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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.

The CONRAD computer code is being developed by the nuclear data group of CEA Cadarache since mid-2000. It was originally designed to analyse neutron-induced reactions in the resonance energy range and then, was extended to higher energies (several MeV) treatment with inclusion of charged-particles penetration factor. In the thermal energy range, a procedure was implemented to manage the so-called Thermal Neutron Constants, especially devoted to the 239Pu, 241Pu, 233U and 235U nuclei. In the resonance range, nuclear models implemented in the CONRAD program rely on R-matrix model fits with in particular, improved treatment of the fission penetration factor and fluctuations of the prompt neutron multiplicity via a two-step (n,γf) process. Above the resonance range, the neutron continuum energy region of the observed cross sections can be analysed either with the TALYS code or with an in-house optical model code named CCCP that is followed by Hauser-Feshbach calculations according to the compound nucleus deexcitation channels. Non-model least squares fitting procedures have been also tested for neutron cross sections adjustment in the continuum energy range, and such, applied in the framework of the standard cross section group at the IAEA (Vienna). This paper will only focus on the evaluation works performed in the resonance range of minor and major actinides with special emphasis on the experimental corrections suitable to reproduce time-of-flight experiments.

The thermal scattering laws (TSL) take into account the crystalline structure and atomic motions of isotopes bound in materials. This paper presents the CINEL code, which was developed to generate temperature-dependent TSL for solid, liquid and free gas materials of interest for nuclear reactors. CINEL is able to calculate TSL from the phonon density of states (PDOS) of materials under the Gaussian-Incoherent approximations. The PDOS can be obtained by using theoretical approaches (e.g., ab initio density functional theory and molecular dynamics) or experimental results. In this work, the PDOS presented in the ENDF/BVIII.0 and NJOY-NCrystal libraries were used for numerical validation purposes. The CINEL results are in good agreement with those reported in these databases, even in the specific cases of TSL with the newly mixed elastic format. The coding flexibility offered by Python using the JupyterLab interface allowed to investigate limits of physical models reported in the literature, such as a four-site model for UO2, anharmonic behaviors of oxygen atoms bound in a Fm3m structure, texture in Zry4 samples and jump corrections in a roto-translational diffusion model for liquid water. The use of graphic processing units (GPU) is a necessity to perform calculations in a few minutes. The performances of the CINEL code is illustrated with the results obtained on actinide oxides having a Fm3m structure (UO2, ThO2, NpO2 and PuO2), low enriched fuel (UMo), cladding (Zry4) and moderators (H2O with a specific emphasis on ice).

A new evaluation of spectrum averaged cross sections (SACS) of 235U, 238U, and 239Pu measured in the 252Cf(sf) reference neutron field is presented and found to be consistent with the original Mannhart SACS evaluation in IRDF-2002. The comprehensive vetted experimental database that includes SACS ratio and absolute SACS measurements of major actinides is being used to update the SACS database employed as input of the new GMApy code to derive the Neutron Standards. An update of the current neutron standards based on Time Projection Chamber (TPC) shape data, a new comprehensive uncertainty quantification, and revised SACS experimental database is proposed which result in a 0.7% increase of the evaluated 239Pu(n,f)/235U(n,f) crosssection ratio in the 1–5 MeV energy region. The increase is due to a 0.3% reduction of the standard 235U(n,f) cross section and a 0.4% increase of the 239Pu(n,f) reference cross section in the 1–5 MeV energy region. Those changes are well within estimated USU fission cross-section uncertainties of 1.2%, but are relevant for the evaluated mean values.

