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The isotopes of Zr with A = [90, 91, 92, 94] make up more than 97% of naturally occurring Zr and are important to many nuclear applications such as nuclear reactors. One of the attractive qualities of naturally occurring Zr isotopes is that they have a low σγ/σt ratio at most neutron energies. Thus, they improve the neutron economy in reactors by preferentially scattering neutrons rather than absorbing them. This same quality also presents a challenge to measuring the capture cross section, σγ, of Zr isotopes. The ENDF/B-VIII.0 library has a relative uncertainty of approximately 10–20% for incident neutron energies < 0.1 MeV and an uncertainty greater than 20% for energies > 0.1 MeV for the majority of natural Zr isotopes. This motivated the Nuclear Criticality Safety Program to embark on a campaign to accurately measure and evaluate these Zr isotopes. In this work, we demonstrate energy-dependent neutron capture cross section measurements for the first enriched sample to be measured: 90Zr.

Over the last decade, efforts were made to improve the performance of the experimental set-ups at the Geel Electron Linear Accelerator (GELINA) neutron time-of-flight facility of the European Commission Joint Research Centre (EC-JRC). These efforts, which result in an improved quality of neutroninduced cross section data for many reaction channels like elastic, inelastic, capture, fission, etc., relate to the accelerator, the measurement setups and the data reduction and analysis procedures. This paper presents a summary of the data produced in the last years at GELINA for nuclear energy applications. Most of the work has been performed as part of the EUFRAT open-access program.

Criticality safety analysis is required at various stages of the back-end of the fuel cycle, i.e. reprocessing, transport, storage and disposal of spent nuclear fuel (SNF). To account for the reduction in reactivity due to fuel burnup, the Burn-Up Credit (BUC) concept was introduced. Evidently, this concept depends on the quality of nuclear data, in particular the absorption cross sections of some key nuclides. A dedicated programme has been established at the GELINA facility of the JRC-Geel to produce accurate cross section data and validate the evaluated nuclear data libraries for neutron interactions with fission fragments that are relevant for a BUC approach. In this work, cross section data for 103Rh and 155Gd are presented and the results are compared with the main evaluation libraries, showing good agreement in the thermal energy region with ENDF/B-VIII.0 and JEFF-3.3, but not with JENDL-4.0 for 103Rh.

JRC Geel operates a neutron time-of-flight facility based on an electron accelerator, GELINA. Experimental setups to determine total and reaction cross sections (capture, elastic and inelastic scattering, fission and charged particle reactions) are available at measurement stations different flight-path length. While most of the experimental work is focussed on nuclear energy applications, regularly cross section of interest for astrophysical problems are measured. Examples of those are capture experiments of nuclei relevant for the s-process such as 89Y or the 16O (n,α) reaction which is the inverse of the 13C(α,n) reaction, an important source of neutrons feeding the s-process.

The zero-power reactor MINERVE (CEA Cadarache) was designed to perform reactivity worth measurements by the oscillation technique. The various experimental programs, undertaken for the last thirty years, involved cylindrical samples with a diameter of about 1 cm and a height ranging from a few cm to 10 cm. Most of the samples are composed of UO2 pellets mixed with a high neutron absorbing nuclide, i.e. fission product, actinide, in a double-sealed Zry-4 container. An experimental program started in 2015 in collaboration with the Joint Research Centre of Geel to study the MINERVE samples at the time-of-flight facility GELINA by the neutron transmission technique. The two main objectives consist of checking both the composition of the MINERVE samples provided by the manufacturer and the quality of the resonance parameters recommended in the evaluated neutron data library JEFF-3.3. The pioneer experiments on MINERVE samples containing 107Ag and 109Ag revealed a substantial Tungsten contamination that was not reported by the manufacturer. Such a Tungsten contamination is related to the manufacturing process of the sample pellets. The observed Tungsten contaminations lead to non-negligible increases of the C/E ratios up to a few percent. second experimental campaign on MINERVE samples containing 99Tc provided useful insight on the quality of the 99Tc resonance parameters measured at the GELINA facility at the end of the 90s. The ongoing program continuing through 2022 will deliver data for samarium (natural, 147, 149, 152), neodymium (natural,143,145), gadolinium (natural, 155), europium (151, 153), rhodium (103), cesium (133), hafnium (180), dysprosium (160, 161, 162, 163, 164) and erbium (168, 170). The present work focuses on the data analysis technique developed for long cylindrical samples with a diameter smaller than the neutron beam, and on the grain size distribution model implemented in the resonance shape analysis REFIT.

A characterisation of cylindrical samples by Neutron Resonance Transmission Analysis (NRTA) at the GELINA facility of JRC Geel (Belgium) is presented. The samples were designed and produced for reactivity worth measurements in the MINERVE reactor of CEA Cadarache (France). NRTA was applied to determine the nuclide composition of UO2, Al2O3 and liquid samples that were doped with silver. The volume number densities of 238U, 107Ag and 109Ag obtained by NRTA are within 2 % fully consistent with the values that are quoted by the manufacturer. In addition, the NRTA data reveal a tungsten contamination which is not reported by the provider. It is shown that such a contamination contributes by up to 5.7 % to the reactivity worth.

