Copper is an important structural material used in nuclear technology, often used as a cover for spent fuel canisters or planned to be used in fusion devices. Despite its significance, there is a lack of integral experiments useful for validating and improving the evaluations of copper nuclear data. To address this gap, a neutron leakage experiment was conducted a few years ago using a point 252Cf(s.f.) neutron source placed inside a large block of copper. In this work a pencil beam transmission-attenuation experiment (a broomstick) employing various thicknesses (5-20 cm) of copper blocks (cylinders of 6 cm in diameter) was undertaken to expand the dataset of available experiments for copper in the fast neutron energy range (1-10 MeV). This type of experiment has the highest sensitivity to the total cross sections, and sensitivities are different from other integral experiments, making it a complementary measurement to already existing integral data. The measurement was performed using stilbene scintillation spectrometry. Measured transmission shows that the current INDEN evaluation, proposed to be adopted for ENDF/B-VIII.1 and JEFF-4 libraries, exhibits excellent agreement with experimental data. The JEFF-3.3 evaluation displays significant discrepancies, consistent with previous results from integral experiments involving copper. In the case of JENDL-5, discrepancies were found in the energy region 1.7-4.9 MeV.

Fluorine is a crucial element for the nuclear industry and technology due to its application in nuclear fuel production as uranium hexafluoride (UF6), its use of Teflon (R) in criticality experiments, and its use in coolant/ fuel materials in several Molten Salt Reactor (MSR) concepts currently being designed. There have been many efforts to improve the neutron reaction cross sections for fluorine. The neutron broad-beam transmission experiment through the PTFE (PolyTetraFluoroEthylene) block at the Dresden AKR-2 reactor has been performed to validate existing evaluated cross-section libraries from 1 to 10 MeV of incident neutron energies. Calculations with studied data sets (ENDF/B-VIII.0, INDEN, JEFF-3.3, and JENDL-5) show satisfactory agreement with the experiment in the energy region above 2.1 MeV within the uncertainty range. The measured neutron transmission shows a systematic disagreement in a calculation to experiment comparison in the energy region 1-2.1 MeV. This behavior is consistent with the measured 252Cf(s.f.) neutron leakage in the PTFE block with dimensions of 60 x 50 x 50 cm. Improvement of evaluated cross sections below 2 MeV of neutron incident energy is needed.

The paper presents a first comprehensive measurement of the total fission gamma spectrum (TFGS) of 252Cf(s.f.) spanning an energy range from 0.2 to 14.3 MeV using stilbene scintillator spectroscopy. The measurement was performed with the 252Cf(s.f.) source placed in a flexo-rabbit ending, with a stilbene detector located at a distance of 100 cm from the source. The room effect was determined by separate measurement with a 50 cm long lead shielding cylinder placed between the source and the detector. The obtained spectrum was corrected for the effect of the 252Cf source structural parts and the flexo-rabbit ending, using correction function obtained by simulation with the MCNP6.2 code. The main benefit of this independent measurement is the wide energy range, from 0.2 up to 14.3 MeV, in a single experiment, whereas existing measurements typically cover narrower intervals. The knowledge of the TFGS is crucial for various applications involving 252Cf(s.f.) sources, including the design and analysis of nuclear systems, the validation of nuclear data, and the estimation of radiation in the nuclear medicine. The comparison of TFGS with the recently evaluated 252Cf(s.f.) prompt fission gamma spectrum (PFGS) indicates that delayed gammas may contribute above 2 MeV up to the highest energies; the current assumption was that delayed gammas contribution above 2 MeV was negligible.

This paper deals with the measurement of Spectrum Averaged Cross Sections in two different neutron fields formed in zero power reactors. The first was Benchmark Neutron Reference Field in the LR-0 reactor, and the second field was in the center of the vertical channel touching the fuel in the VR-1 reactor. The spectrum averaged cross section differs for both cases as the spectra differ, but after normalization to 235U PFNS using calculated correction, both results are in good agreement, thus confirming the spectra in both cases are similar in the 1 – 14 MeV region. A good agreement between lower threshold reactions averaged in actual reactor spectra and prompt fission neutron spectrum of 235U is reported as well.

