect of the time-independent neutrons which are assumed to be room-scattered neutrons. The obtained information is expected to contribute to understanding the quasi-monoenergetic high-energy neutron field and improvements of calibrating neutron detectors in the field.

The ambient dose equivalents H*(10) of photons mixed in the 144, 250 and 565 keV monoenergetic neutron fields were evaluated using measurements from an NaI(Tl) detector and calculations done using the MCNP-ANT code. It was found that H*(10) of the photons produced in the target assembly dominates the dose, particularly near the target. The H*(10) of the photons produced in other materials in the field increases with the increase in distance from the target and could not be neglected at a large distance from the target. The ratios of the H*(10) of the mixed photons to that of the monoenergetic neutrons for 144, 250 and 565 keV neutron fields, were evaluated to be below 5.5, 6.9 and 1.5 %, respectively. The ratios were calculated at calibration points between 100 and 500 cm from the target.

*A highly efficient silicon semiconductor detector with a polyethylene convertor (CH2-SSD) was developed and the neutron fluences were precisely evaluated with it at 1.2 and 2.5 MeV mono-energetic neutron calibration fields at the Facility of Radiation Standard (FRS) in the Japan Atomic Energy Agency (JAEA) using an accelerator. The CH2-SSD was designed to have a higher detection efficiency than that of the conventional one in order to measure the neutron fluencies at the calibration point of more than 100 cm from the target. The NRESP-ANT and PHITS codes were used to calculate the efficiency of the CH2-SSD and its uncertainties. The maximum neutron fluence rates at 100 cm from the target were evaluated to be 1.0×10 3 and 2.0×10 3 cm -2 s -1 in the 1.2 and 2.5 MeV fields, respectively. They are high enough to calibrate dosemeters

An attempt to decrease the anisotropic emissions of neutrons from a cylindrical (241)Am-Be-encapsulated X3 source was conducted with Monte Carlo calculations and experiments. The influence of metal materials and shapes of the external casing to the anisotropy factor were focussed on. Results obtained by calculations using MCNP4C implied that a light and spherical-shaped external casing decreases the anisotropic emission of neutrons. Experimental results using the spherical-shaped aluminium protection case also revealed that the anisotropy factor was close to 1.0 with wide zenith angle ranges.

A quasi-monoenergetic high-energy neutron field produced via 7Li(p,n) reaction consists of a high-energy peak and a continuum down to the low-energy region. The continuum is a background for the measurement for the peak neutrons. Spectral information is necessary to estimate the effect of it. Using a new beam chopping system, the spectral neutron fluence above ∼150 keV was measured by the TOF method. The lower-energy neutrons are time-independent and the unfolding method using the Bonner sphere spectrometer (BSS) is suitable to measure them. However, the common BSS with a 3He or BF3 proportional counter has difficulties in this case: signals caused by high-energy-charged particle are mixed in with the genuine signals caused by thermalized neutrons and the dead-time corrections are difficult in experiments with the pulsed-neutron beam. The gold foil is therefore adopted as a detector element for the BSS. Response matrix of the gold-activation BSS is evaluated by Monte Carlo calculations and response calibrations at a neutron standard facility of the AIST. Measurements for the quasi-monoenergetic high-energy neutron field was performed at the AVF cyclotron facility of TIARA, at the peak neutron energy of 45 MeV.

Development of three calibration fields for 45, 60 and 75 MeV neutrons has been in progress at Takasaki Ion Accelerators for Advanced Radiation Application (TIARA) of JAEA-Takasaki. Achievement of the neutron calibration fields requires establishment of both measurement of neutron fluence and its monitoring technique. In order to measure the neutron fluence of main energy peak especially at a reference point on low fluence condition, a proton recoil counter telescope with high efficiency has been developed. The design of relatively large telescope components brought a high detection efficiency at the reference point with wide irradiation area at a long distance. Effective and precise measurement of the neutron fluence could be performed with uncertainties less than 6.5 % (standard deviation). For detecting neutrons without scattering near the target, a transmission type neutron fluence monitor has been newly developed to measure the neutron fluence directly at the collimator exit. Its performance such as counting rates was tested for various beam intensities.