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.

$^{99}\mathrm{Tc}$ has been the subject of several experimental programs at the JRC-Geel since the late 1990s. This work presents new transmission data measured at the GELINA facility with the time-of-flight technique, whose simultaneous analysis with past transmission experiments provided a revised set of resonance parameters up to 5 keV. An iterative analysis between the resonance and ``continuum'' energy ranges leads to average resonance parameters $[{S}_{0}=0.51(3)$, $\ensuremath{\langle}{\mathrm{\ensuremath{\Gamma}}}_{{\ensuremath{\gamma}}_{0}}\ensuremath{\rangle}=137(7)$ meV, ${D}_{0}=12.8(2)$ eV] and theoretical average capture cross sections consistent with those reported in the literature. At $E=30$ keV, we obtain a capture cross section of 1076(45) mb and a Maxwellian-averaged cross section (MACS) of 1016(43) mb. The agreement with the recommended MACS value of 933(47) mb remains within the quoted uncertainties. This result demonstrates the performances of transmission experiments to provide reliable theoretical capture cross sections for long-lived fission products in the energy range of interest for $s$-process nucleosynthesis calculations.

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

In the measurement of neutron capture cross-sections of fissile isotopes, the fission channel is a source of background which can be removed efficiently using the so-called fission-tagging or fission-veto technique. For this purpose a new compact and fast fission chamber has been developed. The design criteria and technical description of the chamber are given within the context of a measurement of the 233U(n, ������) cross-section at the n_TOF facility at CERN, where it was coupled to the n_TOF Total Absorption Calorimeter. For this measurement the fission detector was optimized for time resolution, minimization of material in the neutron beam and for alpha-fission discrimination. The performance of the fission chamber and its application as a fission tagging detector are discussed., This work was partially supported by the French NEEDS/NACRE Project and by the European Commission within HORIZON2020 via the EURATOM Project EUFRAT.

The joint evaluated fission and fusion nuclear data library 3.3 is described. New evaluations for neutron-induced interactions with the major actinides U , U and Pu , on Am and Na , Ni , Cr, Cu, Zr, Cd, Hf, W, Au, Pb and Bi are presented. It includes new fission yields, prompt fission neutron spectra and average number of neutrons per fission. In addition, new data for radioactive decay, thermal neutron scattering, gamma-ray emission, neutron activation, delayed neutrons and displacement damage are presented. JEFF-3.3 was complemented by files from the TENDL project. The libraries for photon, proton, deuteron, triton, helion and alpha-particle induced reactions are from TENDL-2017. The demands for uncertainty quantification in modeling led to many new covariance data for the evaluations. A comparison between results from model calculations using the JEFF-3.3 library and those from benchmark experiments for criticality, delayed neutron yields, shielding and decay heat, reveals that JEFF-3.3 performes very well for a wide range of nuclear technology applications, in particular nuclear energy., Acknowledgements The authors would like to thank all contributors to the JEFF project that shared their insights and provided their specific input for the present JEFF-3.3 evaluation and its immediate predecessors JEFF-3.2 and JEFF-3.1.2.

International Conference on Nuclear Data for Science and Technology, ND 2019, Bejing, China, 19 May 2019 - 24 May 2019; The European physical journal / Web of Conferences 239, 01006 (2020). doi:10.1051/epjconf/202023901006, Published by EDP Sciences, Les Ulis

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.

This work was partially supported by the French NEEDS/NACRE Project and by the European Commission within HORIZON2020 via the EURATOM Project EUFRAT. The authors would like to acknowledge more specifically JRC Geel for targets preparation and for providing full support for a first test measurement of the fission chamber at GELINA., U-233 is the fissile nuclei in the Th-U fuel cycle with a particularily small neutron capture cross setion which is on average about one order of magnitude lower than its fission cross section. Hence, the measurement of the U-233(n,gamma) cross section relies on a method to accurately distinguish between capture and fission gamma-rays. A measurement of the U-233 alpha-ratio has been performed at the n_TOF facility at CERN using a so-called fission tagging setup, coupling nTOF 's Total Absorption Calorimeter with a novel fission chamber to tag the fission gamma-rays. The experimental setup is described and essential parts of the analysis are discussed. Finally, a preliminary U-233 alpha-ratio is presented., French NEEDS/NACRE Project, European Commission within HORIZON2020 via the EURATOM Project EUFRAT

233U is of key importance among the fissile nuclei in the Th-U fuel cycle. A particularity of 233U is its small neutron capture cross-section, which is on average about one order of magnitude lower than the fission cross-section. The accuracy in the measurement of the 233U capture cross-section depends crucially on an efficient capture-fission discrimination, thus a combined set-up of fission and γ-detectors is needed. A measurement of the 233U capture cross-section and capture-to-fission ratio was performed at the CERN n_TOF facility. The Total Absorption Calorimeter (TAC) of n_TOF was employed as γ-detector coupled with a novel compact ionization chamber as fission detector. A brief description of the experimental set-up will be given, and essential parts of the analysis procedure as well as the preliminary response of the set-up to capture are presented and discussed.

