Particle And Heavy Ion Transport Code System Research Articles

Validation of mammographic x-ray spectra generated using Particle and Heavy Ion Transport code System

A Monte Carlo (MC) code is a robust method to generate a mammographic x-ray spectrum because the geometry of a mammography system can be flexible and directly modeled in MC simulation. However, simulations from MC code need to be validated before it can be reliably used for specific applications. This study aimed to generate and validate the x-ray spectra of relevant anodes used in mammography and breast tomosynthesis using Particle and Heavy Ion Transport code System (PHITS). PHITS version 3.08 was used to generate the x-ray spectra of molybdenum (Mo), rhodium (Rh), and tungsten (W) anodes. The Mo anode spectrum derived using PHITS was compared with those obtained using other MC codes. The generated spectra of all anodes were compared with the literature. Parameters including spectral shape, K characteristic x-ray yield, heel effect, and half-value layer (HVL) were used for a comparative assessment. The differences in these assessment parameters conducted by PHITS and PHITSEGS5 simulations were studied. Regarding the comparative parameters, PHITSEGS5 simulation improved the accuracy of mammographic x-ray generation compared to PHITS simulation; K x-ray and bremsstrahlung yields of the Mo anode spectrum generated by PHITSEGS5 simulation were a better agreement with those generated by other MC code simulations. The PHITSEGS5 spectra overestimated K x-ray and low-energy bremsstrahlung photons in comparison with measured spectra. Subsequently, HVLs calculated from PHITSEGS5 spectra were 1.0% (Mo/Mo) and 7.0% (W/Al) lower than those derived from measured spectra. For Mo and Rh anodes, relative difference of HVLs calculated from PHITSEGS5 spectra and those obtained from literature and measurement were within the TRS 457 acceptance criteria (±0.02 mm Al). The observed difference exceeded the acceptance criteria for W anode. Regarding existed consistency in HVL between simulation and measurement, PHITSEGS5 simulation can be reliably used to generate x-ray spectra of Mo and Rh anodes. However, its accuracy should be improved for generating W anode spectrum.

Monte Carlo study toward the development of a radiation field to simulate secondary neutrons produced in carbon-ion radiotherapy

This study investigates the feasibility of developing a neutron field to simulate secondary neutrons produced in carbon ion radiotherapy (CIRT) with a passive beam using a monoenergetic carbon beam, a neutron-producing target, and neutron-shaping filters via Monte Carlo calculations. The general-purpose particle and heavy-ion transport code system (PHITS) was employed throughout this investigation. After characterizing the targeted neutron field, a neutron-producing target and neutron-shaping filters were selected. A dataset of the neutron energy spectra was compiled using the selected target and filters. Then, we determined the best combination of target and filters and the best angle to the beam axis to reproduce the neutron energy spectrum in the targeted neutron field. The root mean square percentage error and the mean absolute percentage error were used as figures of merit to quantify the agreement with neutron energy spectra in the CIRT treatment room. We found that a monoenergetic carbon beam can sufficiently reproduce the neutron energy spectra in a CIRT treatment room with a passive beam using an iron neutron-producing target doubling as a sharping filter and/or a polyethylene filter and selecting the angle to the beam axis. Therefore, the development of a neutron field to simulate secondary neutrons produced in CIRT with a passive beam is feasible in a non-clinical facility. This simulated neutron field can be useful for not only biological experiments but also physics experiments such as the development and calibration of neutron dosimeters.

