Dynamical fission calculations show that the post-scission configurations resemble two collinear pear-shaped fragments whose juxtaposed surface bulges subside relatively quickly, as the fragments acquire smoother shapes. The associated rapid speed of the healing bulge surface may boost nucleons in the fragment to energies sufficient for emission. The present study explores this mechanism by following the fate of nucleons that are reflected off the inwards moving bulge surface. The simulations suggest that the mechanism may produce high-energy neutrons at the level of a few per cent.

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.

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.

We assess the ability of a light-composite breakup model coupled to pre-compound and compound statistical emission models to describe selected compound nuclear cross sections. The framework employed calculates elastic and nonelastic breakup differential cross sections using the post-form distorted wave Born expressions of the Ichimura-Austern-Vincent approach. This has been incorporated into the EMPIRE nuclear reaction code for the specific case of the deuteron, used in the calculations presented here. The exciton preequilibrium and Hauser-Feshbach compound-nucleus models are used to account for pre-equilibrium and equilibrium emission. The outcome of our analyses yields a satisfactory agreement between theoretical and experimental values in most cases. The analysis also permits a preliminary assessment of the importance of inelastic excitation on the cross-section predictions.

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 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.

We report on a microscopic modeling of the first order of multistep direct emission based on one phonon excitations, given by the QRPA model, and an effective in-medium nucleon nucleon interaction, described within the JLM folding model. The results of our coupled channels framework for deformed target are illustrated in the case of low energy discrete state excitations in 152Sm. Pre-equilibrium predictions and their impact on inclusive and exclusive (n,xn) cross sections and fission cross sections are discussed for actinides. We discuss the importance of i) collective excitations to describe the neutron emission spectra and of ii) spin distribution of the residual nucleus formed after the neutron pre-equilibrium emission, that is a key ingredient to model the residual nucleus decay.

We have conducted a thorough analysis of the potential energy surfaces (PES) in ^236 U and ^233 Th using the Cassini-ovals parameterization within the macro–micro approach. We employed the state-of-the-art immersion water flow (IWF) method to study the saddles on four-dimensional energy grids encompassing reflection-asymmetric shapes. For ^233 Th, we computed the adiabatic potential energy surfaces by minimizing configurations with one blocked neutron within ten levels below and above the Fermi level. Our results show satisfactory agreement with empirical and experimental estimates for both nuclei, specifically regarding the first and second fission barriers. This suggests that our method holds promise in efficiently describing non-compact shapes while reducing the dimensionality of the space without sacrificing accuracy. Interestingly, employing Cassinian oval parameterization fails to reveal a pronounced, hyper-deformed third minimum in the potential energy landscape. Instead, only a shallow third minimum is observed for ^233 Th, while in ^236 U, this minimum ultimately vanishes. This finding holds significant importance when considering the modeling of fission cross-sections.

We have investigated following capture reactions: 197Au(n,g)198Au, 63Cu(n,g)64Cu, 45Sc(n,g)46Sc, 181Ta(n,g)182Ta, 140Ce(n,g)141Ce, 139La(n,g)140La, 176Yb(n,g)177Yb, 55Mn(n,g)56Mn, and 141Pr(n,g)142Pr in a standard 252Cf(s.f.) neutron field. The experimentally derived reaction rates were compared with calculations using state-of-the-art libraries IRDFF-II, ENDF/B-VIII.0, JENDL-5 and JEFF-3.3 performed in MCNP6.2 transport code. The experiment was focused on capture reactions in high energy neutron range. To suppress scattered low energy neutrons, all activation materials were enclosed by a spherical cadmium surface of 30 cm in diameter. Experimentally derived reaction rates of 197Au(n,g)198Au, 63Cu(n,g)64Cu reactions agree with calculations using all state-of-the-art nuclear data libraries. The highest discrepancies were found when evaluating 139La(n,g)140La and 55Mn(n,g)56Mn reactions utilizing all libraries. Experimental reaction rates applying data from the JENDL-5 library agree very well with the results of other reactions except for the 45Sc(n,g)46Sc reaction. 181Ta(n,g)182Ta reaction rate does not agree within uncertainties using any library. The 176Yb(n,g)177Yb reaction rate agrees reasonably well only when performing calculation using JENDL-5 and ENDF/B-VIII.0 libraries, unlike JEFF-3.3 library. 140Ce(n,g)141Ce and 141Pr(n,g)142Pr reaction rates applying JEFF-3.3 and ENDF/B-VIII.0 libraries do not agree within uncertainties.

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.

One of the most important sources of systematic uncertainties in the evaluation of measured cross sections is the absolute normalization of every dataset, which were often performed by measuring simultaneously the reference cross-section of the standard isotope. In other experiments the shape of the cross-section spectrum is normalized using as reference the integral value in a certain energy interval taken from an evaluated library. The choice of the energy interval used as reference has been often left up to the experimentalist criteria, leading to inconsistent normalizations and hardly assessable uncertainties. In this work the experimental datasets of the (n,f) cross section of many actinides are reviewed looking for the best suited energy interval to be recommended for renormalization purposes. Using standard integration intervals, wide enough to get very low statistical uncertainties, should improve the normalization of every experimental dataset, reducing so the associated total uncertainty when making the evaluation. A common integration range from 8 to 10 MeV is proposed for the whole set of actinides needed in fission applications. This energy range, which falls between the second and the third fission-chance thresholds, is characterized by a flat behaviour of the fission cross sections.

