Search

The double differential cross sections (DDXs) of neutron production by 135 and 180 MeV protons were measured to obtain benchmark data for calculating neutron doses in proton therapy. The incident proton energies correspond to those of protons decelerating in a patient. An A-150 tissue-equivalent plastic was used as the target. The measured DDXs were compared with those calculated using PHITS, FLUKA, and Geant4. The calculated DDXs were found to be underestimated compared with the measured ones, especially at forward angles. In proton therapy, exposure evaluations using one of these simulation codes would underestimate the neutron dose.

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

A deep-penetration shielding experiment was performed at the CERN High-energy AcceleRator Mixed-field (CHARM) facility. The protons (24 GeV/c) were injected into a 50-cm-thick copper target and the...

The results of seasonal Al and Mn activity on the surface of the atmosphere above the city of Nur-Sultan are shown from October 2016 to January 2017. The technique of sampling aerosols in various fractions and further analysis of the data are described. Together with scientists from the University of Hiroshima (Hiroshima, Japan) and the University of Tsukuba (Tsukuba, Japan), the composition of aerosols in the air of Astana was monitored. The aim of the project was to develop a methodology for sampling aerosols in various fractions and conduct measurements on an ongoing basis. The studies were carried out using a high-volume air sampler and a cascade of impactors that measure the size distribution and the respirable mass fraction of airborne particles of the environment. Fiberglass filters, which are a commonly used material for sample collection, were also used. The data obtained in the study of aerosol samples using a cascade of impactors allowing the selection of aerosols with sizes up to 0.49 μm showed the content of aluminum isotope in the atmosphere of the city. The smaller the clearance gap, the more particles are trapped. On average, aerosols were taken from more than 200 thousand cubic meters of air.

The influence from uncertainty on light output calibration was studied experimentally for neutron energy spectra up to 300 MeV obtained by an unfolding method using a liquid organic scintillator. The neutrons were generated from the interactions of 100 and 290 MeV/u 28Si ion beams on thick silicon targets at a 75° direction with respect to the beam axis. The light output calibration determined by using Compton edges of γ-rays from 60Co and 241Am-Be sources was used for the unfolding method. The uncertainty of the calibration points at the Compton edges was estimated to be within 3%. Two calibration lines which connected the 3% larger light output point of the Compton edge of the γ-ray from 60Co and that the 3% lower light output point of that from 241Am-Be and vice versa were examined for influence on variation of calibration line. The unfolded spectra using the two calibration lines were compared with one using the calibration line connecting light output points of Compton edges from both γ-ray sources. The comparison indicates the uncertainty of calibration line influences the neutron light output spectrum within order of several %. The all unfolded spectra generally reproduced one by the time-of-flight (TOF) method in the same experiment. The difference between TOF and unfolded spectra were 17 and 8% for 100 and 290 MeV/u data, at maximum.

A measurement of high-energy neutron streaming was performed through a maze at the CERN (Conseil Européen pour la Recherche Nucléaire) High-energy AcceleRator Mixed-field (CHARM) facility. The protons of 24 GeV/c were injected onto a 50-cm-thick copper target and the released neutrons were streamed through a maze with several corridor-legs horizontally designed with the shield walls in the facility. Streaming neutrons were measured by using aluminum activation detectors placed at 10 locations in the maze. From the radionuclide production rate in the activation detectors, the attenuation profile along the maze was obtained for the reaction of 27Al(n,α)24Na. Monte Carlo simulations performed with two codes, the Particle and Heavy Ion Transport System (PHITS) and CERN FLUktuierende KAskade (FLUKA), gave good agreements with the measurements within a factor of 1.7 for the production rates ranging over more than 3 orders of magnitude.

