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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

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.

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.

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

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.

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.

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.

Neutron activation analysis is the reference method used for offline determination of the neutron flux density in defined positions. It can be used in the nuclear energy industry-as well as in medical- or space applications. For accurate neutron flux evaluation, well-known and reliable cross sections are needed. In the thermal and fast energy region, many reliable monitoring reactions exists, however, in case of the epithermal and intermediate energy region, there are practically no dosimetry nuclear reactions sensitive specifically in this energy range. Due to this fact, both new data are being measured and methodologies are under development to describe and test this energy region. It was found that various neutron filters can be used to cut parts of neutron spectra and thus methodology based on spectrum filtering could potentially be employed to survey cross sections of interest. It this paper, the use of 3 different filters – B4C, Cd, and In is studied, on the case of the 55Mn(n,γ) reaction. Measured values of that cross section in the given filtered reference spectra are reported.

Constant improvements of the computational power and methods as well as demands of accurate and reliable measurements for reactor operation and safety require a continuous upgrade of the instrumentation. In particular, nuclear sensors used in nuclear fission reactors (research or power reactors) or in nuclear fusion facilities are operated under intense mixed neutron and gamma-ray fields, and need to be calibrated and modeled to provide selective and accurate neutron and gamma-ray measurements. The French Atomic Energy and Alternative Energies Commission (CEA) and the Jožef Stefan Institute (JSI) have started an experimental program dedicated to a detailed experimental benchmark with analysis using Monte Carlo particle transport calculations and a series of neutron and gamma-ray sensor types used in the JSI TRIGA Mark II reactor. CEA has setup a simplified TRIPOLI-4® modeling scheme of the JSI TRIGA reactor based on the information available in the IRPhEP benchmark in order to facilitate analysis of future neutron and gamma-ray measurements. These allow the CEA to perform a TRIPOLI-4 instrumentation calculation scheme benchmarked with the JSI MCNP model. This paper presents the main results of this CEA calculation scheme application and the analysis of their comparison to the JSI results obtained in 2012 with the MCNP5 & ENDF/B-VII.0 calculation scheme. This paper will conclude with some information about the new experimental program to be carried out in 2022 in the TRIGA reactor core.

The cross section is a fundamental quantity which affects the accuracy of Monte Carlo simulations widely used in nuclear applications. A new dosimetry library IRDFF-II that contains cross section evaluations that include full uncertainty quantification is being developed by the International Atomic Energy Agency and expected to be released in January 2020; a preliminary version IRDFF-1.05 was released in 2014 and is being tested in this work. Validation of the cross-section evaluations proposed for this library is a high priority task. The validation can be realized using integral cross sections measured in standard and/or reference neutron benchmark fields. Integral quantities feature significantly lower uncertainties than differential nuclear data. If the neutron spectrum where the cross section is measured is well characterized, then the Spectrum Averaged Cross Section can be used for validating of existing evaluations.

A recent experimental irradiation and measurement campaign using containers made from boron nitride (BN) at the Jožef Stefan Institute (JSI) TRIGA Mark II reactor in Ljubljana, Slovenia, has shown the applicability of BN for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements. The first part of the paper focuses on the determination of the transmission function of a BN container through Monte Carlo calculations and experimental measurements. The second part presents the process of tayloring the sensitivity of integral activation measurements to specific needs and a selection of suitable radiative capture reactions for neutron spectrum characterization in the epithermal range. A BN container used in our experiments and its qualitative effect on the neutron spectrum in the irradiation position employed is displayed in the Graphical abstract.