Microdosimetric quantities of an accelerator-based neutron source used for boron neutron capture therapy measured using a gas-filled proportional counter

Boron neutron capture therapy (BNCT) is an emerging radiation treatment modality, exhibiting the potential to selectively destroy cancer cells. Currently, BNCT is conducted using a nuclear reactor. However, the future trend is to move toward an accelerator-based system for use in hospital environments. A typical BNCT radiation field has several different types of radiation. The beam quality should be quantified to accurately determine the dose to be delivered to the target. This study utilized a tissue equivalent proportional counter (TEPC) to measure microdosimetric and macrodosimetric quantities of an accelerator-based neutron source. The micro- and macro-dosimetric quantities measured with the TEPC were compared with those obtained via the the particle and heavy ion transport code system (PHITS) Monte Carlo simulation. The absorbed dose from events >20 keV/μm measured free in air for a 1-h irradiation was calculated as 1.31 ± 0.02 Gy. The simulated result was 1.41 ± 0.07 Gy. The measured and calculated values exhibit good agreement. The relative biological effectiveness (RBE) that was evaluated from the measured microdosimetric spectrum was calculated as 3.7 ± 0.02, similar to the simulated value of 3.8 ± 0.1. These results showed the PHITS Monte Carlo simulation can simulate both micro- and macro-dosimetric quantities accurately. The RBE was calculated using a single-response function, and the results were compared with those of several other institutes that used a similar method. However, care must be taken when using such a single-response function for clinical application, as it is only valid for low doses. For clinical dose ranges (i.e., high doses), multievent distribution inside the target needs to be considered.

Analysis Dosage of Boron in BNCT with Simulation Method Using PHITS (Particle and Heavy Ion Transport Code System) Program

Abstrak – Cancer is a dangerous disease which is one of the main causes of death in the world, which is 8.8 million deaths in 2015. A new method has been developed for the treatment of cancer with small side effects, Boron Neutron Capture Therapy (BNCT). BNCT is based on the ability of non-radioactive isotopes Boron-10 to capture thermal neutrons. Boron is the main key of the BNCT process wherein, radiation of cancer cells can be carried out selectively if the cancer contains high concentrations of boron. Boron doses in cancer cells cannot be measured directly, therefore artificial network methods are used to compute doses computationally, namely the Monte Carlo method with the Particle and Heavy Ion Transport Code System (PHITS) program. Boron dose used was 10, 20 and 30 ?g/g. It was found that the greater the boron dose the greater the number of neutrons absorbed. The results showed that the largest absorption dose occurred in the Gross Tumor Volume (GTV) area of ??7.7076E-10 Gy/cm2.
 Key words: Cancer, BNCT, Dose of Boron, PHITS.

Estimation of Residual Radioactivity and Radiation Damage in SiC After Neutron Irradiation

In the process of neutron transmutation doping (NTD), Si, C, and impurities in the SiC become radioactive nuclides by neutron capture reaction (n, $\gamma $ ) and several threshold reactions. The radiations emitted from the residual radionuclides constantly damage the SiC until they completely decay. The radiation damage can create many defects in the SiC, and the defects can reduce the electron mobility and distort the band structure. This radiation damage to the SiC should be evaluated in terms of quality management. In order to quantify the sources of radiation damage, time variations of the residual radionuclides’ radioactivity in 5-in cylindrical SiC ingot after neutron irradiation were calculated using the DCHAIN program linked to the particle and heavy ion transport code system (PHITS). In the calculation, it was assumed that the SiC ingot was irradiated in the NTD1 irradiation hole of the High-Flux Advanced Neutron Application Reactor (HANARO) at the Korea Atomic Energy Research Institute (KAERI). The energy spectrum of neutrons in the HANARO NTD1 irradiation hole calculated by using the Monte Carlo N-particle code version 6.1 (MCNP6.1) code was considered as a neutron source for the calculation. Based on the radioactivity data, the total flux and energy spectra of the electrons and gamma-rays in the SiC were calculated to analyze the properties of the radiation sources. Most of the radiations were emitted from the beta decay of 31Si, but they decayed early, and radionuclides with a relatively long half-life, such as 59Fe and 32P, became the major radiation sources. Finally, the radiation damage was calculated in the displacement per atom (DPA) unit. The time variations of the DPA rate and cumulative DPA in the SiC are presented.

