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Neutron transmission and capture measurements have been performed at the time-of-flight (TOF) facility GELINA of the EC-JRC-Geel, using metallic discs of natural silver with different thicknesses. Resonances of neutron interactions with 107Ag and 109Ag were analysed in the energy region below 1 keV applying the Reich-Moore approximation of the R-Matrix theory. Discrepancies between the experimental data and calculations based on the neutron resonance parameters recommended in the main neutron cross section libraries (JEFF, ENDF/B, JENDL) were observed. Improved resonance parameters were determined by a resonance shape analysis using the REFIT code. The resonance energy, neutron (Γn) and radiation (Γγ) width were adjusted in a simultaneous fit to transmission and capture yield spectra below 100 eV. Above 100 eV fixed average radiation widths ( Γ γ 107 =140 meV, Γ γ 109 = 130 meV) were used and the neutron width Γn was adjusted to only the capture yield data. Capture resonance integral I0 were calculated from the parameters derived in this work (I0,107 = 101.20 b, I0,109 = 1515.24 b) and compared with those calculated using the cross sections recommended in the main libraries. Discrepancies in I0 for 107Ag are mainly due to difference in the parameters of the resonances at 16.4 eV, 41.6 eV, 173.9 eV and 202.8 eV. In case of 109Ag differences are mainly due to the difference in parameters of the resonance at 5.2 eV.

The isotope used in this research was supplied by the United States Department of Energy Office of Science by the Isotope Program in the Office of Nuclear Physics., The neutron capture cross section of 154Gd was measured from 1 eV to 300 keV in the experimental area located 185 m from the CERN n_TOF neutron spallation source, using a metallic sample of gadolinium, enriched to 67% in 154Gd. The capture measurement, performed with four C6D6 scintillation detectors, has been complemented by a transmission measurement performed at the GELINA time-of-flight facility (JRC-Geel), thus minimising the uncertainty related to sample composition. An accurate Maxwellian averaged capture cross section (MACS) was deduced over the temperature range of interest for s process nucleosynthesis modelling. We report a value of 880(50) mb for the MACS at kT = 30 keV, significantly lower compared to values available in literature. The new adopted 154Gd(n,γ ) cross section reduces the discrepancy between observed and calculated solar s-only isotopic abundances predicted by s-process nucleosynthesis models., EUFRAT open access programme of the Joint Research Centre at Geel

Methodologies to derive cross section data from spectrum integrated reaction rates were studied. The Westcott convention and some of its approximations were considered. Mostly measurements without and with transmission filter are combined to determine the reaction cross section at thermal energy together with the resonance integral. The accuracy of the results strongly depends on the assumptions that are made about the neutron energy distribution, which is mostly parameterised as a sum of a thermal and an epi-thermal component. Resonance integrals derived from such data can be strongly biased and should only be used in case no other data are available. The cross section at thermal energy can be biased for reaction cross sections which are dominated by low energy resonances. The amplitude of the effect is related to the lower energy limit that is used for the epi-thermal component of the neutron energy distribution. It is less affected by the assumptions on the shape of the energy distribution. When the energy dependence of the cross section is known and information about the neutron energy distribution is available, a method to correct for a bias on the cross section at thermal energy is proposed. Reactor activation measurements to determine the thermal 241 Am(n, γ ) cross section reported in the literature were reviewed. In case enough information was available, the results were corrected to account for possible biases and included in a least squares fit. These data combined with results of time-of-flight measurements give a capture cross section 720 (14) b for 241 Am(n, γ ) at thermal energy.

Self-interrogation neutron resonance densitometry (SINRD) is a non-destructive analysis technique that can be used to quantify the amount of 239Pu in spent nuclear fuel. It is a passive method that relies on the detection of neutrons, which are emitted by the fuel. The amount of 239Pu is estimated from the ratio of the neutron intensity in the fast energy region and in a region close to the 0.296 eV resonance of 239Pu. The neutron intensity in the resonance region is obtained from a detection system with a high sensitivity to 0.296 eV neutrons. This can be realized by using two neutron detectors with 239Pu as convertor material. One of the detectors is covered by a thin Gd foil and the other by a thin Cd foil. The Gd and Cd foils are referred to as SINRD filters. An approach based on the measurement of a fuel assembly in air and surrounded by a slab of polyethylene was developed at SCK·CEN. This approach foresees the insertion of small neutron detectors in the guide tubes of the assembly, and optimisation studies of SINRD were based on Monte Carlo simulations. Experiments to support the results of such simulations were carried out at the time-of-flight facility GELINA of the Joint Research Centre (JRC) in Geel (Belgium). Transmission measurements were performed to verify the quality of the nuclear data that are used to define the optimum thickness of the SINRD filters. Results of self-indication measurements were used to confirm the basic principle of SINRD, that is, that the best results are obtained with a detector that has a high sensitivity to neutrons with energy close to the energy of a strong resonance of the material under investigation. The results of these experiments are presented in this work.

