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We developed a LOcal-scale High-resolution atmospheric DIspersion Model using Large-Eddy Simulation (LOHDIM-LES) for the safety assessment of nuclear facilities and emergency responses to accidenta...

For validating the number of displacements per atom (dpa) for tungsten under high-energy proton irradiation, we measured displacement cross sections related to defect-induced electrical resistivity changes in a tungsten wire sample under irradiation with 389-MeV protons under 10 K. The Gifford–McMahon cryocooler was used to cool the sample using a conductive coolant via thermal conduction plates of oxygen-free high-conductivity copper and electrical insulation sheets of aluminum nitride ceramic. In this experiment, the displacement cross section was 1612 ± 371 b for tungsten at 389 MeV. A comparison of the experimental displacement cross sections of tungsten with the calculated results obtained using Norgett–Robinson–Torrens (NRT) dpa and athermal recombination-corrected (arc) dpa cross sections indicates that arc-dpa was in better agreement with the experimental data than NRT-dpa; this is similar to the displacement cross sections of copper. From the measurements of damage recovery of the accumulated defects in tungsten through isochronal annealing, which is related to the defect concentration of the sample, approximately 20% of the damage was recovered at 60 K. This trend was similar to those observed in other experimental results for reactor neutrons.

This article presents nuclide-specific organ dose rate coefficients for environmental external exposures due to soil contamination assumed as a planar source at a depth of 0.5 g cm−2in the soil and submersion to contaminated air, for a pregnant female and its fetus at the 24th week of gestation. Furthermore, air kerma free-in-air coefficient rates are listed. The coefficients relate the organ equivalent dose rates (Sv s−1) to the activity concentration of environmental sources, in Bq m−2or Bq m−3, allowing to time-integrate over a particular exposure period. The environmental radiation fields were simulated with the Monte Carlo radiation transport codes PHITS and YURI. Monoenergetic organ dose rate coefficients were calculated employing the Monte Carlo code EGSnrc simulating the photon transport in the voxel phantom of a pregnant female and fetus. Photons of initial energies of 0.015–10 MeV were considered including bremsstrahlung. By folding the monoenergetic dose coefficients with the nuclide decay data, nuclide-specific organ doses were obtained. The results of this work can be employed for estimating the doses from external exposures to pregnant women and their fetus, until more precise data are available which include coefficients obtained for phantoms at different stages of pregnancy.

Double-differential neutron production cross sections (DDXs) for 200-MeV deuteron induced reactions on Li, Be, C, Al, Cu, Nb, In, Ta, and Au were measured at forward emission angles ranging from 0° to 25° by means of a time of flight (TOF) method with EJ301 liquid organic scintillators. The measured DDXs were compared with JENDL-5 and TENDL-2021 nuclear data, and theoretical model calculations using DEURACS and PHITS codes. It was found that the JENDL-5 and DEURACS calculation are in better agreement with the measured DDXs than the PHITS calculation, while TENDL-2021 fails to reproduce both the spectral shape and magnitude of the measured DDXs for all the targets.

For the research and development of accelerator-driven systems (ADSs) and fundamental ADS reactor physics research using the Kyoto University Critical Assembly combined with the fixed-field alternating gradient (FFAG) accelerator, we are conducting a series of experiments on double-differential neutron yields using the FFAG accelerator at Kyoto University. This paper presents an overview of the experiments together with preliminary results.

In this study, the neutron-production double-differential cross sections (DDXs) of nat C, 27Al, nat Fe, and nat Pb bombarded by 20, 34, 48, 63, and 78 MeV protons were measured in the most-forward direction by the time-of-flight (TOF) method using a liquid organic scintillator . Characteristic peak structures were observed in the measured neutron spectra, especially for light nuclei. The experimental data were compared with the results from the Monte Carlo simulation code “Particle and Heavy Ion Transport code System” (PHITS) using the theoretical model Liege Intranuclear Cascade (INCL) or using the Japanese evaluated nuclear data library (JENDL-4.0/HE). While the results of JENDL-4.0/HE were in agreement with the experimental data better than those of INCL, neither could reproduce the measured peak structures for some combinations of the beam and target. For the energy-integrated cross sections, JENDL-4.0/HE reproduced the experimental data within a factor of two in all cases. On the other hand, INCL gave values up to approximately six times larger than the experimental ones.

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.

