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Background: For proper monitoring of the eye lens dose, an appropriate calibration factor of a dosimeter and information about the mean energies of X-rays are indispensable. The scattered X-ray energy spectra should be well characterized in medical practices where eye lenses of medical staffs might be high.Materials and Methods: Scattered X-ray energy spectra were experimentally derived for three different types of X-ray diagnostic and therapeutic equipment, i.e., the computed tomography (CT) scan, the angiography and the fluoroscopy. A commercially available CdZnTe (CZT) spectrometer with a lead collimator was employed for the measurement of scattered X-rays, which was performed in the usual manner.Results and Discussion: From the obtained energy spectra, the mean energies of the scattered X-rays lied between 40 and 60 keV. This also agreed with that obtained by the conventional half value layer method.Conclusion: The scattered X-rays to which medical workers may be exposed in the region around the eyes were characterized by means of spectrometry. The obtained mean energies of the scattered X-rays were found to match the flat region of the dosimeter response.

For proper environmental radiation monitoring, a method to simultaneously determine ambient dose equivalent rate and radioactivity concentration in the air by using a newly developed scintillation spectrometer, namely a CeBr3 spectrometer was investigated. The performance of the proposed method, including energy dependence, angular dependence and the linearity of the spectrometer (i.e. the dose rate dependence of its response), was verified by a series of measurements, conducted according to the procedure of inter-comparison of detectors used for early warning network performed by the European Radiation Dosimetry Group (EURADOS). Measurement results show that the proposed method is suitable for environmental radiation monitoring purposes. After thorough tests, the investigation on obtaining activity concentration in air from the pulse height spectrum of γ-ray was demonstrated in the laboratory by using a point-like sealed 133Ba source to simulate an artificial increase of ambient dose equivalent rate due to a radioactive cloud containing 131I and 133Xe. The photon fluence rate was obtained from the pulse height spectrum by using the unfolding method, and the activity concentration in air for radionuclides of interest could be estimated from the obtained photon fluence rate by applying the conversion coefficient evaluated via a Monte Carlo calculation.

In Japan, the radiation-dose limit for the lens of the eye was revised in April 2021. Consequently, for workers, the numerical values of the equivalent dose to the lens of the eye are equal to those of the effective dose. Radiation workers, radiation safety officers and licensees must comply with regulations related to radiation protection and optimize protection. The new guidelines on dose monitoring of the lens of the eye developed by the Japan Health Physics Society recommend for the dose to be estimated near the eye for accurate estimation, when the dose to the lens approaches or exceeds the management criteria. However, there is limited information regarding the non-uniform exposure of nuclear power plant workers. In this study, the dose equivalents of high-dose-rate workplaces and the personal doses of 88 workers were estimated at four Japanese commercial nuclear power plant sites (RWR: 3 units and BWR: 3 units) and the dose to the lens of the eye and the exposure situations of the workers were analyzed.

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.

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.

This article describes an attempt to estimate the background (BG) pulse height spectrum (PHS) of the CeBr3 scintillation spectrometer due to terrestrial gamma ray components. In radiological emergencies, there are some cases which the prior measurement of the background components can't be made. By subtracting the BG components from the measured gross PHS, timely and accurate information about the artificial radionuclides involved in the emergencies would be effectively drawn to protect the member of the public and first responders. A Monte Carlo calculation was conducted to simulate the PHS by gamma rays from natural radionuclides uniformly distributed in the ground. The calculated PHS were verified to be consistent with measured PHS. Derived dose equivalent rate from the calculated PHS were compared with reference value and found to be in good agreement with each other. The study revealed that applying the calculated PHS to the in-situ gamma ray spectrometry components hardly affected the determination of surface deposition density on soil for radioactive cesium.

In Japan, the possibility to change the current dose limit of the lens of the eye for the radiation workers working in the planned exposure situation (normal controlled situations) to a new ICRP dose limit was discussed. It was further discussed how to appropriately monitor and manage the equivalent dose of the eye lenses for these workers exposed to radiation at their workplaces, such as nuclear and medical facilities. Among the workers exposed to a high-dose radiation at the water storage flange tank deconstructed 90Sr/90Y dominant areas and the nuclear reactor buildings (high dose gamma-ray) of the Fukushima Daiichi Nuclear Power Plant (1F-NPP), Hp(10), Hp(3), and Hp(0.07) at the head and the chest (or the upper arm) were estimated by passive personal dosimeters using thermoluminescence dosimeters (TLDs) and radio photoluminescence glass dosimeters (RPLGDs). The relationship between Hp(10), Hp(3), and Hp(0.07) along with the effects of the sites of wearing dosimeters on the head inside a full-face mask and the chest (or upper arm) were discussed.