We describe the new ENDF/B-VIII.0 evaluated nuclear reaction data library. ENDF/B-VIII.0 fully incorporates the new IAEA standards, includes improved thermal neutron scattering data and uses new evaluated data from the CIELO project for neutron reactions on 1H, 16O, 56Fe, 235U, 238U and 239Pu described in companion papers in the present issue of Nuclear Data Sheets. The evaluations benefit from recent experimental data obtained in the U.S. and Europe, and improvements in theory and simulation. Notable advances include updated evaluated data for light nuclei, structural materials, actinides, fission energy release, prompt fission neutron and γ-ray spectra, thermal neutron scattering data, and charged-particle reactions. Integral validation testing is shown for a wide range of criticality, reaction rate, and neutron transmission benchmarks. In general, integral validation performance of the library is improved relative to the previous ENDF/B-VII.1 library. Fil: Brown, D.A.. Brookhaven National Laboratory; Estados Unidos Fil: Chadwick, M.B.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Capote, R.. International Atomic Energy Agency, Vienna; Austria Fil: Kahler, A.C.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Trkov, A.. International Atomic Energy Agency, Vienna; Austria Fil: Herman, M.W.. Brookhaven National Laboratory; Estados Unidos Fil: Sonzogni, A.A.. Brookhaven National Laboratory; Estados Unidos Fil: Danon, Y.. Rensselaer Polytechnic Institute; Estados Unidos Fil: Carlson, A.D.. National Institute of Standards and Technology; Estados Unidos Fil: Dunn, M.. Spectra Tech; Estados Unidos Fil: Smith, D.L.. Argonne National Laboratory; Estados Unidos Fil: Hale, G.M.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Arbanas, G.. Oak Ridge National Laboratory; Estados Unidos Fil: Arcilla, R.. Brookhaven National Laboratory; Estados Unidos Fil: Bates, C.R.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Beck, B.. Lawrence Livermore National Laboratory; Fil: Becker, B.. Gesellschaft für Anlagen und Reaktorsicherheit; Alemania Fil: Brown, F.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Casperson, R.J.. Lawrence Livermore National Laboratory; Estados Unidos Fil: Conlin, J.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Cullen, D.E.. Lawrence Livermore National Laboratory; Estados Unidos Fil: Descalle, M.A.. Lawrence Livermore National Laboratory; Estados Unidos Fil: Firestone, R.. Lawrence Berkeley National Laboratory; Estados Unidos Fil: Gaines, T.. AWE plc; Reino Unido Fil: Guber, K.H.. Oak Ridge National Laboratory; Estados Unidos Fil: Hawari, A.I.. University of North Carolina; Estados Unidos Fil: Holmes, J.. Naval Nuclear Laboratory; Estados Unidos Fil: Johnson, T.D.. Brookhaven National Laboratory; Estados Unidos Fil: Kawano, T.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Kiedrowski, B.C.. University of Michigan; Estados Unidos Fil: Koning, A.J.. International Atomic Energy Agency, Vienna; Austria Fil: Kopecky, S.. EC-JRC; Bélgica Fil: Leal, L.. Institut de Radioprotection et de Sûreté Nucléaire; Francia Fil: Lestone, J.P.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Lubitz, C.. Naval Nuclear Laboratory; Estados Unidos Fil: Marquez Damian, Jose Ignacio. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro Científico Tecnológico Conicet - Patagonia Norte; Argentina. Comisión Nacional de Energía Atómica. Centro Atómico Bariloche; Argentina Fil: Mattoon, C.M.. Lawrence Livermore National Laboratory; Estados Unidos Fil: McCutchan, E.A.. Brookhaven National Laboratory; Estados Unidos Fil: Mughabghab, S.. Brookhaven National Laboratory; Estados Unidos Fil: Navratil, P.. TRIUMF; Canadá Fil: Neudecker, D.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Nobre, G.P.A.. Brookhaven National Laboratory; Estados Unidos Fil: Noguere, G.. Commissariat A Energie Atomique; Francia Fil: Paris, M.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Pigni, M.T.. Oak Ridge National Laboratory; Estados Unidos Fil: Plompen, A.J.. EC-JRC; Bélgica Fil: Pritychenko, B.. Brookhaven National Laboratory; Estados Unidos Fil: Pronyaev, V.G.. PI Atomstandart at SC Rosatom; Rusia Fil: Roubtsov, D.. Laboratoires Nucleaires Canadiens; Canadá Fil: Rochman, D.. Paul Scherrer Institut; Suiza Fil: Romano, P.. Argonne National Laboratory; Estados Unidos Fil: Schillebeeckx, P.. EC-JRC; Bélgica Fil: Simakov, S.. Karlsruhe Institute of Technology; Alemania Fil: Sin, M.. University of Bucharest; Rumania Fil: Sirakov, I.. Institute For Nuclear Research And Nuclear Energy Bulgarian Academy Of Sciences; Bulgaria Fil: Sleaford, B.. Lawrence Livermore National Laboratory; Estados Unidos Fil: Sobes, V.. Oak Ridge National Laboratory; Estados Unidos Fil: Soukhovitskii, E.S.. Joint Institute for Energy and Nuclear Research; Bielorrusia Fil: Stetcu, I.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Talou, P.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos. Oak Ridge National Laboratory; Estados Unidos Fil: Thompson, I.. Lawrence Livermore National Laboratory; Estados Unidos Fil: van der Marck, S.. NRG; Países Bajos Fil: Welser-Sherrill, L.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Wiarda, D.. Oak Ridge National Laboratory; Estados Unidos Fil: White, M.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Wormald, J.L.. University of North Carolina; Estados Unidos Fil: Wright, R.Q.. Oak Ridge National Laboratory; Estados Unidos Fil: Zerkle, M.. Naval Nuclear Laboratory; Estados Unidos Fil: Zerovnik, G.. EC-JRC; Bélgica Fil: Zhu, Yan. University of North Carolina; Estados Unidos