To devise an activation technique for characterizing mixed radiation fields, secondary particles from a copper target irradiated by 24 GeV/c protons were measured at the CERN High-energy AcceleRator Mixed field facility (CHARM). Activation detector sets consisting of aluminum, niobium, indium, and bismuth, were placed at 30 cm from the target at angles of 15° to 160° with respect to the beam axis. The nuclides generated in these detectors due to irradiation by secondary particles were analyzed by γ -ray spectrometry, and the angular distributions of the production rates were obtained. The results of Monte Carlo calculations using FLUKA code was compared with the experimental results. The calculated results well agreed with the measured data at all angles. The influence of competitive reactions on the measured data were also evaluated by FLUKA. The following nuclear reactions, with low affectivity by competitive reactions, were identified as promising tools for characterizing mixed radiation fields: the 115In( n , n ′ ) 115 m In reaction for detecting neutrons emitted by the evaporation process, the 93Nb( γ , n ) 92 m Nb reaction for verifying the photon distribution generated by neutral-pion decay ( π 0 → 2 γ ), and the 209Bi( p , 4 n )206Po reaction, which detects secondary protons.

The reproduction of neutron fluence derived by the unfolding method was confirmed by simulating an experiment at Cern High energy AcceleRator Mixed field facility (CHARM). Fluences on an NE213 scintillator located at positions surrounded with shields were calculated using PHITS. A neutron light output spectrum and response matrix were calculated according to the calculated fluence. Furthermore, response matrices with simple distributions of neutron incident position and direction on the scintillator were also prepared because a response matrix with guessed distributions is used in measurements. In spite of using response matrices with different distributions, the unfolded fluence agreed with each other, unless the distribution was focused on a position. The agreement of the fluences enables us to measure the fluence at various positions even though the distributions are experimentally unknown. Finally, experimental fluences were obtained under the same conditions, and were compared with the simulation results.

To establish a methodology for neutron spectrum measurement at the CERN High energy AcceleRator Mixed field facility (CHARM), neutron spectra were measured using an NE213 scintillator on top of the CHARM roof shielding where is the CERN Shielding Benchmark Facility (CSBF). The spectra were derived as fluences into the scintillator by the unfolding method using an iterative Bayesian algorithm. The methodology was verified based on the agreement of two spectra measured for different positions and directions of incident neutrons by changing the detector orientation. Since the spectra on the roof-top were obtained within a reasonable beam-time, this methodology is suitable for measuring the spectrum when there is less shielding material. Thus, experimental data for neutron transition can be obtained as a function of shielding thickness using this facility.

The spatial distributions of neutrons and gamma rays in the epithermal mode of the Kyoto University Reactor were measured using a MAGAT-type polymer gel detector doped with LiCl at 6Li concentrations of 0, 10, 100 ppm. Reasonable distributions for thermal neutrons and gamma rays were obtained for 0 and 100ppm.

The CERN High Energy AcceleRator Mixed field facility (CHARM) is located in the CERN Proton Synchrotron (PS) East Experimental Area. The facility receives a pulsed proton beam from the CERN PS with a beam momentum of 24 GeV/c with 5 ⋅ 10 11 protons per pulse with a pulse length of 350 ms and with a maximum average beam intensity of 6.7 ⋅ 10 10 p/s that then impacts on the CHARM target. The shielding of the CHARM facility also includes the CERN Shielding Benchmark Facility (CSBF) situated laterally above the target. This facility consists of 80 cm of cast iron and 360 cm of concrete with barite concrete in some places. Activation samples of bismuth and aluminium were placed in the CSBF and in the CHARM access corridor in July 2015. Monte Carlo simulations with the FLUKA code have been performed to estimate the specific production yields for these samples. The results estimated by FLUKA Monte Carlo simulations are compared to activation measurements of these samples. The comparison between FLUKA simulations and the measured values from γ -spectrometry gives an agreement better than a factor of 2.

The light output due to cosmic-ray muons was determined for an EJ301 scintillator of 12.7 cm in diameter and length. Muons were measured with the setup of the EJ301 scintillator put between two thin plastic scintillators. Coincidence analysis using the signals from these plastic scintillators revealed a broad peak in the distribution of the EJ301 scintillator as measured with a charge-integrating analog-to-digital converter (ADC). The mode of the distribution was obtained by fitting analysis. To convert the corresponding ADC value into light output, continuous-energy protons were measured, with a resulting light output of the mode became to be 20.5 ± 1.0 MeVee. By measuring cosmic-ray muons with this setup, we have shown that we can obtain a calibration point in the higher light output region.

