The paper presents a validation of the evaluated fluorine total cross section using a broomstick experiment conducted at the VR-1 reactor. Precise knowledge of fluorine cross sections is crucial for nuclear applications. A collimated neutron beam was directed through PTFE (polytetrafluoroethylene) cylinders of various thicknesses, and the transmitted fast neutron spectra were measured using a stilbene scintillation spectrometer. The setup's sensitivity to total cross sections enables a robust comparison with evaluated nuclear data files. Last updates to the fluorine cross section within the inden collaboration framework, to be adopted by endf/b-viii.1 and jeff-4.0, have shown improvements in criticality benchmarks and integral neutron spectrum measurements using PTFE. The measurements revealed that current standard libraries (endf/b-viii.1 and jeff-3.3) underestimate the transmitted neutron flux by approximately 10-20% in the 1.0-2.0 MeV energy range, indicating an overestimation of the total cross section. The new inden f19f4t4_tot6 evaluation, which incorporates a reduction in the total cross section below 2 MeV, corrects this discrepancy, bringing the calculated transmitted spectra within the 1 sigma experimental uncertainty of the measured data.

The aim of this paper is to evaluate the thermal neutron scattering properties of zirconium hydride (ZrHx) with particular emphasis on the epsilon-phase (x = 2) and the delta-phase (x = 1.5). The methodology involved performing density functional theory calculations using the VASP code to obtain the force constants. Subsequently, the Phonopy and NJOY codes were used to analyse the lattice dynamics and thermal scattering properties. The main results of this investigation include a detailed study of the atomic structures, mechanical properties, phonon density of states, thermal scattering cross sections and benchmark results. These results are compared with other calculations from the literature and experimental measurements. The assessment at the Jozef Stefan Institute (JSI) generally reduces the discrepancies in benchmark results, but certain inconsistencies remain. The addition of the data for zirconium bound in ZrH to the analysis has no appreciable effect on the benchmark results.

The TRIGA Mark II research reactor at the Jo & zcaron;ef Stefan Institute is a key facility in the field of nuclear research, characterized by its versatility and applicability in a wide range of scientific disciplines. This document highlights its operational history, contributions to nuclear safety, education and various scientific applications, including advances in reactor and radiation physics, neutron activation analysis, environmental science and even contributions to the fight against the COVID-19 pandemic. It highlights the reactor's significant role in fostering international collaborations, improving computer modeling techniques for nuclear research, and providing invaluable educational experiences. The great versatility and applicability of the JSI TRIGA reactor is emphasized by its adaptability to various research needs and its ability to enable groundbreaking studies in both fundamental and applied sciences.

The WIMSD-5B transport code is a deterministic tool for nuclear reactor core design and fuel management. It can efficiently handle pin-cell and supercell models and calculate homogenized cross sections, which are essential for reactor physics calculations. It is used by core design packages such as the CORD-2 package, developed at the Jo & zcaron;ef Stefan Institute, and SEANAP developed by Universidad Polit & eacute;cnica de Madrid (UPM). The WLUP update project https://www-nds.iaea.org/wimsd demonstrated the way to update the WIMS-D libraries with different evaluated nuclear data libraries, including ENDF libraries up to version ENDF/B-VII.1. Using an updated version of the procedure, anew WIMS-D library based on the ENDF/B-VIII.0 data was developed to improve the accuracy of core design calculations. Several improvements to the library were made and the effects of each individual improvement was demonstrated using a 3x3 supercell benchmark model that is representative of atypical pressurized water reactor. Finally, the performance of the library over a diverse set of neutron transport problems was tested for, to ensure no regressions were introduced.