A portable thyroid dose monitoring system has been developed at the Japan Atomic Energy Agency (JAEA) to assess the thyroid equivalent dose for workers and members of the public in a high dose rate environment. The background (B.G.) photon correction is required for an accurate measurement in a high dose rate environment at an early stage after a nuclear accident. The responses of the system to the B.G. photons were calculated with the simplified cylindrical phantoms made of PMMA or the voxel phantoms using the PHITS code. Then they were compared each other among the respective age groups. It was found that the cylindrical phantom with an appropriate size and properly arrangement can mimic the human body in the high dose rate caused by 131I sources. As such we successfully developed the B.G. photon correction method using the cylindrical phantoms; one of those is used as an effective way to calibrate the monitoring system by inserting a133Ba point source into it. As an additional phantom is not required, this method has a great advantage in the practical operation of the B.G. photon correction.

We performed a field study of eye lens dosimetry for workers involved in the decommissioning operation at the Fukushima Daiichi Nuclear Power Plant. In this study, humanoid phantoms equipped with different personal dosemeters were placed at selected locations in the workplace. The experiment showed that Hp(3) at the head is about 20% higher than Hp(10) (or Hp(3)) at the trunk. This level of dose gradient is generally interpreted as being “almost uniform” in radiological control; therefore, tasks conducted in open areas with such relatively small dose gradients (~1.2) will not require specific monitoring with eye lens dosemeters, except when the eye lens dose approaches the dose limit.

In case of a severe nuclear accident, radioiodine monitoring of the thyroid must be carried out for a large number of residents and emergency workers. In the present study, we developed a portable thyroid dose monitoring system consisting of two independent gamma-ray spectrometers and a radiation shield, which could be used under high background radiation conditions. Two types of prototype thyroid monitors were realized; one utilizing CdZnTe detectors for emergency workers and the other employing LaBr3(Ce) detectors for members of the public. As a result of the performance test, it was confirmed that both monitors effectively assess thyroid equivalent doses of

We conducted an eye lens dosimetry study on workers, engaged in recovery and decommissioning operations at Fukushima Daiichi Nuclear Power Station. To investigate the optimal positioning of eye lens dosemeters and the shielding effect of full face mask respirators, photon irradiation experiments to various eye lens dosemeters attached to a head phantom were performed at the secondary photon calibration fields of the Japan Atomic Energy Agency. As a result, no marked difference was observed in readings of the dosemeters attached to the different positions on the head phantom. Moreover, two types of full face mask respirators provided an insignificant shielding effect from photons of 83–662 keV. Comparative measurements between AP and ROT geometries suggested that head-mounted dosemeters would generally produce a small angular variation in readings specifically in multidirectional fields as compared to the torso-mounted dosemeters.

We have developed a portable thyroid monitoring system to measure thyroid doses for members of the public in a nuclear emergency. It is assumed that a large number of these monitoring systems are deployed and maintained for emergency situations. Hence, the calibration performed should be performed as simple as possible. We proposed a new calibration method based on standard voxel phantoms using a point source. By applying this method to the thyroid monitoring system, the counting efficiencies were evaluated. Although this monitor is properly calibrated, any differences between the phantoms used in the calibration and the monitored subjects can affect the measurement accuracy. Therefore, the influence of the anatomical features of the subjects on the detection efficiencies was investigated in detail by Monte Carlo calculations. The effects of the body size on the counting efficiencies of the thyroid monitoring system were estimated using several voxel phantoms. Consequently, the uncertainties in the counting efficiencies due to the body size were found to be within 25%.

An improved analytical method was developed for determining of low levels of radiostrontium in environmental freshwater samples. Emphasis was placed to the in situ pre-concentration of radiostrontium with Powdex resin in large volumes (100-300 L) of freshwater samples from many locations without using of deleterious substances such as NaOH and mineral acids. Measuring electric conductivity (EC) of water samples enabled the estimation of the amount of Powdex resin required for quantitative recovery of Sr from the large water samples in the field. The Powdex resin that adsorbed Sr was brought back to the laboratory, and Sr adsorbed in the resin was eluted by 8 M HNO3 together with Sr carrier added. Strontium was radiochemically separated by the cation exchange method for β counting after removal of most of the Ca using Ca(OH)2 precipitation. Through the procedure the Sr chemical yield was 88% on average. This analytical method was verified by analyzing 170 L of water samples with different salinity values, to which a known amount of (90)Sr was added. The detection limits of (90)Sr activities obtained using the 170 L water samples was estimated to be approximately 0.1 mBq L(-1) for a counting time of 100 min. The method was also applied to environmental samples collected from Ibaraki and Fukushima prefectures; their (90)Sr activities ranged from 0.16 to 0.93 mBq L(-)(1).

