Comparison of external dose estimates using different retrospective dosimetry methods in the settlements located near Semipalatinsk Nuclear Test Site, Republic of Kazakhstan.

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.

Internal organ dose rate conversion coefficients of Japanese macaques to 134Cs,137Cs and 131I†.

The purpose of this study was to estimate the internal dose of radiation in Japanese macaques (aka Nihonzaru or 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 ß-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.

Microdistribution of internal radiation dose in biological tissues exposed to 56Mn dioxide microparticles.

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.

Tooth enamel ESR dosimetry for Hiroshima 'black rain' zone residents.

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.'

Estimation of 'dose-depth' profile in the surface layers of a quartz-containing tile from the former Hiroshima University building indicates the possible presence of beta-irradiation from residual radioactivity after A-bombing.

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.

External dose estimates of laboratory rats and mice during exposure to dispersed neutron-activated 56Mn powder.

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.

Development of a real-time neutron beam detector for boron neutron capture therapy using a thin silicon sensor.

We have developed a new real-time neutron detector, which is able to measure a direct neutron beam of boron neutron capture therapy. The detector consists of both a 40-µm-thick pn diode and around 0.09-µm-thick LiF neutron converter. Experimental results indicate that this neutron detector can measure neutron flux up to 1 × 109 (cm-2 s-1), separately from gamma rays around 500 mGy/h. The measured depth distribution of neutron flux in an acrylic block is in agreement with the activation results of gold.

Internal doses in experimental mice and rats following exposure to neutron-activated 56MnO2 powder: results of an international, multicenter study.

The experiment was performed in support of a Japanese initiative to investigate the biological effects of irradiation from residual neutron-activated radioactivity that resulted from the A-bombing. Radionuclide 56Mn (T1/2 = 2.58 h) is one of the main neutron-activated emitters during the first hours after neutron activation of soil dust particles. In our previous studies (2016-2017) related to irradiation of male Wistar rats after dispersion of 56MnO2 powder, the internal doses in rats were found to be very inhomogeneous: distribution of doses among different organs ranged from 1.3 Gy in small intestine to less than 0.0015 Gy in some of the other organs. Internal doses in the lungs ranged from 0.03 to 0.1 Gy. The essential pathological changes were found in lung tissue of rats despite a low level of irradiation. In the present study, the dosimetry investigations were extended: internal doses in experimental mice and rats were estimated for various activity levels of dispersed neutron-activated 56MnO2 powder. The following findings were noted: (a) internal radiation doses in mice were several times higher in comparison with rats under similar conditions of exposure to 56MnO2 powder. (b) When 2.74 × 108 Bq of 56MnO2 powder was dispersed over mice, doses of internal irradiation ranged from 0.81 to 4.5 Gy in the gastrointestinal tract (small intestine, stomach, large intestine), from 0.096 to 0.14 Gy in lungs, and doses in skin and eyes ranged from 0.29 to 0.42 Gy and from 0.12 to 0.16 Gy, respectively. Internal radiation doses in other organs of mice were much lower. (c) Internal radiation doses were significantly lower in organs of rats with the same activity of exposure to 56MnO2 powder (2.74 × 108 Bq): 0.09, 0.17, 0.29, and 0.025 Gy in stomach, small intestine, large intestine, and lungs, respectively. (d) Doses of internal irradiation in organs of rats and mice were two to four times higher when they were exposed to 8.0 × 108 Bq of 56MnO2 (in comparison with exposure to 2.74 × 108 Bq of 56MnO2). (e) Internal radiation doses in organs of mice were 7-14 times lower with the lowest 56MnO2 amount (8.0 × 107 Bq) in comparison with the highest amount, 8.0 × 108 Bq, of dispersed 56MnO2 powder. The data obtained will be used for interpretation of biological effects in experimental mice and rats that result from dispersion of various levels of neutron-activated 56MnO2 powder, which is the subject of separate studies.

A simulation study on beam property of 124Sb-Be isotope-based neutron for BNCT.

The 9Be(γ,n) neutrons with the energies at 21-24 keV generated by 1.691 MeV photons from 124Sb was investigated as a source of epithermal neutrons for BNCT, using PHITS code. A beam shaping assembly composed of 13 mm thick Be target, a gamma ray shield made of 30 cm thick Bi, and a reflector of 30 cm thick Pb satisfied the beam requirements of IAEA-TECDOC-1223. The needed 124Sb activity was estimated in the order of 1016-1017 Bq. Feasibility of BNCT using 124Sb-Be neutrons would be influenced by the capability of periodic supply of short-lived 124Sb (half-life: 60 days) with such high activity.

