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The reactor baffle is an important component of a nuclear reactor that fixes the location of the fuel assembly in the reactor core. The main types of baffles are called light or heavy. The light baffle has mostly the form of steel plates with outer space filled by water, and the heavy baffle is mostly a forged steel element. Both concepts have advantages as well as disadvantages. In the case of the light baffle, one does not need to solve the issue of void swelling, but the neutron economy is not ideal, while the heavy baffle has a good neutron economy, but void swelling is an issue. This paper deals with the effect of the heavy VVER-1000 baffle on criticality. Criticality was measured using a well-defined core composed of 6 fuel assemblies moved to a simulator of the VVER-1000 internals, which is located at the LR-0 reactor. The experiments confirm the fact that the water filling the cooling channels in the baffle has a strong neutron absorbing effect. The keff calculated using the ENDF/B-VIII.0 library significantly underpredicts the experiment, whereas calculations using a new evaluation of 56Fe by the IAEA (INDEN collaboration) give a better agreement. Generally, the presented results are suitable for validation of iron cross sections.

International audience; The production of useful and high-quality nuclear data requires measurements with high precision and extensive information on uncertainties and possible correlations. Analytical treatment of uncertainty propagation can become very tedious when dealing with a high number of parameters. Even worse, the production of a covariance matrix, usually needed in the evaluation process, will require lenghty and error-prone formulas. To work around these issues, we propose using random sampling techniques in the data analysis to obtain final values, uncertainties and covariances and for analyzing the sensitivity of the results to key parameters. We demonstrate this by one full analysis, one partial analysis and an analysis of the sensitivity to branching ratios in the case of (n,n’γ) cross section measurements.

International audience; The necessary improvement of evaluated nuclear data for nuclear applications development is possible through new and high-quality measurements, often combined with appropriate nuclear-reaction modelling. In particular, improving inelastic cross-section evaluations requires new and high-quality data. We measure (n, n’γ) cross-sections using prompt γ-ray spectroscopy and neutron energy determination by time-of-flight. To extract, from these partial data, the total inelastic cross-section, we rely on theoretical model as well as nuclear structure data such as γ ray emission probabilities. This structure information, tabulated in databases, comes with uncertainty. This directly affects the precision of our results, regardless of how good the measurement is. In this paper, we will present the issue of limited precision structure data and its impact on nuclear reaction data quality in the case of neutron inelastic scattering measurements. We will also discuss how to foresee and mitigate the issue.

To probe the neutron inelastic scattering off 232Th, an experiment took place at the EC-JRC Geel conducted with the experimental setup GRAPhEME to detect emitted γ-rays. The prompt γ-ray spectroscopy method was used and 70 experimental 232Th(n, n’γ) cross sections were obtained from the experimental data. Combining these cross sections, nuclear-structure data available in databases and hypotheses to complete the latter, neutron inelastic level production cross sections in 232Th and the total inelastic cross section were calculated. For the first time, the total inelastic cross section of an actinide nucleus was derived on the total neutron energy range from experimental data only. Comparisons of (n, n’) cross section data with evaluated data reveal a good agreement between them all above 300 keV of neutron energy. TALYS calculations are compatible but lower than the evaluated data.

Neutron reaction data for the set of major chromium isotopes were reevaluated from the thermal energy range up to 20 MeV. In the low energy region, updates to the thermal values together with an improved $R$-matrix analysis of the resonance parameters characterizing the cluster of large $s$-wave resonances for $^{50,53}$Cr isotopes were performed. In the intermediate and high energy range up to 20 MeV, the evaluation methodology used statistical nuclear reaction models implemented in the EMPIRE code within the Hauser-Feshbach framework to evaluate the reaction cross sections and angular distributions. Exceptionally, experimental data were used to evaluate relevant cross sections above the resonance region up to 5 MeV in the major $^{52}$Cr isotope. Evaluations were benchmarked with Monte Carlo simulations of a small suite of critical assemblies highly sensitive to Chromium data, and with the Oktavian shielding benchmark to judge deep penetration performance with a 14-MeV D-T neutron source. A significant improvement in performance is demonstrated compared to existing evaluations., Submitted to Nuclear Data Sheets. 41 pages, 44 Figures, 13 Tables

