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Measurement of radioactive fragment production excitation functions of lead by 400 MeV/u carbon ions Tatsuhiko Ogawa, Mikhail N. Morev, Takeshi Iimoto and Toshiso Kosako a University of Tokyo, 2-11-16, Yayoi, Bunkyo-ku, Tokyo, 113-8656, Japan; b Japan Atomic Energy Agency, 2-4 Shirakata, Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195, Japan; c Science and Engineering Center for Nuclear and Radiation Safety, d.2/8, korp. 5, ul. Malaya Krasnoselskaya, Moscow, 107140, Russia

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 (60)Co and (137)Cs 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.

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.

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.

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 800 MeV/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.

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.

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.

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.

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 m 3 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.

A novel neutron dose measurement method that flexibly responds to variations in the neutron field is being developed by Japan Atomic Energy Agency. This is an implementation of the multi-detector method (first introduced in 1960s) for neutron dose evaluation using a convex hull in the response space defined for multiple detectors. The convex hull provides a range of possible neutron dose corresponding to the incident neutron spectrum. Feasibility of the method was studied using a simulated response of mixed gas proportional counter. Monochromatic neutrons are shown to be fundamentally suitable for mapping the convex. The convex hull can be further reduced taking into consideration a priori information about physically possible incident neutron spectra, for example, theoretically derived moderated neutron spectra originated from a fission neutron source.

49cm厚みの普通コンクリートを用いた遮へい実験を東京大学高速中性子源炉「弥生」を用いて行った。遮へい実験は重コンクリートの壁に囲まれた体系で行い，重コンクリートからの後方散乱が中性子線束に影響しうる状況で行った。計測は197Au(n，γ)，カドミウムで包んだ金の197Au(n，γ)そして115In(n，n′)の反応を用いた箔放射化法により行った。一方で中性子輸送計算を行って普通コンクリート内の中性子スペクトルを求め，中性子捕獲断面積と掛け合わせることで，実験の反応率の再現を試みた。計算結果と実験結果の比較は良好な一致が見えたが，最深部でのみ顕著な不一致が確認された。計算において実験を正しく模擬するには，普通コンクリート後方の重コンクリートの組成が重要であると考えられた。そこでまず壁から散乱される中性子束を無視して計算を行ったところ，計算は実験を0.1倍過小評価した。次に，重コンクリートの組成として文献値を用いたところ，計算値は5倍の過大評価となった。最後に，重コンクリートの成分として施設設計の値を使用したところ，過大評価は2倍に収まった。したがって，重コンクリートで囲まれた体系において中性子の遮へい実験を行う上で，遮へい体の最深部における反応率を計算で正しく再現するには，重コンクリートの組成を計算に正しく反映させることが重要であると判明した。

In Thailand, radioactive materials are widely used for the industry, medicine, research and development fields. Reported here are background and recent developments in the national nuclear legislation system, including regulation of radiation safety and current status of nuclear program in Thailand. Under the Atomic Energy for Peace Act, the Thai Atomic Energy Commission (Thai AEC) is authorized to approve regulations respecting, the conversion, enrichment, processing, reprocessing, possession, import, export, use, packaging, transport, management and storage of nuclear materials. The most recent developments are related to the New Ministerial Regulation on Licensing Requirements Procedures and Nuclear Material, By-Product or Atomic Energy Processing B. E 2550 (A. D. 2007) issued under the Atomic Energy for Peace Act, B. E. 2504 (A. D. 1961). Currently, the Thai Cabinet is discussing the draft new Atomic Energy for Peace Act which is to revise the Act. The draft Act is to sets forth criteria for protecting individuals, society and the environment from radiation hazards with the perspective for anticipated nuclear power sector development in Thailand.

An approach how and what to discuss the decommissioning of large accelerator facilities has been overviewed through a case study on Stanford Linear Accelerator Center (SLAC) in USA, which is also under discussion in Japan. SLAC has its own framework on radioactive waste management, however it is still involved in difficulties about the treatment of large amount of low-level radioactivated materials, such as shielding blocks or soil surrounding the underground accelerator. Introduction of disposition criterion of solid material was deferred by US regulatory authority. SLAC is also focusing on the environmental protection using Environmental Restoration Project due to United States Environmental Protection Agency (USEPA). Remedies for the circumstance water or land which are affected by the SLAC's past operations are started. Every kind of pollution including radioactive contaminations are targets of remedies. It means they address the inclusive management of various risks, such as chemical hazardous risk and radiological consequence risk. It is helpful for other accelerator facilities over the world which program decommissioning.