Double-differential neutron production cross sections (DDXs) for deuteron-induced reactions on Li at 200 MeV were measured for emission angles ranging from 0◦ to 25◦ in steps of 5◦ by means of a time of flight (TOF) method with EJ301 liquid organic scintillators at the Research Center for Nuclear Physics (RCNP), Osaka University. The measured DDXs were compared to theoretical model calculations with the DEURACS and PHITS codes and TENDL-2017 nuclear data. It was found that the DEURACS calculation is in better agreement with the measured DDXs than the PHITS calculation, while TENDL-2017 fails to reproduce both the spectral shape and magnitude of the measured DDXs for all angles.

To validate the Monte Carlo codes for prediction of radiation damage in metals irradiated by > 100 MeV protons, we developed a proton irradiation device with a Gifford–McMahon (GM) cryocooler to cryogenically cool two 0.25-mm-diameter wire samples of aluminum and copper. By using this device, the defect-induced electrical resistivity changes related to the displacement cross section of copper and aluminum were measured under irradiation with 200-MeV protons at 5 K at the beamline of the cyclotron facility at RCNP, Osaka University. After irradiation to a 3.89 × 1018 proton/m2 flux, the damage rate of the aluminum sample was 1.30 × 10−31 Ωm3/proton at 185 MeV and that of copper was 3.60 × 10−31 Ωm3/proton at 196 MeV. Based on measurements of recovery of the accumulated defects in aluminum and copper through isochronal annealing, which is related to the defect concentration in the sample, about 50% of the damage remained at 40 K, with the same tendency observed in other experimental results for reactor neutron, fusion neutron, and 125-MeV proton irradiations. A comparison of the measured displacement cross sections with the calculated results of the NRT-dpa and the athermal-recombination-corrected displacement damage (arc-dpa) cross sections indicates that arc-dpa with the defect production efficiencies provided by Almazouzi for aluminum and Nordlund for copper provide better quantitative descriptions of the displacement cross section than NRT-dpa.

The personal dose equivalent was calculated for the public (newborns; 1-, 5-, 10-, and 15-year-old children; and adults) in an environment contaminated with radioactive cesium (134Cs and 137Cs) distributed in a soil at specific depths of 0.0, 0.5, 2.5, 5.0, 10.0, and 50.0 g/cm2, and the conversion coefficients for converting activity concentration to personal dose equivalent rate were provided for each age group. Monte Carlo calculations were performed using pediatric and adult computational phantoms incorporated into a particle and heavy ion transport code system. Compared with the effective dose and ambient dose equivalent at a height of 100 cm above the ground, the personal dose equivalent was found to provide a good estimate as a measurable quantity for the effective dose and did not exceed the ambient dose equivalent even in the environmental radiation fields, while the personal dose equivalent values increased for younger subjects. The weighted-integral method to obtain the personal dose equ...

Quasi-monoenergetic high-energy neutron fields induced by 7 Li(p,n) reactions are used for the response evaluation of neutron-sensitive devices. The quasi-monoenergetic high-energy field consists of high-energy monoenergetic peak neutrons and unwanted continuum neutrons down to the low-energy region. A two-angle differential method has been developed to compensate for the effect of the continuum neutrons in the response measurements. In this study, the two-angle differential method was demonstrated for Bonner sphere detectors, which are typical examples of moderator-based neutron-sensitive detectors, to investigate the method's applicability and its dependence on detector characteristics. Experiments were performed under 96–387 MeV quasi-monoenergetic high-energy neutron fields at the Research Center for Nuclear Physics (RCNP), Osaka University. The measurement results for large high-density polyethylene (HDPE) sphere detectors agreed well with Monte Carlo calculations, which verified the adequacy of the two-angle differential method. By contrast, discrepancies were observed in the results for small HDPE sphere detectors and metal-induced sphere detectors. The former indicated that detectors that are particularly sensitive to low-energy neutrons may be affected by penetrating neutrons owing to the geometrical features of the RCNP facility. The latter discrepancy could be consistently explained by a problem in the evaluated cross-section data for the metals used in the calculation. Through those discussions, the adequacy of the two-angle differential method was experimentally verified, and practical suggestions were made pertaining to this method.

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.

