Particle And Heavy Ion Transport Code System Research Articles

Optimization of a slab geometry type cold neutron moderator for RIKEN accelerator-driven compact neutron source

We have optimized a cold neutron moderator to be operated at the RIKEN accelerator-driven compact neutron source (RANS) to broaden cold neutron applications and provide more choices for RANS users. We selected a safe and easy to manage material, mesitylene, as the RANS cold moderator. An efficient moderator system was designed by studying and optimizing a coupled cold neutron moderator of mesitylene at 20 K with a polyethylene (PE) pre-moderator at room temperature in the slab geometry with Particle and Heavy Ion Transport code System (PHITS) simulations. The parameters of mesitylene and PE thickness, the reflector, and the shielding configuration were studied to increase cold neutron intensities within [0.1 meV, 10 meV] and thermal neutron intensities within [10 meV, 100 meV]. Consequently, an integrated cold neutron intensity of 1.15 ×103n/cm2∕μA at 2 m from the neutron-producing target was finally achieved, which was 12 times higher than that of the current RANS moderator with 4 cm PE according to simulation results. The cold and thermal neutron intensities of the optimized cold neutron moderator was compared with a cold neutron source based on solid methane, which has a beryllium target and the same proton energy as RANS. The results showed attractive application prospect of mesitylene as cold neutron moderator material.

Simulation code for estimating external gamma-ray doses from a radioactive plume and contaminated ground using a local-scale atmospheric dispersion model.

In this study, we developed a simulation code powered by lattice dose-response functions (hereinafter SIBYL), which helps in the quick and accurate estimation of external gamma-ray doses emitted from a radioactive plume and contaminated ground. SIBYL couples with atmospheric dispersion models and calculates gamma-ray dose distributions inside a target area based on a map of activity concentrations using pre-evaluated dose-response functions. Moreover, SIBYL considers radiation shielding due to obstructions such as buildings. To examine the reliability of SIBYL, we investigated five typical cases for steady-state and unsteady-state plume dispersions by coupling the Gaussian plume model and the local-scale high-resolution atmospheric dispersion model using large eddy simulation. The results of this coupled model were compared with those of full Monte Carlo simulations using the particle and heavy-ion transport code system (PHITS). The dose-distribution maps calculated using SIBYL differed by up to 10% from those calculated using PHITS in most target locations. The exceptions were locations far from the radioactive contamination and those behind the intricate structures of building arrays. In addition, SIBYL's computation time using 96 parallel processing elements was several tens of minutes even for the most computationally expensive tasks of this study. The computation using SIBYL was approximately 100 times faster than the same calculation using PHITS under the same computation conditions. From the results of the case studies, we concluded that SIBYL can estimate a ground-level dose-distribution map within one hour with accuracy that is comparable to that of the full Monte Carlo simulation.

Measurement of thick target neutron yield at 180°for a mercury target induced by 3-GeV protons

A thick target neutron yield (TTNY) for a mercury target at an angle of 180°from the incident beam direction was measured with the time-of-flight method using a 3-GeV proton beam at the Japan Proton Accelerator Research Complex (J-PARC). Comparison of the experimental results with a Monte Carlo particle transport simulation by the Particle and Heavy Ion Transport code System (PHITS) shows that the default spallation model of PHITS, the Liège intranuclear cascade model version 4.6 coupled with the generalized evaporation model (INCL4.6/GEM) reproduces the measured TTNY fairly well, whereas apparent overestimation is observed for the former default model, the Bertini intranuclear cascade model coupled with GEM (Bertini/GEM). These trends are consistent with the experimental results of neutron-induced reaction rates obtained using indium and niobium activation foils.