We present a framework for propagating uncertainties from nuclear data to integral parameters of the Krsko PWR core design calculations and for calculating their sensitivity profiles to nuclear data. This non-intrusive method is intended for multi-step calculations using codes where nuclear data sensitivity capabilities based on adjoint calculation are not available. In our approach, incident neutron reaction cross-sections and v are randomly sampled based on covariance information in the source evaluated nuclear data files and perturbed core design calculations are run. The uncertainties of the critical boron concentration, of the axial difference in relative power and of their change in the reactor cycle are calculated as the standard deviation of the core design output runs. A linear regression framework is designed to compute sensitivity profiles from the nuclear data perturbations used for uncertainty propagation. After validation against the results of the Serpent Monte Carlo code for the TMI pincell criticality benchmark, we calculated the propagated uncertainty of U-235, U-238, O-16, Fe-56 and Pu-239 to the selected reactor core parameters and the sensitivity of the Krsko critical boron concentration at hot zero power to the fission cross section of U-235. The contribution of individual nuclides to the overall uncertainties was estimated with first-order Sobol indices.

The Fusion Evaluated Nuclear Data Library (FENDL) provides essential nuclear data for fusion research and applications. The release of version FENDL3.2b underwent comprehensive verification and validation (V&V) for neutrons, which has been reported in the associated FENDL reference paper. Since this release, a few evaluated nuclear data (ENDF) files for tungsten isotopes and 232Th in the neutron sublibrary and 7Li in the proton sublibrary have been updated. Also, technical fixes were applied to several files in the deuteron sublibrary. The processing of ENDF files to application files in ACE format is a crucial step for applications. Several important patches have been incorporated into the NJOY2016 processing code for the generation of improved application files from the FENDL ENDF files, impacting damage cross sections, heating, treatment of the unresolved resonance range, and charged-particle processing. The updated ENDF and ACE files have been released as FENDL3.2c. This paper describes these developments and provides the results of limited V&V. Along with the development of the nuclear data library, also the data governance of the FENDL project has been improved so that all updates (including processing) are fully traceable and FENDL users can perfectly reproduce application files on their own computing infrastructure.

A novel experimental setup has been developed for improving nuclear data through fission reactor rate measurements using a borated liquid filter. The experimental device involves simulating a temperature shift of the thermal peak in the neutron flux spectrum by varying the boric acid concentration in a specially designed container filled with borated water. Monte Carlo neutron transport simulations were conducted to optimize the filter geometry, evaluate fission chamber response, and determine the most suitable beam port location for implementation at the Jožef Stefan Institute TRIGA Mark II reactor. Among the options, the thermal column beam port was selected for its favorable neutron spectrum and reduced radiation streaming, despite the radial port offering more uniform flux conditions. The device enables measurements using dual fission chamber setup for comparative measurements under similar irradiation conditions, with an additional channel for dosimetry. Results show that the thermal spectrum can be shifted to correspond to an equivalent temperature from 30°C to 100°C. Boron concentration will be precisely controlled via a combination of volumetric analysis and neutron absorption measurements using an external neutron source. Construction of the experimental system is underway, with the first measurements expected in early 2026.

The Fusion Evaluated Nuclear Data Library (FENDL) is a comprehensive and validated collection of nuclear cross section data coordinated by the International Atomic Energy Agency (IAEA) Nuclear Data Section (NDS). FENDL assembles the best nuclear data for fusion applications selected from available nuclear data libraries and has been under development for decades. FENDL contains sub-libraries for incident neutron, proton, and deuteron cross sections including general purpose and activation files used for particle transport and nuclide inventory calculations. We describe the history, selection of evaluations for the various sub-libraries (neutron, proton, deuteron) with the focus on transport and reactor dosimetry applications, the processing of the nuclear data for application codes, and the development of the TENDL-2017 library which is the currently recommended activation library for FENDL. We briefly describe the IAEA IRDFF library as the recommended library for dosimetry fusion applications. We also present work on validation of the neutron sub-library using a variety of fusion relevant computational and experimental benchmarks. A variety of cross section libraries are used for the validation work including FENDL-2.1, FENDL-3.1d, FENDL-3.2, ENDF/B-VIII.0, and JEFF-3.2 with the emphasis on the FENDL libraries. The results of the experimental validation showed that the performance of FENDL-3.2b is at least as good and in most cases better than FENDL-2.1. Future work will consider improved evaluations developed by the International Nuclear Data Evaluation Network (INDEN). Additional work will be needed to investigate differences in gas production in structural materials. Covariance matrices need to be updated to support the development of fusion technology. Additional validation work for high-energy neutrons, protons and deuterons, and the activation library will be needed.