The energy spectrum of prompt neutrons emitted in fission (PFNS) plays a very important role in nuclear science and technology. A Coordinated Research Project (CRP) “Evaluation of Prompt Fission Neutron Spectra of Actinides”was established by the IAEA Nuclear Data Section in 2009, with the major goal to produce new PFNS evaluations with uncertainties for actinide nuclei. The following technical areas were addressed: (i) experiments and uncertainty quantification (UQ): New data for neutron-induced fission of 233U, 235U, 238U, and 239Pu have been measured, and older data have been compiled and reassessed. There is evidence from the experimental work of this CRP that a very small percentage of neutrons emitted in fission are actually scission neutrons; (ii) modeling: The Los Alamos model (LAM) continues to be the workhorse for PFNS evaluations. Monte Carlo models have been developed that describe the fission phenomena microscopically, but further development is needed to produce PFNS evaluations meeting the uncertainty targets; (iii) evaluation methodologies: PFNS evaluations rely on the use of the least-squares techniques for merging experimental and model data. Considerable insight was achieved on how to deal with the problem of too small uncertainties in PFNS evaluations. The importance of considering that all experimental PFNS datamore » are “shape” data was stressed; (iv) PFNS evaluations: New evaluations, including covariance data, were generated for major actinides including 1) non-model GMA evaluations of the 235U(nth,f), 239Pu(nth,f), and 233U(nth,f) PFNS based exclusively on experimental data (0.02 ≤ E ≤ 10 MeV), which resulted in PFNS average energies E of 2.00±0.01, 2.073±0.010, and 2.030±0.013 MeV, respectively; 2) LAM evaluations of neutron-induced fission spectra on uranium and plutonium targets with improved UQ for incident energies from thermal up to 30 MeV; and 3) Point-by-Point calculations for 232Th, 234U and 237Np targets; and (v) data testing: Spectrum averaged cross sections (SACS) calculated for the evaluated 235U(nth,f) PFN field agree within uncertainties with evaluated SACS experimental data. Despite the observed reduction of the PFNS E by about 30 keV for neutron-induced fission of 233U, 235U, and 239Pu, the criticality benchmark outcomes suggested that new evaluations can achieve the same (or better) integral performance with respect to existing evaluations, but the strong compensating effects observed need to be addressed. Summarizing, this project has significantly improved PFNS evaluations and evaluation methodology, provided new PFNS data for applications, and also highlighted the areas for future research« less

A new reactor neutron reference field is proposed, which can be used for measurements of spectrum averaged cross sections (SACS). The reference field is identified in the center of a dry irradiation channel surrounded by six 3.3% enriched fuel assemblies assembled in the zero-power LR-0 reactor with negligible burn-up. The neutron field can be defined as a reference one, because of the well-defined neutron spectrum and neutron flux, well-characterized criticality and core power distribution, and reproducibility of results. Extensive experimental validation was undertaken. A new reference field can be used for the measurement of reaction rates and SACS used in nuclear data validation. It is shown that above 6 MeV the spectrum is undistinguishable from the thermal-neutron induced 235U prompt fission neutron spectrum (PFNS). Thus, when a reaction has threshold over this limit, the SACS in the thermal-neutron induced 235U fission neutron field can be derived from reaction rates obtained in this reference neutron field. A comprehensive set of SACS measurements with uncertainties below 5% in the reference field is presented.

An international effort has produced evaluations of the neutron data standards. Evaluations were obtained for the cross section standards: the H(n,n), 6Li(n,t), 10B(n,067), loB(vx), natc(n,n,) Au(n ...

With the need for improving existing nuclear data evaluations, (e.g., ENDF/B-VIII.0 and JEFF-3.3 releases) the first step was to evaluate the standards for use in such a library. This new standards evaluation made use of improved experimental data and some developments in the methodology of analysis and evaluation. In addition to the work on the traditional standards, this work produced the extension of some energy ranges and includes new reactions that are called reference cross sections. Since the effort extends beyond the traditional standards, it is called the neutron data standards evaluation. This international effort has produced new evaluations of the following cross section standards: the H(n,n), Li-6(n,t), B-10(n, alpha), B-10(n,alpha(1)gamma), C-nat(n,n), Au(n,gamma), U-235(n,f) and U-238(n,f). Also in the evaluation process the U-238(n,gamma) and Pu-239(n,f) cross sections that are not standards were evaluated. Evaluations were also obtained for data that are not traditional standards: the Maxwellian spectrum averaged cross section for the Au(n,gamma) cross section at 30 keV; reference cross sections for prompt gamma-ray production in fast neutron-induced reactions; reference cross sections for very high energy fission cross sections; the Cf-252 spontaneous fission neutron spectrum and the U-235 prompt fission neutron spectrum induced by thermal incident neutrons; and the thermal neutron constants. The data and covariance matrices of the uncertainties were obtained directly from the evaluation procedure. Correction in: Nuclear Data Sheets, Volume 163, January 2020, Pages 280-281, DOI:10.1016/j.nds.2019.12.008

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; A new suite of evaluations for 54,56,57,58 Fe has been developed in the framework of the CIELO international collaboration. New resolved resonance ranges were evaluated for 54 Fe and 57 Fe, while modifications were applied to resonances in 56 Fe. The low energy part of the 56 Fe file is almost totally based on measurements. At higher energies in 56 Fe and in the whole fast neutron range for minor isotopes the evaluation consists of model predictions carefully adjusted to available experimental data. We also make use of the high quality and well experimentally-constrained dosimetry evaluations from the IRDFF library. Special attention was dedicated to the elastic angular distributions, which were found to affect results of the integral benchmarking. The new set of iron evaluations was developed in concert with other CIELO evaluations and they were tested together in the integral experiments before being adopted for the ENDF/B-VIII.0 library.