We designed a neutron spectrometer to measure low-energy neutron beam for boron neutron capture therapy (BNCT) without exchanging thicknesses of neutron moderator. The spectrometer consists of a boron carbide block as a neutron moderator and nine thin neutron silicon sensors with variable detection efficiency. Using the Monte Carlo technique, we simulated the spectrometer’s neutron energy response functions. It represented sharp rise in the response functions with various lower limits of neutron energies distributed over a wide range of neutron energy from a thermal energy to a few 100 keV. The lower limit of neutron energy was dependent on a depth in the boron carbide block. By applying thicker enriched 6LiF neutron converter and covering the deep positioned neutron sensors with polyethylene plates, neutron detection efficiencies were increased to be able to measure the BNCT neutron beam in a short time.

Eight ancient bimetal swords held by Hiroshima University, Japan were analyzed non-destructively through γ-ray radiography and X-ray fluorescence (XRF). 137 Cs and 60 Co γ-ray irradiation sources were used to obtain transmission images of swords. A scanning radiography method using a 60 Co γ-ray source was developed. XRF was used for qualitative elemental analysis of the swords. The presence of iron cores in the hilts of some swords had been observed and it was assumed that the cores were a ritual symbol or had a functional purpose. However, our work reveals that these swords were originally bronze-hilted iron swords and that the rusty blades were replaced with bronze blades to maintain the swords’ commercial value as an antique. Consequently, the rest of the iron blade was left in the hilt as an iron tang. The junction of the blade and the guard was soldered and painted to match the patina color. XRF analysis clearly showed that the elemental Sn/Cu ratios of the blades and the hilts were different. These findings are useful for clarifying the later modifications of the swords and are important for interpreting Bronze Age and Iron Age history correctly.

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.

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.

Cumulative dose of external and internal exposures following the Fukushima Daiichi Nuclear Power Plant (FNPP) accident was estimated for Japanese macaques (Macaca fuscata) around FNPP. Conversion factors for Japanese macaques modeled as ellipsoids were estimated for external exposure from contaminated ground and internal exposure uniformly distributed in the body. Conversion factors for seven radionuclides, namely, tellurium-129 (129Te), 129mTe, iodine-131 (131I), 132Te, 132I, cesium-134 (134Cs) and 137Cs were calculated using the PHITS code. The estimated factors for the seven radionuclides were consistent with those in ICRP Publication 108, using an effective radius for comparison. The external, internal and total exposures for 13 macaques in Namie Town were estimated by applying the calculated factors. The estimated cumulative exposures for the periods from the accident occurred to the sampling date, ranged from 0.26 to 1.6 Gy. The average exposure was 0.64 Gy in averaged over the 11 sampled macaques except for the 2 macaques which might be born after the FNPP accident.

The

In this study, experimental measurements were performed on the spectra of neutrons which penetrate concrete and steel of various thicknesses values when a proton beam of 24 GeV/c was incident on a copper target at the CHARM facility in the East Hall of the CERN Proton Synchrotron (PS) The thicknesses of concrete and steel ranged up to 360 cm and 80 cm, respectively. To measure the neutron spectra, an NE213 scintillator was positioned on the top roof of the shielding structure as the neutron detector. The light output distributions of the detector were converted into the neutron energy spectra using the unfolding method with a calculated response matrix after removing the γ -ray and charged particle events by pulse-shape discrimination and veto counter signals, respectively. The neutron spectra were in agreement with the results obtained using the Monte Carlo simulation code, PHITS, within a factor of 1.4 except for the case of steel 80 cm. The attenuation profiles for concrete and steel were consistent with previous foil activation results within the respective uncertainties.

Double-differential neutron yields from a water phantom bombarded with 290-MeV/nucleon and 430-MeV/nucleon carbon ions were measured at emission angles of 15°, 30°, 45°, 60°, 75°, and 90°, and angular distributions of neutron yields and doses around the phantom were obtained. The experimental data were compared with results of the Monte-Carlo simulation code PHITS. The PHITS results showed good agreement with the measured data. On the basis of the PHITS simulation, we estimated the angular distributions of neutron yields and doses from 0° to 180° including thermal neutrons.

