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Graphite is a candidate of moderator in innovative nuclear reactors such as high-temperature gas-cooled reactors and molten salt reactors. Scattering of thermal neutrons by a moderator material has a significant impact on the reactor core design. To contribute to the development of the innovative nuclear reactors, thermal neutron scattering law data for nuclear-grade graphite were evaluated. Inelastic scattering component due to lattice vibration was evaluated based on phonon density of states computed with first-principles lattice dynamics simulations. The simulations were performed for ideal crystalline graphite. The evaluated inelastic scattering component well reproduced the experimental double-differential cross sections of scattered neutrons from nuclear-grade isotropic graphite measured at Materials and Life science experimental Facility (MLF) of J-PARC. Coherent elastic scattering component due to crystal structure was evaluated based on the neutron scattering and transmission experiments performed at MLF of J-PARC. Regarding comparison with the neutron transmission experiment, it was found that the quantification of small-angle neutron scattering due to structures larger than crystal, such as pores in graphite, is important.

Japanese Evaluated Nuclear Data Library version 5 (JENDL-5) was released in 2021. JENDL-5 is intended to extend its generality from JENDL-4.0 by covering a wide variety of nuclear data for applications not only to nuclear design and decommissioning, but also to other radiation-related fields. Overview of JENDL-5 and a plan for the next of JENDL-5 are presented.

Evaluation of covariance for JENDL was virtually started after the release of JENDL-3.2. The covariance data were obtained for 16 nuclides and compiled to the JENDL-3.2 Covariance File. At the time of the JENDL-4.0 development, covariances were much enhanced especially for actinides; covariance data were given for 99 nuclides in total. The latest version JENDL-5 includes covariance data for 105 nuclides by adding new evaluations for light nuclides and structure materials. An overview of the covariance evaluation for JENDL is presented.

The neutron capture cross sections of 237Np, 243Am and 241Am in the keV energy region were measured by the time-of-flight method at the Japan Proton Accelerator Research Complex (J-PARC). Characterization of the minor actinide samples were made to reduce the uncertainties of the cross section. A neutron beam filter system was developed and installed in ANNRI to solve the double-bunch mode issue of J-PARC. The pulse-height weighting technique was employed to determine the neutron capture cross sections. The neutron capture cross sections of 237Np, 243Am and 241Am were determined with higher accuracies than the past experiments.

It was tried to get useful information feedback on cross section and covariance by application of extended bias factor method with experimental benchmarks’ C/E, sensitivity coefficient and covariance data. As a result, it was succeeded to distinguish “the case unrecommended by specialists” as outlier case. And it was succeeded to get the hints that there is room for improvement in cross section or covariance of some nuclides by investigating outlier cases judged as recommended case by specialists.

In resonance analyses, experimental uncertainties affect the accuracy of resonance parameters. The resonance analysis code REFIT can consider the statistical uncertainty of the experimental data when evaluating the resonance parameter uncertainty. However, since the systematic uncertainties are not independent at each measured energy, they must be treated differently from the statistical uncertainty. In the present study, we developed a new method to incorporate the systematic uncertainty coming from sample thickness into the uncertainty of resonance parameters. We applied this method to transmission of natural zinc measured at ANNRI of MLF in J-PARC and derived the systematic uncertainty of resonance parameters. We found that some of resonance parameters have larger systematic uncertainties than the statistical ones.

Toward deuteron nuclear data evaluation, we have been developing a code system dedicated for deuteron-induced reactions, called DEURACS. In this paper, we review the recent progress in the development of DEURACS. To meet increasing and diversifying demands for deuteron nuclear data, DEURACS has been developed aiming at accurate and comprehensive prediction of deuteron-induced reactions. We present the results of some comparisons between the DEURACS calculations and experimental data for a variety of deuteron-induced reactions at incident energies up to 200 MeV. Double-differential cross sections for emission of neutrons or light ions up to A = 4 and cross sections for production of residual nuclei are analyzed. Neutron yields from deuteron bombardment on thick targets are also analyzed by the Monte Carlo transport simulation based on the deuteron nuclear data provided with DEURACS. Through comparison with experimental and other calculated data, validation of the present modeling in DEURACS is discussed.