We completed a series of measurements on mono-energetic neutron energy spectra of the 7 Li(p,n) reaction with 80–389-MeV protons in the 100-m time-of-flight (TOF) tunnel at the Research Center for Nuclear Physics cyclotron facility. For that purpose, we measured neutron energy spectra of the 80-, 100- and 296-MeV proton incident reactions, which had not been investigated in our previous studies. The neutron peak intensity was 0.9–1.1×10 10 neutrons/sr/µC in the incident proton energy region of 80–389 MeV, and it was almost independent of the incident proton energy. The contribution of peak intensity of the spectrum to the total intensity integrated with energies above 3 MeV varied between 0.38 and 0.48 in the incident proton energy range of 80–389 MeV. To consider the correction required to derive a response in the peak region from the measured total responses of neutron monitors in the 100-m TOF tunnel, we proposed the subtraction method using energy spectra between 0° and 25°. The normalizing factor k against 25° neutron fluence to equalize it to 0° neutron fluence in the continuum region ranges from 0.74 to 1.02 depending on the incident proton energy and angle measured. Even without the TOF method, the subtraction method with the k factor almost decreases the response in the continuum region of a neutron spectrum against the total response of neutron monitors.

To estimate effective doses for members of the public exposed to external radiation from radioactive cesium (134Cs and 137Cs) deposited on the ground by the Fukushima nuclear accident, we calculate the conversion coefficients for converting activity concentration to effective dose rate by using the Particle and Heavy Ion Transport Code System. The data were produced from different age groups within the public (newborns; 1-, 5-, 10-, and 15-year-old children; and adults) for the situations in which radioactive cesium is distributed uniformly in the soil over a planar area and at specific depths of 0.0, 0.5, 2.5, 5.0, 10.0, and 50.0 g/cm2. On the basis of the results, we also derive the conversion coefficients for exponentially distributed volumetric sources. In addition, we obtain the conversion coefficients that give the effective dose accumulated over the first and second months, the first year, and over a lifetime (50 years) because of the contamination remaining on the ground. These calculations indica...

With the aim of providing critical nuclear data of primary knock-on atoms (PKAs) created from proton-induced spallation reactions, a new detection system was designed and dedicated to the PKA measurement that requires lower measurement threshold energies and superior mass resolution than the conventional experimental setups. Such requirements can be fulfilled by employing the TOF-E and dE-E methods, and this new PKA detection system, which consists of two fast timing detectors and one dE-E gas ionization detector. The design of the detectors and the experimental setup, along with the estimated system performed are briefly described in this paper.

A shielding benchmark experiment has been performed using a quasi-monoenergetic 7 Li(p,n) neutron source with the peak energies of 244 and 387 MeV at the Research Center for Nuclear Physics (RCNP) of Osaka University, in order to assess the accuracy of nuclear data and calculation codes used in high-energy accelerator shielging design. Energy spectra behind bulk shields of 10- to 100-cm-thick iron, 25- to 300-cm-thick concrete and their composite are measured using a NE213 organic liquid scintillator with a diameter and thickness of 25.4 cm each with a time-of-flight and an unfolding method. The neutron attenuation lengths are illustared for iron and concrete as a function of the incident energy.

Shielding benchmark experiments are useful to verify the accuracy of calculation methods for the radiation shielding designs of high-energy accelerator facilities. In the present work, the benchmark experiments were carried out for 244- and 387-MeV quasi-monoenergetic neutron field at RCNP of Osaka University. Neutron dose rates through the test shields, 100-300 cm thick concrete and 40-100 cm thick iron, were measured by four kinds of neutron dose equivalent monitors, three kinds of wide-energy range monitors applied to high-energy neutron fields above 20 MeV and a conventional type rem monitor for neutrons up to 20 MeV, placed behind the test shields. Measured dose rates were compared one another. Measured results with the wide-energy range monitors were in agreement one another for both the concrete and the iron shields. For the conventional type rem monitor, measured results are smaller than those with the wide-energy range monitors for the concrete shields, while that are in agreements for the iron shields. The attenuation lengths were obtained from the measurements. The lengths from all the monitors are in agreement on the whole, though some differences are shown. These results are almost same as those from others measured at several hundred MeV neutron fields.