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.

Recently a series of test irradiations was performed at the JSI TRIGA Mark II reactor for the Fission Track-Thermoionization Mass Spectrometry (FT-TIMS) method, which requires a well thermalized neutron spectrum for sample irradiation. For this purpose the Monte Carlo N-Particle Transport Code (MCNP5) was used to computationally support the design of an irradiation device inside the TRIGA model and to support the actual measurements by calculating the neutron fluxes inside the major ex-core irradiation facilities. The irradiation device, filled with heavy water, was designed and optimized inside the Thermal Column and the additional moderation was placed inside the Elevated Piercing Port. The use of the device improves the ratio of thermal neutron flux to the sum of epithermal and fast neutron flux inside the Thermal Column Port by 390% and achieves the desired thermal neutron fluence of 1015 neutrons/cm2 in irradiation time of 20 h.

Issues in evaluation methodology of the prompt fission neutron spectra (PFNS) and neutron multiplicity for the thermal-neutron-induced fission of the {sup 235}U are discussed. The inconsistency between the experimental differential and integral data is addressed. By using differential data as ”shape data” good consistency was achieved between available sets of differential data. Integral dosimetry data have been used to define the PFNS slope at high outgoing neutron energies, where the quality of the differential data is poor. The inclusion into the fit of measured integral (spectrum-averaged) cross sections had a very small impact in the region where differential PFNS data are abundant and accurate, but removed the discrepancy with integral data at higher neutron emission energies. All experimental data are consistently fitted giving a PFNS average energy of 2.008 MeV. The impact on criticality prediction of the newly evaluated PFNS was tested. The highly enriched {sup 235}U solution assemblies with high leakage HEU-SOL-THERM-001 and HEU-SOL-THERM-009 benchmarks are the most sensitive to the PFNS. Criticality calculations for those solutions show a significant increase in reactivity if the average neutron energy of the fission neutrons is reduced from the ENDF/B-VI.5 value of 2.03 MeV. The proposed reduction of the PFNS average energymore » by 1.1% can be compensated by reducing the average number of neutrons per fission ν{sup ¯} at the thermal energy to the Gwin et al. measured value. The simple least-squares PFNS fit was confirmed by a more sophisticated combined fit of differential PFNS data for {sup 233,235}U, {sup 239}Pu and {sup 252}Cf nuclides with the generalised least-squares method using the GMA and GANDR codes.« less

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.

Evaluations are being done 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 cross sections. Evaluations are also being done for data that are not traditional standards including: the Au(n,γ) cross section at energies below where it is considered a standard; reference cross sections for prompt gamma-ray production in fast neutron-induced reactions; reference cross sections for very high energy fission cross sections; the 235U thermal neutron fission spectrum and the 252Cf spontaneous fission neutron spectrum and the thermal constants.

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.

In August 2016, following the recent effort within the Collaborative International Evaluated Library Organization (CIELO) pilot project to improve the neutron cross sections of 235 U, Oak Ridge National Laboratory (ORNL) collaborated with the International Atomic Energy Agency (IAEA) to release a resonance parameter evaluation. This evaluation restores the performance of the evaluated cross sections for the thermal- and above-thermal-solution benchmarks on the basis of newly evaluated thermal neutron constants (TNCs) and thermal prompt fission neutron spectra (PFNS). Performed with support from the US Nuclear Criticality Safety Program (NCSP) in an effort to provide the highest fidelity general purpose nuclear database for nuclear criticality applications, the resonance parameter evaluation was submitted as an ENDF-compatible file to be part of the next release of the ENDF/B-VIII.0 nuclear data library. The resonance parameter evaluation methodology used the Reich-Moore approximation of the R -matrix formalism implemented in the code SAMMY to fit the available time-of-flight (TOF) measured data for the thermal induced cross section of n+235 U up to 100 eV. While maintaining reasonably good agreement with the experimental data, the validation analysis focused on restoring the benchmark performance for 235 U solutions by combining changes to the resonance parameters and to the prompt resonance v below 100 eV.