The CIELO collaboration has studied neutron cross sections on nuclides that significantly impact criticality in nuclear technologies - 235,238U, 239Pu, 56Fe, 16O and 1H - with the aim of improving the accuracy of the data and resolving previous discrepancies in our understanding. This multi-laboratory pilot project, coordinated via the OECD/NEA Working Party on Evaluation Cooperation (WPEC) Subgroup 40 with support also from the IAEA, has motivated experimental and theoretical work and led to suites of new evaluated libraries that accurately reflect measured data and also perform Fil: Chadwick, M.B.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Capote, R.. International Atomic Energy Agenc; Austria Fil: Trkov, A.. Brookhaven National Laboratory; Estados Unidos Fil: Herman, M.W.. Brookhaven National Laboratory; Estados Unidos Fil: Brown, D.A.. European Commission, Joint Research Center; Bélgica Fil: Hale, G.M.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Kahler, A.C.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Talou, P.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Plompen, A.J.. European Commission, Joint Research Center; Bélgica Fil: Schillebeeckx, P.. European Commission, Joint Research Center; Bélgica Fil: Pigni, M.T.. Oak Ridge National Laboratory; Estados Unidos Fil: Leal, L.. Institut de Radioprotection et de Surete Nucleaire; Francia Fil: Danon, Y.. Rensselaer Polytechnic Institute; Estados Unidos Fil: Carlson, A.D.. National Institute of Standards and Technology; Estados Unidos Fil: Romain, P.. Commissariat a L'Energie Atomique; Francia Fil: Morillon, B.. Commissariat a L'Energie Atomique; Francia Fil: Bauger, E.. Commissariat a L'Energie Atomique; Francia Fil: Hambsch, F.-J.. European Commission, Joint Research Center; Bélgica Fil: Kopecky, S.. European Commission, Joint Research Center; Bélgica Fil: Giorginis, G.. European Commission, Joint Research Center; Bélgica Fil: Kawano, T.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Lestone, J.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Neudecker, D.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Rising, M.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Paris, M.. Los Alamos National High Magnetic Field Laboratory; Estados Unidos Fil: Nobre, G.P.A.. Brookhaven National Laboratory; Estados Unidos Fil: Arcilla, R.. Brookhaven National Laboratory; Estados Unidos Fil: Cabellos, O.. Organisation for Economic Co-operation and Development. Nuclear Energy Agency; Francia Fil: Hill, I.. Organisation for Economic Co-operation and Development. Nuclear Energy Agency; Francia Fil: Dupont, E.. Organisation for Economic Co-operation and Development. Nuclear Energy Agency; Francia Fil: Koning, A.J.. International Atomic Energy Agency; Austria Fil: Cano Ott, D.. Centro de Investigaciones Energéticas Medioambientales y Tecnólogicas; España Fil: Mendoza, E.. Centro de Investigaciones Energéticas Medioambientales y Tecnólogicas; España Fil: Balibrea, J.. Centro de Investigaciones Energéticas Medioambientales y Tecnólogicas; España Fil: Paradela, C.. European Commission. Joint Research Center; Bélgica Fil: Durán, I.. Universidad de Santiago de Compostela; España Fil: Qian, J.. China Nuclear Data Center; China Fil: Ge, Z.. China Nuclear Data Center; China Fil: Liu, T.. China Nuclear Data Center; China Fil: Hanlin, L.. China Institute of Atomic Energy; China Fil: Ruan, X.. China Institute of Atomic Energy; China Fil: Haicheng, W.. China Institute of Atomic Energy; China Fil: Sin, M.. Bucharest University; Rumania Fil: Noguere, G.. Commissariat a L'Energie Atomique. Nuclear Energy Division; Francia Fil: Bernard, D.. Commissariat a L'Energie Atomique. Nuclear Energy Division; Francia Fil: Jacqmin, R.. Commissariat a L'Energie Atomique. Nuclear Energy Division; Francia Fil: Bouland, O.. Commissariat a L'Energie Atomique. Nuclear Energy Division; Francia Fil: De Saint Jean, C.. CEA, Nuclear Energy Division; Francia Fil: Pronyaev, V.G.. PI Atomstandart at SC Rosatom; Rusia Fil: Ignatyuk, A.V.. Institute of Physics and Power Engineering; Rusia Fil: Yokoyama, K.. Japan Atomic Energy Agency; Japón Fil: Ishikawa, M.. Japan Atomic Energy Agency; Japón Fil: Fukahori, T.. Japan Atomic Energy Agency; Japón Fil: Iwamoto, N.. Japan Atomic Energy Agency; Japón Fil: Iwamoto, O.. Japan Atomic Energy Agency; Japón Fil: Kunieda, S.. Japan Atomic Energy Agency; Japón Fil: Lubitz, C.R.. Knolls Atomic Power Laboratory; Estados Unidos Fil: Salvatores, M.. Idaho National Laboratory; Estados Unidos Fil: Palmiotti, G.. Idaho National Laboratory; Estados Unidos Fil: Kodeli, I.. Jozef Stefan Institute; Eslovenia Fil: Kiedrowski, B.. University of Michigan; Estados Unidos Fil: Roubtsov, D.. Canadian Nuclear Laboratories; Canadá Fil: Thompson, I.. Lawrence Livermore National Laboratory; Estados Unidos Fil: Quaglioni, S.. Lawrence Livermore National Laboratory; Estados Unidos Fil: Kim, H.I.. Korean Atomic Energy Research Institute; Corea del Sur Fil: Lee, Y.O.. Korean Atomic Energy Research Institute; Corea del Sur Fil: Fischer, U.. Karlsruhe Institute of Technology; Alemania Fil: Simakov, S.. Karlsruhe Institute of Technology; Alemania Fil: Dunn, M.. Oak Ridge National Laboratory; Estados Unidos Fil: Guber, K.. Oak Ridge National Laboratory; Estados Unidos Fil: Marquez Damian, Jose Ignacio. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro Científico Tecnológico Conicet - Patagonia Norte; Argentina. Comisión Nacional de Energía Atómica. Centro Atómico Bariloche; Argentina Fil: Cantargi, Florencia Olga. Comisión Nacional de Energía Atómica. Centro Atómico Bariloche; Argentina Fil: Sirakov, I.. Institute for Nuclear Research and Nuclear Energy; Bulgaria Fil: Otuka, N.. International Atomic Energy Agency; Austria Fil: Daskalakis, A.. Naval Nuclear Laboratory; Estados Unidos Fil: McDermott, B.J.. Naval Nuclear Laboratory; Estados Unidos Fil: van der Marck, S.C.. Nuclear Research and Consultancy Group; Países Bajos