Neutron resonance densitometry has been proposed as a nondestructive analytical method for quantifying special nuclear material (SNM) in the rock- and particle-like debris that is to be removed from the Fukushima Daiichi nuclear power plant. The method is based on neutron resonance transmission analysis (NRTA) and neutron resonance capture analysis combined with prompt-γ-ray analysis (NRCA/PGA). Although quantification of SNM will predominantly rely on NRTA, this will be hampered by the presence of strong neutron-absorbing matrix materials, in particular 10B. Results obtained with NRCA/PGA are used to improve the interpretation of NRTA data. Prompt γ rays originating from the 10B(n, αγ) reaction are used to assess the amount of 10B. The 478 keV γ rays from 10B, however, need to be measured under a high-radiation environment, especially because of 137Cs. To meet this requirement, we developed a well-shaped γ-ray spectrometer consisting of one cylindrical and four rectangular-cuboid LaBr3 scintillators combined with a fast data-acquisition system. Furthermore, to improve the gain stability of the main detector, a special high-voltage divider was developed. Because of the reduction in gain shift, a 3.8% resolution at 662 keV was obtained for long-term measurements. By using the data-acquisition system, which consists of eight 250 MHz digitizers, input signals of over 500 kHz per channel were recorded. The work reported herein demonstrates that, with such a spectrometer, the impact of the Compton edge of 662 keV γ rays from 137Cs is significantly reduced, which allows the 10B amount to be determined with greater sensitivity.

The methodology to derive cross-section data from measurements in a cold neutron beam was studied. Mostly, capture cross-sections at thermal energy are derived relative to a standard cross-section, e.g. the cross-section of the 1 H(n,γ), 14 N(n,γ), or 197 Au(n,γ) reaction, and proportionality between the standard and the measured cross-section, evaluated at different energies in the sub-thermal region, is often assumed. Due to this assumption the derived capture cross-section at thermal energy can be biased by more than 10%. Evidently the bias depends on how much the energy dependence of the cross-section deviates from a direct proportionality with the inverse of the neutron speed. The effect is reduced in case the cross-section is not derived at thermal energy but at an energy close to the average energy of the cold neutron beam. Nevertheless, it is demonstrated that the bias can only be avoided in case the energy dependence of the cross-section is known and proper correction factors are applied. In some cases the results are also biased when the attenuation of the neutron beam within the sample is neglected in the analysis. Some of the cross-section data reported in the literature suffer from such bias effects. Hence, the results have to be corrected using the correction factors presented in this paper.

Principles to determine resonance parameters and their covariance from experimental data are discussed. Different methods to propagate the covariance of experimental parameters are compared. A full Bayesian statistical analysis reveals that the level to which the initial uncertainty of the experimental parameters propagates, strongly depends on the experimental conditions. For high precision data the initial uncertainties of experimental parameters, like a normalization factor, has almost no impact on the covariance of the parameters in case of thick sample measurements and conventional uncertainty propagation or full Bayesian analysis. The covariances derived from a full Bayesian analysis and least-squares fit are derived under the condition that the model describing the experimental observables is perfect. When the quality of the model can not be verified a more conservative method based on a renormalization of the covariance matrix is recommended to propagate fully the uncertainty of experimental systematic effects. Finally, neutron resonance transmission analysis is proposed as an accurate method to validate evaluated data libraries in the resolved resonance region.

Neutron elastic scattering measurements were carried out at the nELBE neutron time-of-flight facility at a 6 m flight path. Energies below 2 MeV were studied using a setup consisting of eight 6Li-glass detectors placed at nominal angles of 15∘ and 165∘ with respect to the incident neutron beam. A deuterated polyethylene sample with 99.999% enrichment in deuterium was used. These angles were chosen since an earlier study showed that the ratio of the differential cross section at these angles is the most sensitive to differences in evaluated files and model calculations. Accurate 165∘/15∘ angle ratios were obtained. Above 1 MeV these are somewhat larger than given by ENDF/B-VII. Simultaneously the early day experiments using a proportional counter to infer angular distributions from deuterium recoil pulse height distributions are being studied through a new experiment with such a device at the Physikalisch-Technische Bundesanstalt (PTB). At 500 keV this experiment favors ENDF/B-VII over JENDL-4.0, while at lower energies agreement with the data is similar.