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.

Neutron-sensitive glass dosimeters were successfully made from reagents of NaPO 3 , Al(PO 3 ) 3 , enriched 10 or 11 B 2 O 3 and AgCl. The glass dosimeters had good RPL characteristics: satisfactory linearity of 0.01–500 mGy and low variation in sensitivity. The RPL characteristics for gamma-rays were compatible with those of a commercial glass dosimeter. A pair of the 10 or 11 B-containing glass dosimeters was effectively used for the thermal neutron dosimetry. In 565 keV and 5.0 MeV monoenergetic neutron calibration fields, albedo neutrons from the ISO water slab phantom were successfully detected with the glass dosimeter pairs.

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.

In Japan, several monoenergetic and quasi-monoenergetic neutron reference fields are currently operating at several research institutes over a wide energy range 8 keV–390 MeV. These neutron fields are utilized for characterizing and developing neutron detectors as well as for neutron-induced cross section studies, shielding benchmark experiments and irradiation testing of micro-electronic devices. This paper reports the current status of these monoenergetic and quasi-monoenergetic neutron reference fields.

A phoswich-type neutron detector was developed in order to measure high-energy cosmic ray neutron spectra in aircraft. The neutron detector consists of an EJ309 organic liquid scintillator that is 121.7mm in diameter and 121.7mm in length and is covered with a 15mm thick EJ299-13 outer plastic scintillator. The neutron response functions of the detector are required for the unfolding method to obtain the energy spectrum. The neutron response functions were created based on MCNPX simulations using an anticoincidence mode with the experimental light-output correlations with particle energies, uniformity of light collection and energy resolutions. The light-output correlation with particle energy, the uniformity of light collection and the energy resolutions were evaluated based on experiments. Measurements of neutron response functions were performed using four quasi-monoenergetic neutron beams from 40 to 80 MeV p-Li reactions to verify the calculated results. The calculated response functions show good agr...

A new measurement system to determine the neutron-capture cross-sections of a nucleus at 0.008≤ E n nat Pb using pulsed neutrons from 7 Li(p,n) 7 Be at 12≤ E n ≤103 keV. Prompt γ-rays from the reaction were detected by means of a highly sensitive anti-Compton NaI(Tl) spectrometer combined with a time-of-flight method. The obtained result demonstrated good sensitivity of the new system to determine the neutron-capture cross-sections of a nucleus at kiloelectron volt neutron energy.

Precise calibration of monitors and dosimeters for use with high energy neutrons necessitates reliable and accurate neutron fluences being evaluated with use of a reference point. A highly efficient Proton Recoil counter Telescope (PRT) to make absolute measurements with use of a reference point was developed to evaluate neutron fluences in quasi-monoenergetic neutron fields. The relatively large design of the PRT componentry and relatively thick, approximately 2 mm, polyethylene converter contributed to high detection efficiency at the reference point over a large irradiation area at a long distance from the target. The polyethylene converter thickness was adjusted to maintain the same carbon density per unit area as the graphite converter for easy background subtraction. The high detection efficiency and thickness adjustment resulted in efficient absolute measurements being made of the neutron fluences of sufficient statistical precision over a short period of time. The neutron detection efficiencies of the PRT were evaluated using MCNPX code at 2.61×10 −6 , 2.16×10 −6 and 1.14×10 −6 for the respective neutron peak energies of 45, 60 and 75 MeV. The neutron fluences were determined to have been evaluated at an uncertainty of within 6.5% using analysis of measured data and the detection efficiencies. The PRT was also designed so as to be capable of simultaneously obtaining TOF data. The TOF data also increased the reliability of neutron fluence measurements and provided useful information for use in interpreting the source of proton events.

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。

簡易型放射線測定器に利用できる半導体放射線検出素子について，その光子検出性能を校正場での実測とモンテカルロシミュレーション計算で明らかにした。低価格の検出素子として，Si p-i-nフォトダイオード及びCdTe検出器を選択した。周辺線量当量H*(10)に対するエネルギー特性を60keVから2MeVのエネルギー領域で評価した。CdTe検出器はエネルギーの増加とともに感度が大きく減少することがわかった。一方，Siダイオードは150keV以上のエネルギーで感度がほぼ平坦になるという利点を有する。自然放射線に対する検出素子の感度は，4～6cpmである。放射線レベルの検出下限レベルを評価したところ，原子力発電所や燃料加工施設などの事故時に放射線レベルの上昇を把握できることがわかった。

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.