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.

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.

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.

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.

The CERN High energy AcceleRator Mixed field (CHARM) facility is situated in the CERN Proton Synchrotron (PS) East Experimental Area. The facility receives a pulsed proton beam from the CERN PS with a beam momentum of 24 GeV/c with 5centerdot10¹¹ protons per pulse with a pulse length of 350 ms and with a maximum average beam intensity of 6.7centerdot10¹⁰ protons per second. The extracted proton beam impacts on a cylindrical copper target. The shielding of the CHARM facility includes the CERN Shielding Benchmark Facility (CSBF) situated laterally above the target that allows deep shielding penetration benchmark studies of various shielding materials. This facility has been significantly upgraded during the extended technical stop at the beginning of 2016. It consists now of 40 cm of cast iron shielding, a 200 cm long removable sample holder concrete block with 3 inserts for activation samples, a material test location that is used for the measurement of the attenuation length for different shielding materials as well as for sample activation at different thicknesses of the shielding materials. Activation samples of bismuth, aluminium and indium were placed in the CSBF in September 2016 to characterize the upgraded version of the CSBF. Monte Carlo simulations with the FLUKA code have been performed to estimate the specific production yields of bismuth isotopes (²⁰⁶ Bi, ²⁰⁵ Bi, ²⁰⁴ Bi, ²⁰³ Bi, ²⁰² Bi, ²⁰¹ Bi) from ²⁰⁹ Bi, ²⁴ Na from ²⁷ Al and ¹¹⁵ m I from ¹¹⁵ I for these samples. The production yields estimated by FLUKA Monte Carlo simulations are compared to the production yields obtained from γ-spectroscopy measurements of the samples taking the beam intensity profile into account. The agreement between FLUKA predictions and γ-spectroscopy measurements for the production yields is at a level of a factor of 2., 4th International Workshop on Accelerator Radiation Induced Activation (ARIA 2017), 22–24 May 2017, Lund, Sweden

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.

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.

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.

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.

The energy spectra of neutrons were measured by a time-of-flight method for 120 GeV protons on thick graphite, aluminum, copper, and tungsten targets with an NE213 scintillator at the Fermilab Test Beam Facility. Neutron energy spectra were obtained between 25 and 3000 MeV at emission angles of 30°, 45°, 120°, and 150°. The spectra were parameterized as neutron emissions from three moving sources and then compared with theoretical spectra calculated by PHITS and FLUKA codes. The yields of the theoretical spectra were substantially underestimated compared with the yields of measured spectra. The integrated neutron yields from 25 to 3000 MeV calculated with PHITS code were 16–36% of the experimental yields and those calculated with FLUKA code were 26–57% of the experimental yields for all targets and emission angles.

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 (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 β-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 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.

In order to investigate the accuracy of high-energy radiation transport codes based on experimental data, the collaboration; Japanese and American Study of Muon Interaction and Neutron detection (JASMIN), has been formed among many institutions and universities of Japan and U.S.A. A series of experiments have been conducted at the Fermilab Pbar target station, the NuMI beam line and the Fermilab Test Beam Facility (FTBF). In this paper the research activities are reviewed.

Systematics of neutron energy spectra for 120 GeV protons on C, Al, Cu and W targets are presented for data measured using the time-of-flight technique, room-scattering subtraction, and experimentally determined neutron detection efficiency. The measured neutron energy spectra exhibited low-energy angular-independent and high-energy angular-dependent components, below and above about 50 MeV, respectively. The neutron yields of both components increased linearly with target thickness. We obtained neutron yields for one interaction length target and determine the normalization factors from the ratio of the target thickness to the interaction length. We observed a non-linear relationship between neutron yield and target mass number.