Photo-induced reaction cross section data are of importance for a variety of current or emerging applications, such as radiation shielding design and radiation transport analyses, calculations of absorbed dose in the human body during radiotherapy, physics and technology of fission reactors (influence of photo-reactions on neutron balance) and fusion reactors (plasma diagnostics and shielding), activation analyses, safeguards and inspection technologies, nuclear waste transmutation, medical isotope production and astrophysical applications. To address these data needs the IAEA Photonuclear Data library was produced in 1999, containing evaluated photo-induced cross sections and neutron spectra for 164 nuclides which were deemed relevant for the applications. Since the release of the IAEA Photonuclear Data Library however, new experimental data as well as new methods to assess the reliability of experimental cross sections have become available. Theoretical models and input parameters used to evaluate photo-induced reactions have improved significantly over the years. In addition, new measurements of partial photoneutron cross sections using mono-energetic photon beams and advanced neutron detection systems have been performed allowing for the validation of the evaluations and assessments of the experimental data. Furthermore, technological advances have led to the construction of new and more powerful gamma-beam facilities, therefore new data needs are emerging. We report our coordinated efforts to address these data needs and present the results of the new up-to-date evaluations included in the new updated IAEA Photonuclear Data Library consisting of 219 nuclides. The new library includes 188 new evaluations produced by the CRP evaluators, and one evaluation taken from the JENDL/PD-2016 library, while 20 evaluations were retained from the previous 1999 IAEA Photonuclear Data Library. In most of the cases, the photon energy goes up to 200 MeV. A total of 55 nuclides are new in this library reflecting the progress in measurements but also the developing data needs. In this paper we discuss the new assessment method and make recommendations to the user community in cases where the experimental data are discrepant and the assessments disagree. In addition, in the absence of experimental data, we present model predictions for photo-induced reaction cross section on nuclides of potential interest to medical radioisotope production.

Templates of uncertainties expected in specific measurement types were recently developed. One aim of these templates is to help evaluators in identifying (1) missing or suspiciously low uncertainties and (2) missing correlations between uncertainties of the same and different experiments, when estimating covariances for experimental data employed in their evaluations. These templates also provide realistic estimates of standard deviations and correlations for a particular uncertainty source and measurement type that can be used by evaluators in situations where they are not supplied by the experimenters. This information allows for a more comprehensive uncertainty analysis across all measurements considered in an evaluation and, thus, more realistic evaluated covariances. Here, we extend a template that is applicable to uncertainties expected in neutron-induced fission, (n,f), cross-section measurements. It is applied to improving covariances of 239Pu(n,f) cross-section measurements in the database underlying the Neutron Data Standards evaluations. This particular example was chosen since this evaluation is primarily based on experimental information. Also, some uncertainties of individual 239Pu(n,f) cross-section experiments in this database were suspected to be underestimated. The evaluated uncertainties obtained after updating the covariances in the database by means of the template indeed do increase compared to their original values. Even more importantly, the evaluated mean values change noticeably. These modified cross sections impact application calculations significantly, as is demonstrated by employing them in simulations of the effective neutron multiplication factor for a few selected critical assemblies. However, this updated evaluated 239Pu(n,f) cross section should not be interpreted as the final one that should replace values of the current Neutron Data Standards project. Evaluations for the Neutron Data Standards of the 239Pu(n,f) cross section must be linked to many other observables included in the associated database, most notably to cross sections for 235U(n,f), but also to those for 10B(n,α), 6Li(n,t), 238U(n,f), and 238U(n,γ), because of included measurements of the 239Pu(n,f) cross section that appear as ratios to these reactions. Some of these other reactions are correlated to further observables in the database. Hence, updating uncertainties of data sets of any of these observables can potentially impact the 239Pu(n,f) cross section. Uncertainties for all measurements of these linked physical observables have to be updated before a comprehensive evaluation of the 239Pu(n,f) cross section and its corresponding uncertainties can be provided.

The spectrum averaged cross section (SACS) in a standard neutron field is a preferable tool for cross section validation. The presented work uses only neutron standard, i.e.

GRAPhEME is a γ-spectrometer developed by CNRS/IPHC Strasbourg (France), in collaboration with EC-JRC Geel (Belgium) and IFIN-HH Bucharest (Romania). With its 6 High Purity Planar Germanium detectors and one fission chamber, GRAPhEME, installed at the EC-JRC GELINA facility, was optimized for measurements of accurate (n, xnγ) cross sections on actinides. The experimental methodology is based on the prompt γ-ray spectroscopy coupled to time-of-flight measurements. In this paper, we present an overview of fifteen years of experiments with GRAPhEME at EC-JRC GELINA facility, illustrated by main achievements to highlight the performances reached by our spectrometer. Beyond the experimental work, a close collaboration with theoreticians has emerged allowing the use of the data produced with GRAPhEME to test and constraint nuclear reaction codes like TALYS, CoH and EMPIRE. In a near future, GRAPhEME will be available to start measurement campaigns at the new neutron beam facility SPIRAL2/NFS. There, studies of (n, 2n) and (n, 3n) reactions will be possible and will complete the work done at EC-JRC GELINA on (n, n ) reactions. Despite the amount of cross section data provided by GRAPhEME up to now, the prompt γ-ray spectroscopy method presents some weaknesses that our collaboration tries to overcome. This goes through new calculation schemes based on theoretical modeling constrained on experimental data to infer the total (n, xn) cross section, new instrument to measure conversion electrons but also by being proactive in dissemination activities to make the nuclear structure community aware of our needs about new accurate nuclear structure information on actinides.