Isotope-production cross sections of residual nuclei in proton- and deuteron-induced reactions on 93Zr at 50 MeV/u

The isotope-production cross sections in p- and d-induced reactions on 93Zr at approximately 50 MeV/nucleon were measured by using the inverse-kinematics method at RIKEN RI Beam Factory. The measured data were compared with the previous experimental 93Zr + p, d at 105 and 209 MeV/nucleon data. This comparison represents that the isotopic distribution of production cross sections at 51 MeV p-induced reaction is appreciably different from those at 105 and 209 MeV. On the other hand, these three data sets show that the shape of isotopic distribution is similar in the case of the d-induced reaction. Also, the measured production cross sections were compared with the theoretical model calculations with Particle and Heavy Ion Transport code System (PHITS) version 3.10 in order to investigate the reproducibility of the models implemented in PHITS. The calculations well reproduced the experimental data even in such low incident energy, while several discrepancies were still seen as in the p- and d-induced reactions at 105 and 209 MeV/nucleon.

Neutron production from thick LiF, C, Si, Ni, Mo, and Ta targets bombarded by 13.4-MeV deuterons

Double-differential thick target neutron yields from LiF, C, Si, Ni, Mo, and Ta targets bombarded by 13.4-MeV deuterons were measured by using an EJ-301 liquid organic scintillator at the Center for Accelerator and Beam Applied Science, Kyushu University. The measured (d, xn) spectra were compared with the (t, xn) spectra measured by the other group at the same incident energy per nucleon (6.7 MeV/u) and theoretical model calculations by Particle and Heavy Ion Transport code System (PHITS) and DEUteron-induced Reaction Analysis Code System (DEURACS). Some bumps are observed in the (d, xn) spectra for low-Z target elements, while no specific structure was seen in the (t, xn) spectra. The PHITS calculation, in which the intra-nuclear cascade of Liége (INCL) and generalized evaporation model (GEM) were used, generally overestimates neutron spectra while the DEURACS calculation agrees with experimental ones fairly well.

Total cross section model with uncertainty evaluated by KALMAN

We propose a new total cross section model including its uncertainty to estimate systematic uncertainty in particle transport simulations. Medians and widths of parameters in the model, which describe the model uncertainty in the simulation, were determined by using the KALMAN code based on the Bayes' theorem. We implemented the new model in the particle and heavy ion transport code system (PHITS), and then performed shielding calculations with neutron beams. By using analysis of variance (ANOVA), the systematic uncertainty associated with the model uncertainty can be estimated from variance in calculated results.

Measurement of Single-Event Upsets in 65-nm SRAMs Under Irradiation of Spallation Neutrons at J-PARC MLF

A neutron irradiation test of static random access memories (SRAMs) was performed using a spallation neutron source at Materials and Life Science Experimental Facility (MLF) in the Japan Proton Accelerator Research Complex (J-PARC). The probability of neutron-induced single-event upsets (SEUs) was measured for 65-nm bulk and silicon on thin buried oxide (SOTB) SRAMs under neutron irradiation at the BL10 experimental facility. The measured SEU data were compared with the previous data of the same SRAMs which were measured at other irradiation facilities having different neutron spectra. The differences in the operating voltage dependence of the measured SEU probabilities are discussed with particular attention to the impact of irradiation side on SEUs. The particle and heavy ion transport code system (PHITS) simulation based on the simple sensitive volume model qualitatively reproduced the operating voltage dependence seen in the measured ratio of SEUs for the Bulk SRAM between the resin side and board side irradiations under different neutron fields.

Monte Carlo simulation of X-ray room shielding in diagnostic radiology using PHITS code

This research was conducted to evaluate scatter distribution at an X-ray room for diagnostic radiography using the Particle and Heavy Ion Transport code System (PHITS). The X-ray room was simulated...

