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For correct assessment of health risks after low-dose irradiation, calculation of radiation exposure estimates is crucial. To verify the calculated absorbed doses, instrumental methods of retrospective dosimetry are used. We compared calculated and instrumental-based estimates of external absorbed doses in the residents of Dolon, Mostik and Cheremushki villages, Kazakhstan, affected by the first nuclear weapon test performed at the Semipalatinsk Nuclear Test Site (SNTS) on August 29, 1949. The ‘instrumental’ doses were retrospectively estimated using the Luminescence Retrospective Dosimetry (LRD) and Electron Spin Resonance (ESR) methods. Correlation between the calculated individual cumulative external absorbed whole-body doses based on typical input data and ESR-based individual doses in the same people was strong (r = 0.782). It was even stronger between the calculated doses based on individual questionnaires’ input data and the ESR-based doses (r = 0.940). Application of the LRD method is useful for validation of the calculated settlement-average cumulated external absorbed dose to air. Reconstruction of external exposure can be supplemented with the data from later measurements of soil contamination with long-lived radionuclides, such as, 137Cs. Our results show the reliability of the calculational method used for the retrospective assessment of individual external doses.

The purpose of this study was to estimate the internal dose of radiation in Japanese macaques (aka Nihonzaruor snow monkey) due to the Fukushima nuclear power plant accident. Images of a male Japanese macaque weighing ~10 kg were acquired using a multi-slice computed tomography (CT) scan with a 64-row segment detector. The CT images were used to create voxel phantoms of the bones, bone marrow, brain, eyes, heart, lungs, stomach, liver, spleen, pancreas, kidneys, intestines, bladder, testes, thyroid and miscellaneous tissue. The Particle and Heavy Ion Transport System (PHITS) Monte Carlo code was used to calculate the internal exposure rate conversion factors for 134Cs, 137Cs and 131I isotopes for the created voxel phantoms with a statistical precision higher than 1%. The PHITS-calculated energy deposits were compared with those for rhesus monkeys. The results showed that the fractions of energy deposits for β-radiation in different organs were almost identical between the two species. For γ-radiation, there was excellent agreement in the self-absorption rate with the approximate curve of the Japanese macaque, with an average deviation of 2%. The maximum deviation of 12% was for the kidney, which has two organs, so the error with the approximate curve is slightly larger due to the energy loss created between organs.

Estimates of external absorbed dose in experimental animals exposed to sprayed neutron-activated 56Mn powder are necessary for comparison with internal absorbed doses estimated under the same exposure conditions, which is required for a correct interpretation of the observed biological effects. It has been established that the measured dose of external absorbed dose as a result of gamma irradiation range 1–15 mGy, which is order of magnitude less than the maximal dose of internal gamma and beta irradiation of the whole body of the same experimental animals irradiated under the same conditions: according to the available literature data, the maximal values ​​of absorbed dose of internal gamma-beta irradiation of the whole body are in the range of 330 mGy–1200 mGy for mice and 100 mGy–150 mGy for rats. It is concluded that under the conditions of experiments with dispersed neutron-activated powder 56MnO2, internal gamma-beta irradiation of experimental animals is the main factor of radiation exposure compared to external gamma irradiation.

The problem of differentiating between primary irradiation and exposure due to residual radioactivity following A-bombing (including beta-exposure), is the subject of special attention and discussions in order to understand the health effects following the Hiroshima and Nagasaki A-bombings, especially among newcomers to cities soon after the detonations. In this work, the method of single quartz grain luminescence retrospective dosimetry was applied for a retrospective estimation of the ‘dose-depth’ profile in a quartz-containing tile extracted from the building of former Hiroshima University (HU), which was a ‘witness’ of the Hiroshima atomic bombing on the 6 August 1945. It has been shown that results of retrospective estimates of the ‘dose-depth’ profile using the method of optically stimulated luminescence (OSL) from inclusions of quartz grains in very thin layers of the sample, in combination with the calculations of the ‘dose-depth’ profile using the Monte Carlo method, indicates the possible presence of beta irradiation of thin layers of the sample located near the surface of the tile facing the air, where there is no electronic equilibrium from gamma radiation.

Electron spin resonance (ESR) dosimetry was applied to human tooth enamel in order to obtain individual absorbed doses for victims of the Hiroshima bomb who lived in the ‘black rain’ area. The so-called ‘black rain’ fell in the form of precipitation on the western part of Hiroshima city and the northwestern suburbs within a few hours after the explosion of the atomic bomb on 6 August 1945, and exposed the population in this area. Only three tooth samples were collected from this area. Since the teeth were located at positions 1, 2 and 4, only the lingual portion was used for the analysis. The results showed that the excess dose after subtracting natural radiation for one (position 4; hh1) was background, for the second (position 2; hh2) it was 133 mGy, and for the other (position 1; hh3) it was 243 mGy. Based on these results, we further investigated the radiation dose attributed to dental X-rays and head CT scan. Such dose of the hh3 radiographic examination was estimated to be 57–160 mGy, which implies an additional exposure around 135 mGy after subtraction. On the other hand, the dose data of hh1 after subtracting dental X-rays was negative. This may mean that such additional doses are an overestimation. In addition, the effect of sunlight should be considered, which is the same direction of overestimation. As a result, the residual dose of 140 mGy suggests the inclusion of radiation from the ‘black rain.’