An IAEA project to update the Neutron Standards is near completion. Traditionally, the Thermal Neutron Constants (TNC) evaluated data by Axton for thermal-neutron scattering, capture and fission on four fissile nuclei and the total nu-bar of 252 Cf(sf) are used as input in the combined least-square fit with neutron cross section standards. The evaluation by Axton (1986) was based on a least-square fit of both thermal-spectrum averaged cross sections (Maxwellian data) and microscopic cross sections at 2200 m/s. There is a second Axton evaluation based exclusively on measured microscopic cross sections at 2200 m/s (excluding Maxwellian data). Both evaluations disagree within quoted uncertainties for fission and capture cross sections and total multiplicities of uranium isotopes. There are two factors, which may lead to such difference: Westcott g-factors with estimated 0.2% uncertainties used in the Axton's fit, and deviation of the thermal spectra from Maxwellian shape. To exclude or mitigate the impact of these factors, a new combined GMA fit of standards was undertaken with Axton's TNC evaluation based on 2200 m/s data used as a prior. New microscopic data at the thermal point, available since 1986, were added to the combined fit. Additionally, an independent evaluation of TNC was undertaken using CONRAD code. Both GMA and CONRAD results are consistent within quoted uncertainties. New evaluation shows a small increase of fission and capture thermal cross sections, and a corresponding decrease in evaluated thermal nubar for uranium isotopes and 239 Pu.

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 latest release of the EMPIRE-3.1 system (codename Rivoli) is being used in the advanced modeling of neutron induced reactions on the 238 U nucleus with the aim of improving our knowledge of neutron scattering. The reaction model includes: (i) a new rotational-vibrational dispersive optical model potential coupling the low-lying collective bands of vibrational character observed in even-even actinides, (ii) the Engelbrecht-Weidenmuller transformation allowing for inclusion of compound-direct interference effects enhanced by a dispersive treatment of the optical model potential, (iii) a multi-humped fission barrier with absorption in the secondary well as described within the optical model for fission, and (iv) a modified Lorentzian model (MLO) of the radiative strength function. Impact of the advanced modeling on elastic and inelastic scattering cross section is being assessed by both comparison with selected microscopic experimental data and integral criticality benchmarks (e.g. FLATTOP, JEMIMA and BIGTEN assemblies). Benchmark calculations provide feedback to improve the reaction modeling and reduce both model and model-parameters uncertainties. Additionally, neutron scattering yields on 238 U measured accurately at RPI by the time-of-flight technique at 29, 60, 112 and 153 degrees have been used as a further constraint on the incident energy dependence of elastic and inelastically scattered neutrons. Improvement of scattering cross sections in existing libraries is discussed.

A comprehensive formalism to calculate fission cross sections based on the extension of the optical model for fission is presented. It can be used for description of nuclear reactions on actinides featuring multi-humped fission barriers with partial absorption in the wells and direct transmission through discrete and continuum fission channels. The formalism describes the gross fluctuations observed in the fission probability due to vibrational resonances, and can be easily implemented in existing statistical reaction model codes. The extended optical model for fission is applied for neutron induced fission cross-section calculations on $^{234,235,238}\mathrm{U}$ and $^{239}\mathrm{Pu}$ targets. A triple-humped fission barrier is used for $^{234,235}\mathrm{U}(n,f)$, while a double-humped fission barrier is used for $^{238}\mathrm{U}(n,f)$ and $^{239}\mathrm{Pu}(n,f)$ reactions as predicted by theoretical barrier calculations. The impact of partial damping of class-II/III states, and of direct transmission through discrete and continuum fission channels, is shown to be critical for a proper description of the measured fission cross sections for $^{234,235,238}\mathrm{U}(n,f)$ reactions. The $^{239}\mathrm{Pu}(n,f)$ reaction can be calculated in the complete damping approximation. Calculated cross sections for $^{235,238}\mathrm{U}(n,f)$ and $^{239}\mathrm{Pu}(n,f)$ reactions agree within 3% with the corresponding cross sections derived within the Neutron Standards least-squares fit of available experimental data. The extended optical model for fission can be used for both theoretical fission studies and nuclear data evaluation.

The results of validation of the latest release of International Reactor Dosimetry and Fusion File, IRDFF-1.03, in the standard 252 Cf(s.f.) and reference 235 U(nth ,f) neutron benchmark fields are presented. The spectrum-averaged cross sections were shown to confirm IRDFF-1.03 in the 252 Cf standard spontaneous fission spectrum; that was not the case for the current recommended spectra for 235 U(nth ,f). IRDFF was also validated in the spectra of the research reactor facilities ISNF, Sigma-Sigma and YAYOI, which are available in the IRDF-2002 collection. The ISNF facility was re-simulated to remove unphysical oscillations in the spectrum. IRDFF-1.03 was shown to reproduce reasonably well the spectrum-averaged data measured in these fields except for the case of YAYOI.

