The joint evaluated fission and fusion nuclear data library 3.3 is described. New evaluations for neutron-induced interactions with the major actinides $$^{235}\hbox {U}$$ , $$^{238}\hbox {U}$$ and $$^{239}\hbox {Pu}$$ , on $$^{241}\hbox {Am}$$ and $$^{23}\hbox {Na}$$ , $$^{59}\hbox {Ni}$$ , Cr, Cu, Zr, Cd, Hf, W, Au, Pb and Bi are presented. It includes new fission yields, prompt fission neutron spectra and average number of neutrons per fission. In addition, new data for radioactive decay, thermal neutron scattering, gamma-ray emission, neutron activation, delayed neutrons and displacement damage are presented. JEFF-3.3 was complemented by files from the TENDL project. The libraries for photon, proton, deuteron, triton, helion and alpha-particle induced reactions are from TENDL-2017. The demands for uncertainty quantification in modeling led to many new covariance data for the evaluations. A comparison between results from model calculations using the JEFF-3.3 library and those from benchmark experiments for criticality, delayed neutron yields, shielding and decay heat, reveals that JEFF-3.3 performes very well for a wide range of nuclear technology applications, in particular nuclear energy.

The joint evaluated fission and fusion nuclear data library 3.3 is described. New evaluations for neutron-induced interactions with the major actinides $$^{235}\hbox {U}$$ 235U , $$^{238}\hbox {U}$$ 238U and $$^{239}\hbox {Pu}$$ 239Pu , on $$^{241}\hbox {Am}$$ 241Am and $$^{23}\hbox {Na}$$ 23Na , $$^{59}\hbox {Ni}$$ 59Ni , Cr, Cu, Zr, Cd, Hf, W, Au, Pb and Bi are presented. It includes new fission yields, prompt fission neutron spectra and average number of neutrons per fission. In addition, new data for radioactive decay, thermal neutron scattering, gamma-ray emission, neutron activation, delayed neutrons and displacement damage are presented. JEFF-3.3 was complemented by files from the TENDL project. The libraries for photon, proton, deuteron, triton, helion and alpha-particle induced reactions are from TENDL-2017. The demands for uncertainty quantification in modeling led to many new covariance data for the evaluations. A comparison between results from model calculations using the JEFF-3.3 library and those from benchmark experiments for criticality, delayed neutron yields, shielding and decay heat, reveals that JEFF-3.3 performes very well for a wide range of nuclear technology applications, in particular nuclear energy.

The Neutron Physics Department at Centro Atómico Bariloche developed new models for the interaction of thermal neutrons with water which have been validated against experimental data, including new thermal scattering experiments, and were adopted for the release of ENDF/B-VIII.0. Although the older models are, in general, good for most applications, some discrepancies had appeared in the case of heavy water, and this motivated new measurements that validated the new model. In the case of light water, the new model predicts a reduction of the total cross section around 0.025 eV when the temperature is increased from room temperature. This reduction, that is not predicted by the existing models, and potentially affects the calculation of temperature reactivity coefficients in nuclear reactors, has been traced to a shift in the vibrational frequency spectrum of liquid water. The only experimental data previously available is from an experiment performed at the Demokritos reactor in the ’60s at 293 K and 473 K, which validates the new model when the cross section ratios are computed. In order to verify this effect at a lower temperature range, a transmission experiment was carried out at the VESUVIO spectrometer in the ISIS facility in the UK in June 2018, measuring the total neutron cross section in the range from 283 K to 353 K. Here, we present this new experimental data and its comparison with the models.

The objective of this paper is to present the results of an evaluation of the gadolinium isotopes with the main focus on the isotopes 155Gd and 157Gd. The evaluations were carried out in the resolved resonance region using the Reich-Moore formalism. The originality on the 155Gd and 157Gd evaluations is the addition of new high-resolution capture cross section measurements performed at the neutron time-of-flight, n_TOF facility for enriched samples and the statistical analysis of the resonance parameters. The resonance analysis was performed with the multilevel R-matrix code SAMMY together with the generalized least-squares technique based on the Bayes’ theory.

High temperature promotes the mobility of UO2 grain boundaries, resulting in grain growth. Therefore, grain growth, based on preand post-irradiation grain size measurements, can be used to estimate the in-reactor temperature of UO2 fuel. Using models based on grain growth, fuel temperatures can be estimated for irradiated fuels. Temperatures can also be calculated using models based on thermal conductivity, element linear power, and fuel burnup. The consistency of the models can be verified by comparing the calculated centreline fuel temperatures with those estimated from measured grain sizes. The objective of this work is to assess the consistency of the models that are used in fuel performance codes. Model parameters (e.g., irradiated fuel grain size, linear power) obtained from the Canadian Nuclear Laboratories database are used to establish fuel centerline temperatures using different models. Then the fuel centerline temperatures, obtained in different ways, are compared for consistency.