Manganese-56 (56Mn) was one of the dominant neutron-activated radionuclides during the first hours following the atomic-bombing of Hiroshima and Nagasaki. The radiation spectrum of 56Mn and the radiation emission from excited levels of 56Fe following 56Mn beta-decay include gamma-quanta, beta-particles, Auger electrons and X-rays. The dispersion of neutron activated 56Mn in the air can lead to entering of radioactive microparticles into the lungs. The investigation of spatial microdistribution of an internal dose in biological tissue exposed to 56Mn is an important matter with regards to the possible elevated irradiation of the lung alveoli and alveolar ducts. The Monte Carlo code (MCNP-4C) was used for the calculation of absorbed doses in biological tissue around 56Mn dioxide microparticles. The estimated absorbed dose has a very essential gradient in the epithelium cells of lung alveoli and alveolar duct: from 61 mGy/decay on the surface of simple squamous cells of epithelium to 0.15 mGy/decay at distance of 0.3 μm, which is maximal cell thickness. It has been concluded that epithelial cells of these pulmonary microstructures are selectively irradiated by low-energy electrons: short-range component of beta-particles spectrum and Auger electrons. The data obtained are important for the interpretation of biological experiments implementing dispersed neutron-activated 56Mn dioxide powder.

The aim of this study was to determine parameters for estimating the internal exposure of all organs in mouse experiments from the radioactivity concentration in organs. The estimation of internal exposure rate conversion coefficients and absorbed fractions for 137Cs, 134Cs and 90Sr by the Particle and Heavy Ion Transport code System (PHITS) with a voxel-based mouse phantom is presented. The geometry of the voxel phantom is constructed from computer tomography images of a mouse 9 cm in length weighing 23.9 g. The voxel-based mouse phantom has the following organs: brain, skull, heart, lungs, liver, stomach, spleen, kidneys, bladder, testis and tissue (tissue and other organs). Gamma- and beta-rays from 137Cs, 134Cs and 90Sr sources in each source organ are generated and scored for every target organ. The internal exposure rate conversion coefficients and absorbed fractions are calculated from deposition energies in each target organ from each source organ and are used to generate an internal exposure rate conversion coefficient matrix and an absorbed fraction matrix. The absorbed fractions of beta-rays in the source organs are roughly 0.5–0.8 for 137Cs and 134Cs, and the absorbed fractions of gamma-rays are <0.04 for 137Cs and <0.03 for 134Cs. The internal exposure rate conversion coefficient matrix is defined using the absorbed fractions. The calculated internal exposure rate coefficient matrix is tested under a uniform radioactivity concentration of 1 Bq/kg for 137Cs, 134Cs and 90Sr. The estimated internal exposure rates in the mouse whole body for 137Cs, 134Cs and 90Sr are 3.28 × 10−3, 2.55 × 10−3and 1.20 × 10−2μGy/d, respectively. These values are very similar to those for an ellipsoid frog (31.4 g) and an ellipsoid crab egg mass (12.6 g) reported in ICRP Publication 108.

A method for the determination of 134+137Cs and 90Sr in cesium hot particles (Cs-HPs) originating from the Fukushima Daiichi Nuclear Power Plant accident has been developed. The method depends on a response function that is calculated by PHITS code and fitted to the beta-ray spectrum measured with a Si-detector. The 134+137Cs radioactivity in the Cs-HPs was consistent with that measured by a Ge-detector, thus confirming the reliability of the method. The 90Sr/137Cs ratios, which ranged from 0.001 to 0.0042, were consistent with a 90Sr/137Cs inventory ratio for contaminated soil samples. That is, the extracted Cs-HPs contained 90Sr in the same ratio as that for the 90Sr/137Cs inventory ratio for the contaminated soils. The method is attractive in that the samples are unaltered, andthat no chemical separation techniques are required.

This study aims to demonstrate the feasibility of a method for estimating the strength of a moving brachytherapy source during implantation in a patient. Experiments were performed under the same conditions as in the actual treatment, except for one point that the source was not implanted into a patient. The brachytherapy source selected for this study was 125I with an air kerma strength of 0.332 U (μGym2h-1), and the detector used was a plastic scintillator with dimensions of 10 cm × 5 cm × 5 cm. A calibration factor to convert the counting rate of the detector to the source strength was measured and then the accuracy of the proposed method was investigated for a manually driven source. The accuracy was found to be under 10% when the shielding effect of additional needles for implantation at other positions was corrected, and about 30% when the shielding was not corrected. Even without shielding correction, the proposed method can detect dead/dropped source, implantation of a source with the wrong strength, and a mistake in the number of the sources implanted. Furthermore, when the correction was applied, the achieved accuracy came close to within 7% required to find the Oncoseed 6711 (125I seed with unintended strength among the commercially supplied values of 0.392, 0.462 and 0.533 U).