The behavior of fuel air mixing and combustion continues to be of prime importance in all kinds of combustion systems as this leads to improvements in combustion efficiency, formation and emission of trace pollutants, and size and dimensions of the combustion chamber. A detailed examination of flame requires close examination of the various ongoing physico-chemical properties as well as detailed stable and short lived chemical species in a combustion system. In this paper, an isotope shift/planer laser-induced fluorescence (IS/PLIF) method is developed that is based on the distinct excitation wavelengths of rotational lines of OD radicals that are displaced from the OH radicals due to their isotope shift. We provide visualization of the spatial distributions of fluorescent emissions from OH and OD for their application in unraveling the complex feature of flames from the view of a comprehensive insightful analysis. Note that OH is typically an intermediate specie in combustion reactions. However, OD molecules are produced from the dissociation reaction of deuterium oxide (D2O) upon injection into a flame. The information obtained from the LIF profiles provided useful information in our quest to examine the fate of mixing between fuel and air during combustion.

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-COThe relative sensitivity of the TE(B)-TE chamber is calculated with the same method as for the TE-TE and C-COThe calculated relative sensitivity of the C-COThe 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.

Many benchmark data are required for the improvement of theoretical model implemented in a Monte Carlo code for particle transport. To acquire the benchmark data, we measured energy spectra of protons emitted from graphite, aluminum, and iron targets bombarded with continuous-energy neutrons, which enable simultaneous measurements at the incident energies from 100 to 600 MeV at a time. The neutron flux incident on the target was measured with a 238 U fission ionization chamber. Protons emitted from the target were measured with three ΔE–E detectors consisting of a thin NE102A scintillator and a thick NE213 liquid scintillator. In the analysis, the pulse shape discrimination of the NE213 scintillator enable us to distinguish events for a charged particle stopping in the scintillator from events for a charged particle penetrating the scintillator. Experimental results were compared with calculations by the PHITS code coupled with the JENDL-HE file, the Bertini model implemented in the PHITS code, and the PEANUT model in the FLUKA code.

A methodology for the time-of-flight measurement of the neutron energy spectrum for a high-energy proton-beam-induced reaction was established at the Fermilab Test Beam Facility of the Fermi National Accelerator Laboratory. The 120-GeV proton beam with 3 × 10 5 protons/spill was prepared for event-by-event counting of incident protons and emitted neutrons for time-of-flight energy determination. An NE213 organic liquid scintillator (12.7 cm in diameter by 12.7 cm in length) was employed with a veto plastic scintillator and a pulse-shape discrimination technique to identify neutrons. Raw waveforms of NE213, veto and beam detectors were recorded to discriminate the effects of multi-proton beam events by considering different time windows. The neutron energy spectrum ranging from 10 to 800 MeV was obtained for a 60-cm-long copper target at 90° with respect to the beam axis. The obtained spectrum was consistent with that deduced employing the conventional unfolding technique as well as that obtained in a 40-GeV/c thin-target experiment.

We have studied pulse shape discrimination of sub-MeV neutron from spallation nuclear reactions. Neutron measurements were carried out with a 2-inch NE213 scintillator viewed by a phototube (HAMAMATSU H1949) for the carbon-carbon reaction at 290 MeV/n. Neuron energies were measured with the TOF technique by using a timing pick-up plastic scintillator. Output signals of detectors were digitized and stored with a digitizer (Acqiris DC282). The digital pulse shape discrimination (DPSD) has been investigated through comparison between three different analyzing algorithms. Effects on the neutron-gamma (n/g) discrimination were also investigated after the subtraction of background and abnormal events. Moreover, in order to realize better n/g discrimination the detector system was cooled down to about zero degrees centigrade by Peltier devices to suppress the dark current of the phototube which would be the main cause distorting pulse shapes. Experimental studies were made in terms of the temperature dependence of detector performance including pulse shape.

Neutron-production double-differential cross-sections on carbon–carbon and oxygen–carbon reactions with incident heavy-ion energy of 290 MeV/nucleon were measured by time-of-flight method using liquid organic scintillators. By use of a detection system specialized for low-energy neutrons, the cross-sections were obtained in a wide energy region from several hundred MeV down to 0.6 MeV for the oxygen-ion incidences. The experimental data were compared with the calculation results using the Monte-Carlo simulation code, PHITS. The PHITS results gave an overall agreement with the measured data within a factor of two.