After the coated CFC wall to ITER-Like Wall (Beryllium/Tungsten/Carbon) transition in 2010–11, confirmation of the neutron yield calibration will be ensured by direct measurements using a calibrated 252 Cf neutron source deployed by the in-vessel remote handling boom and Mascot manipulator inside the JET vacuum vessel. The paper describes preliminary calculations and the results of numerical study of the effect of source holder on neutron detector response. The source baton was designed in such a way, that it does not significantly affect the neutron spectrum, angular neutron flux distribution or activation detector response. All effects are approximately equal to or less than 1%. The largest disturbance to the neutron flux angular distribution and to the neutron spectrum arises from the source capsule. Hence one should obtain as much information as possible about the capsule and the 252 Cf source material in order to avoid additional systematic errors.

Activation measurements can be used for neutron spectrum determination in irradiation facilities. It is very important that self-shielding is taken correctly into account. Activation measurements of monitor materials were performed at the CEA EOLE facility. Self-shielding factors for the monitor materials were calculated at the Commissariat a l’Energie Atomique with the TRIPOLI 4.6 code using a detailed whole-core model, and at the Jožef Stefan Institute with the MCNP code using a simplified model. The results were compared to assess the relative merits and the achievable accuracy of the calculations. In the end, the results were used to determine the neutron spectrum of the studied location in the EOLE irradiation facility.

The ENDF/B-VII.1 library is the latest revision to the United Statesʼ Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 423 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [M. B. Chadwick et al. , “ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data,” Nuclear Data Sheets , 112 , 2887 (2011)]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Projectʼs International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unmoderated and uranium reflected 235 U and 239 Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also confirmed for selected actinide reaction rates such as 236 U, 238,242 Pu and 241,243 Am capture in fast systems. Other deficiencies, such as the overprediction of Pu solution system critical eigenvalues and a decreasing trend in calculated eigenvalue for 233 U fueled systems as a function of Above-Thermal Fission Fraction remain. The comprehensive nature of this critical benchmark suite and the generally accurate calculated eigenvalues obtained with ENDF/B-VII.1 neutron cross sections support the conclusion that this is the most accurate general purpose ENDF/B cross section library yet released to the technical community.

The EMPIRE code system is a versatile package for nuclear model calculations that is often used for nuclear data evaluation. Its capabilities include random sampling of model parameters, which can be utilised to generate a full covariance matrix of all scattering cross sections, including cross-reaction correlations. The EMPIRE system was used to prepare the prior covariance matrices of reaction cross sections of 232 Th, 180,182,183,184,186 W and 55 Mn nuclei for incident neutron energies up to 60 MeV. The obtained modelling prior was fed to the GANDR system, which is a package for a global assessment of nuclear data, based on the Generalised Least-Squares method. By introducing experimental data from the EXFOR database into GANDR, the constrained covariance matrices and cross section adjustment functions were obtained. Applying the correction functions on the cross sections and formatting the covariance matrices, the final evaluations in ENDF-6 format including covariances were derived. In the resonance energy range, separate analyses were performed to determine the resonance parameters with their respective covariances. The data files thus obtained were then subjected to detailed testing and validation. Described evaluations with covariances of 232 Th, 180,182,183,184,186 W and 55 Mn nuclei are included into the ENDF/B-VII.1 library release.

The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He, Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl, K, Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides 235,238U and 239Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es, Fm, and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on 239Pu; and (9) A new decay data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide range of MCNP simulations of criticality benchmarks, with improved performance coming from new structural material evaluations, especially for Ti, Mn, Cr, Zr and W. For Be we see some improvements although the fast assembly data appear to be mutually inconsistent. Actinide cross section updates are also assessed through comparisons of fission and capture reaction rate measurements in critical assemblies and fast reactors, and improvements are evident. Maxwellian-averaged capture cross sections at 30 keV are also provided for astrophysics applications. We describe the cross section evaluations that have been updated for ENDF/B-VII.1 and the measured data and calculations that motivated the changes, and therefore this paper augments the ENDF/B-VII.0 publication [M. B. Chadwick, P. Obložinský, M. Herman, N. M. Greene, R. D. McKnight, D. L. Smith, P. G. Young, R. E. MacFarlane, G. M. Hale, S. C. Frankle, A. C. Kahler, T. Kawano, R. C. Little, D. G. Madland, P. Moller, R. D. Mosteller, P. R. Page, P. Talou, H. Trellue, M. C. White, W. B. Wilson, R. Arcilla, C. L. Dunford, S. F. Mughabghab, B. Pritychenko, D. Rochman, A. A. Sonzogni, C. R. Lubitz, T. H. Trumbull, J. P. Weinman, D. A. Br, D. E. Cullen, D. P. Heinrichs, D. P. McNabb, H. Derrien, M. E. Dunn, N. M. Larson, L. C. Leal, A. D. Carlson, R. C. Block, J. B. Briggs, E. T. Cheng, H. C. Huria, M. L. Zerkle, K. S. Kozier, A. Courcelle, V. Pronyaev, and S. C. van der Marck, “ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology,” Nuclear Data Sheets 107, 2931 (2006)].