The CIELO collaboration has studied neutron cross sections on nuclides that significantly impact criticality in nuclear technologies – 16O, 56Fe, 235,8U and 239Pu – with the aim of improving the accuracy of the data and resolving previous discrepancies in our understanding. This multi-laboratory pilot project, coordinated via the OECD/NEA Working Party on Evaluation Cooperation (WPEC) Subgroup 40 with support also from the IAEA, has motivated experimental and theoretical work and led to suites of new evaluated libraries that accurately reflect measured data and also perform well in integral simulations of criticality.

We describe the new ENDF/B-VIII.0 evaluated nuclear reaction data library. ENDF/B-VIII.0 fully incorporates the new IAEA standards, includes improved thermal neutron scattering data and uses new evaluated data from the CIELO project for neutron reactions on 1H, 16O, 56Fe, 235U, 238U and 239Pu described in companion papers in the present issue of Nuclear Data Sheets. The evaluations benefit from recent experimental data obtained in the U.S. and Europe, and improvements in theory and simulation. Notable advances include updated evaluated data for light nuclei, structural materials, actinides, fission energy release, prompt fission neutron and γ-ray spectra, thermal neutron scattering data, and charged-particle reactions. Integral validation testing is shown for a wide range of criticality, reaction rate, and neutron transmission benchmarks. In general, integral validation performance of the library is improved relative to the previous ENDF/B-VII.1 library.

Sensitivity of fast reactor physics coefficients to nuclear data differs from thermal reactors. A conceptual lead-cooled reactor has been modelled and assessed for nuclear data sensitivity. A TSUNAMI sensitivity analysis was performed to estimate uncertainty and sensitivity of the neutron multiplication factor to the nuclear data of all nuclides and reactions relevant to the reactor concept. The results were compared against MCNP6, which also employs an adjoint-based method for sensitivity calculations. Sensitivity of the neutron multiplication factor was found to agree closely between MCNP6 and TSUNAMI, for lead cross sections and other important nuclear data. The TSUNAMI assessment of nuclear data impact was also examined through a brief survey of nuclear data evaluations. The evaluations were found to have differences of interpretation that impact the TSUNAMI calculations. The consequent uncertainties in fuel temperature and coolant-voiding coefficients were also examined in TSUNAMI. Based on the TSUNAMI ranking of important data, a subset of nuclides and reactions were selected for stochastic sampling of cross-section data using NJOY. The sampled cross-section data were then used in SERPENT simulations to assess uncertainties in calculated parameters, which were consistent with the TSUNAMI results. Crown Copyright (C) 2018 Published by Elsevier Ltd. All rights reserved.

In order to improve the design and safety of thermal nuclear reactors and for verification of criticality safety conditions on systems with significant amount of fissile materials and water, it is necessary to perform high-precision neutron transport calculations and estimate uncertainties of the results. These calculations are based on neutron interaction data distributed in evaluated nuclear data libraries. To improve the evaluations of thermal scattering sub-libraries, we developed a set of thermal neutron scattering cross sections (scattering kernels) for hydrogen bound in light water, and deuterium and oxygen bound in heavy water, in the ENDF-6 format from room temperature up to the critical temperatures of molecular liquids. The new evaluations were generated and processable with NJOY99 and also with NJOY-2012 with minor modifications (updates), and with the new version of NJOY-2016. The new TSL libraries are based on molecular dynamics simulations with GROMACS and recent experimental data, and result in an improvement of the calculation of single neutron scattering quantities. In this work, we discuss the importance of taking into account self-diffusion in liquids to accurately describe the neutron scattering at low neutron energies (quasi-elastic peak problem). To improve modeling of heavy water, it is important to take into account temperature-dependent static structure factors and apply Sköld approximation to the coherent inelastic components of the scattering matrix. The usage of the new set of scattering matrices and cross-sections improves the calculation of thermal critical systems moderated and/or reflected with light/heavy water obtained from the International Criticality Safety Benchmark Evaluation Project (ICSBEP) handbook. For example, the use of the new thermal scattering library for heavy water, combined with the ROSFOND-2010 evaluation of the cross sections for deuterium, results in an improvement of the C/E ratio in 48 out of 65 international benchmark cases calculated with the Monte Carlo code MCNP5, in comparison with the existing library based on the ENDF/B-VII.0 evaluation.