International audience; Evaluations of nuclear reaction data for the major uranium isotopes 238 U and 235 U were performed within the scope of the CIELO Project on the initiative of the OECD/NEA Data Bank under Working Party on Evaluation Co-operation (WPEC) Subgroup 40 coordinated by the IAEA Nuclear Data Section. Both the mean values and covariances are evaluated from 10 −5 eV up to 30 MeV. The resonance parameters of 238 U and 235 U were re-evaluated with the addition of newly available data to the existing experimental database. The evaluations in the fast neutron range are based on nuclear model calculations with the code EMPIRE–3.2 Malta above the resonance range up to 30 MeV. 235 U( n ,f), 238 U( n ,f), and 238 U( n , γ ) cross sections and 235 U( n th ,f) prompt fission neutron spectrum (PFNS) were evaluated within the Neutron Standards project and are representative of the experimental state-of-the-art measurements. The Standards cross sections were matched in model calculations as closely as possible to guarantee a good predictive power for cross sections of competing neutron scattering channels. 235 U( n , γ ) cross section includes fluctuations observed in recent experiments. 235 U( n ,f) PFNS for incident neutron energies from 500 keV to 20 MeV were measured at Los Alamos Chi-Nu facility and re-evaluated using all available experimental data. While respecting the measured differential data, several compensating errors in previous evaluations were identified and removed so that the performance in integral benchmarks was restored or improved. Covariance matrices for 235 U and 238 U cross sections, angular distributions, spectra and neutron multiplicities were evaluated using the GANDR system that combines experimental data with model uncertainties. Unrecognized systematic uncertainties were considered in the uncertainty quantification for fission and capture cross sections above the thermal range, and for neutron multiplicities. Evaluated files were extensively benchmarked to ensure good performance in reactor calculations and fusion-related systems. New comprehensive evaluations show excellent agreement with available differential data and integral performance better than current evaluated data libraries, and represent a step forward in a quest for better nuclear data for applications.