The systematic evaluations of spectrum averaged cross sections of dosimetric reactions over a broad range of energies were performed in 252Cf (spontaneous fission) and 235U(nth,f) neutron fields. The neutron sources used in this study were LR-0, VR-1 zero power research light water reactors, LVR-15 10 MW research light water reactor, and 252Cf neutron source with emission specified precisely by the manganese sulphate bath. All spectral averaged cross sections were inferred from measured reaction rates which were derived from gamma spectrometry. These gamma spectrometry measurements were performed using a single detector in all cases. The ratios of 252Cf and 235U spectral averaged cross sections can be used to specify the high energy tail of the 235U prompt fission neutron spectrum as the 252Cf spontaneous fission spectrum is considered as a standard. Furthermore, ratios are independent of cross section uncertainties since uncertainties in the cross sections are eliminated. Theoretical models of fission can be tested based on our measurements. The calculations were performed in MCNP6.2 transport code using different prompt fission neutron spectra and IRDFF-II cross sections for threshold reactions. The ratios are in good agreement using only ENDF/B-VIII.0 235U prompt fission neutron spectrum suggesting it to be harder than in other evaluations.

An effort is now underway to produce a new evaluation of the neutron standards. It is important to maintain experimental programs to increase the quality and extend the database for the neutron cross section standards in order to improve evaluations of them that will be used to convert cross section measurements made relative to those standards. Measurements have been made for most of the standard cross sections since the last evaluation of the standards. The improved database includes the cross sections 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 reactions and ratios among them. The database also includes the 238U(n,γ) and 239Pu(n,f) cross sectionsin addition to the standard cross sections. Those data were included since there are many ratio measurements of those cross sections with the standards and absolute data are available for them.

Bayesian networks were recently suggested as a framework for nuclear data evaluation. Their theory was to some extent described in a recent preprint and some example evaluations were presented. However, due to their newness in the context of nuclear data evaluation and consequently the lack of experience with them within the community makes it dicult to develop trust in the underlying methodology and consequently also the results produced by it. In this contribution, we aim to make a case why evaluators can trust this methodology in principle but will also elaborate on the fact that Bayesian networks are not a silver bullet for evaluation work. On the contrary, evaluators must assess and quantify essential assumptions about nuclear models and experiments with the same dilligence that is already necessary for the application of the wellestablished Generalized Least Squares (GLS) method. We also explain that the increased ease and flexibility to introduce assumptions regarding nuclear models, experiments and their relationships can help an evaluator to rigorously account for assumptions that are very often neglected in evaluations with the GLS method, such as the non-negativity of cross sections, relative experimental normalization uncertainties and the non-linearity in ratios of cross sections. We believe that adopting the Bayesian network paradigm can help both humans to produce evaluations with clearly traceable assumptions and machines to deal with nuclear data more eciently in terms of execution speed and storage size requirements.

The cross section averaged over 235U thermal-neutron induced fission spectrum is a fundamental quantity that can be used in evaluation and validation of nuclear data. Many experiments focused on the determination of Spectrum Averaged Cross Sections (SACS) in 235U(nth,f). Prompt Fission Neutron Spectrum (PFNS) in light water reactors using enriched uranium fuel. In these reactors, there is already some amount of water moderator between the uranium fuel and the irradiated sample. Due to the decrease of hydrogen cross-section with neutron energy, the high energy tail of the reactor spectrum in cores with water moderator may be harder than the pure PFNS. This paper aims to compare the shape of the actual reactor spectrum in various core positions of a research light-water reactor differing each from other by the effective water thickness. The spectrum shape is determined both by calculations and experimentally using various high energy threshold reactions. The impact of the photo-nuclear reactions ({\gamma},n) competing with (n,2n) in production of the same residual nucleus was shown to be less than a percent for most of studied dosimeters. An important exception was found for 197Au(n,2n) 196Au dosimeter irradiated in the outcore channel where a notable photo-neutron contribution to the production of 196Au is caused by the neutron production from the high energy {\gamma}-rays from thermal-neutron capture in 54Fe. The corresponding ENDF/B-VIII.0 data turned out to underestimate such {\gamma}-yield by 40% in comparison with ENDF/B-VI.8. This has improved but however not resolved the disagreement between our measurement and calculations. The remaining deficiency was attributed to the underestimation of the evaluated cross section IAEA/PD-2019 for the 197Au({\gamma},n) cross section near the reaction threshold. The later was confirmed by comparison with existing measured data.