The measured and evaluated excitation functions are fundamental quantities that affect the accuracy of all calculations in nuclear applications. Some cross sections, such as 14N(n,p)14C, have added value for special applications, as these reactions may be responsible for possible contamination in industrial processes such as spent fuel reprocessing. For the validation of the evaluated cross sections, we can rely on the comparison of the calculated spectrum averaged cross sections (SACS) for the given neutron spectrum with the measured SACS value. The benchmark reference neutron field has been identified, characterized, and well validated in the LR-0 special core. A very large set of SACS measurements in the LR-0 reference field is measured with low uncertainty, making it an excellent set for deconvolution codes' validation. The impact of the gamma-induced reaction on the production yield of neutron-induced reactions was investigated for most of the benchmarked reactions. Gamma competition was found to contribute at most 1 % for the 197Au(n,2n) reaction, while being substantially lower for other target isotopes and neutron-induced activation reactions.

The Working Party on International Nuclear Data Evaluation Co-operation Subgroup 47 (WPECSG47) entitled "Use of Shielding Integral Benchmark Archive and Database for Nuclear Data Validation" was organised from 2019 and 2022 with the objectives to promote more systematic and wider use of shielding benchmark experiments in nuclear data (ND) and transport code validation and development, to provide feedback on the Shielding Integral Benchmark Archive and Database (SINBAD), and to promote its further development in coordination with the Expert Group on Physics of Reactor Systems (EGPRS). Altogether 9 meetings, the large majority (8) held remotely, were organised during the past 3 years to discuss the experience on the use of SINBAD, evaluation of new benchmarks and improvements to be contributed to the database which was severely neglected and lacking maintenance over the past ⇠ 10+ years. Several proposals for new or updated benchmark evaluation were presented and discussed, such as FNG copper, LLNL pulsed spheres, CIAE iron sphere, KFK 1977 gamma measurements, Rez Fe sphere, ASPIS, ORNL Oxygen broomstick, TIARA and others. Complementing the database with new features was also discussed, for example providing the nuclear data sensitivity profiles more systematically would facilitate and better guide the use of data. Information on the geometry, (radiation source) and materials available in CAD format is expected to allow an easier and less error prone reference for computational model preparation and a potential input to CAD based workflows. Inputs for various transport codes and other benchmark data from participants have been shared via the NEA GitLab which could hopefully in the future evolve and form a bases for critically checked and validated benchmark data. Future development of SINBAD will be monitored by EGPRS and the newly created SINBAD Task Force.

The cross section averaged over 235U thermal-neutron induced fission spectrum is a fundamental quantity that can be used in evaluation and validation of nuclear data. Many experiments focused on the determination of Spectrum Averaged Cross Sections (SACS) in 235U(nth,f) Prompt Fission Neutron Spectrum (PFNS) in light water reactors using enriched uranium fuel. In these reactors, there is already some amount of water moderator be-tween the uranium fuel and the irradiated sample. Due to the decrease of hydrogen cross-section with neutron energy, the high energy tail of the reactor spectrum in cores with water moderator may be harder than the pure PFNS. This paper aims to compare the shape of the actual reactor spectrum in various core positions of a research light-water reactor differing each from other by the effective water thickness. The spectrum shape is determined both by calculations and experimentally using various high energy threshold reactions. The impact of the photo-nuclear reactions (gamma,n) competing with (n,2n) in production of the same residual nucleus was shown to be less than a percent for most of studied dosimeters. An important exception was found for 197Au(n,2n)196Au dosimeter irradiated in the outcore channel where a notable photo-neutron contribution to the production of 196Au is caused by the neutron production from the high energy gamma-rays from thermal-neutron capture in 54Fe. The corresponding ENDF/B-VIII.0 data turned out to underestimate such gamma-yield by 40 % in comparison with ENDF/B-VI.8. This has improved but however not resolved the disagreement between our measurement and calculations. The remaining deficiency was attributed to the underestimation of the evaluated cross section IAEA/PD-2019 for the 197Au(gamma,n) cross section near the reaction threshold. The later was confirmed by comparison with existing measured data.