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.

Cross sections for neutron interactions with 238U in the energy region from 5keV to 150keV have been evaluated. Average total and capture cross sections have been derived from a least squares analysis using experimental data reported in the literature. The resulting cross sections have been parameterised in terms of average resonance parameters maintaining full consistency with results of optical model calculations by using a dispersive coupled channel optical model potential. The average compound partial cross sections have been expressed in terms of transmission coefficients by applying the Hauser-Feshbach statistical reaction theory including width-fluctuations. A generalized single-level representation compatible with the energy-dependent options of the ENDF-6 format has been applied using standard boundary conditions. The results have been transferred into a full ENDF-6 compatible data file.

A consistent description of the neutron cross sections from thermal energy up to the MeV region is challenging. One of the first steps consists in optimizing the optical model parameters using average resonance parameters, such as the neutron strength functions. They can be derived from a statistical analysis of the resolved resonance parameters, or calculated with the generalized form of the SPRT method by using scattering matrix elements provided by optical model calculations. One of the difficulties is to establish the contributions of the direct and compound nucleus reactions. This problem was solved by using a slightly modified average R-Matrix formula with an equivalent hard sphere radius deduced from the phase shift originating from the potential. The performances of the proposed formalism are illustrated with results obtained for the 238 U+n nuclear systems.

The role of rotational energy is investigated in reference to the dynamics of 16 O+198 Pt →214 Rn∗ reaction using the sticking (I S ) and the non-sticking (I NS ) limits of moment of inertia within the framework of dynamical cluster decay model. The decay barrier height and barrier position get significantly modified for the use of sticking or non-sticking choice, which in turn reproduce the evaporation residue and the fusion-fission cross-sections nicely by the I S approach, while the I NS approach provides feasible addressal of data only for evaporation residue channel. Moreover, the fragmentation path of decaying fragments of 214 Rn∗ compound nucleus gets influenced for different choices of moment of inertia. Beside this, the role of nuclear deformations i.e. static, dynamic quadurpole (β2 ) and higher order static deformation up to β4 are duly investigated for both choices of the moment of inertia.

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 Joint Research Center Site Geel, European Union. Neutron capture and transmission measurements were carried out using metallic natural cerium and vanadium samples. Together with existing data, the measured data will be used for a new evaluation and will be submitted with covariances to the ENDF/B nuclear data library.

This work reports on accurate, high-resolution measurements of the $^{25}$Mg($n, \gamma$)$^{26}$Mg and $^{25}$Mg($n, tot$) cross sections in the neutron energy range from thermal to about 300 keV, leading to a significantly improved $^{25}$Mg($n, \gamma$)$^{26}$Mg parametrization. The relevant resonances for $n+^{25}$Mg were characterized from a combined R-matrix analysis of the experimental data. This resulted in an unambiguous spin/parity assignment of the corresponding excited states in $^{26}$Mg. With this information experimental upper limits of the reaction rates for $^{22}$Ne($\alpha, n$)$^{25}$Mg and $^{22}$Ne($\alpha, \gamma$)$^{26}$Mg were established, potentially leading to a significantly higher ($\alpha, n$)/($\alpha, \gamma$) ratio than previously evaluated. The impact of these results have been studied for stellar models in the mass range 2 to 25 $M_{\odot}$., Comment: Preprint, 6 pages, 5 figures

International audience; 233U plays the essential role of fissile nucleus in the Th-U fuel cycle. A particularity of 233U is its small neutron capture cross-section which is about one order of magnitude lower than the fission cross-section on average. Therefore, the accuracy in the measurement of the 233U capture cross-section essentially relies on efficient capture-fission discrimination thus a combined setup of fission and γ-detectors is needed. At CERN n_TOF the Total Absorption Calorimeter (TAC) coupled with compact fission detectors is used. Previously used MicroMegas (MGAS) detectors showed significant γ-background issues above 100 eV coming from the copper mesh. A new measurement campaign of the 233U capture cross-section and alpha ratio is planned at the CERN n_TOF facility. For this measurement, a novel cylindrical multi ionization cell chamber was developed in order to provide a compact solution for 14 active targets read out by 8 anodes. Due to the high specific activity of 233U fast timing properties are required and achieved with the use of customized electronics and the very fast ionizing gas CF4 together with a high electric field strength. This paper describes the new fission chamber and the results of the first tests with neutrons at GELINA proving that it is suitable for the 233U measurement.

The CIELO collaboration has studied neutron cross sections on nuclides that significantly impact criticality in nuclear technologies – 16 O, 56 Fe, 235,8 U and 239 Pu – with the aim of improving the accuracy of the data and resolving previous discrepancies in our understanding. This multi-laboratory pilot project, coordinated via the OECD/NEA Working Party on Evaluation Cooperation (WPEC) Subgroup 40 with support also from the IAEA, has motivated experimental and theoretical work and led to suites of new evaluated libraries that accurately reflect measured data and also perform well in integral simulations of criticality.