The neutron capture-cross section of 237Np was measured using the Accurate Neutron-Nucleus Reaction Measurement Instrument (ANNRI) with special emphasis in the region of interest for the core design of Accelerator-Driven Systems (ADS), from 0.5 to 500 keV. A neutron time-of-flight method was employed using the NaI(Tl) spectrometer in the ANNRI beamline at the Japanese Proton Accelerator Research Complex (J-PARC) together with the pulse-height weighting technique. The cross-section was determined by normalizing the results to JENDL-4.0 cross-section data at the first resonance of 237Np. In the 0.5 to 500 keV range, the present preliminary results present better agreement with previous experiments of Weston et al [3]. Experimental data from Esch et al [6] is about 15% lower than the present results. In comparision with evaluated data, ENDF/B-VIII.0 offers better agreement from 0.5 to 10 keV than with JENDL-4.0. From 0.5 to 10 keV JENDL-4.0 underestimates the present results by 10-15%. Nontheless, over 10 keV energy JENDL-4.0 shows good agreement up to 500 keV. The present preliminary cross-section has uncertainties of about 5% from 0.5 to 35 keV, a value lower than the uncertainties present in JENDL-4.0 of 6-10%. However, over 35 keV the total uncertainties steadily increase and amount to 10% at 500 keV.

In order to develop an evaluation method of gamma-ray strength function (GSF), neutron capture pulse-height (PH) spectrum of gold was employed, where it was measured with the NaI(Tl) spectrometer of AN-NRI installed at the Material and Life Science Experimental Facility in J-PARC. The neutron capture gamma-ray spectrum of gold was calculated using the nuclear reaction model code CCONE. In order to obtain the information on GSF from the measured data, a gamma-ray response function for the NaI(Tl) spectrometer was calculated by the Monte-Carlo particle-transport simulation code PHITS. As a result, the PH spectrum comparable with measured one was derived by applying the gamma-ray response function to the calculated gamma-ray spectrum. By evaluation with measured PH spectra, we obtained GSF which reasonably explains measured PH spectrum in the continuum region.

Neutron capture and total cross sections of 155Gd and 157Gd were measured with the neutron time-of-flight method in Accurate Neutron-Nucleus Reaction measurement Instrument at Material and Life Science Experimental Facility of the Japan Proton Accelerator Research Complex using the NaI (Tl) spectrometer and Li-glass detectors. Preliminary cross sections were obtained in the neutron energy region from 4 meV to 100 meV. The derived cross sections agree with evaluated values in JENDL 4.0 and the experimental results by Mastromarco but were not consistent with those by Leinweber.

Recent progress and future plan of the JENDL project are summarized. Two special purpose files were released recently. One is Photo-nuclear Data File 2016 (JENDL/PD-2016) which contains the data of photo-nuclear reaction cross sections covering a wide area of the nuclear chart. The other one is the JENDL Activation Cross Section File for Nuclear Decommissioning 2017 (JENDL/AD-2017) which provides production cross sections of radioactive nuclei by neutrons. A special purpose file JENDL/ImPACT-2018 dedicated to transmutation of long-lived fission products will be released in 2019. Regarding the general-purpose file, we are preparing to release the next version, which would be made available by 2022 as JENDL-5. Evaluation efforts on nuclides across wide range of nuclei form light to heavy have being made. The first test version JENDL-5α1 was created in 2018. The evaluated data will be updated on the basis of benchmark tests on reactor criticalities and neutron shieldings.

A research project entitled "study on accuracy improvement of fast-neutron capture reaction data of long-lived MAs for development of nuclear transmutation systems" is ongoing as a joint collaboration, including Tokyo Tech, Japan Atomic Energy Agency and Kyoto University in Japan. This project focuses on neutron capture reaction of minor actinides 237Np, 241Am and 243Am, in the fast neutron energy region. The final goal of this project is to improve the neutron capture cross sections utilizing a high-intensity neutron beam from a spallation neutron source of the Japan Proton Research Complex. The overview of the project and the current progress are presented.

Activity determination of 241,243Am samples has been performed with two separate methods of calorimetry and gamma-ray spectroscopy. Decay heat measurements of the samples were carried out by using a calorimeter, and activities of the samples were accurately determined with uncertainties less than 0.45%. The primary source of uncertainty in the calorimetric method is the accuracy of available half-life data. Gamma-ray detection efficiencies of a HPGe detector were determined with uncertainties of 1.5% by combining measured efficiencies and Monte Carlo simulation. Activities of the samples were determined with uncertainties less than 2.0% by gamma-ray spectroscopy and were concordant with those of the calorimetry.