Multimodal Monte Carlo treatment system capable of microdosimetry with PHITS

Boron neutron capture therapy (BNCT) is a radiation therapy that combines neutrons and boron drugs. An industry-academia-government collaboration team is currently developing the linac-based treatment device and several peripheral devices to establish this therapy. In the project, a demonstration device, “iBNCT001”, for a linac-based neutron source applicable to BNCT treatment is being developed. We are developing Tsukuba Plan, a multimodal Monte Carlo-based treatment planning system, in addition to the linac-based neutron source device. Tsukuba-Plan has adopted a Particle and Heavy Ion Transport code System (PHITS) as the Monte Carlo-based dose calculation engine. PHITS has a microdosimetric function that can compute a lineal energy distribution and determine relative biological effectiveness through combination with the stochastic microdosimetric kinetic model parameters. By utilising these functions, Tsukuba-Plan is expected to estimate the weighted doses administered to the tumour region and normal tissues in the irradiation field more accurately.

Relation between biomolecular dissociation and energy of secondary electrons generated in liquid water by fast heavy ions

In this work, we measured and simulated the dissociation of biomolecules in liquid water induced by secondary electrons ejected from water molecules during fast heavy-ion irradiation. We calculated the energy spectra of secondary electrons generated along carbon ion tracks in liquid water in the Bragg peak region. The calculation was done using the Particle and Heavy Ion Transport code System (PHITS) in carbon track structure mode. This mode enables simulation of inelastic collisions along a carbon ion track based on the cross sections considered in the Monte Carlo code KURBUC. To understand the biomolecular dissociation processes in our previous MeV-SIMS experiments with microdroplet targets of glycine solution, we calculated the collision spectra of secondary electrons produced near liquid surfaces using PHITS. Furthermore, we examined the relationship between the secondary electron energy and formation of positive and negative glycine fragments. The results showed that the formation of methylene amine cations is caused by secondary electrons with energies of 13–100 eV. The formation of glycine-related negative ions such as cyanide anion, formate anion, and deprotonated glycine was found to be caused by low-energy (less than 13 eV) secondary electrons. These ions are known products of dissociative electron attachment.

Encapsulated Gamma Source Contact Dose Conversion Factors: Updating NCRP-40 Guidance.

Secondary electron generation on the surface of encapsulated gamma sources can play a large role in the dose measured near the surface of the encapsulation. The National Council on Radiation Protection and Measurements Report No. 40 contains contact dose rate conversion factors for encapsulated gamma sources, along with recommended secondary electron correction factors. However, secondary electron correction factors were based on experiments performed in the 1930s and 1940s with encapsulated radium sources, and the correction factors for the other sources listed in the report were estimated based on these radium source measurements. Monte Carlo simulations were performed using the Particle and Heavy Ion Transport code System (PHITS) to calculate the contact dose rate conversion factors for each encapsulated gamma source presented in NCRP-40, taking into account the dose from both gamma rays and secondary electrons. These simulations showed that the contact dose rate conversion factors are much lower than those presented in NCRP-40, and the secondary electron contribution was much greater than the values proposed by NCRP-40. The original research used results from encapsulated 226Ra experiments to determine the secondary electron correction factors for NCRP-40. To support the current Monte Carlo calculations, experiments were conducted using an encapsulated 137Cs source, rare earth magnet, and ion chamber detector to show that the secondary electron correction factors presented in NCRP-40 were not applicable to the geometry of tissue in direct contact with the encapsulation. In this work, contact dose conversion factors for common encapsulated radionuclide sources are presented.

Development of INC-ELF version 2 and incorporation into PHITS: Extension to lower incident energies down to 30 MeV

Version 2 of the Intranuclear Cascade with Emission of Light Fragments (INC-ELF) code is developed and incorporated into the Particle and Heavy Ion Transport code System (PHITS). The aim of INC-ELF2 is to expand it to the lower incident-energy range of 30–200 MeV, where the original INC theory is not applicable. The refinement of the code is based faithfully on the results of basic research on direct reactions within the framework of the INC model. INC-ELF2 can deal with the emission of light fragments from nuclear reactions induced by protons and neutrons. Double-differential cross-sections are calculated by a two-step model combining INC-ELF2 and GEM (generalized evaporation model) in PHITS, and they are compared with experimental data to examine the predictive ability. Excellent agreement is confirmed for various reactions over a wide range of target masses and over the incident-energy range of 30–200 MeV.