Cross section data are fundamental quantities which affect the accuracy of all calculations in nuclear applications. A new dosimetry library IRDFF-II that contains cross section evaluations with full uncertainty quantification was developed by the International Atomic Energy Agency and released in January 2020 (https://www-nds.iea.org/IRDFF). A previous version, IRDFF-1.05, was released in 2014 and experimental validation of the newly released cross section by spectrum averaged cross section (SACS) measurements is a high priority task. For such purpose, a neutron dosimeter set containing 5 target foils was activated in 2 independent experiments at the VR-1 reactor of the Czech Technical University in Prague. Care was taken to derive SACS with low uncertainties. New experimental evaluation method is in good agreement with previous approaches based on relative measurements using monitor foils. Good agreement of measured SACS and evaluated IRDFF-II cross sections is observed. Slight overestimation of evaluated ENDF/B-VIII.0 235U(nth,f) PFNS above 10 MeV is discussed.

The activities of the EUROfusion consortiums on the development of high quality nuclear data for fusion applications are presented. The activities, implemented in the Power Plant Physics and Technology (PPPT) programme of EUROfusion, include nuclear data evaluations for neutron and deuteron induced reactions and the production of related data libraries which satisfy the needs for nuclear analyses of the DEMO fusion power plant and the IFMIF-DONES neutron source. The activities are closely linked to the JEFF initiative of the NEA Data Bank. The evaluation work is complemented by extensive benchmark, sensitivity and uncertainty analyses to check the performance of the evaluated cross-section data and libraries against integral experiments.

An international effort has produced evaluations of the neutron data standards. Evaluations were obtained for the cross section standards: the H(n,n), 6Li(n,t), 10B(n,αγ), 10B(n,α), natC(n,n), Au(n,γ), 235U(n,f), and 238U(n,f) reactions. Also in the evaluation process, the 238U(n,γ) and 239Pu(n,f) nonstandard cross sections were evaluated. Many of these are dosimetry cross sections. Evaluations were also obtained for data that are not traditional standards: Maxwellian spectrum averaged cross section for the Au(n,γ) cross section at 30 keV, reference cross sections for prompt γ-ray production in fast neutron-induced reactions, reference cross sections for very high-energy fission cross sections, the 252Cf spontaneous fission neutron spectrum and the 235U thermal fission neutron spectrum, and the thermal constants. The data and covariances were obtained directly from this evaluation procedure as is required by the dosimetry community.

The energy-energy and reaction-reaction covariance matrices were calculated for the n + 56 Fe damage cross-sections by Total Monte Carlo method using the TENDL-2013 random files. They were represented in the ENDF-6 format and added to the unperturbed evaluation file. The uncertainties for the spectrum averaged radiation quantities in the representative fission, fusion and spallation facilities were first time assessed as 5–25%. Additional 5 to 20% have to be added to the atom displacement rate uncertainties to account for accuracy of the primary defects simulation in materials. The reaction-reaction correlation were shown to be 1% or less.

Atomic displacement and gas production cross-sections were obtained for a number of materials to calculate radiation damage and gas production rate in nuclear- and fusion reactors, and neutron spallation sources. An advanced atomistic modelling approach was applied for calculations of the number of stable displacements in materials.

