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Nuclear data processing is important to connect evaluated nuclear data libraries and radiation transport codes. The nuclear data processing code FRENDY version 1 was released in 2019 to generate ACE-formatted cross section files with simple input data. After we released FRENDY version 1, many functions were developed, e.g. neutron multi-group cross section generation, explicit consideration of the resonance interference effect among different nuclides in a material, adaptive setting of the background cross sections, consideration of the resonance upscattering, ACE file perturbation, statistical uncertainty quantification of probability table, and modification of ENDF-6-formatted file. FRENDY version 2 was released including these new functions. It generates GENDF- and MATXS-formatted neutron multi-group cross section files from an ACE-formatted cross section file or an evaluated nuclear data file. This paper explains the features of the new functions implemented in FRENDY version 2 and the verification of the neutron multigroup cross section generation function of this code. [ABSTRACT FROM AUTHOR]

Coherent scattering is the theoretical basis of X-ray diffraction, which is widely used in the field of materials. A functional module to calculate the nondestructive testing of photon-atomic and photon-molecular coherent scattering cross sections is developed in the nuclear data processing code NECP-Atlas. The photon-atomic coherent scattering cross section is processed in NECP-Atlas with the same method as that of NJOY2016. In practice, photon interacts with materials which usually consist of numerous polyatomic molecules. Therefore, the photon-molecular coherent scattering cross sections should be calculated instead of photon-atomic coherent scattering cross sections to acquire more accurate simulated results. First, the independent atom model (IAM) was implemented to calculate the photon-molecular coherent scattering cross sections with the assumption that the atom was isolated. This approximation works well when the momentum transfer of incident photons is large. At small momentum transfer, the interference effects of various atoms cannot be ignored. Then, the molecular interference model was introduced to consider the influence on angular distribution of secondary photons due to the interference effects. The molecular interference functions of water and ethanol were simulated using molecular dynamics simulation, and then applied in the ACE library. The atomic form factors of the atoms in water and ethanol molecules were modified by the molecular interference function. To quantify the influence of molecular interference effects on angular distribution of secondary photons, an imaging system was simulated. Numerical results show that the molecular interference effects have a significantly influence on the angular distribution of photons at small momentum transfer, while at large momentum transfer, the molecular interference effects can be ignored.

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• Capability for processing photonuclear data is developed in nuclear data processing code NECP-Atlas. • The newly developed capability is verified. • NECP-Atlas can provide comprehensive data for photon-neutron coupling transport calculations. In the present work, NECP-Atlas has been developed to feature a capability for processing photonuclear data, providing ACE format library for Monte Carlo codes, alongside multi-group data for deterministic transport calculation codes. The work emphasizes the development of a capability responsible for generating multi-group data, including the multi-group photonuclear reaction cross-sections and scattering matrix. The capability for generating a multi-group scattering matrix, which is lacking in conventional nuclear data processing codes, enables NECP-Atlas to provide comprehensive data for photon-neutron coupling transport calculations. The photonuclear ACE library and multi-group photonuclear absorption cross-sections generated by NECP-Atlas are verified with the results obtained by NJOY21. To verify the photonuclear scattering matrix, the photon and neutron leakage spectra of a one-dimensional spherical shell problem and the photon and neutron dose equivalent rates of an electron linear accelerator are calculated and compared with the results from the Monte Carlo simulations. The numerical results affirm that NECP-Atlas has the capability for processing photonuclear data with high accuracy. [ABSTRACT FROM AUTHOR]