Internal exposure rate conversion coefficients and absorbed fractions of mouse for 137Cs, 134Cs and 90Sr contamination in body.

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 × 10-3 and 1.20 × 10-2 µGy/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.

Determination of radiocesium depth profile by unfolding method with imaging plate.

Radiocesium depth profiles in Fukushima soil cores were derived with an unfolding method by using an imaging plate with a response matrix for 134Cs, 137Cs, and 40K radionuclides calculated by a PHITS simulation. The unfolded depth profiles were validated by comparison with profiles measured with a Ge detector. The unfolded depth profiles agreed well with those measured by the Ge detector. The depth profiles obtained were obtained with a 1 mm depth bin width and are useful for estimating the migration of radiocesium in soil.

Measurement of 90Sr radioactivity in cesium hot particles originating from the Fukushima Nuclear Power Plant Accident.

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.

An in vitro verification of strength estimation for moving an 125I source during implantation in brachytherapy.

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).

Comparison of calculated beta- and gamma-ray doses after the Fukushima accident with data from single-grain luminescence retrospective dosimetry of quartz inclusions in a brick sample.

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.

Computational investigation of suitable polymer gel composition for the QA of the beam components of a BNCT irradiation field.

This study investigated the optimum composition of the MAGAT polymer gel which is to be used in the quality assurance measurement of the thermal neutron, fast neutron and gamma ray components in the irradiation field used for boron neutron capture therapy at the Kyoto University Reactor. Simulations using the PHITS code showed that when combined with the gel, 6Li concentrations of 0, 10 and 100ppm were found to be potentially usable.

Determination of the thermal and epithermal neutron sensitivities of an LBO chamber.

An LBO (Li2B4O7) walled ionization chamber was designed to monitor the epithermal neutron fluence in boron neutron capture therapy clinical irradiation. The thermal and epithermal neutron sensitivities of the device were evaluated using accelerator neutrons from the 9Be(d, n) reaction at a deuteron energy of 4 MeV (4 MeV d-Be neutrons). The response of the chamber in terms of the electric charge induced in the LBO chamber was compared with the thermal and epithermal neutron fluences measured using the gold-foil activation method. The thermal and epithermal neutron sensitivities obtained were expressed in units of pC cm2, i.e., from the chamber response divided by neutron fluence (cm-2). The measured LBO chamber sensitivities were 2.23 × 10-7 ± 0.34 × 10-7 (pC cm2) for thermal neutrons and 2.00 × 10-5 ± 0.12 × 10-5 (pC cm2) for epithermal neutrons. This shows that the LBO chamber is sufficiently sensitive to epithermal neutrons to be useful for epithermal neutron monitoring in BNCT irradiation.

Measurement of the gamma-ray energy spectrum of the educational Kinki University Reactor (UTR-KINKI).

The gamma-ray energy spectrum of the Kinki University Reactor (UTR-KINKI) was estimated from Ge detector measurements combined with Monte Carlo N-particle transport criticality calculations. The gamma rays mainly originated from prompt fission components, although small amounts of gamma rays from (n,γ) reactions, fission product gamma rays, and activation gamma rays were detected. The averaged gamma-ray tissue kerma rate in the irradiation port during UTR-KINKI operation at 1W was calculated as 10.5cGy/h based on the estimated gamma-ray energy spectrum. This value is consistent with a previous measurement with paired ionization chambers and a tissue equivalent gas proportional counter. This result demonstrates the reliability of the estimated gamma-ray energy spectrum.

Internal exposure to neutron-activated 56Mn dioxide powder in Wistar rats: part 1: dosimetry.