In the paper we validate theoretical models of the pulse against experimental data from the Jozef Stefan Institute TRIGA Mark II research reactor. Data from all pulse experiments since 1991 have been collected, analysed and are publicly available. This paper summarizes the validation study, which is focused on the comparison between experimental values, theoretical predictions (Fuchs-Hansen and Nordheim-Fuchs models) and calculation using computational program Improved Pulse Model. The results show that the theoretical models predicts higher maximum power but lower total released energy, full width at half maximum and the time when the maximum power is reached is shorter, compared to Improved Pulse Model. We evaluate the uncertainties in pulse physical parameters (maximum power, total released energy and full width at half maximum) due to uncertainties in reactor physical parameters (inserted reactivity, delayed neutron fraction, prompt neutron lifetime and effective temperature reactivity coefficient of fuel). It is found that taking into account overestimated correlation of reactor physical parameters does not significantly affect the estimated uncertainties of pulse physical parameters. The relative uncertainties of pulse physical parameters decrease with increasing inserted reactivity. If all reactor physical parameters feature an uncorrelated uncertainty of 10 % the estimated total uncertainty in peak pulse power at 3 $ inserted reactivity is 59 %, where significant contributions come from uncertainties in prompt neutron lifetime and effective temperature reactivity coefficient of fuel. In addition we analyse contribution of two physical mechanisms (Doppler broadening of resonances and neutron spectrum shift) that contribute to the temperature reactivity coefficient of fuel. The Doppler effect contributes around 30 %–15 % while the rest is due to the thermal spectrum hardening for a temperature range between 300 K and 800 K.

This paper presents the design and analysis of a liquid neutron filter based on water with variable concentrations of boric acid, intended to simulate the thermal neutron spectrum shift in response to temperature changes. The device is designed to enable precise experimental measurement of fission reaction rates at different effective temperatures, particularly for heavy actinides in irradiated nuclear fuel, where temperature feedback effects are critical during reactor operation. The study investigates the optimal design, including container material, geometry, and irradiation positions within the TRIGA reactor. Calculations were performed using the MCNP code and various geometries were tested to determine the configuration that maximises the accuracy of fission cross-section measurements. The paper also examines the impact of the liquid neutron filter on reactor criticality and identifies the most suitable beam port for device integration. Results show that the filter design can effectively simulate neutron spectral shifts and provide valuable data for improving nuclear data libraries, with manageable effects on reactor operation.

To computationally support hydrogen and helium plasma discharges in the early stages of tokamak operation and to support the commissioning of the neutron detectors during these operational phases, creation of a realistic neutron and gamma ray particle source for Monte Carlo simulations will be needed. One of the most important parts of creating the particle source is calculating the reaction rates of the particle-emitting reactions to determine the emission profile in the plasma and the energy spectra of the emitted particles. In this paper the analysis and evaluation of cross sections for important neutron-emitting reactions, namely, 9 Be(p,n γ ) 9 B, 9 Be( 3 He,n γ ) 11 C, and charged-particle emission reactions 9 Be(p,d)2 α and 9 Be(p, α ) 6 Li that cause neutron emission in the next step of interactions are presented. The reaction cross sections were evaluated based on experimental measurements and empirical models describing the interaction of two charged particles. Evaluation of the associated uncertainties was also performed. The main goal of the work is to propose the newly evaluated cross sections for inclusion in the FENDL nuclear data library, thus making the cross section available to other researchers studying the above listed reactions.