To estimate the beta- and gamma-ray doses in a brick sample taken from Odaka, Minami-Soma City, Fukushima Prefecture, Japan, a Monte Carlo calculation was performed with Particle and Heavy Ion Transport code System (PHITS) code. The calculated results were compared with data obtained by single-grain retrospective luminescence dosimetry of quartz inclusions in the brick sample. The calculated result agreed well with the measured data. The dose increase measured at the brick surface was explained by the beta-ray contribution, and the slight slope in the dose profile deeper in the brick was due to the gamma-ray contribution. The skin dose was estimated from the calculated result as 164 mGy over 3 years at the sampling site.

ABSTRACTDownward migration of 137Cs in soils was studied in three mixed deciduous forests c.a. 40 km northwest of the Fukushima Daiichi Nuclear power plant (FDNPP), Japan. We selected three different types of forest regarding to environmental condition such as slope inclinations and snow coverage conditions at the time of deposition. We examined temporal changes in the vertical distribution of 137Cs from litter layers to 10 cm soil depth for two years (2.3 to 4.3 years after the FDNPP accident in 2011). At all three study sites, the 137Cs in the litter layer had largely migrated to surface soil by 2013. After 2014, about 80% of the 137Cs in forest soils (litter layer to 10 cm soil depth) remained within 0–5 cm soil layer. The vertical distribution had not changed substantially since 2014, suggesting that changes to the downward migration rates of 137Cs in soils drastically decreased with time. In addition, small amounts of migrating 137Cs could not be detected by the present method because there was a large spatial variation in the distribution of soil 137Cs. The results showed similar patterns of soil 137Cs distribution among the three study sites although there were differences in the environmental conditions.

The calculated risk of cancer in humans due to radiation exposure is based primarily on long-term follow-up studies, e.g. the life-span study (LSS) on atomic bomb (A-bomb) survivors in Hiroshima and Nagasaki. Since A-bomb radiation consists of a mixture of γ-rays and neutrons, it is essential that the relative biological effectiveness (RBE) of neutrons is adequately evaluated if a study is to serve as a reference for cancer risk. However, the relatively small neutron component hampered the direct estimation of RBE in LSS data. To circumvent this problem, several strategies have been attempted, including dose-independent constant RBE, dose-dependent variable RBE, and dependence on the degrees of dominance of intermingled γ-rays. By surveying the available literature, we tested the chromosomal RBE of neutrons as the biological endpoint for its equivalence to the microdosimetric quantities obtained using a tissue-equivalent proportional counter (TEPC) in various neutron fields. The radiation weighting factor, or quality factor, Qn, of neutrons as expressed in terms of the energy dependence of the maximum RBE, RBEm, was consistent with that predicted by the TEPC data, indicating that the chromosomally measured RBE was independent of the magnitude of coexisting γ-rays. The obtained neutron RBE, which varied with neutron dose, was confirmed to be the most adequate RBE system in terms of agreement with the cancer incidence in A-bomb survivors, using chromosome aberrations as surrogate markers. With this RBE system, the cancer risk in A-bomb survivors as expressed in unit dose of reference radiation is equally compatible with Hiroshima and Nagasaki cities, and may be potentially applicable in other cases of human radiation exposure.

The Walker circulation is the key component of the atmospheric zonal circulation in the tropics. In this study, it is shown that anomalous Walker circulations associated with two types of the Indian Ocean Dipole (IOD) are remarkably different. During a positive canonical IOD with negative (positive) sea surface temperature (SST) anomalies in the eastern (central to western) tropical Indian Ocean, a single‐cell anomalous Walker circulation forms over the Indian Ocean. On the other hand, a double‐cell anomalous Walker circulation with a rising branch in the central Indian Ocean is formed during a positive IOD Modoki, which is associated with positive (negative) SST anomalies over the central (eastern and western) tropical Indian Ocean. The above anomalous Walker circulations are found to develop as part of positive ocean‐atmosphere feedback. Furthermore, the above difference in the anomalous Walker circulation may affect the biennial tendency of the IOD. Two flavors of the IOD are associated with different Walker circulation cellsOnly the canonical IOD is associated with significant SSH anomaliesBiennial tendency of the IOD may be influenced by its flavor