There were two sources of ionizing irradiation after the atomic bombings of Hiroshima and Nagasaki: (1) initial gamma-neutron irradiation at the moment of detonation and (2) residual radioactivity. Residual radioactivity consisted of two components: radioactive fallout containing fission products, including radioactive fissile materials from nuclear device, and neutron-activated radioisotopes from materials on the ground. The dosimetry systems DS86 and DS02 were mainly devoted to the assessment of initial radiation exposure to neutrons and gamma rays, while only brief considerations were given for the estimation of doses caused by residual radiation exposure. Currently, estimation of internal exposure of atomic bomb survivors due to dispersed radioactivity and neutron-activated radioisotopes from materials on the ground is a matter of some interest, in Japan. The main neutron-activated radionuclides in soil dust were 24Na, 28Al, 31Si, 32P, 38Cl, 42K, 45Ca, 46Sc, 56Mn, 59Fe, 60Co, and 134Cs. The radionuclide 56Mn (T 1/2 = 2.58 h) is known as one of the dominant beta- and gamma emitters during the first few hours after neutron irradiation of soil and other materials on ground, dispersed in the form of dust after a nuclear explosion in the atmosphere. To investigate the peculiarities of biological effects of internal exposure to 56Mn in comparison with external gamma irradiation, a dedicated experiment with Wistar rats exposed to neutron-activated 56Mn dioxide powder was performed recently by Shichijo and coworkers. The dosimetry required for this experiment is described here. Assessment of internal radiation doses was performed on the basis of measured 56Mn activity in the organs and tissues of the rats and of absorbed fractions of internal exposure to photons and electrons calculated with the MCNP-4C Monte Carlo using a mathematical rat phantom. The first results of this international multicenter study show that the internal irradiation due to incorporated 56Mn powder is highly inhomogeneous, and that the most irradiated organs of the experimental animals are: large intestine, small intestine, stomach, and lungs. Accumulated absorbed organ doses were 1.65, 1.33, 0.24, 0.10 Gy for large intestine, small intestine, stomach, and lungs, respectively. Other organs were irradiated at lower dose levels. These results will be useful for interpretation of the biological effects of internal exposure of experimental rats to powdered 56Mn as observed by Shichijo and coworkers.

Triple ionization chamber method for clinical dose monitoring with a Be-covered Li BNCT field.

PURPOSE: Fast neutron, gamma-ray, and boron doses have different relative biological effectiveness (RBE). In boron neutron capture therapy (BNCT), the clinical dose is the total of these dose components multiplied by their RBE. Clinical dose monitoring is necessary for quality assurance of the irradiation profile; therefore, the fast neutron, gamma-ray, and boron doses should be separately monitored. To estimate these doses separately, and to monitor the boron dose without monitoring the thermal neutron fluence, the authors propose a triple ionization chamber method using graphite-walled carbon dioxide gas (C-CO2), tissue-equivalent plastic-walled tissue-equivalent gas (TE-TE), and boron-loaded tissue-equivalent plastic-walled tissue-equivalent gas [TE(B)-TE] chambers. To use this method for dose monitoring for a neutron and gamma-ray field moderated by D2O from a Be-covered Li target (Be-covered Li BNCT field), the relative sensitivities of these ionization chambers are required. The relative sensitivities of the TE-TE, C-CO2, and TE(B)-TE chambers to fast neutron, gamma-ray, and boron doses are calculated with the particle and heavy-ion transport code system (PHITS). METHODS: The relative sensitivity of the TE(B)-TE chamber is calculated with the same method as for the TE-TE and C-CO2 chambers in the paired chamber method. In the Be-covered Li BNCT field, the relative sensitivities of the ionization chambers to fast neutron, gamma-ray, and boron doses are calculated from the kerma ratios, mass attenuation coefficient tissue-to-wall ratios, and W-values. The Be-covered Li BNCT field consists of neutrons and gamma-rays which are emitted from a Be-covered Li target, and this resultant field is simulated by using PHITS with the cross section library of ENDF-VII. The kerma ratios and mass attenuation coefficient tissue-to-wall ratios are determined from the energy spectra of neutrons and gamma-rays in the Be-covered Li BNCT field. The W-value is calculated from recoil charged particle spectra by the collision of neutrons and gamma-rays with the wall and gas materials of the ionization chambers in the gas cavities of TE-TE, C-CO2, and TE(B)-TE chambers (10B concentrations of 10, 50, and 100 ppm in the TE-wall). RESULTS: The calculated relative sensitivity of the C-CO2 chamber to the fast neutron dose in the Be-covered Li BNCT field is 0.029, and those of the TE-TE and TE(B)-TE chambers are both equal to 0.965. The relative sensitivities of the C-CO2, TE-TE, and TE(B)-TE chambers to the gamma-ray dose in the Be-covered Li BNCT field are all 1 within the 1% calculation uncertainty. The relative sensitivities of TE(B)-TE to boron dose with concentrations of 10, 50, and 100 ppm 10B are calculated to be 0.865 times the ratio of the in-tumor to in-chamber wall boron concentration. CONCLUSIONS: The fast neutron, gamma-ray, and boron doses of a tumor in-air can be separately monitored by the triple ionization chamber method in the Be-covered Li BNCT field. The results show that these doses can be easily converted to the clinical dose with the depth correction factor in the body and the RBE.

Neutron relative biological effectiveness in Hiroshima and Nagasaki atomic bomb survivors: a critical review.

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.