Neutron-nucleus cross sections calculated by macroscopic potentials are compared with a microscopic one to study the performance for long-lived fission products. The macroscopic potentials show a good agreement with the microscopic one at higher energies, where neutron experimental data are scarce. Besides it, analyses of differential elastic cross sections at low energies also suggest that the macroscopic potentials are still effective and applicable enough for the long-lived fission products.

In the framework of the OECD/NEA WPEC subgroup 41, ways to improve neutron induced capture cross sections for 241Am and 237Np are being sought. Decay data, energy dependent cross section data and neutron spectrum averaged data are important for that purpose and were investigated. New time-of-flight measurements were performed and analyzed, and considerable effort was put into development of methods for analysis of spectrum averaged data and re-analysis of existing experimental data.

We have developed an integrated code system dedicated for theoretical analysis and prediction of deuteron-induced reactions, which is called DEUteron-induced Reaction Analysis Code System (DEURACS). DEURACS consists of several calculation codes based on theoretical models to describe respective reaction mechanisms and it was successfully applied to (d,xp) and (d,xn) reactions. In the present work, the analysis of (d,xn) reactions is extended to higher incident energy up to nearly 100 MeV and also DEURACS is applied to (d,xd) reactions at 80 and 100 MeV. The DEURACS calculations reproduce the experimental double-differential cross sections for the (d,xn) and (d,xd) reactions well.

The neutron capture cross section of 99Tc was measured using NaI(Tl) detectors of the Accurate Neutron-Nucleus Reaction Measurement Instrument (ANNRI) at the Japan Proton Accelerator Research Complex (J-PARC) in the energy range from thermal to the keV energy region. Preliminary results were presented and compared with previous measurements and evaluations.

In order to analyze the experimental data measured by the Accurate Neutron-Nucleus Reaction Measurement Instrument (ANNRI) at the Japan Proton Accelerator Research Complex/Materials and Life Science Experimental Facility (J-PARC/MLF), it is necessary to take into account the double-bunch structure of the neutron pulse and the energy resolution function for the operational condition of the J-PARC/MLF. The modified REFIT code has been developed to treat the double-bunch neutron pulse and the energy resolution function for J-PARC/MLF. In this study, we applied the modified REFIT code to analyze the new data of the neutron capture cross section of 155Gd and 157Gd recently measured by ANNRI in the J-PARC/MLF, and obtained the resonance parameters of two Gd isotopes. We discussed the differences between the our obtained results and the other libraries.

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.

A comprehensive research program on deuteron nuclear data motivated by development of accelerator-based neutron sources is being executed. It is composed of measurements of neutron and gamma-ray yields and production cross sections, modelling of deuteron-induced reactions and code development, nuclear data evaluation and benchmark test, and its application to medical radioisotopes production. The goal of this program is to develop a state-of-the-art deuteron nuclear data library up to 200 MeV which will be useful for the design of future (d,xn) neutron sources. The current status and future plan are reviewed.

The Hauser-Feshbach codes, EMPIRE, TALYS, CCONE, and CoH3, which are widely utilized in nuclear data evaluations, are compared, with a particular focus on neutron-induced reactions on major actinides. We report the results of Hauser-Feshbach calculations using well-defined input parameters, and discuss the differences among these codes.

A comprehensive approach to determine nuclear data of unstable nuclei will be described. It consists of a measurement of fission and capture cross sections, mass distribution of fission fragments (independent fission yields) and number of prompt fission neutrons by the method of surrogate reactions. A multi-dimensional Langevin model is being developed to estimate the independent fission yields theoretically. Furthermore, the β decay properties of the fission fragments, almost all are neutron-rich unstable nuclei, are investigated systematically by improving the gross theory of β decay, which will yield information on the decay heat and delayed-neutron data

We have so far developed a computational code system dedicated to deuteron-induced reactions in combination with some theoretical models. In our previous works, the code system was successfully applied to systematic analyses of double-differential cross sections (DDXs) of (d,xp) reactions for 12C, 27Al, and 58Ni at incident energies up to 100 MeV. In the present work, we apply the code system to neutron emission from deuteron-induced reactions. Since there is few experimental data of DDXs of (d,xn) reactions, double-differential thick target neutron yields (TTNYs) are calculated and compared with experimental data instead of DDXs. The calculation using the code system reproduces the measured TTNYs for carbon at incident energies up to 50 MeV.

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.