The neutron flux and fission power profiles through a fuel bundle and across a fuel element are important aspects of nuclear fuel analysis in multi-scale/multi-physics modelling of Pressurized Heavy Water Reactors (PHWRs) with advanced fuel bundles. Fuel channels in many existing PHWRs are horizontal. With ageing, pressure tubes creep and fuel bundles in these pressure tubes are eccentrically located, which results in an asymmetric coolant flow distribution between the top and bottom of the fuel bundles. The diametral change of the pressure tube due to creep is not constant along the fuel channel; it reaches a maximum in the vicinity of the maximum neutron flux location. The Cross-sectional asymmetric positioning of fuel bundles in a crept pressure tube contributes to an asymmetric power distribution within a ring of fuel elements. Modern reactor physics lattice codes (such as WIMS-AECL(1)) are capable of predicting the details of power distribution from basic principles. Thermalhydraulics subchannel codes (such as ASSERT-PV) use models to describe inhomogeneous power distribution Within and across fuel elements (e.g., flux tilt model, different powers in different ring elements, or radial power profiles). In this work, physics and thermalhydraulics codes are applied to quantify the effect of eccentricity of a fuel bundle on power variations across it and within a fuel element, and ultimately on the fuel temperature distribution and fuel centerline temperature, which is one of the indicators of fuel performance under normal operating conditions (NOC). Due to fissioning, Xe and Kr gases are produced, which contribute to the internal gas pressure within an element. The increase in gas pressure may jeopardize the fuel element integrity (sheath rupture). On the other hand, fission gas is produced as a function of neutron flux/power and diffuses to the grain boundaries and fuel-pellet interface as a function of fuel temperature. Therefore, it is important to know the fuel temperatures and distributions within the fuel. (C) 2017 Elsevier Ltd. All rights reserved.

The decay of actinides such as Pu-238, results in recoil damage and helium production in spent nuclear fuels. The extent of the damage depends on storage time and spent fuel composition and has implications for the integrity of the fuels. Some advanced nuclear fuels intended for use in pressurized heavy water pressure tube reactors have high initial plutonium content and are anticipated to exhibit swelling and embrittlement, and to accumulate helium bubbles over storage times as short as hundreds of years. Calculations are performed to provide estimates of helium production and fuel swelling associated with alpha decay as a function of storage time. Significant differences are observed between predicted aging characteristics of natural uranium and the advanced fuels, including increased helium concentrations and accelerated fuel swelling in the latter. Implications of these observations for long term storage of advanced fuels are discussed. Crown Copyright (C) 2017 Published by Elsevier Ltd. All rights reserved.

To improve the evaluations in thermal scattering sub-libraries, a set of thermal neutron scattering cross sections (scattering kernels) was developed recently at Centro Atomico Bariloche (CAB) for deuterium and oxygen bound in liquid heavy water, and made available in the ENDF-6 format. These libraries are based on a combination of results of molecular dynamics (MD) simulations and recent experimental data and, when used, result in an improvement of calculations of observables of single neutron scattering experiments, compared to results based on previous evaluations. In this work, we provide additional details on the CAB evaluation of heavy water at room temperature and discuss the important integral characteristics of neutron scattering kernels, such as the cross sections, average scattering cosine, and average secondary energy. Then, the new set of thermal scattering kernels is applied in modelling criticality of the ZED-2 reactor, and the international benchmarks LEU-MET-THERM-003 (ICSBEP handbook) and ZED2-HWR-EXP-001 (IRPhEP handbook) are analyzed in detail. The differences in the estimates of criticality due to changes in the S(alpha,beta)data from the reference one (ENDF/B-VII) to the CAB evaluation are 100-200 pcm; they are comparable but smaller than the ZED-2 benchmark uncertainties (approximate to +/- 300 pcm). Changing the reference evaluation of O-16 to the recently developed one from the CIELO project (WPEC subgroup 40) results in a decrease of criticality by similar to 100 pcm. Using different combinations of the improved nuclear data for deuterium, oxygen, and the CAB TSL model, we obtain biases up to 300-400 pcm in the estimates of criticality of the selected ZED-2 benchmarks. Application of the similar to new evaluations for U-235 and U-238 from the CIELO project improves the estimates of Ice by decreasing the bias by 100 pcm that indicates a need for further investigations of the ZED-2 critical assemblies. Crown Copyright (C) 2017 Published by Elsevier Ltd. All rights reserved.