Contribution of cadmium to the total amount of positron creation in a reactor-based slow positron beamline

In the slow positron beamline at the Kyoto University Research Reactor (KUR), positron creation was enhanced by increasing the gamma-ray intensity at the positron source via the reaction of 113Cd(n,γ)114Cd. To achieve this, a cadmium (Cd) cap was attached to the positron source, surrounding it, and thus, without intentional cooling, the temperature was able to reach near the melting point of Cd via nuclear heating. In this study, the degree to which the Cd cap contributes to the quantity of positron creation was estimated by using the Monte Carlo calculation code PHITS (Particle and Heavy Ion Transport code System), which simulates radiation transportation and interaction with matter. As a result, the number of positrons created was found to become 2.0±0.1 times higher by using the Cd cap at the KUR slow positron beamline.The use of the Cd cap was confirmed to be significantly effective for enhancing positron creation.

Study on gamma-ray shielding properties of lead tellurite glass systems using PHITS

The gamma-ray shielding properties, mass attenuation coefficient, mean free path and half value layer of lead tellurite glass systems were investigated in the photon energy range of 0.015–15 MeV using particle and heavy ion transport code system (PHITS) with the use of narrow beam transmission technique. The gamma shielding efficiency of studied glass systems was compared to that of standard concretes in terms of mean free path and half value layer. The obtained results show that PHITS is very capable for investigation of radiation shielding properties to good accuracy with maximum percentage differences of less than 2% comparing to the results by XCOM database. This will make a useful alternative technique for studying radiation parameters of other glasses, for which no experimental results are available.

Estimation of uncertainty in lead spallation particle multiplicity and its propagation to a neutron energy spectrum

ABSTRACT This paper presents an approach to uncertainty estimation of spallation particle multiplicity of lead (natPb), primarily focusing on proton-induced spallation neutron multiplicity (xpn ) and its propagation to a neutron energy spectrum. The xpn uncertainty is estimated from experimental proton-induced neutron-production double-differential cross-sections (DDXs) and model calculations with the Particle and Heavy Ion Transport code System (PHITS). Uncertainties in multiplicities for (n, xn), (p, xp), and (n, xp) reactions are then inferred from the estimated x pn uncertainty and the PHITS calculation. Using these uncertainties, uncertainty in a neutron energy spectrum produced from a thick natPb target bombarded with 500 MeV proton beams, measured in a previous experiment, is quantified by a random sampling technique, and propagation to the neutron energy spectrum is examined. Relatively large uncertainty intervals (UIs) are observed outside the lower limit of the measurement range, which is prominent in the backward directions. Our findings suggest that a reliable assessment of spallation neutron energy spectra requires systematic DDX experiments for detector angles and incident energies below 100 MeV as well as neutron energy spectrum measurements at lower energies below ∼1.4 MeV with an accuracy below the quantified UIs.

Radiation Dose Analysis in Criticality Accident of Fuel Debris in Water

Removal of fuel debris is regarded as one of the most important operations in the decommissioning of the Fukushima Daiichi nuclear power station (1F-NPS) to decrease long-term risk. To begin the operation, the consequences of possible criticality accidents must be evaluated in advance. In this work, we evaluated radiation doses during possible criticality accidents at 1F-NPS in assumptive fuel debris systems. In particular, the relationship between the water level surrounding the fuel debris and the radiation dose was investigated. This is because the water level surrounding the fuel debris is thought to have an impact on radiation dose during accidents as it affects both the reactivity and shielding of radiation. A combination of space-dependent kinetic analysis and radiation transport analysis was carried out in order to consider the special characteristics of fuel debris systems in water. Instead of traditional point-kinetics analysis, we used the Multi-region Integral Kinetic (MIK) code, which is a unique method based on Monte Carlo neutron transport calculations. The radiation transport calculation code Particle and Heavy Ion Transport Code System (PHITS) was used as well. The analyses revealed that the dose caused by criticality accidents may be the largest in systems in which part of the fuel debris is exposed to the air.