The kernel broadening approach to Doppler broadening has been shown to be equivalent to the convolution of a function related to the unheated cross section and Gaussian distribution function. This convolution may be expressed as the inverse Fourier transform of the product of the Fourier transforms of the convolution arguments. By recognizing that the unheated cross section function is an odd function, the resulting equations are reduced to a Hankel Transform of order α = 1 2 . A discretized version of the Hankel Transform is applied to produce a set of equations that are independent of the interpolation scheme used in the pointwise representation of the cross sections. The heated cross section at a given energy is consequently calculated as the sum of the contributions from each of these pointwise intervals and each contribution is represented as a separate Hankel Transform. • Mathematical representation of Doppler Broadening as Inverse Hankel Transforms. • Preliminary implementation and comparison to analytical and legacy models. • Analytical Benchmarks for pointwise constant, linear and 1 v interpolation intervals. • Verification to kernel-broadening method for various projectile/target reactions. [ABSTRACT FROM AUTHOR]

The nuclear data processing code NECP-Atlas, developed at Xi’an Jiaotong University, has various functions to process the evaluated nuclear data to generate the applied nuclear data library for the nuclear reactor designs. In this paper, the photonrelated data generation capacities were developed in NECPAtlas, including the photon production cross sections and photon irradiation cross sections. In the conventional nuclear data processing code such as NJOY, only the production cross section for the prompt photon released with the neutroninduced reactions, and the photonatom reaction cross section can be produced, while the delayed photon released by the fission products is ignored regardless of its large contribution for the heating of the regions without fissionable materials such as control and reflector assemblies in fast neuron reactors. In this paper, the capacity for the generation of the production cross section for the delayed photon was developed in NECPAtlas. Besides, in the conventional nuclear data processing code, only the neutron irradiation damage cross section can be generated, while in the practical reactors, electron and positron also cause significant irradiation damage. The photonatom reactions, including Compton scattering, pair production and photoelectric effect, emit electron and positron. Therefore, the photon can also cause irradiation damage. To satisfy the demands from various application aspects, the capacity for the calculation of the photon irradiation damage cross sections was also developed in NECPAtlas. Meanwhile, the electron and positron irradiation damage cross section can also be produced, because it is the basis for the calculation of photon irradiation damage cross section. The delayed photon production cross sections were tested on the EBRⅡ fast reactor, and the results were compared with the library of the ERANOS which includes the production cross section of delayed photon. The difference of photon heating calculated based on the two libraries respectively from NECPAtlas and ERANOS is less than 0.93%, which means the correctness of the delayed photon data produced by NECPAtlas. Besides, the numerical results show that if the delayed photon is ignored, the difference of photon heating for the control assembly and reflector assembly can reach 3258% and 2041%, compared with the results of considering the delayed photon. The electron and positron displacement damage cross sections of Fe and Au calculated by NECPAtlas were verified with the results published by Oka Ridge National Laboratory, and the photon displacement cross sections calculated by NECPAtlas were verified with results reported by other researches. The numerical results show good agreement with the reference results.

Materials subjected to neutron irradiation will suffer from a build-up of damage caused by the displacement cascades initiated by nuclear reactions. Previously, the main "measure" of this damage accumulation has been through the displacements per atom (dpa) index, which has known limitations. This paper describes a rigorous methodology to calculate the primary atomic recoil events (often called the primary knock-on atoms or PKAs) that lead to cascade damage events as a function of energy and recoiling species. A new processing code SPECTRA-PKA combines a neutron irradiation spectrum with nuclear recoil data obtained from the latest nuclear data libraries to produce PKA spectra for any material composition. Via examples of fusion relevant materials, it is shown that these PKA spectra can be complex, involving many different recoiling species, potentially differing in both proton and neutron number from the original target nuclei, including high energy recoils of light emitted particles such as α-particles and protons. The variations in PKA spectra as a function of time, neutron field, and material are explored. The application of PKA spectra to the quantification of radiation damage is exemplified using two approaches: the binary collision approximation and stochastic cluster dynamics, and the results from these different models are discussed and compared.

