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.

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.

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.

Design study of multi-imaging plate system for BNCT irradiation field at Kyoto university reactor.

The converter configuration for a multi-imaging plate system was investigated for the application of quality assurance in the irradiation field profile for boron neutron capture therapy. This was performed by the simulation calculation using the PHITS code in the fields at the Heavy Water Neutron Irradiation Facility of Kyoto University Reactor. The converter constituents investigated were carbon for gamma rays, and polyethylene with and without LiF at varied (6)Li concentration for thermal, epithermal, and fast neutrons. Consequently, potential combinations of the converters were found for two components, gamma rays and thermal neutrons, for the standard thermal neutron mode and three components of gamma rays, epithermal neutrons, and thermal or fast neutrons, for the standard mixed or epithermal neutron modes, respectively.

Apparatus development for measurement of (134)Cs and (137)Cs radioactivity of soil contaminated by the Fukushima Daiichi Nuclear Power Plant accident.

We developed an apparatus containing a NaI(Tl) scintillator to measure the (134)Cs and (137)Cs radioactivity of soil contaminated by the Fukushima Daiichi Nuclear Power Plant accident. The unfolding method with the least-squares technique was used to determine the radioactivity. Detector responses for each radionuclide in soil were calculated with EGS5 code for the unfolding method. The radionuclides that were measured were (40)K, (134)Cs, (137)Cs, (208)Tl, (214)Bi, and (228)Ac. The measured spectrum agreed well with the spectrum calculated from the response matrix and measured radioactivities. The unfolding method allows us to use the NaI(Tl) scintillator despite the overlap of peaks.

Mapping of the cumulative ß-ray dose on the ground surface surrounding the Fukushima area.

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 ß-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 ß-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 ß-ray dose was 710 mSv for one year at Akogi-Teshichiro, Namie Town.

Measurement of spatial distribution of neutrons and gamma rays for BNCT using multi-imaging plate system.

Quality assurance of the spatial distributions of neutrons and gamma rays was tried using imaging plates (IPs) and converters to enhance the beam components in the epithermal neutron mode of the Kyoto University Reactor. The converters used were 4mm thick epoxy resin with B4C at 6.85 weight-percent (wt%) (10)B for epithermal neutrons, and 3mm thick carbon for gamma rays. Results suggested that the IP signal does not need a sensitivity correction regardless of the incident radiation that produces it.

Calculation of coincidence summing in gamma-ray spectrometry with the EGS5 code.

Coincidence summing correction factors, including not only γ-γ and X-γ, but also ß-γ and conversion electron-γ, coincidences were calculated by Monte Carlo simulation of decay schemes by using the EGS5 code. A large contribution from ß-rays to the factors was found for 24Na for the n-type detector in the case of close source-to-detector geometry with a point source.

Estimation of ß-ray dose in air and soil from Fukushima Daiichi Power Plant accident.

Following the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident of 2011, which deposited radionuclides across Tohoku and northern Kanto, ß-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 ß-ray dose for (129m)Te, (129)Te, (131)I, (132)Te, (132)I, (134)Cs and (137)Cs in air as a function of altitude and in soil. These calculations of ß-dose rate for each radionuclide were conducted for the conditions following the FDNPP accident, with (137)Cs deposition assumed to be 1000 kBq/m(2). 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 ß-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 ß-ray exposure for small creatures as well as for human skin.

Paddy-field contamination with 134Cs and 137Cs due to Fukushima Dai-ichi Nuclear Power Plant accident and soil-to-rice transfer coefficients.

The transfer coefficient (TF) from soil to rice plants of (134)Cs and (137)Cs in the form of radioactive deposition from the Fukushima Dai-ichi Nuclear Power Plant (FDNPP) accident in March 2011 was investigated in three rice paddy fields in Minami-Soma City. Rice crops were planted in the following May and harvested at the end of September. Soil cores of 30-cm depth were sampled from rice-planted paddy fields to measure (134)Cs and (137)Cs radioactivity at 5-cm intervals. (134)Cs and (137)Cs radioactivity was also measured in rice ears (rice with chaff), straws and roots. The rice ears were subdivided into chaff, brown rice, polished rice and rice bran, and the (134)Cs and (137)Cs radioactivity concentration of each plant part was measured to calculate the respective TF from the soil. The TF of roots was highest at 0.48 ± 0.10 in the field where the (40)K concentration in the soil core was relatively low, in comparison with TF values of 0.31 and 0.38 in other fields. Similar trends could be found for the TF of whole rice plants, excluding roots. The TF of rice ears was relatively low at 0.019-0.026. The TF of chaff, rice bran, brown rice and polished rice was estimated to be 0.049, 0.10-0.16, 0.013-0.017 and 0.005-0.013, respectively.