A better knowledge of $(n,xn)$ reaction cross sections is important for both reaction modeling and energy applications. This article focuses on inelastic scattering of neutrons off $^{238}\mathrm{U}$ for which improvements are needed to better constrain evaluations and solve inconsistencies in nuclear power reactor calculations. A new precise measurement of $(n,xn\ensuremath{\gamma})$ reaction cross sections on $^{238}\mathrm{U}$ has been performed at the GELINA (Geel Electron LINear Accelerator) neutron facility operated by EC-JRC-Geel (Belgium) with the GRAPhEME (GeRmanium array for Actinides PrEcise MEasurements) setup. The prompt $\ensuremath{\gamma}$-ray spectroscopy method coupled to time-of-flight measurements is used to extract $(n,xn\ensuremath{\gamma})$ cross section values which can be further combined to infer the total neutron inelastic scattering cross section. Cross section data for 18 $\ensuremath{\gamma}$ transitions (five never measured before) are presented and compared to the data in the literature. Emphasis is especially given to the uncertainty determination to produce partial cross section data as accurate as possible. Due to intrinsic limitations of the experimental method, the use of additional nuclear structure information coupled with theoretical modeling is required to determine the total $(n,{n}^{\ensuremath{'}})$ cross section over the whole neutron energy range. We have investigated modeling aspects of the $^{238}\mathrm{U}(n,{n}^{\ensuremath{'}}\ensuremath{\gamma})$ cross sections related to the description of compound nucleus and preequilibirum mechanisms as well as the discrete part of nuclear structure. Through comparison between experimental and calculated $(n,{n}^{\ensuremath{'}}\ensuremath{\gamma})$ cross sections, we pinpoint inaccuracies in the description of specific reaction mechanisms and challenge recently implemented models. This helps improving the whole modeling of the $(n,{n}^{\ensuremath{'}})$ reaction.

Jacques Raynal seminal contributions to the optical model development and numerical implementation of the solution of coupled-channel equations have been key to generate nuclear data for applications in the last three decades. We recall some of the interactions we had with Jacques from late nineties up until his death. His ECIS code was for a long time, since his collaboration with Peter Moldauer in early eighties, the only optical model and compound decay code where Engelbrecht–Weidenmuller transformation was implemented to consider the impact of strongly coupled channels with large direct inelastic cross sections on the decay of the compound channel. Such physical effect, which was neglected for a long time, proved very important to describe the observed enhancement (Relative to the bare Hauser–Feschbach calculation) of the inelastic scattering cross section on $$^{238}$$ U nucleus, which is a major fuel component in nuclear power reactors. Jacques also proposed an unique and very clever method to calculate dispersive integrals analytically in 1996. His excellent mathematical background and programming skills set a very high bar for future generations. This contribution represents a small homage to our dear colleague and friend.

Comprehensive calculations of photon-induced reactions on $^{233--238}\mathrm{U}$ targets for incident photon energies from 3 up to 30 MeV are undertaken with the statistical model code empire-3.2 Malta. Results are compared with the experimental data from EXFOR and with the current evaluations. The differences and the similarities between the models and parameters used in calculations of photon- and neutron-induced reactions on the same nuclei are discussed with focus on fission. The role of the extended optical model for fission that includes partial damping in the continuum in improving the description of the measured data is pointed out.

Silicon dioxide is a common material in nature and is also often used in the nuclear industry (concrete, radioactive waste repositories in bedrocks), therefore, properly characterizing neutronic properties is very important for nuclear energy and criticality safety applications. The criticality worth was measured for several core arrangements containing silica sand both inside the core and/or as a reflector by measuring the moderator level. The core arrangements used sand as both lateral and central reflectors in different combinations. The results showed that the new thermal scattering law matrix for silicon dioxide available in ENDF/B-VIII.0 may have a problem for backward neutron scattering.

We measured the neutron capture-to-fission cross-section ratio (α ratio) and the capture cross section of $^{235}$U between 0.2 and 200 eV at the n_TOF facility at CERN. The simultaneous measurement of neutron-induced capture and fission rates was performed by means of the n_TOF BaF$_{2}$ Total Absorption Calorimeter (TAC), used for detection of γ rays, in combination with a set of micromegas detectors used as fission tagging detectors. The energy dependence of the capture cross section was obtained with help of the $^{6}$Li(n,t) standard reaction determining the n_TOF neutron fluence; the well-known integral of the $^{235}$U(n,f) cross section between 7.8 and 11 eV was then used for its absolute normalization. The α ratio, obtained with slightly higher statistical fluctuations, was determined directly, without need for any reference cross section. To perform the analysis of this measurement we developed a new methodology to correct the experimentally observed effect that the probabilities of detecting a fission reaction in the TAC and the micromegas detectors are not independent. The results of this work have been used in a new evaluation of $^{235}$U performed within the scope of the Collaborative International Evaluated Library Organisation (CIELO) Project, and are consistent with the ENDF/B-VIII.0 and JEFF-3.3 capture cross sections below 4 eV and above 100 eV. However, the measured capture cross section is on average 10% larger between 4 and 100 eV.