The mock-up of the EU Test Blanket Module (TBM) based on the Helium-Cooled Lithium-Lead (HCLL) concept was irradiated at Frascatti, Italy, in order to assess the uncertainty in the tritium production rate (TPR) due to the uncertainty in the nuclear data and the computational methods. The benchmark should contribute also to the validation of the new nuclear cross-section and covariance data evaluations. This paper presents the final design of the benchmark and the analysis using the deterministic transport, sensitivity and uncertainty code system. The analysis includes the calculation of the tritium production rate (TPR) in LiPb layers and the neutron reaction rates, which were measured in the experimental set-up. The SUSD3D cross-section sensitivity and uncertainty code together with the 2D/3D deterministic transport code package DOORS is used for the analysis of the experiment. Based on the sensitivity analyses the most important nuclear reactions and energy ranges in the particular reaction rate measurements are identified, as well as the corresponding uncertainties.

In support of the breeder blanket development program, the EU is conducting a dedicated neutronics R&D effort to provide the basis for the design of nuclear tests to be performed in ITER on the Test Blanket Modules (TBMs). It includes the development of computational tools comprising both Monte-Carlo and deterministic transport, sensitivity and uncertainty codes, the generation of high quality neutron cross-section and covariance data libraries. These are validated experimentally in view of their application in the ITER TBM and the DEMO design. To this purpose, two neutronics experiments have been carried out on mock-ups of both the Helium Cooled Pebble Bed (HCPB) and the Helium Cooled Lithium Lead (HCLL) variants of ITER TBMs, at 14-MeV neutron sources. Redundant experimental techniques have been used to measure the resulting tritium production rate and the neutron and gamma ray spectra which are needed to predict the blanket shielding performance, nuclear power production and all nuclear loads. The comparison of experiment and corresponding calculation is obtained with the associated uncertainty margin based on experimental as well as calculational uncertainties. At the same time, suitable nuclear measuring techniques for TBMs in ITER, in particular for the tritium production, are being developed, optimised, and tested in the mock-up experiments. In particular, the present paper summarizes the preliminary results of the latest of such experiments, i.e. the one conducted on a mock-up of the HCLL blanket. It describes also the dedicated neutronics R&D activities, including the efforts to specify the neutronics tests and objectives in TBMs in ITER, considering the various design concepts.

Key reactions have been selected to compare JEFF-3.3 (CIELO 2) and IAEA CIELO (CIELO 1) evaluated nuclear data files for neutron induced reactions on 235U and 238U targets. IAEA CIELO evaluation uses reaction models to construct the evaluation prior, but strongly relied on differential data including all reaction cross sections fitted within the IAEA Neutron Standards project. The JEFF-3.3 evaluation relied on a mix of differential and integral data with strong contribution from nuclear reaction modelling. Differences in evaluations are discussed; a better reproduction of differential data for the IAEA CIELO evaluation is shown for key reaction channels.

Whenever the sample size in neutron activation analysis cannot be made small enough, self-shielding effects need to be taken into account. When several resonance absorbers are present in the sample, resonance interference must also be considered. Estimation of the self-shielding factors by the Monte Carlo technique is too cumbersome for routine application. Various simplified approaches were compared to rigorous Monte Carlo calculations, pointing out their potential limitations. Good results are obtained using self-shielding factors calculated from evaluated nuclear data libraries and tabulated as a function of the dilution cross-section. The dilution cross-section depends on the material composition and the sample dimensions through the equivalence principle, which is well known in reactor physics. Resonance interference is calculated by solving the neutron spectrum slowing-down equation from cross-sections in 640-group representation. The MATSSF code was written for the purpose and is available on request.