核燃料物質に関連する国際的な管理の体制，特に，保障措置に絡む国内外の状況と対応についての歴史的経緯及び現状を整理した。また，天然/劣化ウラン300g，トリウム900g以下の，いわゆる少量の核燃料物質に焦点をあて，それらを使用する際の法的要件と手続きにおける留意事項をまとめ，そのような物質を使用する事業所がとるべき適切な対応について，東京大学の例を挙げつつ，教育プログラムの策定と実施に着目し，考察をした。

A wide range thermal neutron detector was developed based on the Scintillator with Optical Fiber (SOF) detector which has been previously used for thermal neutron monitoring during boron neutron capture therapy irradiation. With this new detector system we intended to address the issues of real-time thermal neutron flux measurement and the simultaneous measurement of a wide range of thermal neutron flux in a BNCT irradiation field which were difficult to implement with the gold wire activation method. The dynamic range of linearity of the SOF detector was expanded by using a plastic scintillator with a rapid decay time. On the other hand, the contribution of gamma rays and fast neutrons in the measured signals were compensated from those obtained by a pair of SOF detectors, one with a 6 LiF neutron converter and the other without. The discrimination level for the measured signals was also optimized to further reduce the contribution of gamma rays and fast neutrons signals. A non-paralyzable system model was applied to correct for the dead-time in the detector system. A good agreement between the thermal neutron flux measured by the gold wire activation method and the paired SOF detector system was observed. However, measurements which would normally take a few days to perform with the gold wire activation method were obtained in just about 15 min using the SOF detector system. We also confirmed the dynamic range of linearity for the SOF detector system to be in the order of magnitude of 10 4 .

Some of technologically enhanced naturally occurring radioactive material (TENORM), such as radon and monazite sand, gradually became a target for radiological protection. In order to regulate TENORM safely and economically, it is essential to consider the characteristics of TENORM such as ubiquity, huge volume, and very low activity levels. In this paper, radiation protection principles and standards for NORM/TENORM are summarized based on the reports published by the international organizations (e. g. ICRP and IAEA) to assist the development of national regulatory framework. The survey results on the present Japanese situations on industries related to NORM/TENORM are provided, and the categorization for NORM/TENORM, which is established by the Radiation Council of Japan, is explained how to control NORM/TENORM based on their exposure doses instead of their activity levels. Finally, basic processes to solve the problems related to NORM/TENORM are discussed.

A simple technique was developed for using an imaging plate (IP) to estimate the individual concentrations of radon progeny in the air. The objective of this technique is to survey concentrations in high-humidity environments, such as caves, utility mains, and underground facilities.An IP called BAS-III, which performs well in extremely humid environments, was selected. The sampling flow rate was 65 liters per minute by a 47 mm-dia. glass-fiber filter (GF/F), and the grab sampling time was determined to be 5-10 minutes. One minute after the air sampling, the surface of the filter was attached directly to the IP in a prepared cassette. This was repeated six times for 10 minutes, for a total exposure of 60 minutes. Six values of Photo-Stimulated Luminescence (PSL) indicating the decay curve of the radon progeny sampled on the filter were analyzed. In environments with extremely high concentrations of radon, more accurate measurements can be obtained by making a simple distinction between PSL values for α and β rays. The lower detection limit for the equilibrium equivalent concentration (EEC) of radon was estimated to be about 20 Bqm-3. This system can be successfully adopted in humid areas.

New in-vivo calibration phantoms (anthropometric phantoms) were developed to meet the needs for Japanese standard phantoms. Two important characteristics of these phantoms were that (1) they were designed using Japanese body size survey data, and (2) they were designed so that they can be adapted to various positions or geometries. The performance of these phantoms was tested with respect to body size, activity distribution along the axis, and counting efficiency. The actual dimensions of the anthropometric phantoms were compared with the survey data. Most items (31 of 47) indicated good agreement between the actual values and the survey data for the adult anthropometric phantom. The activity distribution for the anthropometric phantoms was compared with that for block phantoms that simulate a uniform activity distribution. The anthropometric phantoms have some gaps in their joints. The measurement results, however, indicated that these gaps did not significantly affect the overall accuracy of the measurements. Differences in counting efficiency between the block phantoms and the anthropometric phantoms for the same age were no more than 6%.

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 (241)Am and propose an effective procedure for skull counting with an existing system, taking into account the biokinetic behaviour of (241)Am 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.