Target photons mixed in the 144, 250 and 565 keV mono-energetic neutron calibration fields were measured using a cylindrical NaI(Tl) detector with 7.62 cm both in diameter and in length. The ambient dose equivalent H*(10) of the photons was evaluated by applying the “G(E) function” to the measured pulse height spectrum. Neutrons induce photons by nuclear reactions in the NaI(Tl) detector and affect the pulse height spectrum. In order to eliminate the influence of these neutron events, the time-of-flight technique was applied with operating the accelerator in the pulse mode. The ratios by the ambient dose equivalent H*(10) of the photons to the 144, 250 and 565 keV neutrons were evaluated to be 3.3%, 4.7% and 0.9%, respectively. Although high energy photons ranging from 6 to 7 MeV are emitted by the 19F(p,αγ)16O reactions, the dose of the target photons is low enough to calibrate neutron dosemeters except for ones with high sensitivity to the photons.

For individual monitoring for external exposure of workers, electronic dosimeters have been widely used as supplementary dosimeters in nuclear facilities. In this study, an electronic pocket dosimeter for measuring the Hp(10) and Hp(0.07), which was recently developed by Aloka Co., Ltd., has been examined in terms of (1) energy dependences for gamma- and beta-rays and (2) applicability in actual radiation workplaces. For the latter purpose, performance tests were carried out in hot cells of the Reactor Fuel Examination Facility in the Japan Atomic Energy Agency. From the examinations, the followings were found; (1) the energy responses vary within ±20 % in the intended energy range for both radiations. (2) At the actual radiation workplaces, the readings of the dosimeter tend to show slightly higher Hp(10) and lower Hp(0.07) than those of other types of dosimeters, i.e. glass dosimeters and thermoluminescence dosimeters.

An inter-comparison of the neutron calibration field of the D2O moderated 252Cf source was carried out at JAEA and KAERI. Measuring the neutron spectra at KAERI by means of the method employed at JAEA, we attempted to evaluate the basic parameters for the D2O moderated 252Cf source and validate our method for determination of these parameters. We obtained the differences in the spectrum at each facility and theoretically clarified the effect of the geometries of the source assembly and the dimensions of the irradiation room on the neutron standard field.

The neutron calibration fields using accelerators have been developed at the Facility of Radiation Standards (FRS) of JAEA-Tokai for neutrons below 20 MeV and at Takasaki Ion Accelerators for Advanced Radiation Application (TIARA) of JAEA-Takasaki for those over 20 MeV. At FRS, monoenergetic neutrons are generated by using proton or deuteron beam from a 4 MV Van-de-Graaff (Pelletron) accelerator. Up until the present, developments of the 8, 144, 250, 565 keV, 5.0 and 14.8 MeV fields have been completed. As procedure for the development, measurement of neutron energy spectrum, peak energy and reference fluence, development of monitoring scheme for neutron fluence and establishment of the traceability on neutron fluence to the primary standards were achieved. At TIARA, three neutron fields of 45, 60 and 75 MeV are planned to be established. Quasi-monoenergetic neutrons are generated from 7Li(p,n)7Be reaction by using proton beam from an AVF cyclotron, and led into an irradiation room through a cylindrical c...

241Am-Be中性子線源を用いる中性子校正場は，中性子線量計の校正のために一般に広く用いられている。中性子線量計を適切に校正するためには，中性子校正場に関わる個々のパラメータを正確に決定する必要があるが，その中でも，校正点における中性子フルエンス率は重要なパラメータの一つである。中性子フルエンス率は線源中心から校正点までの距離と線源強度で決定されるが，それ以外に補正項を導入する必要がある。中性子は線源内で等方的に放出されるが，線源物質や線源を封じる壁材料との散乱は不可避であり，この効果によって中性子フルエンスの角度分布に変動が生じる。これを中性子放出の非等方性と呼び，その度合いを非等方係数FI(θ)で評価する。著者らは，(独)日本原子力研究開発機構 東海研究開発センター 原子力科学研究所 放射線標準施設棟で運用されている保護ケース付き241Am-Be中性子線源について，高精度ロングカウンタを用いた中性子放出の角度分布測定を行い，保護ケース付き241Am-Be中性子線源からの中性子放出の非等方係数を決定した。また，手法の妥当性確認のために実施した，ロングカウンタを用いたX3型カプセル外側からの中性子放出の角度分布の測定結果と他の研究者の測定結果及びMCNPによる計算シミュレーションの結果との比較についても報告する。