Neutron resonance densitometry (NRD) is a non-destructive analysis method, which can be applied to quantify special nuclear materials (SNM) in small particle-like debris of melted fuel that are formed in severe accidents of nuclear reactors such as the Fukushima Daiichi nuclear power plants. NRD uses neutron resonance transmission analysis (NRTA) to quantify SNM and neutron resonance capture analysis (NRCA) to identify matrix materials and impurities. To apply NRD for the characterization of arbitrary shaped thick materials, a generalized method for the analysis of NRTA data has been developed. The method has been applied on data resulting from transmission through thick samples with an irregular shape and an areal density of SNM up to 0.253 at/b (≈100 g/cm2). The investigation shows that NRD can be used to quantify SNM with a high accuracy not only in inhomogeneous samples made of particle-like debris but also in samples made of large rocks with an irregular shape by applying the generalized analysis method for NRTA., JRC.D.4-Standards for Nuclear Safety, Security and Safeguards

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 thermal neutron-elastic-scattering cross-section data for {sup 16}O used in various modern evaluated-nuclear-data libraries were reviewed and found to be generally too high compared with the best available experimental measurements. Some of the proposed revisions to the ENDF/B-VII.0 {sup 16}O data library and recent results from the TENDL system increase this discrepancy further. The reactivity impact of revising the {sup 16}O data downward to be consistent with the best measurements was tested using the JENDL-3.3 {sup 16}O cross-section values and was found to be very small in MCNP5 simulations of the UO{sub 2} and reactor-recycle MOX-fuel cases of the ANS Doppler-defect numerical benchmark. However, large reactivity differences of up to about 14 mk (1400 pcm) were observed using {sup 16}O data files from several evaluated-nuclear-data libraries in MCNP5 simulations of the Los Alamos National Laboratory HEU heavy-water solution thermal critical experiments, which were performed in the 1950's. The latter result suggests that new measurements using HEU in a heavy-water-moderated critical facility, such as the ZED-2 zero-power reactor at the Chalk River Laboratories, might help to resolve the discrepancy between the {sup 16}O thermal elastic-scattering cross-section values and thereby reduce or better define its uncertainty, although additional assessment work would bemore » needed to confirm this. (authors)« less

Measurements were performed at the time-of-flight facility GELINA to determine the 238U(n,\( \gamma\)) cross section in the resonance region. Experiments were carried out at a 12.5 and 60m measurement station. The total energy detection principle in combination with the pulse height weighting technique was applied using C6D6 liquid scintillators as prompt \( \gamma\)-ray detectors. The energy dependence of the neutron flux was measured with ionisation chambers based on the 10B(n,\( \alpha\)) reaction. The data were normalised to the isolated and saturated 238U resonance at 6.67 eV. Special procedures were applied to reduce bias effects due to the weighting function, normalization, dead time and background corrections, and corrections related to the sample properties. The total uncertainty due to the weighting function, normalization, neutron flux and sample characteristics is about 1.5%. Resonance parameters were derived from a simultaneous resonance shape analysis of the GELINA capture data and transmission data obtained previously at a 42m and 150m station of ORELA. The parameters of resonances below 500 eV are in good agreement with those resulting from an evaluation that was adopted in the main data libraries. Between 500 eV and 1200 eV a systematic difference in the neutron width is observed. Average capture cross section data were derived from the experimental capture yield in the energy region between 3.5 keV and 90 keV. The results are in good agreement with an evaluated cross section resulting from a least squares fit to experimental data available in the literature prior to this work. The average cross section data derived in this work were parameterised in terms of average resonance parameters and included in a least squares analysis together with other experimental data reported in the literature.

International audience; A resolved resonance evaluation was performed for $^{16}$O in the energy range 0 eV to 6 MeV using the computer code SAMMY resulting in a set of resonance parameters (RPs) that describes well the experimental data used in the evaluation. A RP covariance matrix (RPC) was also generated. The RP were converted to the evaluated nuclear data file format using the R-Matrix Limited format and the compact format was used to represent the RPC. In contrast to the customary use of RP, which are frequently intended for the generation of total, capture, and scattering cross sections only, the present RP evaluation permits the computation of angle dependent cross sections. Furthermore, the RPs are capable of representing the ($n, \alpha$) cross section from the energy threshold (2.354 MeV) of the ($n, \alpha$) reaction to 6 MeV. The intent of this paper is to describe the procedures used in the evaluation of the RP and RPC, the use of the RPC in benchmark calculations and to assess the impact of the $^{16}$O nuclear data uncertainties in the calculate $dk_{eff}$ for critical benchmark experiments.