Neutron energy spectra behind concrete and iron shields were measured for quasi-monoenergetic neutrons above 200 MeV using a Bonner sphere spectrometer (BSS). Quasi-monoenergetic neutrons were produced by the 7 Li(p,xn) reaction with 246-MeV and 389-MeV protons. Shielding materials are concrete blocks with thicknesses from 25 cm to 300 cm and iron blocks with thicknesses from 10 cm to 100 cm. The response function of BSS was also measured at neutron energies from 100 MeV to 387 MeV. In data analysis, the measured response function was used and the pingpong scattering effect between the BSS and the shielding material was considered. The neutron energy spectra behind the concrete and iron shields were obtained by the unfolding method using the MAXED code. Ambient dose equivalents were obtained as a function of a shield thickness successfully.

Double-differential cross sections (DDXs) of deuteron-induced neutron production reactions on Li, Be, and C at 102 MeV were measured at forward angles (≤ 25 ∘ ) by means of a time of flight method with NE213 liquid organic scintillators at the Research Center of Nuclear Physics, Osaka University. The experimental results were compared with model calculations with PHITS and DEURACS. The DEURACS calculation reproduces the experimental DDXs for C at very forward angles than the PHITS one. Moreover, the incident energy dependence of the Li(d,xn) reaction was investigated by adding the DDX data measured previously at 25 and 40 MeV.

Radiation monitoring at high energy proton accelerators poses a considerable challenge due to the complexity of the encountered stray radiation fields. These environments comprise a wide variety of different particle types and span from fractions of electron-volts up to several terra electron-volts. As a consequence the use of Monte Carlo simulation programs like FLUKA is indispensable to obtain appropriate field-specific calibration factors. At many locations of the LHC a large contribution to the particle fluence is expected to originate from high-energy neutrons and thus, benchmark experiments with mono-energetic neutron beams are of high importance to verify the aforementioned detector response calculations. This paper summarizes the results of a series of benchmark experiments with quasi mono-energetic neutrons of 100, 140, 200, 250 and 392 MeV that have been carried out at RCNP - Osaka University, during several campaigns between 2006 and 2014.

Bonner Spheres Spectrometry in its high-energy extended version is an established method to quantify neutrons at a wide energy range from several meV up to more than 1 GeV. In order to allow for quantitative measurements, the responses of the various spheres used in a Bonner Sphere Spectrometer (BSS) are usually simulated by Monte Carlo (MC) codes over the neutron energy range of interest. Because above 20 MeV experimental cross section data are scarce, intra-nuclear cascade (INC) and evaporation models are applied in these MC codes. It was suspected that this lack of data above 20 MeV may translate to differences in simulated BSS response functions depending on the MC code and nuclear models used, which in turn may add to the uncertainty involved in Bonner Sphere Spectrometry, in particular for neutron energies above 20 MeV. In order to investigate this issue in a systematic way, EURADOS (European Radiation Dosimetry Group) initiated an exercise where six groups having experience in neutron transport calculations with the MC codes (MCNP, MCNPX, FLUKA, PHITS, MARS, or GEANT4) calculated the responses of a bare 3 He proportional counter, a 3 He proportional counter embedded in the middle of a 9 inch polyethylene sphere, and a 3 He proportional counter centred in a 9 inch polyethylene sphere containing a lead shell, at neutron energies of 1, 10, 100, and 1000 MeV. In general, calculated responses agreed very well for neutron energies below 20 MeV, whatever MC code used. At higher energies, however, certain differences were observed among the different calculations, which may mainly be attributed to the application of different INC models and their parameters. It was found that up to 1 GeV most of the results ranged between calculations previously published that were obtained with MCNP/LAHET using the Bertini INC model and GEANT4 using the Binary and Bertini INC models. These results indicate that use of different MC codes and INC models for the calculation of BSS response functions may result in an uncertainty of unfolded neutron fluences above 20 MeV of about 20%.

Neutron double-differential cross sections from carbon ion incident on carbon, nitrogen and oxygen targets have been measured for neutron energies down to 0.6 MeV in wide directions from 15 ∘ to 90 ∘ with 100- and 290-MeV/A incident energies at the Heavy Ion Medical Accelerator in Chiba (HIMAC), National Institute of Radiological Sciences. Two sizes of NE213 scintillators were used as neutron detectors in order to enable neutron energy from below one to several hundred MeV. The neutron energy was measured by the time-of-flight technique between the beam pickup detector and an NE213 scintillator. By using the experimental data, the validity of the calculation results by the PHITS code was examined.

A calculation system for the estimation of decontamination effects (CDE) is developed in the present work to aid in the effective planning of decontamination procedures. This system calculates dose...