EMPIRE by Herman et al. (Nucl Data Sheets 108:2655, 2007) is a modular system of nuclear reaction codes, comprising various nuclear models, and designed for calculations over a broad range of energies and incident particles. A projectile can be a neutron, a proton, an ion, or a photon. The energy range extends from the beginning of the unresolved resonance region for neutron-induced reactions and goes up to several hundred MeV for heavy-ion induced reactions. The code accounts for the major nuclear reaction mechanisms, including direct, pre-equilibrium, and compound nucleus ones. Direct reactions are described by a generalized optical model or by a coupled channels approach. The pre-equilibrium mechanism can be treated by several quantum mechanical and semiclassical models. The compound nucleus decay is described by a full featured Hauser-Feshbach model with γ-cascade and width fluctuations. Advanced treatment of the fission channel takes into account transmission through a multiple-humped fission barrier with absorption in the wells. Here we will discuss the general capabilities of the code to model nuclear reactions and provide several examples of its use in modeling pre-compound and compound nuclear decay.

Pre-equilibrium nuclear reactions occur on an intermediate time scale when compared to fast direct reactions and to the slow process of compound nucleus evaporation. They dominate the reaction spectra in a large range of energy, especially for fast incident projectiles. The quantum models developed so far describe the process as multiple collisions of an incident leading particle with the target nucleus. The low-energy excited states of the target can have several components forming a collective state. We present here, preliminary results on the description of such a state in the Random Phase Approximation that considers long and short range residual interactions beyond mean field. We have investigated the contributions of different particle–hole modes to the low-lying excited states of the target nucleus. In addition, a simplified model for the particle–hole basis is studied.

We discuss the design and software implementation of a nuclear data evaluation pipeline applied for a fully reproducible evaluation of neutron-induced cross sections of $^{56}$Fe above the resolved resonance region using the nuclear model code TALYS combined with relevant experimental data. The emphasis is on the mathematical and technical aspects of the pipeline and not on the evaluation of $^{56}$Fe, which is tentative. The mathematical building blocks combined and employed in the pipeline are discussed in detail. A unified representation of experimental data, systematic and statistical errors, model parameters and defects enables the application of the Generalized Least Squares (GLS) and its natural extension, the Levenberg-Marquardt (LM) algorithm, on a large collection of experimental data. The LM algorithm tailored to nuclear data evaluation accounts for the exact non-linear physics model to determine best estimates of nuclear quantities. Associated uncertainty information is derived from a Taylor expansion at the maximum of the posterior distribution. We also discuss the pipeline in terms of its IT (=information technology) building blocks, such as those to efficiently manage and retrieve experimental data of the EXFOR library and to distribute computations on a scientific cluster. Relying on the mathematical and IT building blocks, we elaborate on the sequence of steps in the pipeline to perform the evaluation, such as the retrieval of experimental data, the correction of experimental uncertainties using marginal likelihood optimization (MLO) and after a screening of thousand TALYS parameters -- including Gaussian process priors on energy dependent parameters -- the fitting of about 150 parameters using the LM algorithm. The code of the pipeline including a manual and a Dockerfile for a simplified installation is available at www.nucleardata.com., 45 pages

A recently derived dispersive optical model potential (DOMP) for $^{208}$Pb is extended to consider the non-locality in the real potential and the shell-gap in the definition of the nuclear imaginary potentials near the Fermi energy. The modified DOMP improves the simultaneous description of nucleon scattering on $^{208}$Pb and of the $^{208}$Pb particle-hole bound states. The new potential is shown to give a very good description of nucleon scattering data on near-magic targets $^{206,207}$Pb and $^{209}$Bi., Comment: 9 pages, 8 figures

An IAEA research project was dedicated to the compilation, evaluation and recommendation of cross-section data for the accelerator production of $^{11}$C, $^{13}$N, $^{15}$O, $^{18}$F, $^{64}$Cu, and $^{124}$I positron-emitting radionuclides clinically used for PET imaging, and for the accelerator production of two gamma emitters, $^{81}$Rb and $^{123}$I, used in SPECT imaging. Cross sections for 19 charged-particle induced reactions that can be employed for radionuclide accelerator production were evaluated including uncertainties. The resulting reference cross-section data were obtained from Pad\'e fits to selected and corrected experimental data, and integral thick target yields were subsequently deduced. Uncertainties in the fitted results were estimated via a Pad\'e least-squares method with the addition of a 4\% assessed systematic uncertainty to the estimated experimental uncertainty. Experimental data were also compared with theoretical predictions available from the TENDL library. All of the numerical reference cross-section data with their corresponding uncertainties and deduced integral thick target yields are available on-line at the IAEA-NDS medical portal https://www-nds.iaea.org/medportal/ and also at the IAEA-NDS web page https://www-nds.iaea.org/medical/., Comment: 24 pages, 36 figures, 150 references