Dose Analysis of Boron Neutron Capture Therapy (BNCT) Treatment for Lung Cancer Based on Particle and Heavy Ion Transport Code System (PHITS)

The objectives of this study were to determine the effect of boron concentration on total dose rate for lung cancer treatment, and to determine the effect of boron concentration on the length of irradiation time for lung cancer treatment. This study was computer simulation-based using the Particle and Heavy Ion Transport code System (PHITS) by defining the geometry and components of lung cancer and the surrounding organism as the object being studied and the source of radiation used. The type of phantom used was the ORNL of an adult Asian male. The neutron source used was Kartini Reactor. The independent variable was the boron concentration of 30, 40, 50, 60, and 70 ?g/g cancer tissue and the dependent variables were the dose rate and the irradiation time. The results of this study indicated that the larger the amount of boron concentration that was injected, the higher the rate of total dose the organ received, where the total dose rate for each variation of boron concentration were 1.34 × 10-3 Gy/s, 1.71 × 10-3 Gy/s, 2.07 × 10-3 Gy/s, 2.42 × 10-3 Gy/s, and 2.78 × 10-3 Gy/s, and the larger the amount of boron concentration that was injected, the faster the irradiation time for the treatment of lung cancer was, where the irradiation time required for each variation of boron concentration was 37294 s, 29240 s, 24180 s, 20633 s, and 17996 s.

Conceptual design of a subcritical research reactor for inertial fusion energy with the J-EPoCH facility

The Japan Establishment for a Power-laser Community Harvest (J-EPoCH) is proposed as a next generation laser facility having multi-purpose high repetition laser beams at the maximum rate of 100 Hz. One of the experimental areas is assigned for laser fusion research. The omnidirectional 12 laser beams with 8 kJ and a 5 PW heating laser are arranged. A Large High Aspect Ratio Target (LHART) with the laser beams would yield ~1013 neutrons. The fusion energy corresponds to ~22 J/shot. As one of the J-EPoCH applications, a laser fusion subcritical research reactor with replaceable cores has been conceptually designed based on existing technologies. Moreover,we can conduct a variety of fusion engineering studies: energy conversion, tritium breeding, neutron irradiation effects and other subjects. The feasibility of tritium breeding and preliminary energy conversion studies are proven according to calculations by the Particle and Heavy Ion Transport code System (PHITS).

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 9cm in length weighing 23.9g. 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 1Bq/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 (31.4g) and an ellipsoid crab egg mass (12.6g) reported in ICRP Publication 108.

Estimation of reliable displacements-per-atom based on athermal-recombination-corrected model in radiation environments at nuclear fission, fusion, and accelerator facilities

The displacements-per-atom (dpa) is widely used as an exposure unit to predict the operating lifetime of materials in radiation environments. Because the athermal-recombination-corrected dpa (arc-dpa) model is a more realistic model than the standard Norgett–Robinson–Torrens (NRT) model, new evaluation of radiation damage for various applications at nuclear fission, fusion, and accelerator facilities will be performed using the arc-dpa model as a standard. Determination of the rescaling factor from the NRT-dpa to arc-dpa model for various applications is also important. In this work, the recent arc-dpa model of various materials including C, Al, Si, Fe, Cu, and W are incorporated in the Particle and Heavy Ion Transport code System (PHITS), and the rescaling factors (NRT-dpa/arc-dpa) over a wide energy range are reported. For neutron incidences with the energy spectrum determined in selected nuclear facilities and proton incidences with energies of 600 MeV–50 GeV, the rescaling factor for each material is independent of these irradiation conditions with almost the same value for each material. This value is mainly determined by the saturation in the damage production efficiency in the damage energy region over approximately 1 keV. Our findings will be beneficial for rescaling the NRT-dpa used for radiation damage applications over a wide energy region at various facilities.