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.

A cross section evaluation of neutron induced reactions on Ti-48 is undertaken using the Unified Monte Carlo-B (UMC-B) approach. The evaluation concentrates on estimating the covariance and the use of the UMC-B allows avoiding the deficiencies of linear regression brought by the traditional least squares method. Eight main neutron and charged particle emission reactions from n+Ti-48 in the fast neutron energy region below 20 MeV are studied in this work. The posterior probability density function (PDF) of each neutron cross section is obtained in a UMC-B Bayesian approach by convoluting the model PDFs sampled based on model parameters and the likelihood functions for the experimental data. Nineteen model parameters including level density, pair corrections, optical model and Kalbach matrix element parameter are stochastically sampled with the assumption of normal distributions to estimate the model uncertainty. The Cholesky factorization approach is applied to consider potential parameter correlations. Finally, the posterior covariance matrices are generated using the UMC-B generated weights. The new evaluated results are compared with the CENDL-3.2, ENDF/B-VIII.0, JEFF-3.3, TENDL-2021 and JENDL-5 evaluations and differences are discussed.

We show how the multi-step direct reaction series can be evaluated using Monte Carlo methods. Nucleon-nucleon collisions occur according to the random selection of a nucleon’s attenuation factor along its classical trajectory. The particles and hole excited in a collision are selected from a local Fermi distribution. We assume that the particles continue to propagate and possibly collide again before leaving the nucleus. We assume that holes collide in place, to possibly produce other particles and holes.

Precise normalization of ToF yield measurements of neutron induced fission in fissile targets is challenging, but the appropriate normalization is also of critical importance for nuclear energy and criticality safety, among other applications. A typical normalization relies on the Thermal Neutron Constants (TNC) recommended at the thermal point (fission σf0 and capture σγ0 thermal cross sections) by the Neutron Standards. However, many ToF experiments do not collect data down to the thermal energy and use normalization cross-section integrals defined at different arbitrary energy intervals. Normalization fission cross-section integrals I3 are recommended in between-valleys energy regions 8.1–14.7 eV, 7.8–11 eV, 9–20 eV, and 11.7–19.5 eV for fissile targets 233U, 235U, 239Pu, and 241Pu with values equal to 689.0(10.8), 245.7(4.1), 1059(6), and 1378(33) b⋅eV, respectively. The 235U normalization integral I3 derived in this work of 245.7(4.1) b⋅eV is in good agreement within quoted uncertainties with the Neutron Standards value of 247.5(3.0) b⋅eV.Additional cross-section integrals I1 in the thermal region from 20–60 meV are derived to fix both the normalization and the slope of the fission cross section at the thermal point providing additional constraints for R-matrix evaluations of experimental fission yields. Ratios σf0/I1 and I3/I1 feature very low uncertainty due to the strong positive correlations between the numerator and the denominator and are comprehensively derived for the first time. Integral ratios I3/I1 of 39.31(54), 13.08(20), 41.65(22), and 40.46(85) are recommended as reference for fissile targets 233U, 235U, 239Pu, and 241Pu, respectively. Similarly, ratios σf0/I1 of 30.40(18), 31.22(12), 29.58(7), and 29.95(35) are also recommended as reference for corresponding fissile targets. These recommended ratios can be used by experimentalists to normalize their measured thin-target fission yield data, but also by evaluators to constrain their R-matrix analysis that use multiple sets of measured data renormalized as recommended in this work.An evaluation of thermal (n,f) cross sections σf0 presented in this work for all fissile targets is in excellent agreement with the Thermal Neutron Constants defined in the Neutron Standards within one-sigma uncertainty. Such agreement guarantees the reliability of our new evaluation of reference integrals based on the whole database of consistent experimental TOF data.

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.

Spectrum -averaged cross-section (SACS) ratio measurement showing reduced uncertainties of measured SACS in two independent fission neutron fields is presented. The used prompt fission neutron fields correspond to the 252 Cf(sf) and 235 U( n th , f ) fission neutrons. The employed SACS in the 235 U( n th , f ) prompt fission neutron spectrum (PFNS) were measured using three different light water reactors: LR-0 and VR-1 zero power reactors and the LVR-15 10 MWt reactor. The employed SACS in the 252 Cf(sf) prompt fission neutron field were measured using a certified high -flux neutron source. Existing correlations among measured SACS in the two different neutron fields are estimated and used to reduce the uncertainty of measured SACS ratio for IRDFF dosimetry reactions. The derived set of measured SACS ratios with reduced uncertainty extends previous works into the higher -energy fission neutron range up to 20-30 MeV. The SACS ratio in 252 Cf(sf) and 235 U( n th , f ) PFNS can be used to probe the high-energy tail of the 235 U( n th , f ) fission neutron spectrum as the 252 Cf(sf) reference neutron spectrum is relatively well known. Derived experimental SACS ratio data. featuring low uncertainty are compared to the calculated dosimetry SACS ratio using the IRDFF-II dosimetry cross sections and the 235 U( n th , f ) ENDF / B-VII.1, ENDF / B-VIII.0, or JEFF -3.3 PFNS evaluations. The ENDF / B-VIII.0 evaluation of the 235 U( n th , f ) PFNS agrees well with derived SACS ratio data within quoted uncertainties. Other libraries predict a significantly lower fraction of 235 U( n th , f ) fission neutrons above 11 MeV of the outgoing neutron energy.