High quality nuclear data is the most fundamental underpinning for all neutron metrology applications. This paper describes the release of version II of the International Reactor Dosimetry and Fusion File (IRDFF-II) that contains a consistent set of nuclear data for fission and fusion neutron metrology applications up to 60~MeV neutron energy. The library is intended to support: a) applications in research reactors; b) safety and regulatory applications in the nuclear power generation in commercial fission reactors; and c) material damage studies in support of the research and development of advanced fusion concepts. The paper describes the contents of the library, documents the thorough verification process used in its preparation, and provides an extensive set of validation data gathered from a wide range of neutron benchmark fields. The new \mbox{IRDFF-II} library includes 119 metrology reactions, four cover material reactions to support self-shielding corrections, five metrology metrics used by the dosimetry community, and cumulative fission products yields for seven fission products in three different neutron energy regions. In support of characterizing the measurement of the residual nuclei from the dosimetry reactions and the fission product decay modes, the present document lists the recommended decay data, particle emission energies and probabilities for 68 activation products. It also includes neutron spectral characterization data for 29 neutron benchmark fields for the validation of the library contents. The IRDFF-II library and comprehensive documentation is available online at https://www-nds.iaea.org/IRDFF/. Evaluated cross sections can be compared with experimental data and other evaluations at https://www-nds.iaea.org/exfor/endf.htm. The new library is expected to become the international reference in neutron metrology for multiple applications., Submitted to Nuclear Data Sheets, 109 pages, figures revised

Evaluated nuclear data uncertainties are often perceived as unrealistic, most often because they are thought to be too small. The impact of this issue in applied nuclear science has been discussed widely in recent years. Commonly suggested causes are: poor estimates of specific error components, neglect of uncertainty correlations, and overlooked known error sources. However, instances have been reported where very careful, objective assessments of all known error sources have been made with realistic error magnitudes and correlations provided, yet the resulting evaluated uncertainties still appear to be inconsistent with observed scatter of predicted mean values. These discrepancies might be attributed to significant unrecognized sources of uncertainty (USU) that limit the accuracy to which these physical quantities can be determined. The objective of our work has been to develop procedures for revealing and including USU estimates in nuclear data evaluations involving experimental input data. We conclude that the presence of USU may be revealed, and estimates of magnitudes made, through quantitative analyses. This paper identifies several specific clues that can be explored by evaluators in identifying the existence of USU. It then describes numerical procedures to generate quantitative estimates of USU magnitudes. Key requirements for these procedures to be viable are that sufficient numbers of data points be available, for statistical reasons, and that additional supporting information about the measurements be provided by the experimenters. Realistic examples are described to illustrate these procedures and demonstrate their outcomes as well as limitations. Our work strongly supports the view that USU is an important issue in nuclear data evaluation, with significant consequences for applications, and that this topic warrants further investigation by the nuclear science community., Comment: 39 pages, 16 figures, submitted to Nuclear Data Sheets

The results of systematic evaluations of spectrum averaged cross section (SACS) measurements in the fission neutron fields of 252Cf and 235U are presented. The data form a complete database of high-threshold experimental SACS measured in the same installation under the same conditions and using the same high purity germanium gamma spectrometer. This is crucial to reduce the uncertainty of the ratio and the data scattering and therefore, to minimize discrepancies compared to cross section measured under different conditions in different laboratories. This new dataset complements and extends earlier experimental evaluations. The total emission of the 252Cf neutron source during the experiments varied from 9.5E8 to 4.5E8 neutrons per second. The emission was derived in accordance to the data in the Certificate of Calibration involving absolute flux measurements in a manganese sulphate bath. Concerning 235U fission neutron field, the irradiations were carried out in a specifically designed core assembled in the zero power light water LR-0 reactor. This special core has a well described neutron field. After the irradiation, the low volume irradiated samples to be measured by gamma spectrometry were placed directly on the upper cap of a coaxial high purity germanium (HPGe) detector in a vertical configuration (ORTEC GEM35P4). High volume samples were homogenized and strewn into the Marinelli beaker. The HPGe detector is surrounded by the lead shielding box with a thin inner copper cladding and covered with rubber for suppression of background signal and bremsstrahlung. The experimental reaction rates were derived for irradiated samples from the Net Peak Areas (NPA) measured using the semiconductor HPGe detector. The measured reaction rates are used to derive the spectrum-averaged cross sections. Furthermore, measured reaction rates are also compared with MCNP6 calculations using various nuclear data libraries, in particular IRDFF evaluations.