The accuracy of deuterium nuclear data is important for reactor physics simulations of heavy water (D2O) reactors. The elastic neutron scattering cross section data at thermal energies, σs,th, have been observed to have noticeable impact on the reactivity values in simulations of critical systems involving D2O. We discuss how the uncertainties in the thermal scattering cross sections of 2H(n,n)2H and 16O(n,n)16O propagate to the uncertainty of the calculated neutron multiplication factor, keff, in thermal critical assemblies with heavy water neutron moderator/reflector. The method of trial evaluated nuclear data files, in which specific cross sections are individually perturbed, is used to calculate the sensitivity coefficients of keff to the microscopic nuclear data, such as σs(E) characterized by σs,th. Large reactivity differences of up to ≃ 5–10 mk (500–1000 pcm) were observed using 2H and 16O data files with different elastic scattering data in MCNP5 simulations of the LANL HEU heavy-water solution thermal critical experiments included in the ICSBEP handbook.

In multi-scale / multi-physics modelling of Pressurized Heavy Water Reactors (PHWRs), the neutron flux and fission power profiles through a fuel bundle and along a fuel element are important aspects in the reactor physics part of the simulations. The power profiles change with fuel burn-up. In addition, the skin effect from Pu build-up has been observed at the surface layer of fuel elements as the burn-up progresses. As a result, the element power distribution has a sharp increase near the fuel surfaces due to inhomogeneous distribution of Pu (mostly due to 239Pu in PHWRs). Modern reactor physics (RP) lattice codes (such as WIMS-AECL 3.1) are capable of predicting the details of power distribution at the length scales of ∼ 0.1mm (or less) from basic principles. On the other hand, thermalhydraulics (TH) subchannel codes (such as ASSERT-PV) use models to describe inhomogeneous power distribution along the fuel pins (e.g., a flux tilt model). In this work, we demonstrate how one can combine WIMS-AECL 3.1 with ASSERT-PV for a multi-physics simulation of PHWR fuel bundles with asymmetric power distribution within fuel elements inside pressure tubes. Although the effects of changes in element power distributions with burn-up on critical heat flux (CHF) and dryout power are expected to be small, they need to be accurately quantified for the outer ring elements where the power peak at the surface due to the Pu skin effect and the power gradient within a fuel element are expected to be high. We also discuss RP and TH modelling of the pressure tube (PT) creep effect in PHWRs using detailed models of fuel bundles in WIMS-AECL 3.1 and ASSERT-PV. PT creep results in a larger gap at the top of a bundle in a horizontal channel, so one needs to quantify its impact on the power distribution through the bundle and within fuel pins due to the asymmetry between the top and the bottom of a bundle. As an example, the effect of burn-up and pressure tube creep on the distribution of power within fuel elements, and hence on the CHF and dryout power, will be investigated for a 43-element fuel bundle.

CIELO (Collaborative International Evaluated Library Organization) provides a new working paradigm to facilitate evaluated nuclear reaction data advances. It brings together experts from across the international nuclear reaction data community to identify and document discrepancies among existing evaluated data libraries, measured data, and model calculation interpretations, and aims to make progress in reconciling these discrepancies to create more accurate ENDF-formatted files. The focus will initially be on a small number of the highest-priority isotopes, namely 1H, 16O, 56Fe, 235,238U, and 239Pu. This paper identifies discrepancies between various evaluations of the highest priority isotopes, and was commissioned by the OECD's Nuclear Energy Agency WPEC (Working Party on International Nuclear Data Evaluation Co-operation) during a meeting held in May 2012. The evaluated data for these materials in the existing nuclear data libraries — ENDF/B-VII.1, JEFF-3.1, JENDL-4.0, CENDL-3.1, ROSFOND, IRDFF 1.0 — are reviewed, discrepancies are identified, and some integral properties are given. The paper summarizes a program of nuclear science and computational work needed to create the new CIELO nuclear data evaluations.

Neutron elastic scattering measurements were carried out at the nELBE neutron time-of-flight facility at a 6 m flight path. Energies below 2 MeV were studied using a setup consisting of eight 6Li-glass detectors placed at nominal angles of 15∘ and 165∘ with respect to the incident neutron beam. A deuterated polyethylene sample with 99.999% enrichment in deuterium was used. These angles were chosen since an earlier study showed that the ratio of the differential cross section at these angles is the most sensitive to differences in evaluated files and model calculations. Accurate 165∘/15∘ angle ratios were obtained. Above 1 MeV these are somewhat larger than given by ENDF/B-VII. Simultaneously the early day experiments using a proportional counter to infer angular distributions from deuterium recoil pulse height distributions are being studied through a new experiment with such a device at the Physikalisch-Technische Bundesanstalt (PTB). At 500 keV this experiment favors ENDF/B-VII over JENDL-4.0, while at lower energies agreement with the data is similar.