In the conventional method to generate thermal scattering cross section of moderator materials, only one of the coherent elastic scattering and incoherent elastic scattering is considered in neutronics calculations. For the inelastic scattering, fixed incident energy grid is used in the nuclear data processing codes. The multipoint linearization method is used to refine the incident energy grid for inelastic scattering. We select ZrHx (zirconium hydride) as an example to analyze the effects of the above described treatments on the reactivity of several critical benchmarks. The numerical results show that the incident energy grid has an obvious effect on the effective multiplication factor (keff) of the analyzed reactors; simultaneously considering the coherent and incoherent elastic scattering also affects keff by tens of pcm.

The neutrons generated in fusion plasmas interact with materials via nuclear reactions. The resulting transmutations and atomic displacements have life-limiting consequences for fusion reactor components. A detailed understanding of the production, evolution and material consequences of the damage created by cascades of atomic displacements requires, as a vital primary input, a complete description of the energy-spectrum of initial (prompt) atomic displacement events (the primary knock on atoms or PKAs) produced by direct neutron nuclear interactions. There is also the possibility that the radionuclides produced under transmutation will create further PKAs as they decay, and so the rate of these must also be quantified. This paper presents the latest results from the analysis of PKA spectra under neutron irradiation, focussing particularly on the variation in PKA distributions due to changes in composition under transmutation, but also on the PKA contributions from radioactive decay of materials that become activated under irradiation.

Highlights • A new nuclear data processing code named NECP-Atlas is developed. • Some new processing functions have been implemented into the code. • NECP-Atlas is qualified to generate cross sections for neutron transport calculations. Abstract A new nuclear data processing code called NECP-Atlas is under development at Xi'an Jiaotong University in China. The motivation for the development is to establish a platform to carry out deeper researches on nuclear data processing methods to satisfy the demands on accurate cross sections in the fields of high-fidelity transport simulation and advanced reactor design. The present paper describes the methods used in the current version of NECP-Atlas and demonstrates the performance and accuracy of the code on a variety of benchmarks. At the present time, NECP-Atlas can process ENDF/B-VII.1, ENDF/B-VII.0, CENDL-3.1, JEFF-3.2 and JENDL-4.0 evaluations, and generate WIMS-D and ACE format libraries. The accuracy of the code is comparable with NJOY2016. NECP-Atlas is competent to provide cross sections for deterministic and Monte Carlo transport calculations. [ABSTRACT FROM AUTHOR]

Abstract Some efforts have been made to exactly consider the effect of neutron up-scattering caused by thermal motion of target nuclei and resonance elastic scattering on the multi-group cross sections and scattering matrices. Firstly, the resonance elastic scattering kernel (RESK) formulations for anisotropic scattering up to any Legendre order has been adopted to represent the exact Doppler broadened energy transfer kernels. A semi-analytical integration method is applied to perform the RESK calculations. Combining with the RESK calculations, a linearization algorithm is proposed to generate the RESK interpolation tables. The RESK data can be interpolated precisely based on the interpolation tables to reduce the calculation burden. Secondly, a neutron slowing-down equation solver is developed based on the RESK instead of the conventional asymptotic scattering kernel, where the effect of neutron up-scattering on the neutron energy spectrum can be exactly considered. The accuracy of the multi-group cross sections and scattering matrices are significantly improved when the exact energy spectrum is applied for the group collapsing calculations. All the methods mentioned above have been implemented into a newly developed nuclear data processing code called NECP-Atlas. Numerical results show that the proposed methods are capable of producing accurate multi-group cross sections for downstream calculations; the fuel Doppler coefficients and eigenvalues of the real problems will be improved if the up-scattering effect is incorporated into the multi-group cross sections and matrices. [ABSTRACT FROM AUTHOR]

The probability table method is widely used for continuous energy Monte Carlo calculation codes to treat the self-shielding effect in the unresolved resonance region. The ladder method is used to calculate the probability table. This method generates pseudo resonance structures using random numbers based on the averaged resonance parameters. The probability table affects the ladder number, i.e., the number of pseudo resonance structures. Since the higher the ladder number the longer it takes to create a cross-section library, reducing the value of the ladder number is an effective way to reduce the time spent in the creation of a cross-section library. However, the exact value of the most appropriate ladder number to use has not been found. Currently, users have to use a nuclear data processing system to manually set the ladder number to a large enough value in order to obtain a sufficiently converged probability table. In this study, the probability tables for all nuclides prepared in JENDL-4.0 are used to find the optimal ladder number. The results show that the probability table differences are small enough when the ladder number is 100. The impact of these differences on neutronics calculations is found to be small.