The results of validation of the latest release of International Reactor Dosimetry and Fusion File, IRDFF-1.03, in the standard 252Cf(s.f.) and reference 235U(nth,f) neutron benchmark fields are presented. The spectrum-averaged cross sections were shown to confirm IRDFF-1.03 in the 252Cf standard spontaneous fission spectrum; that was not the case for the current recommended spectra for 235U(nth,f). IRDFF was also validated in the spectra of the research reactor facilities ISNF, Sigma-Sigma and YAYOI, which are available in the IRDF-2002 collection. The ISNF facility was re-simulated to remove unphysical oscillations in the spectrum. IRDFF-1.03 was shown to reproduce reasonably well the spectrum-averaged data measured in these fields except for the case of YAYOI.

This Meeting was organized to implement the recommendation of the second Research Coordinated Meeting (RCM) of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) “Primary Radiation Damage Cross Sections” to analyse the accuracy and consistency of the radiation damage-relevant nuclear data in the major nuclear data evaluations with the eventual goal of identifying the most reliable data and provide quantitative uncertainty estimates. Participants have considered the status of the primary nuclear data, such as recoils spectra in the latest releases of ENDF, JEFF, JENDL, FENDL, ROSFOND and TENDL, the ways of deriving the damage quantities KERMA, NRTor arc-dpa and gas production cross sections as well as the recipes for an assessment of their uncertainties. This report contains the contemporary view of the Meeting participants on these issues in the form of a consolidated set of statements, recommendations and individual summaries.

The International Reactor Dosimetry and Fusion File (IRDFF) is an extension of the International Reactor Dosimetry File (IRDF-2002) to cover fission, fusion and accelerator driven applications up to an energy of 60 MeV and now includes 75 reactions.The Nuclear Data Section of IAEA initiates a Coordinated Research Project (CRP) with the main goal to validate and test the IRDFF library.This paper will describe the scope and present status of IRDFF as well as the objectives of the CRP, i.e. the overview of experimental data required for validation.We inform the community about this activity and draw the attention of researchers who can make a contribution.

The International Network of Nuclear Reaction Data Centres (NRDC) provides nuclear reaction data services to users through collection and compilation of experimental nuclear reaction data in the EXFOR database. The database includes neutron-induced, charged-particle-induced, and photonuclear data for projectile energies up to 1 GeV. Sophisticated search options and user-friendly retrieval interface for downloading data in different formats have been developed at IAEA Nuclear Data Section. Additional output options such as data plotting capabilities are provided as well. The database is constantly revised and extended and at present contains about 20,000 experimental works accumulated in its 40 years of history. The paper will present several recent IAEA NDS activities related to the development of the EXFOR database and retrieval system.

The Nuclear Data Section of IAEA assembles, develops and disseminates nuclear data for basic science and practical applications. This paper gives an overview of the NDS activities aimed at the development and maintenance of general purpose databases, such as collection of experimental and evaluated reaction cross sections data, as well as those for the specific energy and material applications, namely for ion material beam analyses, neutron dosimetry, fission and fusion neutronics. The latest upgrades, developments and existing weaknesses are highlighted.

The high-flux test module (HFTM) of the International Fusion Materials Irradiation Facility IFMIF will be exposed to an intense flux of neutrons with energies of up to approximately 55 MeV. The threshold of tritium-producing reactions in HFTM and the specimen materials is typically between 10 and 20 MeV, therefore the assessment of the tritium inventory becomes important for the handling of the specimen cells and specimens after the irradiation cycle. In this work, we calculate the tritium production in the specimen cells by neutrons with the McDeLicious code and activation nuclear data from the libraries EAF-2007 and IEAF-2007. The results differ by roughly a factor of two. We discuss ambiguities in the relevant activation data and conclude that better experimental verification of cross sections of isotopes most important for the tritium production in the test cells is desirable.

To satisfy growing needs in nuclear data at intermediate energies, the Proton Activation Data File (PADF) has been prepared. It contains 418,575 excitation functions of nuclear reactions for 2,355 target nuclei from Mg to Ra at proton energies up to 150 MeV. The data are available in the PADF for stable and unstable target nuclei including isomeric targets with the half-life more than one second. Cross-sections included in the PADF were obtained by using the TALYS code, the modified ALICE code, and the available experimental data.