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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 order to perform an integral evaluation of 233U nuclear data, measurements of substitution reactivity worth between 233U sample and high enriched uranium (HU) sample in the Kyoto University Critical Assembly (KUCA) were carried out. In this study, the experimental core was consisted of 24 fuel elements and 1 special fuel element with the 233U or HU sample. The fuel element was consisted of 31 unit cells and sandwiched by a upper and a lower polyethylene reflector. The unit has one enriched uranium plate of 1/16” thickness, three polyethylene plates of 1/8” thickness. The special fuel element was loaded at the center of the core. The substitution reactivity worth was defined as difference between the excess reactivity of the core in which the 233U sample or the HU sample inserted the core. The excess reactivities ware measured by the positive period method. As the result, the substation reactivity was obtained 0.0146±0.0006 %∆k/k. A numerical result by MVP3 with JENDL-4.0 was 0.0141±0.0002 %∆k/k and a C/E was 0.964 ± 0.041. On the other hand, the calculated substitution reactivity worth with JENDL-5 was 0.0168±0.0002 %∆k/k and a C/E was 1.151±0.047.

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.

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.

To develop the 100-400 MeV quasi-monoenergetic neutron field, we measured neutron and unexpected gamma-ray energy spectra of the 7Li(p,n) reaction with 80-389 MeV protons in the 100-m time-of-flight (TOF) tunnel at the Research Center for Nuclear Physics (RCNP) cyclotron facility. Neutron energy spectra with energies above 3 MeV were measured by the TOF method, which had been reported in our previous papers, and photon energy spectra with energies above 0.1 MeV were measured by the automatic unfolding function of the radiation dose monitor DARWIN. For neutron spectra, the contribution of peak intensity to the total intensity integrated with energies above 3 MeV varied between 0.38 and 0.48 in the proton energy range of 80–389 MeV. For gamma-ray spectra, highenergetic gamma-rays at around 70 MeV originated from the decay of π0 were observed with proton energies higher than 200 MeV. For the 246-MeV proton incident reaction, the contribution of gamma-ray dose to neutron dose is negligible because the ratio of gamma-ray dose to neutron dose is 0.014.

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 7Li(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.

This paper describes the results of neutron spectrometry and dosimetry measurements using an extended range Bonner Sphere Spectrometer (ERBSS) with 3He proportional counter performed in quasi-mono-energetic neutron fields at the ring cyclotron facility of the Research Center for Nuclear Physics (RCNP), Osaka University, Japan. Using 100 MeV and 296 MeV proton beams, neutron fields with nominal peak energies of 96 MeV and 293 MeV were generated via 7Li(p,n)7Be reactions. Neutrons produced at 0° and 25° emission angles were extracted into the 100 m long 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 ERBSS data were unfolded using the MSANDB unfolding code. At high energies, the neutron spectra were also measured by means of the TOF method using NE213 organic liquid scintillators. The results are discussed in terms of ambient dose equivalent, H*(10), and compared with the readings of other instruments operated during the experiment.

To improve accuracy of neutron nuclear data on minor actinides, a Japanese nuclear data project entitled “Research and development for Accuracy Improvement of neutron nuclear data on Minor ACtinides (AIMAC)” has been implemented. Several independent measurement techniques were developed for improving measurement precision at J-PARC/MLF/ANNRI and KURRI/LINAC facilities. Effectiveness of combining the independent techniques has been demonstrated for identifying bias effects and improving accuracy, especially in characterization of samples used for nuclear data measurements. Capture cross sections and/or total cross sections have been measured for Am-241, Am-243, Np-237, Tc-99, Gd-155, and Gd-157. Systematic nuclear data evaluation has also been performed by taking into account the identified bias effect. Highlights of the AIMAC project are outlined.

In this study, we carried out Monte Carlo simulations to obtain the energy resolution of the neutron flux for TOF measurements in the KURRI-LINAC pulsed neutron facility. The simulation was performed on the moderated neutron flux from the pac-man type moderator at the energy range from 0.1 eV to 10 keV. As the result, we obtained the energy resolutions (ΔE/E) of about 0.7% to 1.3% between 0.1 eV to 10 keV. The energy resolution obtained from Monte Carlo simulation agreed with the resolution using the simplified evaluation formula. In addition, we compared the energy resolution among KURRI-LINAC and other TOF facilities, the energy dependency of the energy resolution with the pac-man type moderator in KURRI-LINAC was similar to the J-PARC ANNRI for the single-bunch mode.

A shielding benchmark experiment has been performed using a quasi-monoenergetic 7Li(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.

The neutron capture cross section measurement on 91Zr was performed at neutron TOF beam line ANNRI installed at J-PARC/MLF. Prompt capture gamma rays from the sample were detected with an array of large Ge detectors at a distance of 21.5 m from the spallation neutron source by the time-of-fligh (TOF) method. The neutron capture gamma-ray pulse-height spectra from the 182-eV p-wave resonance and the 292-eV s-wave resonance were obtained by gating on the TOF regions, respectively. Though the decay patterns of primary transitions from the capture state were quite different between resonances, the prominent characteristics common to both resonances was the very strong ground-state transition from the 935-keV state. Therefore, a ground-state transition method was applied to obtain the capture yield, so that the background components due to impurities were successfully eliminated. The preliminary result of the neutron capture cross section for 91Zr up to 5 keV is presented.

Improvement of accuracy of neutron nuclear data for minor actinides (MAs) and long-lived fission products (LLFPs) is required for developing innovative nuclear system transmuting these nuclei. In order to meet the requirement, the project entitled as “Research and development for Accuracy Improvement of neutron nuclear data on Minor ACtinides (AIMAC)” has been started as one of the “Innovative Nuclear Research and Development Program” in Japan at October 2013. The AIMAC project team is composed of researchers in four different fields: differential nuclear data measurement, integral nuclear data measurement, nuclear chemistry, and nuclear data evaluation. By integrating all of the forefront knowledge and techniques in these fields, the team aims at improving the accuracy of the data. The background and research plan of the AIMAC project are presented.

A shielding experiment using activation detectors was performed at the CERN High-energy AcceleRator Mixed-field (CHARM) facility and the CERN Shielding Benchmark Facility (CSBF). The protons (24 GeV/c) were bombarded into a 50-cm-thick copper target, and the released neutrons were transmitted through various shields located vertically upward from the target. Ordinary concrete slabs of 40- to 160-cm thicknesses and steel slabs of 20- to 80-cm thicknesses were installed by changing the material and the thickness of the shield. Activation detectors of bismuth, indium, aluminum, and graphite were variously placed in these shields and activated by high energy neutrons. From the radionuclide production rate in the activation detectors, the attenuation profiles through the various shielding materials were obtained for the reactions of 209Bi(n,xn)210-xBi(x = 4–9), 115In(n,n')115mIn, 27Al(n,α)24Na, and 12C(n,2n)11C. Estimated attenuation lengths of high energy neutrons through concrete and steel were compared with cited data and discussed. Monte Carlo simulations using FLUKA, PHITS, and GEANT4 were also performed as benchmark calculations, and they agreed with the experimental data, generally within a factor of 2. [ABSTRACT FROM AUTHOR]