An experimental program, called AMSTRAMGRAM, was recently conducted in the Minerve low power reactor operated by CEA Cadarache within the frame of the CHANDA initiative (Solving CHAllenges in Nuclear Data). Its aim was to measure the integral capture cross section of 241 Am in the thermal domain. Motivation of this work is driven by large differences in this actinide thermal point reported by major nuclear data libraries. The AMSTRAMGRAM experiment, that made use of well characterized EC-JRC americium samples, was based on the oscillation technique commonly implemented in the Minerve reactor. First results are presented and discussed in this article. A preliminary calculation scheme was used to compare measured and calculated results. It is shown that this work confirms a bias previously observed with JEFF-3.1.1 (C/E-1 = −10.5 ± 2%). On the opposite, the experiment is in close agreement with 241 Am thermal point reported in JEFF-3.2 (C/E-1 = 0.5 ± 2%).

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.

A new set of measurements for the total and capture cross section determination of W isotopes was done using GELINA (GEel LINear Accelerator), a neutron Time-Of-Flight facility at the Institute for Reference Materials and Measurements (IRMM). Measuring stations at different flight path lengths were used in order to cover a broad neutron energy range with high resolution demands. Experimental techniques adopted for both transmission and capture measurements are well established using a (6)Li glass detector and C(6)D(6) scintillation arrays as detections systems respectively., As target samples highly enriched (182,183,184,186)W metallic discs were used.

The international High Priority Request list notes: “In the nuclear industry hafnium is used as neutron absorbing material to regulate the fission process. Interpretation of critical experiments with UOx fuel conducted by CEA in the AZUR zero-power reactors has shown systematic underestimation of the reactivity worth that may be attributed to an overestimated natural hafnium capture cross section in the epi-thermal energy range” To service the request for improved resonance data a PhD project has:a) Improved REFIT R-matrix evaluation code. b) Obtained hafnium oxide samples enriched in Hf176, 177, 178, 179 isotopes. c) Gained support from NUDAME and EUFRAT projects. d) Prepared enriched and natural Hf samples. e) Performed capture and transmission Time of Flight measurements at the GELINA linear accelerator. f) Analysed the capture counts to generate yields using AGS and AGL codes. g) Used REFIT to perform least squares analysis of GELINA measurements. (Included previous ORNL, Harwell and RPI transmissions and capture yields.) h) Submitted results to EXFOR. i) Included resolved resonance parameters in JEFF evaluations taking the resolved range to over 1 keV. j) Tested evaluations with suitable benchmarks. k) Passed resolved resonance data to CEA Cadarache for unresolved analysis. Resultant Hf evaluations will be included in JEFF3.2.

Inelastic scattering cross sections for sodium are of interest to the development of sodium cooled fast reactors. A recent OECD-NEA subgroup analysed the sensitivity of reactor parameters to cross sections and accordingly determined target uncertainties for the nuclear data. Comparing these target uncertainties with the current status of nuclear data uncertainties and covariance data resulted in a list of target priorities. Among these features sodium inelastic scattering for which a target uncertainty of about 5% was established, approximately two to three times as good as the uncertainty for current evaluated data files for this isotope (see OECD-NEA High Priority Request List, http://www.nea.fr/html/dbdata/hprl/hprlview.pl?ID=448). At IRMM, the GAINS gammaarray for inelastic neutron scattering was developed with the purpose of measuring cross sections with uncertainties at or below the target uncertainties for nuclides like Na using the (n, n′γ)technique. Measurements were performed at the GELINA facility at a 200 m flight path with eight high purity germanium detectors. The sample was an 80 mm diameter metallic disk prepared at IRMM by cutting, pressing, rolling and punching. For the experimental work, a careful review was made of the gamma-efficiency calibrations and the flux normalization in order to investigate in detail the corrections and the final uncertainties that may realistically be achieved. An elaborate account will be presented of the data analysis and checks that have been made and implications for earlier work by our group will be discussed. First results will be shown for the gamma-production cross sections of the main transitions in the energy range from threshold to 10 MeV.

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.

Cross sections for neutron-induced reactions of $^{241}\text{Am}$ in the resonance region have been evaluated. Results of time-of-flight cross section experiments carried out at the GELINA, LANSCE, ORELA and Saclay facilities have been combined with optical model calculations to derive consistent cross sections from the thermal energy region up to the continuum region. Resolved resonance parameters were derived from a resonance shape analysis of transmissions, capture yields, and fission yields in the energy region up to 150 eV using the refit code. From a statistical analysis of these parameters, a neutron strength function (${10}^{4}{S}_{0}=1.01\ifmmode\pm\else\textpm\fi{}0.12$), mean level spacing (${D}_{0}=0.60\ifmmode\pm\else\textpm\fi{}0.01$ eV) and average radiation width ($\ensuremath{\langle}{\mathrm{\ensuremath{\Gamma}}}_{{\ensuremath{\gamma}}_{0}}\ensuremath{\rangle}=43.3\ifmmode\pm\else\textpm\fi{}1.1$ meV) for $s$-wave resonances were obtained. Neutron strength functions for higher partial waves ($lg0$) together with channel and effective scattering radii were deduced from calculations based on a complex mean-field optical model potential, applying an equivalent hard-sphere scattering radius approximation.