The 8 and 27 keV monoenergetic neutron calibration fields have been developed by using (45)Sc(p, n)(45)Ti reaction. Protons from a 4-MV Pelletron accelerator are used to bombard a thin scandium target evaporated onto a platinum disc. The proton energies are finely adjusted to the resonance to generate the 8 and 27 keV neutrons by applying a high voltage to the target assemblies. The neutron energies were measured using the time-of-flight method with a lithium glass scintillation detector. The neutron fluences at a calibration point located at 50 cm from the target were evaluated using Bonner spheres. A long counter was placed at 2.2 m from the target and at 60 degrees to the direction of the proton beam in order to monitor the fluence at the calibration point. Fluence and dose equivalent rates at the calibration point are sufficient to calibrate many types of the neutron survey metres.

The ambient/personal dose equivalent per fluence for D(2)O moderated (252)Cf neutron source was determined by measurement. An appropriate subtraction of the scattered neutrons is required for the accurate measurement of direct neutrons. A cubic shadow object was used for the subtraction of the scattered neutrons from the surroundings. The scattered neutrons to be subtracted vary with the position of the shadow object due to the large volume of the source. Using the Monte Carlo code MCNP-4C, the optimum positions of the shadow object were surveyed for subtracting the scattered neutrons. The energy spectra of direct neutrons were measured in the optimum position. The dosimetric parameters for the D(2)O moderated (252)Cf neutron source were reasonable, taking into account the uncertainties of the parameters.

The spectral neutron fluence was measured by a combination of the TOF method (above keV region) and the Bonner unfolding method (whole energy region), for a 45-MeV quasi-monoenergetic high-energy neutron field. The extrapolated TOF-based spectrum information was used as a default spectrum for the Bonner unfolding method. The Bonner measurements were performed using the 3He-counter based Bonner sphere spectrometer (BSS) and the gold-activation BSS. While both unfolding reproduced reasonable solutions, difference in amount of the low-energy neutrons was observed between the types of BSS.

Several intercomparison exercises were organised by the International Atomic Energy Agency (IAEA) on the determination of operational quantities at the regional or interregional basis. In East Asia region, a third phase of the intercomparison finished in mid 2004. It was organised within the frame of the Regional Cooperation Agreement (RCA) as a follow-up to previous exercises carried out during 1990-1992 and 1995-1996. The results of this intercomparison for the determination of operational quantities were satisfactory for all Member States. The laboratories demonstrated a good performance in quantities tested. The purpose of this paper is to present the results of the RCA/IAEA intercomparison and the future of RCA activities in support of assessment of occupational exposure by organising intercomparison runs.

A phoswich detector composed of a thin plate CsI(Tl) scintillator and a plastic scintillator (BC-400) has been designed and evaluated in order to improve the sensitivity in the low-energy region of a large-area plastic scintillation detector. This newly designed phoswich detector can be applied to both gross gamma measurement and energy spectrometry for low-energy gamma-ray emitters. Judging by estimations of minimum detectable activity, the lower measurable energy of a large-area plastic scintillation detector can be expanded down to a few tens of keV by adding a thin plate CsI(Tl) scintillator.

The moderator structure for a neutron spectrometer was optimized with the Monte Carlo code MCNP-4B. The spectrometer consists of a cylindrical moderator and a position-sensitive thermal neutron detector and obtains energy spectra from thermal neutron distribution along its cylindrical axis. The structure of the moderator was improved by using a low hydrogen density material on one end and a high hydrogen density on the other, and inserting a neutron absorber that eliminates thermal neutron diffusion. This design improves the energy resolution of the spectrometer, especially for low-energy neutrons from a few to 100 keV. The designed spectrometer can be applied to the measurement of energy spectra over a neutron energy range from a few keV to 20 MeV.

