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Reaction rate ratios are integral responses that are used within the criticality experiments field because they contain spectral information. While these types of measurements have been utilized for nuclear data validation with historic experiments, few experiments of this type have been utilized for recent experiments, as few exist. This work focuses on measured reaction rate ratios for two nearly bare plutonium critical assemblies with different geometries: one that is cube like (with a Pu mass of 40 kg) and one that is slab like (with a Pu mass of 109 kg). Irradiations were performed with both configurations in which foils were placed near the center of the assembly. Plutonium, highly enriched uranium, depleted uranium, and Au foils were included in the irradiation and counted via high-purity germanium detectors. From these measurements, reaction rate ratios were calculated. Measured and simulated values and uncertainties are presented for the reaction rate ratios. Ratios utilizing the following reactions are given in this work: 197Au(n, $$\gamma $$ γ), 197Au(n,2n), 235U(n,fission), 238U(n,fission), 238U(n,2n), 238U(n, $$\gamma $$ γ), and 239Pu(n,fission). Uncertainties for the measured reaction rate ratios ranged from 4% to 7%, and the contribution of various parameters to this uncertainty was investigated. The results are compared to historical experiments and should be used for nuclear data validation for future nuclear data library releases. These measurements are part of the EUCLID (Experiments Underpinned by Computational Learning for Improvements in Nuclear Data) project, which utilizes measurement responses in addition to keff (such as these reaction rate ratios) to help reduce uncertainties in 239Pu nuclear data. [ABSTRACT FROM AUTHOR]

The accurate and efficient solution of the neutron transport equation is essential in nuclear reactor physics for understanding reactor kinetics, stability, and safety. This study investigates the application of the Krylov-Schur method to three-dimensional neutron transport criticality calculations within the Marvin code, comparing its performance to the traditional Power Iteration (PI) method. Using the Takeda benchmark problems, the Krylov-Schur solver demonstrated high accuracy in eigenvalue calculations and neutron flux distributions, closely matching reference Monte Carlo (MC) results. Additionally, the parallel efficiency of the Krylov-Schur method was evaluated, showing significant speed-up and better scalability compared to the PI method, particularly in large-scale computations. However, the method requires a larger memory footprint due to the need to store multiple Krylov subspace vectors and Schur decompositions. Despite this, the findings highlight the Krylov-Schur method's robustness and computational efficiency, making it a promising tool for neutron transport simulations in complex reactor configurations. Future work will focus on investigating the subtraction of high-order eigenvalues and eigenvectors using the Krylov-Schur method to further enhance neutron transport simulations. • Applies Krylov-Schur in 3D reactor criticality calculations, providing robust performance in complex simulations. • Compares Krylov-Schur and PI methods for accuracy and computational efficiency in criticality calculations. • Demonstrates Krylov-Schur's superior parallel scalability for large-scale neutron transport simulations. • Highlights Marvin code as a discrete ordinates solver optimized for serial and parallel computations. [ABSTRACT FROM AUTHOR]

The safe and secure management of civil separated plutonium is a UK government and NDA priority. One potential solution to address this considers the manufacture of a modified version of mixed oxide (MOX) fuel, comprising PuO 2 dispersed within a UO 2 matrix and doped with a suitable neutron absorbing element to maintain criticality control. As an initial step to understand whether an industrially-relevant, proven MOX fuel fabrication process could offer a potential route to the production of a Pu-disposition matrix based on MOX, a series of Gd-doped UO 2 pellets were prepared by Orano at the CDA workshop of the MELOX facility in France. Characterisation was performed to quantify the density, morphology (grain size and porosity), Gd distribution and Gd incorporation mechanism. It was found that the materials produced were highly reproducible and similar in density and morphology, irrespective of the variables investigated, and similar to unirradiated UOX and MOX fuel. Gd was distributed in a similar manner to the distribution of PuO 2 in unirradiated MIMAS (MIcronisation of a MASter Blend) MOX fuel and evidence for the existence of a solid solution between Gd 2 O 3 and UO 2 was ascertained, which could be viewed as favourable from a GDF post-closure criticality control perspective. The source of the powder had the greatest effect on the final characteristics of the Pu-disposition MOX pellets, due to sintering reactivity; however, these differences were minor. These results are a promising step towards the full-scale manufacture of ceramics suitable for the immobilisation and disposition of separated PuO 2 in a GDF, should policy dictate. [ABSTRACT FROM AUTHOR]

The present study focuses on the influence of neutron scattering cross section and angular distribution when employing the Thermal Scattering Law (TSL) on criticality calculation benchmarks, in comparison with the Free Gas Model (FGM). The benchmarks sensitive to beryllium oxide (HMT-027) and graphite (HMT-026) as well as some simplified Pressurized Water Reactor (PWR) benchmarks are interpreted using the OpenMC code to assess the influence of TSL. The results indicate that the influence of TSL mainly attributes to both the total scattering cross section and the secondary angular distribution characterized by the average cosine of the scattering angle. The HMT-027 and HMT-026 series benchmarks respectively show the more significant influence of one of these two effects. For the simplified PWR benchmark, employing the TSL of H in H 2 O weakens the neutron moderation effect, which results in a reduction of around 100 pcm in the calculated k eff , while the combined influence of cross section and secondary angular distribution is negligible for the TSL of UO 2 fuel. • The impact of TSL on criticality calculations is systematically investigated. • Both scattering cross section and angular distribution of TSL influence k eff. • The impact of TSL for BeO on k eff can be up to 1700 pcm for the HMT-027 benchmarks. • The integral impact of TSL for UO 2 is negligible in typical PWR. • TSL for H 2 O weakens neutron moderation and burnup-dependently reduces k eff. [ABSTRACT FROM AUTHOR]