Neutron activation dosimetry is the primary method for the determination of the neutron flux or fluence, and in general, it is sensitive to the thermal and resonance energy ranges (radiative capture reactions-(n,gamma)\documentclass[12pt]{minimal} \usepackage{amsmath} \usepackage{wasysym} \usepackage{amsfonts} \usepackage{amssymb} \usepackage{amsbsy} \usepackage{mathrsfs} \usepackage{upgreek} \setlength{\oddsidemargin}{-69pt} \begin{document}$$(n,\gamma )$$\end{document} reactions) and the fast energy range (threshold reactions). However, there are very few nuclear reactions which are sensitive specifically to neutrons in the intermediate-epithermal-energy region. This energy region, along with the fast energy range, will become particularly important in the development and deployment of new reactor technologies (Generation IV reactors and Small Modular Reactors-SMRs), which are currently being championed as technologies enabling a meaningful contribution to decarbonization and the fight against climate change, as well as nuclear fusion. The epithermal neutron energy range is also of particular importance for Boron Neutron Capture Therapy (BNCT), a neutron-based cancer therapy, particularly effective for the treatment of head and neck cancer, malignant meningioma, melanoma and hepatocellular carcinoma. This work investigates and demonstrates the applicability of a particular set of (n,gamma)\documentclass[12pt]{minimal} \usepackage{amsmath} \usepackage{wasysym} \usepackage{amsfonts} \usepackage{amssymb} \usepackage{amsbsy} \usepackage{mathrsfs} \usepackage{upgreek} \setlength{\oddsidemargin}{-69pt} \begin{document}$$(n,\gamma )$$\end{document} reactions in conjunction with boron-based neutron filters to achieve sensitivity in the epithermal energy region, and discusses avenues for future research in this context.

The integral experiments covering the neutron leakage from geometrically simple assemblies with a 252Cf source inside are very valuable tools usable in validation of transport cross section data, since geometric uncertainties play a much smaller role in simple geometric assemblies than in complex assemblies as for example reactor pressure vessel geometry. Since 252Cf(s.f.) is standard neutron source, the uncertainties connected with the source neutron spectrum can be even neglected. The paper refers on validation efforts of neutron leakage from stainless steel block 50 x 50 x50 cm in Research Center Rez. Both the neutron leakage flux at a distance of 1 m from the center of the cubical assembly using stilbene spectrometry and the activation rates at different positions of the assembly were evaluated. In addition to experiments, main sources of uncertainty were identified and evaluated. The results of the stilbene measurements are consistent with the activation measurements results.

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 study emphasizes the intricate nature of predicting reactivity loss in nuclear reactor depletion calculations, revealing that accurate results depend on a multitude of factors beyond just cross sections. The research, which utilizes models from the Three Mile Island core and Krsko Nuclear Power Plant, shows that while computational codes such as Serpent, OpenMC, and WIMS-D can consistently reproduce reactivity loss, the precision of these predictions is significantly influenced by factors like fission yields, energy released per fission, and the specific nuclear data libraries employed. The findings demonstrate that the ENDF/B-VIII.0 library tends to produce a more negative burnup profile, leading to stronger reactivity gradients during burnup. However, the ENDF/B-VIII.1 library improves upon this by reducing the reactivity gradient, aligning more closely with the predictions from ENDF/B-VII.1, a library that has been well-regarded by the industry for its consistency and reliability.

The feasibility and the limitations of using nuclear energy as a dispatchable power source for covering the daily fluctuations of the solar electricity are examined. In particular, the perspective of electricity production in Slovenia until 2050 is investigated, focussing on the projected rapid increase in the solar electricity production and the resulting large-scale annual and daily fluctuations. However, the study is relevant for any electrical grid where photovoltaics is likely to become one of the main sources of electricity. A simulation study based on a nonlinear pressurized water reactor (PWR) model with 2-point neutron kinetics controlled by two groups of control rods using a new simplified control approach is presented. The relevant nonlinearities and the controllability issue affecting this control configuration are discussed. It is shown that a nuclear power plant can be used to compensate for some of the expected power fluctuations using historical data of the solar power production in Slovenia.