[Show abstract][Hide abstract] ABSTRACT: The neutron streaming for torus duct was investigated experimentally and analytically by using actual scale of experimental setups and 252Cf neutron source, in order to obtain the basic data of the shielding design such as the nuclear fuel facilities. The torus ducts with 1 m of radius of curvature and 3 cm and 5 cm of the duct diameters were used. Neutron doses were measured by REM counter and solid state track detector (SSTD). The MCNP4A calculation was done for the comparison with the experimental results. In the case where the arrangement of the source and the detector is comparatively able to foresee through the duct, the doses were high as a matter of course. In the case where the other side could not be foreseen directly through the torus duct, the dose became 4E-2 as compared with the straight duct. The dose for 3 cm of the duct diameter was generally smaller than that of 5 cm. The doses for 3 cm increased only 7% as compared with the bulk case (90 cm thickness, source position: disk, detector position: center).
Preview · Article · Aug 2014 · Journal of Nuclear Science and Technology
[Show abstract][Hide abstract] ABSTRACT: aNuclear facilities with strong radioactivity need massive concrete shields. In the view of shielding design, there arise some difficulties in the estimation of radiation doses from a gap or a void between the hatch and the wall. So, there is a need of experimental data and calculational procedures on such geometry for a reasonable shielding design. The experiments were carried out by 252Cf neutron source. A gap between the shielding hatch and the wall, and the offset of the shielding hatch were simulated by piling up the concrete slabs to the height of 30 cm. The thickness of the slabs are 10 cm each. Neutron dose was measured by REM-counter and CR-39 plate. To simulate various types of the shielding hatch, gap width and offset length of each gap were changed.
Preview · Article · Aug 2014 · Journal of Nuclear Science and Technology
[Show abstract][Hide abstract] ABSTRACT: A measurement of neutron dose rate on iron-polyethylene shielding structure was carried out by 252Cf source. Simulated geometry was slit-like opening of polyethylene in iron slab and polyethylene slab shielding.These experiment was done at research facility of Hazama Co,. Iron slab and polyethylene slab thickness were 10 cm each. A gap of the polyethylene was simulated. Neutron REM-counter, polyethylene covered BF3 counter (STUDSVIK 2202-D), was used for measurement of streaming neutron dose equivalent. The solid state track detector (SSTD), allyl-diglycol-carbonate, were used for measurement of fast neutron dose equivalent in the range of 170Kev to 15Mev.The experimental data was obtained against gap width, source location and detector location.Obtained data shows strong correlation between dose rate and above parameters.These data was investigated in the view of to make use of actual facility design and compared with calculation such as MCNP4B.From the result of gap streaming experiment and calculation, we obtained allowable gap width as 6mm for this case (10cm polyethylene thickness).
No preview · Article · Aug 2014 · Journal of Nuclear Science and Technology
[Show abstract][Hide abstract] ABSTRACT: Monitoring preparation for internal contamination with actinides (e.g. Pu and Am) is required to assess internal doses at
nuclear fuel cycle-related facilities. In this paper, the authors focus on skull counting in case of single-incident inhalation
of 241Am and propose an effective procedure for skull counting with an existing system, taking into account the biokinetic behaviour
of 241Am in the human body. The predicted response of the system to skull counting under a certain counting geometry was found to
be only ∼1.0 × 10−5 cps Bq−1 1y after intake. However, this disadvantage could be remedied by repeated measurements of the skull during the late stage
of the intake due to the predicted response reaching a plateau at about the 1000th day after exposure and exceeding that in
the lung counting. Further studies are needed for the development of a new detection system with higher sensitivity to perform
reliable internal dose estimations based on direct measurements.
No preview · Article · Jun 2014 · Radiation Protection Dosimetry
[Show abstract][Hide abstract] ABSTRACT: Measurements for internal dose assessment are required to be conducted based on the distribution of radionuclides in the body,
which may change depending on the lapsed time. In this study, a biokinetic analysis code, which can be used in practical radiation
control is developed, and the results of 60Co and 137Cs biokinetics are visualised as examples by drawing the depositions for each organ and tissue in a figure of the body as
a function of lapsed time. In addition, based on visualised biokinetics, precautions for in vivo measurements are also discussed. These discussions led to the conclusion that the information of visualised biokinetics is
useful for actual measurements in practical radiation control.
No preview · Article · Jun 2013 · Radiation Protection Dosimetry
[Show abstract][Hide abstract] ABSTRACT: Reaction rate distributions were measured inside a 60-cm thick concrete pile placed at the lateral position of a thick (stopping length) iron target that was bombarded with heavy ions, 400 MeV/u C and 800 MeV/u Si. Foils of aluminum and gold, as well as gold, tungsten and manganese covered with cadmium were inserted at various locations in the concrete pile to serve as activation detectors. Features of reaction rate distribution, such as the shape of the reaction rate profile, contribution of the neutrons from intra-nuclear cascade and that from evaporation to the activation reactions are determined by the analysis of measured reaction rates. The measured reaction rates were compared with those calculated with radiation transport simulation codes, FLUKA and PHITS, to verify their capability to predict induced activity. The simulated reaction rates agree with the experimental results within a factor of three in general. However, systematic discrepancies between simulated reaction rates and measured reaction rates attributed to the neutron source terms are observed.