Photoactivation of nuclear isomer (115m)In with a halflife of 4.48 h occurs by (60)Co gamma-ray irradiation. This is because the resonance gamma-ray absorption occurs at 1078 keV level for stable (115)In, and that energy gamma-rays are produced by Compton scattering of (60)Co primary gamma-rays. In this work, photoactivation of (115m)In was applied to estimate the dose rate distribution around a (60)Co irradiation source utilizing a standard dose rate taken by alanine dosimeter. The (115m)In photoactivation was measured at 10 to 160 cm from the (60)Co source. The derived dose rate distribution shows a good agreement with both alanine dosimeter data and Monte Carlo simulation. It is found that angular distribution of the dose rate along a circumference at radius 2.8 cm from the central axis shows +/- 10% periodical variation reflecting the radioactive strength of the source rods, but less periodic distribution at radius 10 and 20 cm. The (115m)In photoactivation along the vertical direction in the central irradiation port strongly depends on the height and radius as indicated by Monte Carlo simulation. It is demonstrated that (115m)In photoactivation is a convenient method to estimate the dose rate distribution around a (60)Co source.

Electron spin resonance (ESR) dosimetry of teeth is used extensively for dose estimation following exposure to radiation. The population inhabiting the northeast region of India is prone to different cancers of the head and neck, and their prevalence is several times the national average. The objective of this study was to determine the role of radiation in the causation of this high cancer incidence by performing ESR spectroscopic measurements of tooth samples collected from the general population living in and around the city of Silchar. Nineteen tooth samples were used, and the age of the patients was 13-60 years. The excess dose, determined by subtraction of the natural background dose from the dose absorbed by the enamel, was found to the extent of 123 +/- 43 mGy. However, the individual excess dose was found to be higher in subjects who consumed lime (5/6) than in non-lime-consuming subjects (2/13). It is not entirely clear if radiation is the cause of this excess cancer in this region of India. Therefore there is a need for wider studies including consideration of tobacco consumption as well as a larger number of samples for tooth enamel dosimetry.

Microdosimetry study has been carried out at the education and research mini-reactor of Kinki University (UTR-KINKI) using a tissue equivalent gas proportional counter (TEPC). The microdosimetric single event spectra for 0.5, 1, 2, 3 and 5 microm site sizes were obtained in the lineal energy range from 1 to 1000 keV/microm. Neutron and gamma-ray fractional doses were estimated from the single event spectra. The neutron dose fraction was varied from 35 to 55% for 0.5 to 5 microm site size. The averaged lineal energy, y(D), for each site size was likewise estimated and found to be dependent on the site size. The averaged lineal energy for neutron was slightly larger than that of the fission neutrons from (252)Cf, and the averaged lineal energy for gamma-ray had similar site-size-dependence of 25 keV gamma-rays and 250 kV X-rays. Relative biological effectiveness was found to be 4.1 +/- 0.13 for UTR-KINKI using Tilikidis's 2 Gy-response function. The estimated RBE for UTR-KINKI neutrons is quite close to the previous biological experimental value of 4.3 +/- 0.6 for micronucleated cells in gill cell of Medaka and 4.6 +/- 0.5 for induction of lymphocyte apoptosis in the thymus of ICR mice.

Microdosimetric single event spectra as a function of depth in a phantom for the 290 MeV/nucleon therapeutic carbon beam at HIMAC were measured by using a tissue equivalent proportional counter (TEPC). Two types of geometries were used: one is a fragment particle identification measurement (PID-mode) with time of flight (TOF) method without a backward phantom, and the other is an in-phantom measurement (IPM-mode) with a backward phantom. On the PID-mode geometry, fragments produced by carbon beam in a phantom are identified by the DeltaE-TOF distribution between two scintillation counters positioned up- and down-stream relative to the tissue equivalent proportional counter (TEPC). Lineal energy distributions for carbon and five ion fragments (proton, helium, lithium, beryllium and boron) were obtained in the lineal-energy range of 0.1-1000 keV/microm at eight depths (7.9-147.9 mm) in an acrylic phantom. In the IPM-mode geometry, the total lineal energy distributions measured at eight depths (61.9-322.9 mm) were compared with the distributions in the PID-mode. Both spectra are consistent with each other. This shows that the PID-mode measurement can be discussed as the equivalent of the phantom measurement. The dose distribution of the carbon beam and fragments were obtained separately. In the depth dose curve, the Bragg peak was observed. Relative biological effectiveness (RBE) for the carbon beam in the acrylic phantom was obtained based on a biological response function as a lineal-energy. The RBE of carbon beam had a maximum of 4.5 at the Bragg peak. Downstream of the Bragg peak, the RBE rapidly decreases. The RBE of fragments is dominated by Boron particles around the Bragg peak region.