The shielding effect for neutrons and gamma rays provided by a louver-type steel structure inserted into the second leg of a concrete duct was measured and analyzed. The louver is an assembly of steel plates that are stacked at the same interval to reduce the radiation streaming while keeping air flowing through the duct. The experiment was carried out at the Japan Research Reactor-4 (JRR4) using a large concrete duct that was temporarily installed for this experiment.Experimental data for the shielding effect of the louver were obtained through the use of thermoluminescent dosimeters (CaSO{sub 4} and BeO) and an ionization chamber for the gamma dose. A rem counter was used to obtain neutron dose, while a solid-state track detector was used for fast neutron dose. Finally, indium activation foil was used to obtain the thermal neutron flux. A NaI(Tl) scintillation spectrometer was used for the measurement of gamma rays from activated foil.The measured data were compared to that derived from numerical analyses. Numerical analyses included the use of the conventional S{sub n} transport code DOT3.5, the Monte Carlo code MCNP4A, and calculations with empirical formulas.MCNP4A provided satisfactory estimates for all cases. If proper calculations were carried out, then DOT3.5 provided more » acceptable estimates except for the thermal neutrons in spite of the limitations of the code's two-dimensional geometrical modeling. Calculations by hand using simple empirical formulas with modifications, like that for the angular flux correction, also could provide fairly accurate estimates. « less

Trace level uranium in wastewater is a particularly important issue for material accountancy and radiation protection. It is difficult to detect trace level uranium without concentrating wastewater. This study is a discussion of particle-induced X-ray emission (PIXE) analysis that is used to find concentration of uranium easily. In the procedure discussed here, the quantity of solution and backing material are examined to determine how they affect the practicality of taking measurements. Our study resulted in findings that show higher sensitivity, shorter measurement time and wider range of concentration than those obtained by usual methods.

The concentrations of 222Rn (radon) and its progeny with surrounding environmental gamma-dose rates were measured simultaneously inside and outside of buildings at 10 locations around Taipei and Hualien in Taiwan. For summer radon in Taiwan, indoor concentrations were estimated to be about 20 Bq m−3 with about 90 nSv h−1 of environmental gamma, and outdoors, about 10 Bq m−3 with about 70 nSv h−1. The equilibrium factors were calculated to be 0.2–0.3 indoors and 0.3–0.4 outdoors. Indoor radon concentration had a weak positive correlation with gamma-dose rate. Since there is a possibility that high radon concentrations exist indoors during the cool season in Taiwan because of extremely low ventilation rates in the dwellings, a winter survey in January through February will be needed for future estimation of the annual effective dose.

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.

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,...

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.Results of the experiments shows that the dose rate goes up with increment of gap width when the offset was fixed. The experimental data were also compared with the calculations by MCNP4B. We obtain the guide of design for radiation shielding hatch.

152Eu activities induced by the atomic explosion in Nagasaki, have been experimentally estimated in two roof tiles in order to improve the accuracy of the former data about positions distant from the epicentre (at 1100 m) as the slant distance. Several aspects causing estimation errors have been discussed in more detail: i.e. influence of the natural background nuclides such as 210 Bi, 210 Pb, 212 Bi and 138 La, and half-life selection of 152 Eu. As a result, the value of 52 Eu/Eu (i.e. ratio of the 152 Eu activity to total Eu) has been estimated to be 0.336 ± 0.032 (Sample A) and 0.491 ± 0.040 (Sample B) Bq.mg-' as the values at the time of the bombing. In a comparison between the experimental values and the Dosimetry System 1986 (DS86) calculation, the ratio between DS86 calculations and the present work is estimated to be 0.40-0.58. More detailed discussion on the dose estimation of thermal neutrons might be needed for the improvement of the DS86 system.

配管中の油中気泡割合を測定する方法として, 小型の中性子源と減速材型中性子検出器の組合せによるオンライン型ボイド率測定法を開発した。線源には高速中性子源である252Cfを使用し, 測定時のバックグラウンドを低減するために, 減速材厚みや測定部周囲の遮蔽等を検討した。静止状態でボイド率の校正を行い, さらにシャシーダイナモ上で実車によるAT油中のボイド率を実測し, 重量測定法との比較により本測定法の有効性を検証した。

Shielding characteristics of iron, lead, ordinary concrete, heavy concrete, graphite, marble, water, and paraffin were calculated for monoenergetic source neutrons with energies