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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

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.

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.

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 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.

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.

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,...

Reaction rate distribution in concrete with/without boron dopant up to a thickness of 60 cm was measured using Yayoi fast reactor located at University of Tokyo. The 7 reaction rates such as (197)Au(n,gamma), (59)Co(n,gamma), (115)In(n,n'), (55)Mn(n,gamma), (23)Na(n,gamma), (94)Zr(n,gamma) and (96)Zr(n,gamma) were measured at 12 different depths, and the reduction of the reaction rate as a result of boron doping was quantitatively analysed. These reaction rates were also used to determine epithermal neutron spectrum shape parameter. Monte Carlo simulations of the experimental setup were performed using the MCNP-5 code. Simulated depth profiles of reaction rates and the epithermal neutron spectrum shape parameter agreed with the experimental results with fair accuracy. This experimental results provide useful data to benchmark the accuracy of neutron transport codes in the prediction of transmission and neutron spectrum distortion in boron-doped concrete.

A simple system to evaluate the 222 Rn (radon) exhalation rate from soil has been improved. A sampling cuvette of 2.1 L is placed so that it covers the targeted ground soil, and radon emanating from the soil accumulates within the cuvette for 24 h. Its internal radon concentration is measured by the combination of an activated charcoal (PICO-RAD) and a liquid scintillation counting system. This study shows variations of the conversion factor (CF: unit Bq m −3 /cpm) of PICO-RAD. The range of CF due to temperature (10–30 °C) was between −21% and +69%, and this due to humidity (30–90%) was between 0% and −15%. Humidity and radon concentration in the cuvette covering soil tended to saturate in a few hours. The above information was used to correct the CF for the evaluation. The improved system shows high reliability and can be easily applied to natural environments.

Publisher Summary The airborne 222 Rn (radon) progeny is a primary natural radioactive source generating an effective dose. It is important to identify high concentration areas for the radon progeny and to estimate these concentrations in air. However, most of general commercial instruments for measuring radon progeny concentrations using a precise electric detector are not perfect, because they are not suitable for use in high humidity environments that can reach up to 100% such as caves, utility mains, spas, and underground facilities. A new technique for estimating the airborne concentration of individual radon progeny has been developed. This method uses an imaging plate (IP) of BAS-III as a radiation detector. This system has a strong advantage due to the special IP feature for measurements in environments with extremely high humidity. In addition, the mobility to sample the environmental data is high because of the simple construction and light weight of the IP and cassette. An electric device is not needed to record the alpha information from the sample filter into the IP.

Most commercially based activated charcoal radon collectors were designed for use in indoor environments. However, at present, they are often used for research in radon surveys in unique environments, such as in the bathrooms, underground areas, mines, caves and tunnels. In these environments, the relative humidity would be around 100%, and a change in the sensitivity of cpm(Bq m−3)−1(radon) would occur. For this study, the reduction in the sensitivity of activated charcoal radon collector due to environmental humidity was investigated, and the data correction was discussed. Here, ST-100 (Pico-Rad) was selected as an example of a familiar activated charcoal radon collector. According to our performance test, the humidity of 90% (20 °C) resulted in a 15% reduction of the sensitivity for 24 h collection. The ST-100 user should discuss the necessity of data correction by comparing the change of sensitivity with other levels of estimation errors.

The 152Eu activity depth profile of a granite pillar of the Motoyasu bridge located 132 m from the Hiroshima atomic bomb hypocenter was assessed. The pillars each measured 82 cm in depth, 82 cm in width and 193 cm in height. One of the pillars was bored and 6.8-cm-diameter core samples were removed and cut into 2-cm-thick disks. Two gamma rays of 152Eu, 122 keV and 344 keV, in each disk were measured using a low background, gamma-ray spectrometer, and the activity distribution was determined as a function of depth in the granite. A concentration of stable Eu in the granite was determined by activation analysis. The specific radioactivity of 152Eu and 154Eu at the pillar surface was determined to have been 117 and 24 Bq per mg Eu, respectively, at the time of detonation. The value of 152Eu agrees within 20% of that calculated by Loewe. The depth profile of 152Eu in granite demonstrates a distinct difference from the estimates made only by thermal neutrons. Present data provide valuable information for the analysis of the neutron spectrum of the Hiroshima atomic bomb and its intensity.

To investigate the neutron dose and spectra around a fast reactor from the point of view of radiation protection and shielding, neutron measurements were conducted at the reactor top of JOYO, a Jap...

A cylindrical BF3 counter with cylindrical polyethylene moderators of 1-, 3-, 6- and 10-cm wall thickness were developed for neutron spectrometry, having response functions calculated by the time-dependent multi-group Monte Carlo code, TMMCR, and evaluated by a time-of flight experiment in the energy region from epithermal to about 50 keV. This system of neutron detectors was used to measure the environmental neutron spectrum and dose equivalent around a 252Cf fission neutron source, a fast neutron source reactor, a cyclotron and an electron synchrotron. The neutron spectrum and dose obtained by this detector system showed very goad agreement with those measured by a commercially available neutron rem counter and calculated by our MMCR Monte Carlo code. As a result of these comparisons, this detector system has been established as a standard spectrometer and dosimeter for environmental measurements around nuclear facilities.