The effects of solar radiation on the ESR signals in human tooth enamel were investigated. Enamel samples were exposed to the sunlight in Hiroshima, Japan, on sunny days for a total of about 228 hours over one year. The intensity of the illuminating sunlight was measured with a light meter, which then was converted to the intensity of the solar radiation. Three types of signals caused by the solar radiation were identified. The signal named S1 is identical to the gamma ray radiation-induced axially symmetrical signal and contains two signals S1( perpendicular) and S1(//) corresponding to the perpendicular and parallel components of the g-tensor characterized by g( perpendicular) = 2.0018 and g(//) = 1.9975 respectively. In addition, two other peaks named S2 and S3 are induced by the solar radiation. S2 is very near the inherent signal with g = 2.0052, possibly created by the same paramagnetic centers as the natural signal, and S3 is a weaker signal with g = 2.010. On increasing the amount of solar radiation S1 increases linearly, but S2 and S3 reach saturation. The average effect of the solar radiation on the S1( perpendicular) signal is estimated by dose equivalent gamma ray irradiation as 7.8 +/- 0.5 mGy (MJ m(-2))(-1), which corresponds to 19.6 +/- 1.3 mGy h(-1) at the latitude of Hiroshima. Signals S2 and S3 may be used to recognize the effect of solar radiation on the enamel.

A characteristic hot-filament type X-ray generator was constructed for irradiation of cultured cells. The source provides copper K, iron K, chromium K, molybdenum L, aluminium K and carbon K shell characteristic X-rays. When cultured mouse m5S cells were irradiated and frequencies of dicentrics were fitted to a linear-quadratic model, Y = alphaD + betaD2, the chromosomal effectiveness was not a simple function of photon energy. The alpha-terms increased with the decrease of the photon energy and then decreased with further decrease of the energy with an inflection point at around 10 keV. The beta-terms stayed constant for the photon energy down to 10 keV and then increased with further decrease of energy. Below 10 keV, the relative biological effectiveness (RBE) at low doses was proportional to the photon energy, which contrasted to that for high energy X- or gamma-rays where the RBE was inversely related with the photon energy. The reversion of the energy dependency occurred at around 1-2 Gy, where the RBE of soft X-rays was insensitive to X-ray energy. The reversion of energy-RBE relation at a moderate dose may shed light on the controversy on energy dependency of RBE of ultrasoft X-rays in cell survival experiments.

In order to improve the accuracy of the tooth enamel EPR dosimetry method, EPR spectra recording conditions were optimized. The uncertainty of dose determination was obtained as the mean square deviation of doses, determined with the use of a spectra deconvolution program, from the nominal doses for ten enamel samples irradiated in the range from 0 to 500 mGy. The spectra were recorded at different microwave powers and accumulation times. It was shown that minimal uncertainty is achieved at the microwave power of about 2 mW for a used spectrometer JEOL JES-FA100. It was found that a limit of the accumulation time exists beyond which uncertainty reduction is ineffective. At an established total time of measurement, reduced uncertainty is obtained by averaging the experimental doses determined from recorded spectra following intermittent sample shaking and sample tube rotation, rather than from one spectrum recorded at longer accumulation time. The effect of sample mass on the spectrometer's sensitivity was investigated in order to find out how to make appropriate corrections.

The dependence of boron-dose enhancer (BDE) characteristics on dose protocol and 10B concentration was evaluated for BNCT using near-threshold 7Li(p,n)7Be direct neutrons. The treatable protocol depth (TPD) was utilized as an evaluation index. MCNP calculations were performed for near-threshold 7Li(p,n)7Be at a proton energy of 1.900 MeV and for a polyethylene BDE. The effect of dose protocol on BDE characteristics was reflected in terms of the optimum BDE thickness needed for maximum TPD which was found to be independent of the treatable dose but was observed to vary for different combinations of the tolerance doses for heavy charged particles and gamma rays. For the 10B concentration dependence, the TPD was increased by increasing the T/N ratio, i.e., the ratio of the 10B concentration in the tumour (10BTumour) to that in the normal tissue (10BNormal), and by increasing 10BTumour and 10BNormal at constant T/N ratio. It was found that the use of BDE becomes unnecessary from the viewpoint of increasing the TPD, when 10BTumour is over a certain level which is decided by the conditions of the dose protocol.

Following whole-body irradiation of ICR mice with various doses of fission neutrons or X-rays, the frequency of micronuclei (MNs) in peripheral blood reticulocytes was measured at 12h intervals beginning immediately after irradiation and ending at 72h after irradiation. The resulting time-course curve of MN frequency had a clear peak 36h after irradiation, irrespective of the type of radiation applied and the dose used. The MN frequency, averaged as the unweighted mean over the experimental time course, showed a linear increase with increasing dose of either fission neutrons or X-rays. The linear response to X-rays supports reported conclusion that induction of MN formation in reticulocytes is a dose-rate independent phenomenon. The relative biological effectiveness (RBE) of fission neutrons to X-rays for MN induction was estimated to be 1.9 ± 0.3. This value is considerably lower than the RBE value of 4.6 ± 0.5 reported for the same fission neutrons for induction of lymphocyte apoptosis in the thymus of ICR mice that represents dose-rate independent, one-track event. Based on these results, we propose that MNs increased in reticulocytes after irradiation mostly represent acentric fragments caused by single chromosome breaks, and that some confounding factor is operating in erythroblasts for the formation of aberrations from non-rejoining DNA double-strand breaks more severely after high-LET radiation than after low-LET radiation.