Temperature-dependent dynamic structure factors S(Q, ω) for liquid water have been calculated using a composite model, which is based on the decoupling approximation of the mean square displacement of the water molecules into diffusion and solid-like vibrational parts. The solid-like vibrational part Svib(Q, ω) is calculated with the phonon expansion method established in the framework of the incoherent Gaussian approximation. The diffusion part Sdiff(Q, ω) relies on the Egelstaff–Schofield translational diffusion model corrected for jump diffusions and rotational diffusions with the Singwi–Sjölander random model and Sears expansion, respectively. Systematics of the model parameters as a function of temperature were deduced from quasi-elastic neutron scattering data analysis reported in the literature and from molecular dynamics (MD) simulations relying on the TIP4P/2005f model. The resulting S(Q, ω) values are confronted by means of Monte Carlo simulations to inelastic neutron scattering data measured with IN4, IN5, and IN6 time-of-flight spectrometers of the Institut Laue-Langevin (ILL) (Grenoble, France). A modest range of temperatures (283–494 K) has been investigated with neutron wavelengths corresponding to incident neutron energies ranging from 0.57 to 67.6 meV. The neutron-weighted multiphonon spectra deduced from the ILL data indicate a slight overestimation by the MD simulations of the frequency shift and broadening of the librational band. The descriptive power of the composite model was suited for improving the comparison to experiments via Bayesian updating of prior model parameters inferred from MD simulations. The reported posterior temperature-dependent densities of state of hydrogen in H2O would represent valuable insights for studying the collective coupling interactions in the water molecule between the inter- and intramolecular degrees of freedom. [ABSTRACT FROM AUTHOR]

A characterisation of cylindrical samples by Neutron Resonance Transmission Analysis (NRTA) at the GELINA facility of JRC Geel (Belgium) is presented. The samples were designed and produced for reactivity worth measurements in the MINERVE reactor of CEA Cadarache (France). NRTA was applied to determine the nuclide composition of UO2, Al2O3 and liquid samples that were doped with silver. The volume number densities of 238U, 107Ag and 109Ag obtained by NRTA are within 2 % fully consistent with the values that are quoted by the manufacturer. In addition, the NRTA data reveal a tungsten contamination which is not reported by the provider. It is shown that such a contamination contributes by up to 5.7 % to the reactivity worth.

The evaluation of the 56Fe neutron induced reactions is currently ongoing at the CEA-Cadarache using the code CONRAD, with the goal to cover the whole energy range from the Resolved Resonance Region to the continuum and estimate the corresponding uncertainties and covariance matrices. Some first results and issues occurred from this work are presented and discussed here, more specifically on the analysed transmission and capture data in the Resolved Resonance Region, as well as the optical and statistical model calculations in the fast neutron energy range.

Neutron capture cross section measurements have been performed at the time-of-flight facility GELINA of the EC-JRC-Geel. Prompt gamma rays, originating from a natural silver sample, were detected by a pair of C6D6 liquid scintillation detectors. The total energy detection principle in combination with the pulse height weighting technique has been used. In this contribution the experimental details together with the data reduction process are described. In addition, first results of calculations with REFIT are presented to verify the quality of recommended cross section data in the resolved resonance region.

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.

Evaluations are being done for the H(n,n), 6Li(n,t), 10B(n,αγ), 10B(n,α), C(n,n), Au(n,γ), 235U(n,f) and 238U(n,f) standard cross sections. Evaluations are also being done for data that are not traditional standards including: the Au(n,γ) cross section at energies below where it is considered a standard; reference cross sections for prompt gamma-ray production in fast neutron-induced reactions; reference cross sections for very high energy fission cross sections; the 235U thermal neutron fission spectrum and the 252Cf spontaneous fission neutron spectrum and the thermal constants.

New results are presented of the 234U neutron-induced fission cross section, obtained with high accuracy in the resonance region by means of two methods using the 235U(n,f) as reference. The recent evaluation of the 235U(n,f) obtained with SAMMY by L. C. Leal et al. (these Proceedings), based on previous n_TOF data [1], has been used to calculate the 234U(n,f) cross section through the 234U/235U ratio, being here compared with the results obtained by using the n_TOF neutron flux.

A set of neutron transmission and capture experiments based on the Time Of Flight (TOF) technique, were performed in order to determine the 241Am capture cross section in the energy range from 0.01 eV to 1 keV. The GELINA facility of the Institute for Reference Materials and Measurements (IRMM) served as the neutron source. A pair of C6D6 liquid scintillators was used to register the prompt gamma rays emerging from the americium sample, while a Li-glass detector was used in the transmission setup. Results from the capture and transmission data acquired are consistent with each other, but appear to be inconsistent with the evaluated data files. Resonance parameters have been derived for the data up to the energy of 100 eV.

Several studies based on the JEFF-3.1.1 nuclear data library show a systematic overestimation of the critical keff for core configurations of MOX fuel assemblies. The present work investigates possible improvements of the C/E results by using new evaluations for Am-241, Pu-239 and Pu-240.