Evaluation of the properties for quasi-monoenergetic neutron calibration fields of high energies more than 20 MeV at TIARA is proceeding for development of the field. Among the properties needed for the development as the standard calibration field, we report on measurement of the neutron beam profile using an imaging plate with a polyethylene converter and on estimation of the contribution of scattered neutrons into the irradiation field based on pulse height distribution at various off-beam positions measured using an organic liquid scintillation detector.

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.

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.

A new cerium-doped LaCl3 (Ce) scintillator is evaluated with respect to the application in environmental gamma-ray dosimetry and spectrometry . The scintillator is very attractive for gamma-ray spectrometry in the case of high count rate, because it has excellent energy resolution and fast decay time. The performance characteristics of a scintillator with a 25.4 mm×25.4 mm LaCl3(Ce) crystal are studied and compared to those of a NaI(Tl) scintillator with the same size crystal. Acquired pulse height spectra are converted to dose rates by using the G ( E ) function method. Though the LaCl3(Ce) crystal itself produces a rather high background in the crystal itself, the scintillator provides good energy information and dose-rate readings from low to high-level (several mGy/h) by subtracting the self-background. The properties of LaCl3(Ce) scintillator suggest that the scintillator could be a promising candidate for monitoring at high dose levels as in emergencies, as well as at ordinary levels of background radiation.

The first investigation of a radiation detector made of a diamond single crystal grown by a chemical vapor deposition (CVD) method was successfully carried out. The diamond single crystal, having a size of 2.0×2.0×0.7 mm 3 , was grown by the CVD method and then applied to the detector. The CVD method was able to reduce nitrogen and boron impurities in diamond crystals more effectively than a high-pressure and high-temperature (HP/HT) method. Consequently the detector had high-energy resolution of 0.4% for 5.486 MeV α particles from an 241 Am source. The behavior of charge carriers in the detector could not be adequately investigated because the detector had a partial depression layer and strong rectification. Ionization energy of the CVD diamond was determined to be 16.1 eV; this value was higher than that of the HP/HT type IIa diamond, i.e., 13.1 eV. This result shows the potential of CVD diamond single crystals because an ideal diamond should have higher ionization energy than the HP/HT type IIa diamond that had the influence of impurities.

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.

Pulse height distributions were measured using a CdZnTe detector inside a lead collimator to investigate main source producing high dose rates above the shield plugs of Unit 3 at Fukushima Daiichi Nuclear Power Station. It was confirmed that low energy photons are dominant. Concentrations of Cs-137 under 60 cm concrete of the shield plug were estimated to be between 8.1E+9 and 5.7E+10 Bq/cm2 from the measured peak count rate of 0.662 MeV photons. If Cs-137 was distributed on the surfaces of the gaps with radius 6m and with the averaged concentration of 5 points, 2.6E+10 Bq/cm2 , total amount of Cs-137 is estimated to be 30 PBq.

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.

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.

A compact single ion irradiation system has been developed to examine energetic particle effects on materials and devices. The system has been constructed by use of commercially available and inexpensive standard components and can be easily recomposed according to beam requirements for various irradiation experiments. The beam adjustment was automatically performed in obedience to a computer program based on a modified SIMPLEX method. The beam performances of the system, i.e., pulse width, the number of ions included in one pulse and beam size were examined by use of a microchannel plate, a SI-SSD, a CCD image sensor and a CR-39 track detector, respectively. A beam of a single ion, 1 nsec in pulse width and several /spl mu/m in diameter has been successfully produced by this system.

Characteristics of a synthetic diamond radiation detector in high-temperature environments up to 196°C were investigated. The radiation detector was made of a high-purity type IIa diamond single crystal synthesized by a high-pressure and high-temperature (HP/HT) method. In order to suppress a strong increase in leakage current in a high-temperature environment, a surface barrier structure composed of Schottky and ohmic contacts was adopted. The detector operated as an energy spectrometer for 5.486 MeV α particles up to 156°C. The peak due to the α particles obtained at 68°C slightly shifted to the higher-energy side compared with the same peak obtained at 25°C; furthermore, the energy resolution of these peaks was improved with the increase in the detector temperature. Effects of trapping and detrapping on holes were suppressed by thermal excitation, and this resulted in an improvement in collection efficiency of charge carriers. With a further increase in the detector temperature, loss of charge carriers became more significant. Finally, operation of the detector was prevented at 196°C by a polarization phenomenon caused by accumulation of space charge in the detector.