Neutron dose is transferred to biological materials through the recoil protons produced by elastic scattering. When a low-velocity proton collides with the atoms or molecules of a target, it changes to a hydrogen atom by electron capture; this hydrogen atom then changes to a proton by losing the electron. Because the hydrogen atom has a different ionization cross section from that of a proton, the charge exchange processes need to be considered to calculate stopping power for low energy protons. The proton neutralization effect has been simulated by using a proton track structure code developed by taking into account charge exchange processes. The microdosimetric spectrum for 1 MeV neutrons was calculated by assuming a continuous slowing down approximation (csda) and the results of the proton track code. It was found that hydrogen atoms after proton neutralized by electron capture contribute about 24% to neutron dose.

The relative biological effectiveness (RBE) of various energy neutrons produced from a Schenkel-type accelerator at the Research Institute for Radiation Biology and Medicine, Hiroshima University (HIRRAC), compared with 60Co gamma-ray radiation was determined. The neutron radiations and gamma-ray radiation produced good linear changes in the frequency of micronuclei induced in the root-tip cells of Allium cepa onion irradiated as dry dormant seeds (seed assay) and seedlings (seedling assay) with varying radiation doses. Therefore the RBE for radiation-induced micronuclei can be calculated as the ratio of the slopes of the fitted linear dose response for the neutron radiations and the 60Co gamma-ray radiation. The RBE values by seed assay and seedling assay decreased to 174 +/- 7, from 216 +/- 9, and to 31.4 +/- 1.0, from 45.3 +/- 1.3 (one standard error), respectively, when neutron energies increased to 1.0 MeV, from 0.2 MeV, in the present study. Furthermore, the ratio of the micronucleus induction rates of seed assay to seedling assay by gamma-ray radiation was much lower than that by neutron radiations.

Residual 152Eu activities induced by neutrons from the Nagasaki atomic bomb were measured for nine mineral samples located up to 1,061 m in the slant range and one control sample at 2,850 m from the hypocenter. A chemical separation to prepare europium-enriched samples was performed for all samples, and gamma ray measurements were carried out with a low background well-type germanium detector. In this paper, the measured specific activities of 152Eu are compared with activation calculations based on the DS86 neutron fluence and the 93Rev one. The calculated-to-measured ratios are also compared with those of 60Co and 36Cl. The present results indicate that the measurements agree to the calculation within a factor of three as observed in the nuclear tests at Nevada. The activation level of environmental neutrons and the detection limit for 152Eu are also discussed.

The relative biological effectiveness (RBE) of mixed neutron and gamma-ray radiation emitted at a 252Cf source at the Research Institute for Radiation Biology and Medicine, Hiroshima University, compared with 60Co gamma-ray radiation was determined. The tissue-absorbed dose contribution of the accompanying gamma radiation was about 35.7% to the total tissue-absorbed dose from the 252Cf mixed radiation. The 252Cf mixed radiation and 60Co gamma rays produced approximate linear changes in the frequency of micronuclei induced in root-tip cells of Allium cepa L. onion seedlings after irradiation as dry dormant seeds with varying absorbed doses in onion seeds. Therefore, the RBE for radiation-induced micronuclei was calculated as the ratio of the slopes for the 252Cf mixed radiation and the 60Co gamma rays. The deduced RBE value of 252Cf mixed radiation to 60Co gamma rays to induce micronuclei in dry dormant onion seed cells was about 90.5 +/- 3.6 (+/- 1sigma); the RBE of neutrons from the 252Cf mixed radiation was about 150 +/- 6 (+/- 1sigma). Furthermore, the sensitivity ratio of the induction rate of micronuclei in dry dormant seeds to that in seedlings by neutrons from 252Cf mixed radiation was significantly different from that by 60Co gamma rays. From these results, we concluded that the repair efficiency of DNA damage induced by neutrons may be different from that by gamma rays.

Residual 60Co activity in five steel samples induced by neutrons from the Nagasaki atomic bomb has been measured within about 1000 m from the hypocenter. The chemical separation of cobalt and nickel from steel samples was performed, and cobalt-enriched samples were prepared for all samples. Gamma-ray measurements were carried out with a low-background well-type germanium detector. The gamma-ray spectra for five samples were compared with the spectrum of a control sample to ensure that the observed 60Co was actually induced by A-bomb neutrons. The activation of cobalt by environmental neutrons was also investigated. It has been shown that the present 60Co data are consistent with earlier Hashizume's data.