To support the US Department of Energy Nuclear Criticality Safety Program, neutron-induced cross section experiments were performed at the Geel Electron Linear Accelerator of the Institute for Reference Material and Measurements of the Joint Research Centers, European Union. Neutron capture and transmission measurements were carried out using a metallic calcium sample. The measured data will be used for a new calcium evaluation, which will be submitted with covariances to the ENDF/B nuclear data library., JRC.D.4-Standards for Nuclear Safety, Security and Safeguards

We generated a preliminary set of resonance parameters for 182,183,184,186W in the neutron energy range of thermal up to several keV. The evaluation methodology uses the Reich-Moore approximation to fit with the R-matrix code SAMMY, the high-resolution measurements performed in 2010 and 2012 at the Geel linear accelerator facility. For 183W, the transmission data and capture cross sections calculated with the set of resonance parameters are compared with the experimental values, and some of the average properties of the resonance parameters are discussed. In the analyzed energy range, this work almost doubles the existing resolved resonance evaluations in the ENDF/BVII. 1 library. A preliminary analysis of the performance of the calculated cross sections based on Lead slowing-down benchmarks is presented and briefly discussed., JRC.D.4-Standards for Nuclear Safety, Security and Safeguards

The neutron inelastic scattering cross-section of 23 Na has been measured in response to the relevant request of the OECD–NEA High Priority Request List, which requires a target uncertainty of 4% in the energy range up to 1.35 MeV for the development of sodium-cooled fast reactors. The measurement was performed at the GELINA facility with the Gamma Array for Inelastic Neutron Scattering (GAINS), featuring eight high purity germanium detectors. The setup is installed at a 200 m flight path from the neutron source and provides high resolution measurements using the (n,n′γ)-technique. The sample was an 80 mm diameter metallic sodium disk prepared at IRMM. Transitions up to the seventh excited state were observed and the differential gamma cross-sections at 110° and 150° were measured, showing mostly isotropic gamma emission. From these the gamma production, level and inelastic cross-sections were determined for neutron energies up to 3838.9 keV. The results agree well with the existing data and the evaluated nuclear data libraries in the low energies, and provide new experimental points in the little studied region above 2 MeV. Following a detailed review of the methodology used for the gamma efficiency calibrations and flux normalization of GAINS data, an estimated total uncertainty of 2.2% was achieved for the inelastic cross-section integrals over the energy ranges 0.498–1.35 MeV and 1.35–2.23 MeV, meeting the required targets.

In the resonance region nuclear resonance parameters are mostly obtained by a least square adjustment of a model to experimental data. Derived parameters can be mutually correlated through the adjustment procedure as well as through common experimental or model uncertainties. In this contribution we investigate four different methods to propagate the additional covariance caused by experimental or model uncertainties into the evaluation of the covariance matrix of the estimated parameters: (1) including the additional covariance into the experimental covariance matrix based on calculated or theoretical estimates of the data; (2) including the uncertainty affected parameter in the adjustment procedure; (3) evaluation of the full covariance matrix by Monte Carlo sampling of the common parameter; and (4) retroactively including the additional covariance by using the marginalization procedure of Habert et al., JRC.D.4-Standards for Nuclear Safety, Security and Safeguards

A comprehensive set of Monte Carlo simulations was performed with the MCNP5 code to provide a generic characterisation of the neutron and photon fluxes for time-of-flight measurements at all flight paths of the GELINA facility. Simulations were performed for the direct flux configuration (DFC, 10 keV–20 MeV) and the moderated flux configuration (MFC, 10 meV–1 MeV). Fluxes and flux energy spectra were obtained for both neutrons and photons. For neutrons, additionally, detailed resolution functions and figures of merit were obtained. The validity of the approach for the photon spectra is shown by comparison with a dedicated measurement. Also, a verification is presented of the validity of the neutron resolution function by comparison with measured capture and transmission data for 103 Rh and 56 Fe in the incident neutron energy range from 70 eV to 50 keV. This comprehensive overview will facilitate the planning and analysis of measurements at the GELINA facility with an improved knowledge of its physical characteristics.

The influence of target characteristics on results of neutron-induced cross-section measurements is discussed. The basic principles of total and reaction cross-section experiments are described. The discussion shows that each application needs targets with specific requirements, which are characterized for quantities, such as the total number of nuclei per unit area, effective area and homogeneity. The result of such a characterization can have a strong impact on the total uncertainty of the quantities deduced from the measured data. Based on the measurement principles and on practical experience, recommendations for specific cross-section measurements are presented. These recommendations refer to both the target properties and to the methods used for target characterization. In addition, a characterization method based on the use of neutron resonances is presented. This method can be used to determine the presence and quantity of contaminants and impurities which have a strong impact on the results of cross-section measurements.

It has been suggested that the capture and scattering cross sections of natural cadmium are not well described by the resonance parameters that are given in the evaluated data files. In particular, doubts on the parameters of the first resonance of 113 Cd at 0.178 eV have been raised. This resonance is of high importance in the interpretation in many integral experiments, such as neutron activation analysis, in which cadmium foils are used to shield from thermal neutrons. A new set of experiments has been designed and performed at the neutron time-of-flight facility GELINA, to determine the total cross section and to extract a set of resonance parameters. The covariance information of the experimental data is propagated and the correlation between the resonance parameters is derived. The obtained parameters are then compared to the data available in the literature. Finally a set of criticality experiments from the international handbook of evaluated critical safety benchmark experiments is used to quantify the influence of the change in the resonance parameters.