A neutronics experiment on a mock-up of the EU Test Blanket Module (TBM), helium cooled lithium lead concept, is in preparation with the objective to validate the capability of the neutronics codes and nuclear data to predict nuclear responses, such as the tritium production rate (TPR), with qualified uncertainties. Three independent measurements of the TPR will be performed using Li 2 CO 3 pellets. Other measurement techniques have been developed using thermo-luminescence detectors, and diamond detectors covered with 6 LiF. Neutron flux spectra will also be measured from fast energies down to thermal energies, relevant for TPR. Comparison of measured quantities ( E ) with the same calculated quantities ( C ) will be provided, together with the related uncertainties. The paper presents the results of development of the measurement techniques and their relevance for tritium measurements in TBM in ITER. It presents also the pre-analyses conducted to optimise the mock-up configuration so that the neutron spectra are as similar as possible to those in the TBM in ITER. Sensitivity/uncertainty assessments of the TPR show that the calculation uncertainty due to the uncertainties of the neutron cross sections amounts to a few %, depending on position. The largest uncertainties are due to the elastic scattering (n,2n), and (n,3n) reactions on Pb.

The nuclear reaction code EMPIRE has been extended to provide evaluation capabilities for neutron cross section covariances in the thermal, resolved resonance, unresolved resonance and fast neutron regions. The Atlas of Neutron Resonances by Mughabghab is used as a primary source of information on uncertainties at low energies. Care is taken to ensure consistency among the resonance parameter uncertainties and those for thermal cross sections. The resulting resonance parameter covariances are formatted in the ENDF-6 File 32. In the fast neutron range our methodology is based on model calculations with the code EMPIRE combined with experimental data through several available approaches. The model-based covariances can be obtained using deterministic (Kalman) or stochastic (Monte Carlo) propagation of model parameter uncertainties. We show that these two procedures yield comparable results. The Kalman filter and/or the generalized least square fitting procedures are employed to incorporate experimental information. We compare the two approaches analyzing results for the major reaction channels on {sup 89}Y. We also discuss a long-standing issue of unreasonably low uncertainties and link it to the rigidity of the model.

As a follow-up of the work presented at the ND-2007 conference in Nice, additional fast reactor benchmarks were analyzed. Adjustment to the cross sections in the keV region was necessary. Evaluated neutron cross section data files for 180,182,183,184,186W isotopes were produced. Covariances were generated for all isotopes except 180W. In the resonance range the retro-active method was used. Above the resolved resonance range the covariance prior was generated by the Monte Carlo technique from nuclear model calculations with the Empire-II code. Experimental data were taken into account through the GANDR system using the generalized least-squares technique. Introducing experimental data results in relatively small changes in the cross sections, but greatly constrains the uncertainties. The covariance files are currently undergoing testing.

Following the outcomes and the recommendations of the Working Party on Evaluation Cooperation (WPEC) SG-26 the impact of the uncertainties and energy correlations in the prompt neutron fission spectra (PFNS) was studied in several critical benchmarks, both thermal and fast. Two different approaches to calculate the sensitivity of keff with respect to the fission neutron spectra, including the new so-called normalized sensitivity method, were used as a means to check and assure the mathematical correctness of the derived fission spectrum covariance matrices. A new Monte Carlo (MC) method has been proposed and applied to produce the covariance matrices of the PFNS for the neutron induced fission on 235U, 238U and 239Pu nuclei. The MC method was validated by comparison with the matrix derived analytically using the so-called File-30 formalism in ENDF terminology. Several sets of covariance matrices of the PFNS of 235U, 238U and 239Pu nuclei were derived from the uncertainties assigned to the model parameters of the Watt and the recently proposed Kornilov PFNS parameterizations.

An evaluation for neutron induced reactions of 232 Th has been performed in the unresolved resonance region from 4 to 100 keV. A generalized single-level representation compatible with the energy-dependent options of the ENDF-6 format has been used. The average partial cross sections have been expressed in terms of transmission coefficients by applying the Hauser–Feschbach statistical reaction theory including width-fluctuations. The evaluation is based on a combined analysis of experimental average cross section data (including the most recent capture cross section data obtained at the time-of-flight facilities GELINA and n_TOF), and results of dispersive coupled-channel optical model calculations. Experimental data on transmission and self-indication measurements as well as integral benchmark calculations have been used for validation. The evaluation also includes covariance information.