Yoshihiro Nakane, Masayuki Hagiwara, Yosuke Iwamoto, Hiroshi Iwase, Daiki Satoh, Tatsuhiko Sato, Hiroshi Yashima, Tetsuro Matsumoto, Akihiko Masuda, Tomoya Nunomiya, Yukio Sakamoto, Hiroshi Nakashima, Atsushi Tamii, Kichiji Hatanaka and Takashi Nakamura Japan Atomic Energy Agency, 2-4 Shirakata, Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195, Japan; High Energy Accelerator Research Organization (KEK), 1-1 Oho, Tsukuba-shi, Ibaraki-ken, 305-0801, Japan; Research Reactor Institute, Kyoto University, 2 Asashiro-Nishi, Kumatori-cho, Sennan-gun, Osaka, 590-0494, Japan; National Institute of Advanced Industrial Science and Technology, 1-1-1 Higashi, Tsukuba-shi, Ibaraki-ken, 305-8561, Japan; Fuji Electric Co., Ltd., 1 Fuji-machi, Hino-shi, Tokyo, 191-8502, Japan; ATOX Co., Ltd., 1201 Takada, Kashiwa-shi, Chiba, 277-0861, Japan; Research center for Nuclear Physics, 10-1 Mihogaoka, Ibaraki-shi, Osaka University, Osaka, 567-0047, Japan; Tohoku University, 6-6 Aoba, Aramaki, Aoba-ku, Sendai, 980-8578, Japan

National Institute of Advanced Industrial Science and Technology, 1-1-1 Umezono, Tsukuba, Ibaraki, 305-8568 Japan; Tokyo Institute of Technology, 2-12-1 O-okayama, Meguro-ku, Tokyo, 152-8350 Japan; High Energy Accelerator Research Organization (KEK), 1-1 Oho, Tsukuba, Ibaraki, 305-0801 Japan; Japan Atomic Energy Agency, 2-4 Shirakata, Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195, Japan; e Kyoto University, 2-1010 Asayo-nishi, Kumatori-cho, Sennan-gun, Osaka, 590-0494 Japan; f German Reserach Center for Enviromental Health, 85764 Neuherberg, Germany; g Research Center for Nuclear Physics, Osaka University, 10-1 Mihogaoka, Ibaraki, Osaka, 567-0047 Japan; h Tohoku University, 6-3 Aoba, Aramaki, Aoba-ku, Sendai, Miyagi, 980-8578 Japan

Nobuhiro Shigyo, Yusuke Uozumi, Haruhiko Uehara, Tomoya Nishizawa, Keiichi Hirabayashi, Daiki Satoh, Toshiya Sanami, Yusuke Koba, Masashi Takada and Naruhiro Matsufuji Kyushu University, 744 Motooka, Nishi-ku, Fukuoka, 819-0935, Japan; Japan Atomic Energy Agency, 2-4 Shirakata, Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195, Japan; High Energy Accelerator Research Organization, 1-1 Oho, Tsukuba, Ibaraki, 305-0801, Japan; National Institute for Radiological Sciences, 4-9-1 Anagawa, Inage-ku, Chiba, 263-8555, Japan

aHigh Energy Accelerator Research Organization, 1-1 Oho, Tsukuba-shi, Ibaraki-ken, 305-0801, Japan; bJapan Atomic Energy Agency, 2-4 Shirakata, Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195, Japan; cNational Institute of Advanced Industrial Science and Technology, 1-1-1 Higashi, Tsukuba-shi, Ibaraki-ken, 305-8561, Japan; dResearch Reactor Institute, Kyoto University, 2-1010 Asashiro-nishi,Kumatori-cho, Sennan-gun, Osaka-hu, 590-0494, Japan; eResearch Center for Nuclear Physics,10-1 Mihogaoka, Ibaraki-shi, Osaka-hu. 567-0047, Japan; fShimizu Corporation,4-17, Etchujima 3-Chome, Koto-ku, Tokyo-to, 135-8530, Japan

Neutron activation cross sections for Bi and Co at 386 MeV were measured by activation method. A quasi-monoenergetic neutron beam was produced using the (7)Li(p,n) reaction. The energy spectrum of these neutrons has a high-energy peak (386 MeV) and a low-energy tail. Two neutron beams, 0° and 25° from the proton beam axis, were used for sample irradiation, enabling a correction for the contribution of the low-energy neutrons. The neutron-induced activation cross sections were estimated by subtracting the reaction rates of irradiated samples for 25° irradiation from those of 0° irradiation. The measured cross sections were compared with the findings of other studies, evaluated in relation to nuclear data files and the calculated data by Particle and Heavy Ion Transport code System code.