No preview · Article · Jan 2012 · Nuclear Instruments and Methods in Physics Research Section B Beam Interactions with Materials and Atoms
[Show abstract][Hide abstract] ABSTRACT: Personal neutron dosimeters are necessary for radiation protection of nuclear workers in high energy accelerator facilities. To improve sensitivity for high energy neutrons above 10 MeV, a new personal neutron dosimeter was designed using a solid state nuclear track detector (CR-39) covered with Multi-Layer radiator consisting of polyethylene, polyamide and iron/aluminum sheets. In this work, we investigated various radiator combinations and thicknesses in order to detect high energy neutrons above 10 MeV. Energy response of each radiator was calculated using the multi-purpose Monte Carlo simulation code, Particle and Heavy Ion Transport System (PHITS). To evaluate the accuracy of our calculations, the energy response of each radiator was measured in a mono-energetic neutron field at the Facility of Radiation Standard (FRS) of the Japan Atomic Energy Agency (JAEA). The energy response of each radiator calculated by PHITS was in good agreement with our experimental results. The results obtained in this research indicate that PHITS is a useful software tool for the design of new personal neutron dosimeters using CR-39 and using Multi-layer radiator approach.
No preview · Article · Dec 2011 · Radiation Measurements
[Show abstract][Hide abstract] ABSTRACT: Spallation and neutron capture reaction rate distributions were measured using activation detectors inside a 90-cm thick ordinary concrete pile exposed to a field of secondary particles escaping a thick (stopping length) iron target bombarded with various intermediate energy ions, 230 MeV/u He, 400 MeV/u C, and 800MeV/u Si. Activation detectors of aluminum, bismuth, gold, and gold covered with cadmium were inserted at various depths in the concrete pile. In addition, the distributions of activation reaction rate were simulated by FLUKA and PHITS Monte-Carlo codes. Generally, comparison of measured and calculated reaction rates show agreement within a factor of two. The experimental data will be useful for benchmarking Monte-Carlo radiation transport simulation code capabilities in estimating radioactivity induced in accelerator radiation shielding.
No preview · Article · Sep 2011 · Nuclear Instruments and Methods in Physics Research Section B Beam Interactions with Materials and Atoms
[Show abstract][Hide abstract] ABSTRACT: A dose evaluation using multiple radiation detectors can be improved by the convex optimisation method. It enables flexible dose evaluation corresponding to the actual radiation energy spectrum. An application to the neutron ambient dose equivalent evaluation is investigated using a mixed-gas proportional counter. The convex derives the certain neutron ambient dose with certain width corresponding to the true neutron energy spectrum. The range of the evaluated dose is comparable to the error of conventional neutron dose measurement equipments. An application to the neutron individual dose equivalent measurement is also investigated. Convexes of particular dosemeter combinations evaluate the individual dose equivalent better than the dose evaluation of a single dosemeter. The combinations of dosemeters with high orthogonality of their response characteristics tend to provide a good suitability for dose evaluation.
[Show abstract][Hide abstract] ABSTRACT: The International Commission on Radiological Protection has recommended that cosmic radiation exposure of crew in commercial
jet aircraft be considered as occupational exposure. In Japan, the Radiation Council of the government has established a guideline
that requests domestic airlines to voluntarily keep the effective dose of cosmic radiation for aircraft crew below 5 mSv y–1. The guideline also gives some advice and policies regarding the method of cosmic radiation dosimetry, the necessity of explanation
and education about this issue, a way to view and record dose data, and the necessity of medical examination for crew. The
National Institute of Radiological Sciences helps the airlines to follow the guideline, particularly for the determination
of aviation route doses by numerical simulation. The calculation is performed using an original, easy-to-use program package
called ‘JISCARD EX’ coupled with a PHITS-based analytical model and a GEANT4-based particle tracing code. The new radiation
weighting factors recommended in 2007 are employed for effective dose determination. The annual individual doses of aircraft
crew were estimated using this program.
No preview · Article · May 2011 · Radiation Protection Dosimetry
[Show abstract][Hide abstract] ABSTRACT: Neutron-induced reaction rate depth profiles inside concrete shield irradiated by intermediate energy neutron were calculated using a Monte-Carlo code and compared with an experiment. An irradiation field of intermediate neutron produced in the forward direction from a thick (stopping length) target bombarded by 400 MeV nucleon(-1) carbon ions was arranged at the heavy ion medical accelerator in Chiba. Ordinary concrete shield of 90 cm thickness was installed 50 cm downstream the iron target. Activation detectors of aluminum, gold and gold covered with cadmium were inserted at various depths. Irradiated samples were extracted after exposure and gamma-ray spectrometry was performed for each sample. Comparison of experimental and calculated shows good agreement for both low- and high-energy neutron-induced reaction except for (27)Al(n,X)(24)Na reaction at the surface.