This work was carried out in view of the possible use of diamond detectors as high resolution neutron spectrometers for the ITER project. An MCNP5(X) based computational tool has been developed to simulate the fast neutron response of diamond detectors. The source neutrons are generated by a source routine, developed earlier, that includes deuteron beam energy loss, angular straggling, and two-body relativistic kinematics. The diamond detector routine calculates a pulse height spectrum that is built up by elastic and inelastic scattering, (n,a), (n,p), and (n,d) reaction channels. A combination of nuclear data from ENDF/B-VII.0, TENDL-2010, and ENSDF is used. The simulated spectra are compared with measured spectra. It is shown that the simulation tool allows an interpretation of most of the characteristic features in the spectrum. This is an important step towards the use of diamond detectors for spectral analysis and fluence measurements. © 2001 Elsevier Science. All rights reserved. © 2012 Elsevier B.V.

Resonance parameters for neutron induced reactions in 106,108,110,111,112,113,114,116Cd have been evaluated. The parameters are the result of an analysis of experimental data available in the literature together with a parameter adjustment to transmission and capture data obtained at the time-of-flight facility GELINA. The parameters derived from the GELINA data are in reasonable agreement with those quoted in the literature. From the analysis of the GELINA transmission data a thermal neutron total cross section equal to 2450 ± 40 b has been deduced for natCd at 300 K. This value is in agreement with results of previous measurements which have been performed at thermal reactor beams using different techniques. It differs by about 1.5% from the value 2413 b which was recently deduced from an adjustment to results of an integral experiment. The GELINA transmission and capture data in the low energy region are not fully consistent with resonance parameters recommended in evaluated data files. The impact of the resonance parameters obtained in this work on cadmium transmission factors and on the interpretation of an integral experiment is discussed.

International audience; Significant efforts have been made over the lastfew years in the French Alternative Energies and Atomic Energy Commission (CEA) to adopt multi-step Monte Carlo calculation schemes in the investigation and interpretation of the response of nuclear reactor instrumentation detectors (e.g.miniature ionization chambers - MICs and self-powered neutron or gamma detectors - SPNDs and SPGDs). The first step consists of the calculation of the primary data, i.e. evaluation of the neutron and gamma flux levels and spectra in the environment where the detector is located, using a computational model of the complete nuclear reactor core and its surroundings. Thesedata are subsequently used to define sources for the following calculation steps, in which only a model of the detector under investigation is used. This approach enables calculations withsatisfactory statistical uncertainties (of the order of a few %) within regions which are very small in size (the typical volume of which is of the order of 1 mm$^3$). The main drawback of a calculation scheme as described above is that perturbation effects on the radiation conditions caused by the detectors themselves are not taken into account. Depending on the detector, the nuclear reactor and the irradiation position,the perturbation in the neutron flux as primary data may reach 10 to 20%. A further issue is whether the model used in the second step calculations yields physically representative results. This is generally not the case, as significant deviations may arise, depending on the source definition. In particular, as presentedin the paper, the injudicious use of special ptions aimed at increasing the computation efficiency (e.g. reflective boundary conditions) may introduce unphysical bias in the calculated flux levels and distortions in the spectral shapes. This paper presents examples of the issues described above related to a case study on the interpretation of the signal from differenttypes of SPNDs, which were recently irradiated in the Jozef Stefan Institute TRIGA Mark II reactor in Ljubljana, Slovenia, and provides recommendations on how they can be overcome.The paper concludes with a discussion on the renormalization of the results from the second step calculations, to obtain accurate physical values.

Absolute average capture cross sections of gold, thorium, tantalum, molybdenum, copper and strontium in 252 Cf spontaneous fission neutron spectrum were simulated for two types of experiment setups preformed by Z. Dezso and J. Csikai and by L. Green. The experiments were simulated with MCNP5 using cross section data from the ENDF/B-VII.0 library. The determination of neutron backscattering was calculated with the use of neutron flagging. Correction factors to experimentally measured values were determined to obtain average cross sections in a pure 252 Cf spontaneous fission spectrum. Influence of concrete wall thickness, air moisture and room size on the average cross section was analyzed. Correction factors amounted to about 30%. Corrected values corresponding to average cross sections in a pure 252 Cf spectrum were calculated for 197 Au, 232 Th, 181 Ta, 98 Mo, 65 Cu and 84 Sr. Average cross sections were also calculated with the RR_UNC software using IRDFF-v.1.05 and ENDF/B-VII.0 libraries. The revised average radiative capture cross sections are 75.5±0.1 mb for 197 Au, 87.0±1.6 mb for 232 Th , 98.0±4.5 mb for 181 Ta, 21.2±0.5 mb for 98 Mo, 10.3±0.3 mb for 63 Cu, and 34.9±6.5 mb for 84 Sr.