Monte Carlo neutron transport codes can be used for high-fidelity predictions of the performance of nuclear systems. However, validation against experiments is required in order to establish the credibility in the results and identify the inaccuracies due to the used calculation scheme and associated databases. The International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP) contains criticality safety benchmarks derived from experiments that have been performed at various nuclear critical facilities around the world and are very valuable for validation purposes. The main objective of this work is the identification and modelling of experimental benchmarks included at ICSBEP in support of the validation of Monte Carlo neutron transport calculations when applied to fast systems, and in particular, KENO-VI and associated AMPX-formatted continuous-energy libraries from SCALE package. In such systems, the predicted k-eff values can be very sensitive to the treatment of nuclear data in the Unresolved Resonance Region (URR). Consequently, benchmarks with intermediate and fast spectra are identified and modelled with KENO-VI. Then, calculated results with and without probability tables in the URR are compared with each other in order to identify the most sensitive configurations to the URR. As a result of the proposed study, recommendations are given about the benchmarks that should be modelled and analysed to qualify the processed continuous-energy libraries before their use in Monte Carlo transport codes for practical fast reactor applications.

The processing code SPECTRA-PKA produces energy spectra of primary atomic recoil events (or primary knock-on atoms, PKAs) for any material composition exposed to an irradiation spectrum. Such evaluations are vital inputs for simulations aimed at understanding the evolution of damage in irradiated material, which is generated in cascade displacement events initiated by PKAs. These PKA spectra present the full complexity of the input (to SPECTRA-PKA) nuclear data-library evaluations of recoil events. However, the commonly used displacements per atom (dpa) measure, which is an integral measure over all possible recoil events of the displacement damage dose, is still widely used and has many useful applications – as both a comparative and correlative quantity. This paper describes the methodology employed that allows the SPECTRA-PKA code to evaluate dpa rates using the energy-dependent recoil (PKA) cross section data used for the PKA distributions. This avoids the need for integral displacement kerma cross sections and also provides new insight into the relative importance of different reaction channels (and associated different daughter residual and emitted particles) to the total integrated dpa damage dose. Results are presented for Fe, Ni, W, and SS316. Fusion dpa rates are compared to those in fission, highlighting the increased contribution to damage creation in the former from high-energy threshold reactions. [ABSTRACT FROM AUTHOR]

Materials subjected to neutron irradiation will suffer from a build-up of damage caused by the displacement cascades initiated by nuclear reactions. Previously, the main “measure” of this damage accumulation has been through the displacements per atom (dpa) index, which has known limitations. This paper describes a rigorous methodology to calculate the primary atomic recoil events (often called the primary knock-on atoms or PKAs) that lead to cascade damage events as a function of energy and recoiling species. A new processing code SPECTRA-PKA combines a neutron irradiation spectrum with nuclear recoil data obtained from the latest nuclear data libraries to produce PKA spectra for any material composition. Via examples of fusion relevant materials, it is shown that these PKA spectra can be complex, involving many different recoiling species, potentially differing in both proton and neutron number from the original target nuclei, including high energy recoils of light emitted particles such as α -particles and protons. The variations in PKA spectra as a function of time, neutron field, and material are explored. The application of PKA spectra to the quantification of radiation damage is exemplified using two approaches: the binary collision approximation and stochastic cluster dynamics, and the results from these different models are discussed and compared. [ABSTRACT FROM AUTHOR]

