EASY—a tool for activation calculations

The European Activation System, EASY, includes as the source of nuclear data the European Activation File (EAF). A new version of EAF, EAF-2007, has been developed that contains cross-section data for deuteron- and proton-induced reactions in addition to the traditional neutron-induced data. The main reason for the addition of these data to EAF is to enable activation calculations to be performed for the International Fusion Materials Irradiation Facility, which is a planned materials test facility, and for other accelerator-driven devices with an incident upper energy limit of 60 MeV. EAF-2007 has benefited from the generation and maintenance of comprehensive activation files and the development of the processing code SAFEPAQ-II. Cross-section validation exercises against both experimental data and systematics, which were started on the EAF-4 file (1995), enable a comprehensive assessment of the data. Although EAF-2007 is the best-validated activation neutron cross-section library in the world, currently <3% of all the reactions can be compared with experimental information and sometimes then only for a very limited, and not always relevant, energy range. As with EAF-2001, EAF-2003, and EAF-2005, results of integral experiments have been used to correct, adjust, and validate data. This can be done using SAFEPAQ-II by inputting the measured effective cross sections. Validation using integral data has been performed by means of direct comparison with measurements of various materials under relevant neutron spectra. A tool has been recently developed that is of importance now that libraries contain so much calculated data. Statistical Analysis of Cross Sections is used to look for trends in the library data for a particular reaction type, and this has proved efficient in identifying reactions with data that need correction or improvement. This method has been used with EAF-2005 and EAF-2007 and is a valuable additional validation method.

For validating radioactive inventories calculated by the European Activation System (EASY), a series of integral activation experiments in a white fast neutron field ranging up to 20 MeV was conducted within a task of theEuropean Fusion Technology Programme. The irradiated samples were investigated by gamma ray spectrometry. Specific activities in becquerels per kilogram of sample material were derived for many product nuclides, including, in some cases, products of sequential charged particle induced reactions and of exotic neutron reactions such as (n,2p) and (n,pa). Experiments on vanadium alloys, nickel, copper, lithium orthosilicate, Eurofer-97 steel and tungsten are described. The experimental results are given together with sufficient details of each irradiation to allow comparisons with future EASY or other calculations. Also, calculation-over-experiment ratios for EASY-99 calculations and, in cases of single-pathway products, experimental one-group production cross sections are given. Some results given earlier are presented in revised form. For a few product nuclides, results are new and were not presented in our earlier unpublished reports.

Improved activation tests of fusion structural materials with a deuteron–beryllium neutron source

Activation experiment with tungsten in fusion peak neutron field

Measurement and analysis of radioactivity induced in yttrium and lead in fusion peak neutron field

Impurities and evaluation of induced activity of SiC f/SiC composites

In the Cavity Fission Spectrum Standard Neutron Field of the BR-1 Reactor at the Belgian CEN/SCK Laboratory, a series of cross section measurements was made in which the neutron fluence rate was determined by a transfer measurement from the known fluence rate in the vicinity of a calibrated 252Cf neutron source at the U.S. National Bureau of Standards. The absolute calibration of the 252Cf neutron source was established by means of a manganese bath intercomparison with the standard neutron source NBS-I. Fission cross sections are reported for 238U, 235U, 233U, 239Pu, 240Pu, 237Np and 232Th. Also included in the measurements were the 115In(n,n’) cross section and the 58Ni(n,p) cross section.

The ENDF formats and libraries are decided by the Cross Section Evaluation Working Group (CSEWG), a cooperative effort of national laboratories, industry, and universities in the US and Canada, and are maintained by the National Nuclear Data Center (NNDC). Earlier versions of the ENDF format provided representations for neutron cross sections and distributions, photon production from neutron reactions, a limited amount of charged-particle production from neutron reactions, photo-atomic interaction data, thermal neutron scattering data, and radionuclide production and decay data (including fission products). Version 6 (ENDF-6) allows higher incident energies, adds more complete descriptions of the distributions of emitted particles, and provides for incident charged particles and photo-nuclear data by partitioning the ENDF library into sublibraries. Decay data, fission product yield data, thermal scattering data, and photo-atomic data have also been formally placed in sublibraries. In addition, this rewrite represents an extensive update to the Version V manual.

A new evaluation of spectrum averaged cross sections (SACS) of 235U, 238U, and 239Pu measured in the 252Cf(sf) reference neutron field is presented and found to be consistent with the original Mannhart SACS evaluation in IRDF-2002. The comprehensive vetted experimental database that includes SACS ratio and absolute SACS measurements of major actinides is being used to update the SACS database employed as input of the new GMApy code to derive the Neutron Standards. An update of the current neutron standards based on Time Projection Chamber (TPC) shape data, a new comprehensive uncertainty quantification, and revised SACS experimental database is proposed which result in a 0.7% increase of the evaluated 239Pu(n,f)/235U(n,f) crosssection ratio in the 1–5 MeV energy region. The increase is due to a 0.3% reduction of the standard 235U(n,f) cross section and a 0.4% increase of the 239Pu(n,f) reference cross section in the 1–5 MeV energy region. Those changes are well within estimated USU fission cross-section uncertainties of 1.2%, but are relevant for the evaluated mean values.