The energy spectrum of fission neutrons in the biological irradiation field of the Kinki University reactor, UTR-KINKI, has been determined by a multi-foil activation analysis coupled with artificial neural network techniques and a Au-foil activation method. The mean neutron energy was estimated to be 1.26 +/- 0.05 MeV from the experimentally determined spectrum. Based on this energy value and other information, the neutron dose rate was estimated to be 19.7 +/- 1.4 cGy/hr. Since this dose rate agrees with that measured by a pair of ionizing chambers (21.4 cGy/hr), we conclude that the mean neutron energy could be estimated with reasonable accuracy in the irradiation field of UTR-KINKI.

The aim of the present study was to examine the excess relative risk for leukemia mortality and all cancers, except leukemia, among Hiroshima atomic-bomb survivors by applying ABS93D and ABS2000D. Particular attention was given to any difference in the neutron-dose estimates between the two dosimetry systems. The study subjects were 51,532 atomic-bomb survivors registered in a database of the Research Institute for Radiation Biology and Medicine of Hiroshima University (RIRBM). The results obtained by both dosimetry systems are similar: the excess relative risk per Sv for leukemia mortality and all cancers except leukemia is significantly higher compared to the control group. In addition, the difference in the excess relative risks between the two systems is not significant. Therefore, it is indicated that a modification of the neutron-dose estimates would not markedly change the conclusions about the cancer mortality risk.

An 800-kb region on chromosome 8q21, which complements the phenotype of cells from Nijmegen breakage syndrome patients, is a candidate for the locus of the underlying gene, termedNBS1.The sequence of this 800-kb region of DNA indicated that the size of this segment is 755,832 bp with an additional 36-kb gap. From this region, we identified four genes includingNBS1,a gene coding for a 27-kDa vitamin D-dependent calcium-binding protein (27-kDa calbindin), the mitochondrial 2,4-dienoyl-CoA reductase gene, and a novel gene,C8orf1/hT41.All four genes were aligned in a 250-kb centromeric portion of the region, and no gene was found in the remaining telomeric portion containing 500 kb. The genomic organization of theC8orf1/hT41andNBS1genes has been analyzed using the computer programs GRAIL 2 and GENSCAN. They predicted and successfully found more than 93% of the exons, even a small 54-bp exon, indicating that one or more exons in any gene can be identified by these programs. GENSCAN was more efficient at locating the four genes than GRAIL 2 and identified 15 of the 16 exons of theNBS1gene. This 800-kb region contained repetitive sequences, including 179 copies of theAlusequence (1 copy/4.2 kb), 123 copies of the L1 sequence (1 copy/6.1 kb), 107 copies of the LTR sequence (1 copy/7.1 kb), and 63 copies of the MER sequence (1 copy/12 kb). There was a slight but not significant difference in the repetitive content of the gene-rich region and the remaining noncoding region. Our results indicate that computer-assisted methods are useful and powerful for identifying exons of both known and novel genes.

In this report, we have reviewed the basic features of the accident processes and radioactivity releases that occurred in the Chernobyl accident (1986) and in the Fukushima-1 accident (2011). The Chernobyl accident was a power-surge accident that was caused by a failure of control of a fission chain reaction, which instantaneously destroyed the reactor and building, whereas the Fukushima-1 accident was a loss-of-coolant accident in which the reactor cores of three units were melted by decay heat after losing the electricity supply. Although the quantity of radioactive noble gases released from Fukushima-1 exceeded the amount released from Chernobyl, the size of land area severely contaminated by 137Cesium (137Cs) was 10 times smaller around Fukushima-1 compared with around Chernobyl. The differences in the accident process are reflected in the composition of the discharged radioactivity as well as in the composition of the ground contamination. Volatile radionuclides (such as 132Te-132I, 131I, 134Cs and 137Cs) contributed to the gamma-ray exposure from the ground contamination around Fukishima-1, whereas a greater variety of radionuclides contributed significantly around Chernobyl. When radioactivity deposition occurred, the radiation exposure rate near Chernobyl is estimated to have been 770 μGy h−1 per initial 137Cs deposition of 1000 kBq m−2, whereas it was 100 μGy h−1 around Fukushima-1. Estimates of the cumulative exposure for 30 years are 970 and 570 mGy per initial deposition of 1000 kBq m−2 for Chernobyl and Fukusima-1, respectively. Of these exposures, 49 and 98% were contributed by radiocesiums (134Cs + 137Cs) around Chernobyl and Fukushima-1, respectively.