[Show abstract][Hide abstract] ABSTRACT: Activities on radon management strategy of international organisations (International Atomic Energy Agency, International Commission on Radiation Protection, etc.) should be carefully and continuously traced to discuss how to control radon in various environments, for example, dwellings, workplace, underground, caves, mines, hot springs, disposal facilities and so on. It is more reasonable in parallel to set radon reference level by effective dose criteria of Sv y(-1) as well as by radon concentration in air of Bq m(-3). How to investigate radon concentration in each environment, and how to make decisions on needed action for radiation protection from natural radon,--these should be discussed for each environmental situation on a case-by-case basis. International discussion as well as domestic discussion is continuously needed, not only among the radon specialists and regulators, but also including stakeholders who are the main users of regulation and guidance.
[Show abstract][Hide abstract] ABSTRACT: In diagnostic radiology, the tube voltage [peak kilovoltage (kV(p))] is one of the most important parameter affecting both radiation exposure and image contrast. So, an accurate kV(p) meter is necessary to control kV(p) in the medical radiography practice with the overall uncertainty less than ± 5 % according to IEC 61676. Therefore, both invasive and non-invasive calibration methods of kV(p) meter were established and applied to different kinds of commercial quality control instruments for diagnostic radiology. Calibration of kV(p) meter by the invasive method is the most accurate (with uncertainty of 1.67 %, k=2); however, the non-invasive method also provides good results (with uncertainty of 3.12 %, k=2). Due to their detailed design, the commercial kV(p) meters have various responses with X-ray beam, so the working regime of a particular device type must be appropriately selected with a specific X-ray machine used for calibration of kV(p) meter.
No preview · Article · Mar 2011 · Radiation Protection Dosimetry
[Show abstract][Hide abstract] ABSTRACT: A new experimental approach based on the stacked foil method to determine fragmentation cross-section from its threshold to the initial projectile energy as a continuous function of energy is proposed. This method was tested against fragmentation reaction product yields calculation by FLUKA Monte-Carlo radiation transport simulation code. Excitation functions were obtained by applying the method to the spatial distribution of fragments in a thick carbon target irradiated by 500 MeV/u 56Fe ions calculated with FLUKA and compared with cross-sections built in to FLUKA. Fairly good agreement of two excitation functions is found indicating validity of the method. Possible corrections and errors are discussed.
[Show abstract][Hide abstract] ABSTRACT: A pre-etching step is applied with carbon dioxide (hereafter, CO2 pre-etching) of a solid state nuclear track detector (SSNTD) to improve the sensitivity for protons detection. In this study, we quantitatively estimated the sensitization of SSNTD with CO2 pre-etching to improve the sensitivity of a personal neutron dosimeter. We measured the etch pit size and the critical angle using a 1.7 MV tandem accelerator at the University of Tokyo to evaluate the proton sensitivity. In addition, we calculated the neutron energy response of SSNTD with a polyethylene radiator with the particle and heavy ion transport system multipurpose three-dimensional Monte Carlo code. To evaluate the accuracy of these calculations, neutron energy responses of SSNTD with and without CO2 pre-etching were measured at the Facility of Radiation Standards of the Japan Atomic Energy Agency. The neutron energy responses calculated by the above Monte Carlo code agreed with the experimental results to within 35%. The results obtained in this research indicate that CO2 pre-etching at a pressure of 0.6 MPa for 3 days increases the radiator effect by up to 250% for the 5 and 14.8 MeV neutrons.
[Show abstract][Hide abstract] ABSTRACT: The precise energy-dependent response curve of the imaging plate to low-energy photons was obtained by reciprocally combining a calculation and an experiment. The calculation was carried out using the Monte Carlo particle transportation code EGS5. The experiment was carried out using monochromatic photons at the photon factory in KEK. Initially, there was a discrepancy between the calculation and the experiment because of light attenuation in the phosphor layer in the read-out process. However, by setting the attenuation coefficient of this light attenuation to 0.013 μm-1 and decreasing the imaging plate (IP) response exponentially with depth from the phosphor surface, it was indicated that the calculation result basically follows the experimental result. By using the response curve obtained in this paper, radiation dose can be evaluated accurately in the dose estimation using IP.
[Show abstract][Hide abstract] ABSTRACT: Radon adsorption by activated charcoal collectors such as PicoRad radon detectors is known to be largely affected by temperature and relative humidity. Quantitative models are, however, still needed for accurate radon estimation in a variable environment. Here we introduce a temperature calibration formula based on the gas adsorption theory to evaluate the radon concentration in air from the average temperature, collection time, and liquid scintillation count rate. On the basis of calibration experiments done by using the 25 m³ radon chamber available at the National Institute of Radiological Sciences in Japan, we found that the radon adsorption efficiency may vary up to a factor of two for temperatures typical of indoor conditions. We expect our results to be useful for establishing standardized protocols for optimized radon assessment in dwellings and workplaces.
No preview · Article · Oct 2010 · Journal of Environmental Radioactivity