There is a long-standing controversy on nuclear data uncertainty assessment for general purpose nuclear data libraries. On the one hand, nuclear data users would like the libraries to predict uncertainties for selected integral quantities consistent with the integral experimental uncertainties, while on the other hand, doing so could make evaluations dependent on selected integral datasets breaking the general applicability of the library to any existing or future applications. This article studies the hypothesis that certain correlations between nuclear data, which come from the immutable nature of the reactor physics in the integral experiment used as benchmarks, and can be estimated almost independently of the choice of selected integral experiments, nuclear data library, or evaluation methodology. This article reports the findings of an international computational inter-comparison study carried out under the auspices of the Working Party on International Nuclear Data Evaluation Co-operation. The participants represented 5 different organizations, on three different continents and used different initial nuclear data libraries and different calculation methodologies. This study focused on estimating the correlation coefficients between fission, capture and nu-bar for Pu-239 which would arise in the final evaluated nuclear data library if a plutonium metal fast-neutron-spectrum critical experiment with typical integral-measurement uncertainty of 100 pcm was used in the validation and feedback to compile the nuclear data library. The additional knowledge of the correlation coefficients can effectively reduce the propagated uncertainty on criticality experiments from the extended library. This exercise helped to improve understanding the different approaches used, to identify weaknesses and provide indications where further work is required to develop a scientifically rigorous method. This article does not aim at recommending these methods as standards. It aims at honoring the effort of Massimo Salvatores, who was instrumental in pushing studies that should lead to development of uncertainty estimation and by participating in them.

The set of 233U resonance parameters of the ENDF/B-VIII.0 nuclear data library was adopted from the previous ENDF/B-VII.1 evaluation using the external levels to update the thermal values. Adoption of IAEA 2017 thermal standards ( σ f = 533.0 ± 2.2 b, σ c = 44.9 ± 0.9 b, and ν ‾ tot = 2.487 ± 0.011 ) and of the IAEA-recommended thermal-neutron induced prompt fission neutron spectrum (PFNS) with average PFNS energy of 2.030 ± 0.013 MeV requires a re-evaluation of 233U neutron cross sections in the resolved resonance region. A newly produced evaluation is being tested on benchmarks carefully selected from the Handbook of International Criticality Safety Benchmark Experiments (ICSBEP) which are highly sensitive to 233U data. An important goal of this work was to eliminate the strong negative gradient of the calculated effective multiplication factors with respect to the epithermal fission fraction observed in the validation of the ENDF/B-VIII.0 library for those assemblies. A significant improvement in integral performance of critical 233U solutions is observed for the newly proposed evaluation. Further work addressing the fast neutron region is needed.

The new paradigm of nuclear reaction evaluation postulates that differential and integral experiments along with the reaction models should be used in concert to produce the evaluated data files. The results of a previous assimilation project are summarized as a proof of principle that adjusting reaction model parameters to the results of integral experiments is feasible. The new paradigm requires vast modernization of the nuclear data infrastructure. Such modernization is actually carried out by several national and international efforts. We describe those dedicated to: (i) handling of differential and integral data, (ii) providing and operating the sensitivity profiles, (iii) ensuring reproducibility of the evaluations, and (iv) enabling automated verification of the entire library. We also point to the importance of the nuclear reaction modeling and discuss advantages and disadvantages of the new paradigm.

It is shown here that 14-MeV D+T LLNL pulsed-sphere experiments bring complementary information into the process of validating nuclear data compared to experiments that are traditionally used for this purpose—such as critical assemblies. To be more specific, the 14-MeV D+T LLNL pulsed-sphere neutron-leakage spectra enable to validate scattering and fission nuclear data up to 15 MeV (compared to approximately up to 5 MeV when using criticality experiments) and employ to this end simple compound targets containing only few isotopes. Sensitivity profiles of the spectra to nuclear data are calculated in order to understand in detail which isotopes, observables, and energy ranges of nuclear data contribute significantly to their simulation. These profiles are presented for a few selected spheres containing 16O, 12C, 56Fe, and 239Pu. It is shown that the neutron-leakage spectra of spheres containing light elements are mostly sensitive to elastic- and inelastic-scattering cross sections on discrete levels and corresponding angular distributions. Spheres of structural materials are sensitive to elastic- and inelastic-scattering cross sections, including scattering on discrete levels and the continuum, and double-differential cross sections. Actinide spheres are also strongly sensitive to the fission observables, in particular to the total-fission neutron spectrum. Thin spheres (in which neutrons experience on average less than one scatter) are mostly sensitive to data near the elastic peak, in the energy range from 12–15 MeV, while thicker ones can be sensitive to data at lower incident-neutron energies due to multiple-scattering effects. This information is brought together with simulations of 71 pulsed-sphere neutron-leakage spectra using the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear-data libraries. This analysis highlights ENDF/B-VIII.0 data that could be further investigated for potential shortcomings (6Li, 12C, 16O, 24-26Mg, 27Al, 48Ti, 56Fe, and 208Pb) or are likely reliable (1,2H, 7Li, 9Be, 14N, 235,238U, and 239Pu) as indicated by validating with LLNL pulsed-sphere experiments.