A large amount of the fission products released by the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident on 11 March 2011 was deposited in a wide area from Tohoku to northern Kanto. A map of the estimated cumulative β-ray dose (70 μm dose equivalent) on the soil surface for one year after the FDNPP accident has been prepared using previously reported calculation methods and the 2-km mesh survey data by MEXT. From this map of estimated dose, areas with a high cumulative β-ray dose on the soil surface for one year after the FDNPP accident were found to be located in the Akogi-Teshichiro to Akogi-Kunugidaira region in Namie Town, and in the southern Futaba Town to the northern Tomioka Town region. The highest estimated cumulative β-ray dose was 710 mSv for one year at Akogi-Teshichiro, Namie Town.

The present study continues our previous research on investigating the biological effects of low-level gamma radiation in rice at the heavily contaminated Iitate village in Fukushima, by extending the experiments to unraveling the leaf proteome. 14-days-old plants of Japonica rice (Oryza sativaL. cv. Nipponbare) were subjected to gamma radiation level of upto 4 µSv/h, for 72 h. Following exposure, leaf samples were taken from the around 190 µSv/3 d exposed seedling and total proteins were extracted. The gamma irradiated leaf and control leaf (harvested at the start of the experiment) protein lysates were used in a 2-D differential gel electrophoresis (2D-DIGE) experiment using CyDye labeling in order to asses which spots were differentially represented, a novelty of the study. 2D-DIGE analysis revealed 91 spots with significantly different expression between samples (60 positive, 31 negative). MALDI-TOF and TOF/TOF mass spectrometry analyses revealed those as comprising of 59 different proteins (50 up-accumulated, 9 down-accumulated). The identified proteins were subdivided into 10 categories, according to their biological function, which indicated that the majority of the differentially expressed proteins consisted of the general (non-energy) metabolism and stress response categories. Proteome-wide data point to some effects of low-level gamma radiation exposure on the metabolism of rice leaves.

This study sought to demonstrate the feasibility of estimating the source strength during implantation in brachytherapy. The requirement for measuring the strengths of the linked sources was investigated. The utilized sources were 125I with air kerma strengths of 8.38–8.63 U (μGy m2 h–1). Measurements were performed with a plastic scintillator (80 mm × 50 mm × 20 mm in thickness). For a source-to-source distance of 10.5 mm and at source speeds of up to 200 mm s–1, a counting time of 10 ms and a detector-to-needle distance of 5 mm were found to be the appropriate measurement conditions. The combined standard uncertainty (CSU) with the coverage factor of 1 (k = 1) was ∼15% when using a grid to decrease the interference by the neighboring sources. Without the grid, the CSU (k = 1) was ∼5%, and an 8% overestimation due to the neighboring sources was found to potentially cause additional uncertainty. In order to improve the accuracy in estimating source strength, it is recommended that the measurment conditions should be optimized by considering the tradeoff between the overestimation due to the neighboring sources and the intensity of the measured value, which influences the random error.

Following the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident of 2011, which deposited radionuclides across Tohoku and northern Kanto, β-ray dose evaluation has been performed to estimate radiation exposure for small creatures like insects as well as human skin. Using the Monte Carlo radiation transport code MCNP-4C, we calculated the β-ray dose for 129mTe, 129Te, 131I, 132Te, 132I, 134Cs and 137Cs in air as a function of altitude and in soil. These calculations of β-dose rate for each radionuclide were conducted for the conditions following the FDNPP accident, with 137Cs deposition assumed to be 1000 kBq/m2. Beta-ray dose rate was found to be ∼10-fold (resp. 5-fold) higher than the γ-ray dose rate in the soil (resp. on the ground surface) at ∼20 days after deposition, and ∼4-fold (resp. 1.7-fold) higher after 6 months or more. For convenience, the height dependence of the ratio for 0, 10, 30, 90, 180 and 365 days after deposition was obtained by a fitting function. The cumulative 70 µm β-ray dose at 30, 60 and 90 days after deposition was estimated to be 35, 45 and 53 mGy for the ground surface, and 61, 79 and 92 mGy in the soil, respectively. These results can be used to estimate the external β-ray exposure for small creatures as well as for human skin.

The aim of this study is to demonstrate the feasibility of estimating the strength of the moving radiation source during patient implantation. The requirement for the counting time was investigated by comparing the results of the measurements for the static source with those for the source moving at 2, 5, 10 and 20 cm s−1. The utilized source was 125I with an air-kerma strength of 0.432 U (μGym2h–1). The detector utilized was a plastic scintillation detector (8 cm × 5 cm × 2 cm in thickness) set at 8 cm away from the needle to guide the source. Experiments were conducted in order to determine the most desirable counting time. Analysis using the maximum of the measured values while the source passed through the needle indicated that the results for the moving source increased more than those for the static source as the counting time decreased. The combined standard uncertainty, with the coverage factor of 1, was within 4% at the counting time of 100 ms. This investigation supported the feasibility of the method proposed for estimating the source strength during the implantation procedure, regardless of the source speed. The method proposed is a potential option for reducing the risk of accidental replacements of sources with those of incorrect strengths.