Energy responses of three types of Japanese neutron dose-equivalent (DE) meters were evaluated by Monte Carlo simulations and measurements. The energy responses were evaluated for thermal neutrons, monoenergetic neutrons with energies up to 15.2 MeV, and also for neutrons from such radionuclide sources as 252Cf and 241Am–Be. The calculated results were corroborated with the measured ones. The angular dependence of the response and the DE response were also evaluated. As a result, reliable energy responses were obtained by careful simulations of the proportional counter, moderator and absorber of the DE meters. Furthermore, the relationship between pressure of counting gas and response of the DE meter was discussed. By using the obtained responses, relations between predicted readings of the DE meters and true DE values were studied for various workplace spectra.

A Monte Carlo code system, MCNP-ANT, has been developed to simulate the transport of charged particles (

Dose conversion coefficients from neutron fluence to operational quantities have been changed with the revision of radiation protection ordinances in Japan in April 2001. The influences of the revision were examined for the popularly used neutron dosemeters. It was shown, as a result, that the relative energy response to dose equivalent was not so changed by the revision.In addition, various dosemeters were examined on the ratios between the reading and true dose equivalent, when they were used in the various neutron fields. The percentages of the cases whose ratios were within 1-2 were 23-77% and 9-82% in 180 fields, for the representative rem counters and personal dosemeters, respectively, when the dosemeters were calibrated with 252Cf source.

A portable, light-weight long counter (LC) with small dimensions was developed. This LC consists of a (3)He thermal neutron counter, a cylindrical moderator and outer shields. It was designed to have an almost flat response in a neutron energy range of 0.4 eV to 5 MeV. The portable LC has a radius of 11 cm and a length of 39 cm. Its weight was successfully reduced to 15 kg. Polystyrene was employed instead of polyethylene for the front part of the moderator in order to increase the sensitivity to low-energy neutrons. The response function calculated using the MCNP code was consistent with the results of experiments using monoenergetic neutron calibration fields.

In general, radioactivity measurements for volume samples are performed by the γ-ray spectrometry method with germanium detectors. In the method, peak efficiency curves must be obtained for each kind of sample in advance. For the reason, usually, many standard volume sources have been made which have different shapes, densities, matrix compositions and so on. This paper describes about a simple method for determination of γ-ray efficiency curves by the combination of Monte Carlo simulations and a single-point calibration using a standard point source. This method enables us to determine the efficiency curves for various types of samples and detectors easily and precisely.

Counting efficiencies of a whole-body counter installed in the Japan Atomic Energy Research Institute were evaluated using Monte Carlo simulation and measurement to validate the calculation method of the counting efficiency curves of the whole-body counter for the water-filled block-shape phantoms containing radionuclides. Consequently, the counting efficiency curves in the photon energy range of 300–1,461 keV were obtained using the Monte Carlo simulation. The calculations are in fairly good agreement with the measurements in the photon energy range of 662–1,461 keV. It was also found that the self-scattering effect of photons in the phantoms accounts for about 7% of the counting efficiency for 662 keV photons.

Backscattered photons from slab phantoms were analyzed to investigate the effects of phantom size and material on calibration of dosimeters. The backscatter factor estimated by measurement with an ionization chamber was compared for three different phantoms ; a 40× 40×15 cm3 polymethyl methacrylate (PMMA) phantom, a 30×30×15 cm3 PMMA phantom and a water-filled phantom recommended by the International Organization for Standardization (ISO) . Monte Carlo calculations using EGS4 code were also performed to examine the measured results and to analyze the effect of phantom size and material on the dosimeter calibration. It was found that the ISO phantom could give up to 10% higher value of calibration factor than the 40×40×15 cm3 PMMA phantom, now commonly used in Japan, for photons of 60 keV.