Isotope production in proton-, deuteron-, and carbon-induced reactions on Nb93 at 113 MeV/nucleon

Isotope-production cross sections for p-, d-, and C-induced spallation reactions on Nb93 at 113 MeV/nucleon were measured using the inverse-kinematics method employing secondary targets of CH2, CD2, and C. The measured cross sections for Mo90, Nb90, Y86,88 produced by p-induced reactions were found to be consistent with those measured by the conventional activation method. We performed benchmark tests of the reaction models INCL-4.6, JQMD, and JQMD-2.0 implemented in the Particle and Heavy Ion Transport code System (PHITS) and of the nuclear data libraries JENDL-4.0/HE, TENDL-2017, and ENDF/B-VIII.0. The model calculations also showed generally good agreement with the measured isotope-production cross sections for p-, d-, and C-induced reactions. It also turns out that, among the three nuclear data libraries, JENDL-4.0/HE provides the best agreement with the measured data for the p-induced reactions. We compared the present Nb93 data with the Zr93 data, that were measured previously by the same inverse kinematics method (Kawase et al., Prog. Theor. Exp. Phys. 2017, 093D03 (2017)2050-391110.1093/ptep/ptx110), with particular attention to the effect of neutron-shell closure on isotope production in p- and d-induced spallation reactions. The isotopic distributions of the measured production cross sections in the Zr93 data showed noticeable jumps at neutron number N=50 in the isotopic chains of ΔZ=0 and -1, whereas no such jump appeared in isotopic chain of ΔZ=0 in the Nb93 data. From INCL-4.6 + GEM calculations, we found that the jump formed in the evaporation process is smeared out by the intranuclear cascade component in Nb91 produced by the Nb93(p,p2n) and (d,d2n) reactions on Nb93. Moreover, for Nb93, the distribution of the element-production cross sections as a function of the change in proton number ΔZ is shifted to smaller ΔZ than for Zr93, because the excited Nb prefragments generated by the cascade process are more likely to emit protons than the excited Zr prefragments, due to the smaller proton-separation energies of the Nb isotopes.

Evaluation of scintillating-fiber detector response for 14 MeV neutron measurement

A scintillating-fiber (Sci-Fi) detector has been employed to measure 14 MeV neutrons for the triton burnup study in the first deuterium plasma campaign of the Large Helical Device (LHD). The pulse-height spectra of the Sci-Fi detector are used to choose a suitable threshold for the discrimination of 14 MeV neutrons from a mix-radiation field of low-energy neutrons and gamma-rays. The measured pulse-height spectra of the Sci-Fi detector have two components with different decay slopes from the LHD experiment. To study the pulse-height property of the Sci-Fi detector, the pulse-height spectra on different energy neutrons have been measured by using the accelerator-based neutron source with d-D, p-Li, and d-Li reactions. Meanwhile, the simulations of the detector response have been performed by using the Particle and Heavy Ion Transport code System (PHITS). In the LHD experiment, the first decay component of the pulse-height spectra in low-pulse-height region has been found to correspond to the signals induced by 2.45 MeV neutrons and gamma-rays. In addition, the high-pulse-height region has been confirmed by both the accelerator experiment and the PHITS calculation to correspond to the recoil-proton edge induced by triton burnup 14 MeV neutrons. The detection efficiency of 14 MeV neutrons for the Sci-Fi detector calculated by the PHITS code agrees well with the detection efficiency of 14 MeV neutrons for the Sci-Fi detector evaluated in the LHD experiment. The Sci-Fi detector can work as a standard detector for the 14 MeV neutron measurement with a suitable threshold.