18F is an important radioisotope used in nuclear medicine for diagnostic purposes. It is produced most often via 18O (p,n)18F reaction. During its production, the oxygen contained in the H218O molecules placed in the cooled target is irradiated by accelerated protons. The production of each 18F atom is accompanied by anisotropic high-energy neutron emission, which is a problem for ensuring radiation safety. These neutrons are responsible for the high radiation dose in a cyclotron pit during 18F production as well as for activation of cyclotron components or might be responsible even for target material changes due to their high penetration properties. A precise characterization of a neutron field on the surface of the liquid water target, used for 18F production, is presented. The neutron flux characterization is performed both by using large-area activation detectors and stilbene detectors at 1 m from the target in a forward direction. The experiments show significant discrepancies between calculated and measured reaction rates in higher threshold reactions 58Ni(n,x)57Co and 51V (n,α)48Sc, confirming that the current neutron production reaction models are not suitable for the characterization of the secondary neutrons leaking from the target.

Comprehensive calculations of cross sections for neutron-induced reactions on 232–237U targets are performed in the 10 keV–30 MeV incident energy range with the code EMPIRE–3.2 Malta. The advanced modelling and consistent calculation scheme are aimed at improving our knowledge of the neutron scattering and emission cross sections, and to assess the consistency of available evaluated libraries for light uranium isotopes. The reaction model considers a dispersive optical potential (RIPL 2408) that couples from five (even targets) to nine (odd targets) levels of the ground-state rotational band, and a triple-humped fission barrier with absorption in the wells described within the optical model for fission. A modified Lorentzian model (MLO) of the radiative strength function and Enhanced Generalized Superfluid Model nuclear level densities are used in Hauser-Feschbach calculations of the compound-nuclear decay that include width fluctuation corrections. The starting values for the model parameters are retrieved from RIPL. Excellent agreement with available experimental data for neutron emission and fission is achieved, giving confidence that the quantities for which there is no experimental information are also accurately predicted. Deficiencies in existing evaluated libraries are highlighted.

A corrected description of the interpolation scheme for the neutron data standards is given. It is emphasized that the energies (nodes) listed are where the evaluated cross sections are shown. A numerical example that shows how the high-resolution 235U(n,f) data should be averaged to be compared with lower resolution cross sections is given.

The OECD-NEA High Priority Request List (HPRL) is a point of reference to guide and stimulate the improvement of nuclear data for nuclear energy and other applications, and a tool to bridge the gap between data users and producers. The HPRL is application-driven and the requests are submitted by nuclear data users or representatives of the user's communities. A panel of international experts reviews and monitors the requests in the framework of an Expert Group mandated by the NEA Nuclear Science Committee Working Party on International Nuclear Data Evaluation Cooperation (WPEC). After approval, individual requests are classified to three categories: high priority requests, general requests, and special purpose requests (e.g., dosimetry, standards). The HPRL is hosted by the NEA in the form of a relational database publicly available on the web. This paper provides an overview of HPRL entries, status and outlook. Examples of requests successfully completed are given and new requests are described with emphasis on updated nuclear data needs in the fields of nuclear energy, neutron standards and dosimetry., submitted to EPJ Web of Conferences for the proceedings of the ND2019 International Conference on Nuclear Data for Science and Technology (Beijing ,China, May 19 - 24, 2019)

Background: The $^{7}\mathrm{Li}(p,n)^{7}\mathrm{Be}$ reaction is one of the most used nuclear reaction for accelerator-based neutron sources. There are few experimental cross section data in the double-value energy region and they are discrepant, as are the reaction yields.Purpose: We derive the $^{7}\mathrm{Li}(p,n)^{7}\mathrm{Be}$ reaction cross section, and measure with small uncertainty the $^{197}\mathrm{Au}(n,\ensuremath{\gamma})^{198}\mathrm{Au}$ spectrum-averaged cross section at neutron energy around 30 keV.Method: By irradiating Li metal targets over the proton energy range of 1879 to 1960 keV, thick target yields were measured using the generated $^{7}\mathrm{Be}$ activity. Based on the theoretical description of the reaction yield, accelerator parameters and reaction cross sections are derived. Gold foils were activated with the neutron field generated by the $(p,n)$ reaction on a Li target at a proton energy of about a half keV above the reaction threshold.Results: The thick target yield is well reproduced when the Breit-Wigner single-resonance formula for $s$-wave particles is used to describe the reaction cross section. The ratio between neutron and proton widths was found to be equal to ${\mathrm{\ensuremath{\Gamma}}}_{n}/{\mathrm{\ensuremath{\Gamma}}}_{p}=5.4\sqrt{1\ensuremath{-}{T}_{th}/{T}_{p}}$. The detailed balance principle is used to obtain the cosmologically important time-reversed $^{7}\mathrm{Be}(n,p)^{7}\mathrm{Li}$ reaction cross section. The measured $^{197}\mathrm{Au}(n,\ensuremath{\gamma})^{198}\mathrm{Au}$ spectrum-averaged cross section agrees with the value calculated from the ENDF/B-VIII.0 library.Conclusions: We demonstrated the feasibility of deriving the $^{7}\mathrm{Li}(p,n)^{7}\mathrm{Be}$ reaction cross section from the thick target yield. Using the ratio between neutron and proton widths obtained in this work reduces the uncertainty in calculating the reaction cross section to a factor of 2.3.