A spherical neutron counter for spectroscopy and dosimetry has been developed. The counter consists of a spherical polyethylene moderator and three slender /sup 3/He position-sensitive proportional counters. Rough measurements of neutron fluence spectra and ambient dose equivalents, H*(10), have been done in the moderated-neutron field of JAERI. Characteristics of the counter are examined and compared with those of a Bonner sphere neutron spectrometer.

A quasi-monoenergetic neutron field generated in the 7Li(p,n) reaction consists of a high-energy monoenergetic peak and a continuum to the low-energy region. The continuum is a background for the response measurements of neutron detectors for the high-energy-peak neutrons and the effect of the continuum should be properly considered. A spectral fluence of the continuum is typically measured by time-of-flight (TOF) methods with an organic liquid scintillation detector. The lowest energy of this method is limited to several MeV because of a high repetition rate of proton beams generated by the AVF cyclotron. Recently, a new beam chopping system available at the TIARA AVF cyclotron facility achieved 1/80 reduction of the beam repetition rate and the lower limit energy of the TOF method has decreased to 20 keV. A Li-glass scintillation detector was also used for measurements of keV region, which is below the working range of the organic liquid scintillation detector. The spectral fluence of the quasi-monoenergetic high-energy neutron field measured by the TOF method using the two scintillation detectors is reported in this presentation.

A spherical neutron counter for spectrometry and dosimetry has been developed. The counter consists of a spherical polyethylene moderator and three slender 3He position-sensitive proportional counters inserted into the moderator. The thermal neutron distribution in the moderator, which is a function of the energy of the incident neutrons is obtained with the proportional counters. Characteristics of the counter have been studied with experiments and Monte Carlo simulations. Also shown are energy spectra of a 252Cf neutron source that are obtained by an unfolding method using the thermal neutron distributions as response functions.

Glow curves of LiF TLDs (TLD-100) were measured for 10-40 keV monoenergetic photons from synchrotron radiation, and the energy and dose responses were evaluated based on the glow curve data. The ratio of measured to calculated energy response was almost unity for the lower temperature region of the glow curve including the main peak, and was about 2.0 for the higher one. Dose responses to 10 keV 30 keV X rays, and 60 Co gamma rays showed different supralinearity. The maximum value of supralinearity decreased with decreasing photon energy, that is, with increasing the average linear energy transfer of incident photons

A parallel plate free-air ionization chamber was filled with air, nitrogen and argon, and the ionization current was measured for 30-keV monoenergetic incident photons, using a synchrotron radiation source. The photon intensity was measured using a total absorption calorimeter placed behind the chamber. Because the W value for each gas is accurately known, photon mass energy-absorption coefficients for each gas were measured. The measured coefficients are in agreement with published tables by Hubbell within 2%.

A simplified method in which the equivalent doses in tissue or organs and effective dose for external photon exposure according to the ICRP 1990 Recommendations could be approximated by a single empirical expression is proposed. In this method, the position of an organ is assumed to be the main factor that determine its dose for external photon irradiation. Many organ depths with different body size for ages 1, 5, 10 years and adults were determined using the cross-sections of mathematical organ descriptions, and effective depth for approximating effective dose was obtained with the organ depths assigned to tissue weighting factors.Compared with the precised data for adults by the Monte Carlo method, the simple expression gives good approximates for most of deep organs and effective dose within about 20% except some organs, especially for lateral geometry. It was found that irradiation geometry affects the variations in age-dependent effective dose, and that the effective depths for ages 1, 5, 10 years, normalized to the adults depth, are directly proportional to the cubic root of the body weight. The usefulness of the concept of effective depth was also discussed with regard to the specified point of effective does in the body, obtained from the combination of the depths for the three typical geometries.

A total absorption calorimeter was developed to measure the absolute intensity of monoenergetic X-ray beams from 10 keV to 70 keV of synchrotron radiation. Experiments with synchrotron radiation have demonstrated that the heat power above about 1 μW due to monoenergetic synchrotron radiation is measured with an accuracy of about 1%.