Development of capillary plate neutron detector filed with liquid scintillator by using recoiled-particle trajectory analyses

New 14 MeV neutron detector using capillary plate with liquid scintillator has been developed aiming at the triton burn-up study on the deuterium plasma experiment of the Large Helical Device (LHD). A capillary plate with 25 mm in diameter, 10 mm in thickness, and 10 μm pitch honeycomb structure is used as a detector head. The scintillating light trajectory is measured by an Electron-Multiplying CCD (EMCCD) via relay lenses. The trajectories for neutrons and gamma-rays are simulated by the Particle and Heavy Ion Transport code System (PHITS). Maximum trajectory length is estimated to be 2 mm and 0.1 mm for 14 MeV neutron and 2.45 MeV neutron, respectively. The trajectory of a proton recoiled by a neutron is linear, and the trajectory length depends on the recoiled proton energy. On the other hand, the trajectory of an electron produced by gamma-ray is not straight due to the random walk of the electron in the scintillator. Therefore, we can identify 14 MeV neutrons from 2.45 MeV neutrons and gamma-rays by the trajectory length and the shape. The developed detector was installed in the basement of the LHD torus hall just under the vertical collimator embedded in the floor concrete for the second deuterium plasma campaign of LHD in 2018. We identify trajectory which is considered due to the proton recoiled by14 MeV neutrons on the LHD deuterium discharge.

Estimation of Muon-Induced SEU Rates for 65-nm Bulk and UTBB-SOI SRAMs

Negative and positive muon-induced single-event upset (SEU) rates were estimated for 65-nm bulk and ultrathin body and thin buried oxide silicon-on-insulator (UTBB-SOI) static random access memories (SRAMs). The SEU cross sections for muon incidence on the two SRAMs were experimentally characterized and compared. The experimental results showed that the negative muon SEU cross sections for the bulk SRAM are significantly larger than those for the UTBB-SOI. Estimation of muon SEU rates at ground level was performed using the experimental results and the Monte Carlo simulation with the Particle and Heavy Ion Transport code System (PHITS). The estimated muon SEU rates were compared with the measured neutron SEU rates. The contribution of muons was found to be considerably smaller than that of neutrons. Attenuation effect of muons and neutrons in a five-story building was also investigated by particle transport simulation with PHITS. The muon SEU rate on the first floor was estimated to be at most 10% of the neutron SEU rate on the same floor.

Monte Carlo studies for the radiation shielding in the bunker of the electron beam accelerator using PHITS

Particle and Heavy Ion Transport Code System (PHITS) is a Monte Carlo simulation program used for transport and collision of nearly all particles over wide energy range. One of its useful features for accelerator facility design is the calculation of dose distribution. This paper describes how the geometry of the bunker for low electron beam accelerator in Nuklear Malaysia (NM) was modelled and used to calculate the equivalent dose caused by the radiation for specific region in the bunker by using PHITS. The dose distribution and energy deposition were obtained. Finally, the simulation results are compared to the measured value from the dose measurement testing inside the bunker facility and found both results are in a good agreement.

Comparative neutron-shielding properties of metal oxide/HDPE composites using a Monte Carlo Code of PHITS

Development of efficient neutron-shielding materials has become one of the most important issues in radiation safety. In order to estimate shielding performances of the materials of interest, simulations of neutron-shielding characteristics using well-proved methods are necessary. In this work, a Monte Carlo code of Particle and Heavy Ion Transport code System (PHITS) was used to simulate neutron-shielding properties of HDPE composites that were added with either B2O3, Sm2O3, Gd2O3, or CdO as a filler at various contents. The energy of the incoming neutrons was fixed at 0.025 eV with the beam direction pointing at the right angle to the material’s surface. The results showed that Gd2O3/HDPE composites had the lowest neutron transmission ratio compared with other fillers at the same contents, implying the highest neutron-shielding ability. Furthermore, the simulations also showed that the values of neutron transmission ratios of all fillers decreased with increasing filler’s contents. Other information such as half-value-layer (HVL) and tenth-value-layer (TVL) were also simulated and reported in this work. We encouraged further investigation of this work by comparing the simulated results with the actual experiment in order to validate the accuracy of the simulation.