Evaluation of a back-illuminated solid state detector with thin dead layer for super heavy element research

The precise energy measurement of charged particles using a silicon semiconductor detector (SSD) is very important in studies of the characteristics of superheavy elements (SHEs). It is necessary to evaluate the dead layer thickness of the SSD to measure the recoil energy of charged particles, such as SHEs, and α particles or fission fragments that are emitted upon the successive decay from the SHEs. Since the energy resolution of the SSD depends on the thickness of the dead layer, a back-illuminated SSD with a thin dead layer has been introduced for SHE research. To evaluate the dead layer thickness of the SSD, energy spectra were first investigated by irradiating 241Am α particles at incident angles of 0° and 45° for two types of SSD. The dead layer thicknesses of the SSDs were deduced by a computer simulation using the particle and heavy ion transport code system (PHITS).

Evaluation for gamma-ray rejection ability affecting neutron discrimination property in scintillating-fiber type of fast neutron detector

Scintillating-fiber (Sci-Fi) detector has been employed to measure 14 MeV neutrons in the magnetic confinement fusion experiment. Design of the Sci-Fi detector reduces response to gamma-rays, compared to bulk scintillators, while at the same time enabling directional responses to energetic neutrons. Ability of gamma-ray rejection is an important parameter to judge the discrimination property for a fast neutron detector. In order to investigate the gamma-ray rejection abilities in the case of 2.45 MeV neutron and higher-energy neutrons, the pulse height spectra (PHS) of four different-head detectors in different diameter and with/without the aluminum (Al) matrix have been measured by using the accelerator-based neutron source with d–D reaction. Four different-head detectors show the different pulse height properties and the gamma-ray rejection abilities, which impact upon the neutron discrimination abilities. Meanwhile, simulations of detector responses have been performed using the Particle and Heavy Ion Transport code System (PHITS) to analyze the experimental results. Neutron discrimination abilities of small diameter Sci-Fi is better than that of large diameter Sci-Fi. Al matrix is beneficial to enhance 6.6–10 times difference of the gamma-ray rejection affecting the neutron discrimination ability in the case of Sci-Fi with/without Al matrix. It is also found that the neutron discrimination abilities of the Sci-Fi without the Al matrix and a plastic scintillator are the same.

Development of a predictive capability of short-pulse laser-driven broadband x-ray radiography

High intensity, short-pulse laser interaction with a solid metal target produces broadband hard x-rays potentially for various applications of x-ray radiography. Here experimental benchmarking of numerical modelling for short-pulse laser-driven broadband x-ray radiography is presented. Angular dependent x-ray spectra are first calculated with a hybrid particle-in-cell code, Large Scale Plasma (LSP), using fast electron parameters inferred from an analysis of measured bremsstrahlung signals. Subsequently, a calculated x-ray spectrum in the direction of radiography is used in photon transport calculations using a Monte Carlo code, Particle and Heavy Ion Transport code System (PHITS), to simulate a radiographic image including a modelled 3D test object, an x-ray attenuation filter and an image plate detector. Simulated radiographic images are compared with measurements obtained in an experiment using a 50-TW Leopard short-pulse laser at the University of Nevada Reno. Results show that simulations reproduce the experimental images well for three different attenuation filters (plastic, aluminium, and brass), while 1D transmission profiles for the plastic and aluminium filters are quantitatively in good agreement. The modelling approach established in this work could be used as a predictive tool to simulate radiographic images of complex 3D solid objects at any arbitrary angular position or to optimize experimental components such as the source spectrum, x-ray attenuation filters and a detector type depending on a radiographic object without carrying out radiographic experiments.

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.&#x0D; Key words: Cancer, BNCT, Dose of Boron, PHITS.&#x0D;

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.