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The neutron yields caused by spontaneous α -decay of actinides and subsequent ( α ,xn) reactions were re-evaluated for the reactor fuel materials UO 2 , UF 6 , PuO 2 and PuF 4 . For this purpose, the most recent reference data for decay parameters, α -particle stopping powers and ( α ,xn) cross sections were collected, analysed and used in calculations. The input data and elaborated code were validated against available thick target neutron yields in pure and compound materials measured at accelerators or with radioactive sources. This paper provides the specific neutron yields and their uncertainties resultant from α -decay of actinides 241 Am, 249 Bk, 252 Cf, 242,244 Cm, 237 Np, 238−242 Pu, 232 Th and 232−236,238 U in oxide and fluoride compounds. The obtained results are an update of previous reference tables issued by the Los Alamos National Laboratory in 1991 which were used for the safeguarding of radioactive materials by passive non-destructive techniques. The comparison of the updated values with previous ones shows an agreement within one estimated uncertainty ( ≈ 10 % ) for oxides, and deviations of up to 50% for fluorides. [ABSTRACT FROM AUTHOR]

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 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. Additional six reference fields were assessed (four from plutonium critical assemblies, two measured fields for thermal-neutron induced fission on 233U and 239Pu targets) but not used for validation due to systematic discrepancies in C/E reaction rate values or lack of reaction-rate experimental data. Another ten analytical functions are included that can be useful for calculating average cross sections, average energy, thermal spectrum average cross sections and resonance integrals. The IRDFF-II library and comprehensive documentation is available online at www.nds.iaea.org/IRDFF/. Evaluated cross sections can be compared with experimental data and other evaluations at www.nds.iaea.org/exfor/endf.htm. The new library is expected to become the international reference in neutron metrology for multiple applications. [ABSTRACT FROM AUTHOR]

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), 6 Li(n,t), 10 B(n, α ), 10 B(n, α 1 γ ), nat C(n,n), Au(n, γ ), 235 U(n,f) and 238 U(n,f). Also in the evaluation process the 238 U(n, γ ) and 239 Pu(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, γ ) cross section at 30 keV; reference cross sections for prompt γ -ray production in fast neutron-induced reactions; reference cross sections for very high energy fission cross sections; the 252 Cf spontaneous fission neutron spectrum and the 235 U 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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]