The International Reactor Dosimetry File (IRDF)-2002 released in 2004 by the IAEA (see http://www-nds.iaea.org/irdf2002/) contains cross-section data and corresponding uncertainties for 66 dosimetry reactions. New cross-section evaluations have become available recently that re-define some of these dosimetry reactions including: (1) high-fidelity evaluation work undertaken by one of the authors (KIZ); (2) evaluations from the US ENDF/B-VII.0 and candidate evaluations from the US ENDF/B-VII.1 libraries that cover reactions within the International Evaluation of Neutron Cross-Section Standards; (3) European JEFF3.1 library; and (4) Japanese JENDL-4.0 library. Additional high-threshold reactions not included in IRDF-2002 (e.g.,59Co(n,3n) and 209Bi(n,3n)) have been also evaluated to characterize higher-energy neutron fields. Overall, 37 new evaluations of dosimetry reactions have been assessed and intercomparisons made with integral measurements in reference neutron fields to determine whether they should be adopted to update and improve IRDF-2002. Benchmark calculations performed for newly evaluated reactions using the ENDF/B-VII.0 235U thermal fission and 252Cf spontaneous fission neutron spectra show that calculated integral cross sections exhibit improved agreement with evaluated experimental data when compared with the equivalent data from the IRDF-2002 library. Data inconsistencies or deficiencies of new evaluations have been identified for 63Cu(n,2n), 60Ni(n,p)60m+g Co, 55Mn(n,γ), and 232Th(n,f) reactions. Compared with IRDF-2002, the upper neutron energy boundary was formally increased from the actual maximum energy of typically 20 MeV up to 60 MeV by using the TENDL-2010 cross sections and covariance matrices. This extension would allow the updated IRDF library to be also used in fusion dosimetry applications. Uncertainties in the cross sections for all new evaluations are given in the form of relative covariance matrices. Newly evaluated excitation functions should be considered as suitable candidates in the preparation of an improved version of the IRDF that was planned to be released for testing in December 2011.

To calculate the neutron kinetic energy release parameters (kerma factors) for an element consisting of several isotopes, it is necessary to use separate neutron cross-section and radioactive decay data for each isotope. In the case of chlorine, cross sections for natural chlorine are available, but cross sections for the individual isotopes are not. Kerma factors for chlorine were calculated using natural chlorine cross sections and weighting the contribution of each isotope to the decay data, by using the fractional contribution as a variable parameter. The kerma factors calculated at particular neutron energies are found to be sensitive to this parameter and dependent on that energy. The purpose of this Note is to encourage experimental determination of the relevant neutron cross sections of isotopically pure samples of chlorine. 2 figures, 3 tables.

Recent multi-messenger detection of the binary neutron star merger (GW170817) energized the astrophysical community and encouraged further research for determination of nuclear physics observables. Comprehensive studies of atomic nuclei in the cosmos provide an opportunity for investigating these astrophysical phenomena and acquiring complementary information on stellar nucleosynthesis processes that can be verified using the latest nuclear data. Evaluated Nuclear Data File (ENDF) libraries contain complete collections of reaction cross sections over the energy range relevant to astrophysics, fission yields and decay data. These data collections have been used worldwide in nuclear science, industry and national security applications. There is great interest in exploring the ENDF/B-VIII.0 and TALYS Evaluated Nuclear Data Library (TENDL-2015) for nuclear astrophysics purposes and comparing findings with the Karlsruhe Astrophysical Database of Nucleosynthesis in Stars (KADoNiS). The Maxwellian-averaged cross sections (MACS) and astrophysical reaction rates have been calculated using the ENDF/B-VIII.0 and TENDL-2015 evaluated data sets. The calculated cross sections were combined with the solar system abundances and fitted using the classical model of stellar nucleosynthesis. Astrophysical rapid- and slow-neutron capture, r- and s-process, respectively, abundances were obtained from present data and compared with available values. Further analysis of MACS reveals potential evaluated libraries data deficiencies and a strong need for new measurements. The current results demonstrate a large nuclear astrophysics potential of evaluated libraries and mutually beneficial relations between nuclear industry and research e orts.

The reconstruction of fields and adjustment of the parameters of reactor models are increasingly prioritized for reactor physics because the efficiency and safety of the control of a nuclear plant depend on solving them. The results of reconstruction and parameter adjustments are largely determined by the probability distributions of the computed and measured data. The probability distribution of the computed data in turn depends on the properties of the neutron-physical model and probabilistic characteristics of the parameters of this model. The present work is devoted to the search for and investigations of efficient algorithms for calculating the covariation functions of the neutron flux density. A model of a stationary subcritical reactor with an internal neutron source is studied. Linear perturbation theory, the theorem on the spectral decomposition of operators, and the theory of stochastic processes are the tools used for performing the analysis. The uncertainty of the nuclear data, the technological parameters of the core, and the inaccuracies in the model approximations give rise to a significant error in neutron-physical calculations. For this reason, reactor models must be verified and adjusted on the basis of measurements in order to make the control and design of reactor setups more reliable and efficient. The measurements and calculations can be compared with one another correctly only if their errors are taken into account. The problem of quantitative assessment of the computational uncertainty in the model parameters is solved by means of the theory of sensitivity and uncertainty analysis [1‐3]. Methods for evaluating the uncertainty of linear and nonlinear functions of the neutron field are now adequately developed [4, 5]. Less attention is given to the evaluation of the uncertainty of distributed characteristics: the neutron and temperature fields, distribution of the nuclear concentration, energy release fields, and so on. The probabilistic characteristics of the distributed parameters are used in the algorithms for reconstruction, adjustment, and optimal filtration of computational models [6‐9]. Of special value for these algorithms are the second moments of the function of state of a model ‐ the covariation matrices, to whose study the work is devoted. The aim of the present work is to study the structure of and search for efficient algorithms for calculating the covariation matrices of the neutron field on the basis of the multigroup diffusion approximation. The analysis is performed for a stationary subcritical reactor with a distributed internal neutron source.

The 20Ne+ 159Tb reaction studied with the KX-ray method at 15 [formula omitted

Hydrogen effects in neutron activation analysis