This paper describes the results of neutron spectrometry and dose measurements using a Bonner Sphere Spectrometer (BSS) at the ring cyclotron facility of the Research Center for Nuclear Physics (RCNP), Osaka University, Japan. Quasi-monoenergetic neutron fields were generated using the 7Li (p,n)7Be reaction and 246 and 389 MeV protons. Neutrons produced at 0° and 30° emission angles were extracted into a time-of-flight (TOF) tunnel, and the energy spectra were measured at a distance of 35 m from the target. To deduce the corresponding neutron spectra from thermal to the nominal maximum energy, the BSS data were unfolded using the MSANDB code and response functions were calculated by Monte Carlo (MC) methods. These spectra are compared to spectral measurements using NE213 organic liquid scintillators applying the TOF method. The results are discussed in terms of ambient dose equivalent H* (10) and compared with the readings of other instruments operated during the experiment.

The production cross sections of 10 Be, 26 Al and 36 Cl produced from Cu and Y targets irradiated with 300 MeV protons were measured by accelerator mass spectrometry. Using these cross section values in conjunction with those of the spallation products we confirm that nuclear fragmentation contributes to production of 10 Be in the 300 MeV proton-induced reactions for Cu and Y. These cross sections obtained are compared to cross sections of 10 Be, 26 Al and 36 Cl from 300 MeV proton-induced reactions for other target elements. We also confirm that the cross section values of 10 Be and 36 Cl for Cu and Y are consistent with the tendency of those cross section values to decrease with increasing target mass.

At the Fukui Prefectural Hospital Proton Therapy Center, neutron doses behind the concrete shields and at the maze were measured with three types of radiation monitors (DARWIN, Wendi-2, and a rem meter) along with solid-state nuclear track detectors. The measured data were compared with estimations of analytical models and the Monte Carlo code, Particle and Heavy-Ion Transport code System (PHITS). The analytical model, using the parameters employed in the shielding design of the facility, gave considerably larger values than the measured data. This means that the facility was designed with a sufficient margin of safety. The results calculated by PHITS were less than those of the analytical model and were about three times larger than the measured data. From a perspective that seeks conservative estimation with less margin, the Monte Carlo simulation is a good tool for shielding design of accelerator-based proton treatment facilities.

Response measurements of a Bonner sphere spectrometer (BSS) were performed for high-energy neutrons up to 400 MeV in quasi-monoenergetic neutron fields at the Research Center for Nuclear Physics (RCNP), Osaka University. The quasi-monoenergetic neutron field was generated by a 7Li(p,n) reaction with the resulting spectra consisting of a monoenergetic high-energy peak followed by an unwanted continuum tail down to the low-energy region. Two-angle differential measurements for a variety of Bonner spheres were performed to obtain monoenergetic responses by subtracting the tail contributions. Discussions were made on data analysis for tail subtraction. While each Bonner sphere had a different energy response, good agreements were seen in the responses between the MCNPX calculations and the measurements with a suitable subtraction. Response calculations were also made for the whole spectra obtained by unfolding the measured BSS data, with results that reproduce well the measured responses.

The neutron energy spectra of carbon, iron, and gold targets with 140 MeV protons at 180 were measured at the RCNP cyclotron facility using the time-of-flight technique. The results were compared with results calculated using the evaluated nuclear data file (the LA150 library) and the intranuclear cascade models (JAM and JQMD) implemented in the PHITS code. The results using LA150 and JQMD generally agreed with the experimental data except for the carbon target, but the JAM could not reproduce high-energy neutrons above 10 MeV because of the lack of pre-equilibrium reactions and the reflection and refraction of a nucleon with other nucleons in the nucleus, for example. Experimental neutron energy spectra at 180 were composed of two evaporation Maxwellian-type components (in two neutron energy regions of 1 – 4 MeV and 4 – 10 MeV) and a preequilibrium (in the neutron energy region above 10 MeV) Maxwellian-type component All experimental spectra were fitted well by the sum of three Maxwellian-type equations across the entire energy region of the neutrons.