The Generalized Nuclear Data Structure (GNDS) has been being designed as the replacement of the conventional ENDF-6 format for the evaluated nuclear data libraries. The specification for the GNDS format was released by OECD/NEA in 2020. And the latest evaluated nuclear data libraries ENDF/B-VIII.0 and TENDL2019 were released in the ENDF-6 format and the GNDS format, simultaneously. The capability of processing the GNDS format evaluated nuclear data libraries is developed in the nuclear data processing code NECP-Atlas. To be compatible with the GNDS and ENDF-6 formats, the memory management method and the data structure in NECP-Atlas are modified to support more data beyond the ENDF-6 format. At the same time, a module called GNDS-Toolkit is developed to read, parse and convert the GNDS/XML format libraries. The ENDF/B-VIII.0 library in GNDS/XML format is processed into ACE format and multi-group libraries, and used in criticality calculations, lattice physics calculations and shielding calculations. The results show that the GNDS format libraries can be exactly processed by NECP-Atlas. • The capability of processing the GNDS format is developed in NECP-Atlas. • Root_Lib is constructed to be compatible with the ENDF-6 and GNDS formats. • The neutronics calculation results based on GNDS agree well with ENDF-6. [ABSTRACT FROM AUTHOR]

Highlights: [•] ERANOS libraries are produced based on four modern nuclear data sets. [•] The MOZART MZA/MZB benchmarks are analyzed with these li- braries. [•] Results are generally acceptable in an academic context, but for highly accurate applications data adjustment is required. [•] Some discrepancies between the calculations and the benchmark results remain and cannot be readily explained. [•] Successful generation of ECCO libraries and covariance data for ERA- NOS. [Copyright &y& Elsevier]

Abstract: This paper discusses the uncertainty analysis of the criticality of the prototype Fast Breeder Reactor Monju with JENDL-4.0 nuclear data. The JENDL-4.0 nuclear data are more physically correct and complete than previous JENDL releases. The well-established ERANOS code was used. Since no cross section libraries based on JENDL-4.0 exist for ERANOS, these libraries were generated as part of the work. To validate the newly made libraries, benchmark calculations were performed. A model of the Monju reactor was prepared, based exclusively on publicly available information. The criticality of May 2010 of Monju was adequately calculated. A perturbation analysis identified the most important differences between JENDL-3.3 and JENDL-4.0. A strong contribution of 241Am was found. In the subsequent uncertainty analysis, it was found that the overall uncertainty is roughly the same between JENDL-3.3 and JENDL-4.0. This result was not as expected. Two main causes have been identified: firstly, in JENDL-4.0 covariance data has been added for isotopes and energy ranges for which there previously was no covariance data. Secondly, for some isotopes the covariances in JENDL-4.0 are larger than in JENDL-3.3. For several main reactions and isotopes, the uncertainty is largely reduced between JENDL-3.3 and JENDL-4.0. But for many isotopes with a non-negligible contribution the uncertainty is increased in JENDL-4.0. The balance is a slight increase in the overall uncertainty when JENDL-4.0 is used. [Copyright &y& Elsevier]

• Code-to-experiment verification of FRENDY was carried out. • ACE files of JENDL-4.0 and ENDF/B-VII.1 were processed by FRENDY. • Various criticality benchmarks of the HTTR were used for the verification. • Good agreement between calculation results and experiment results was achieved. • FRENDY could be another choice to generate ACE files besides NJOY. Japan Atomic Energy Agency has developed a new nuclear data processing code, named FRENDY, to generate ACE-formatted files from various evaluated nuclear libraries. A code-to-experiment verification of FRENDY processing was carried out in this study with criticality benchmark assessments of the high temperature engineering test reactor. The ACE files of the JENDL-4.0 and ENDF/B-VII.1 was generated successfully by FRENDY. These ACE files have been used in MCNP6 transport for various benchmark problems of the high temperature engineering test reactor. As a result, the k eff and reaction rate obtained by MCNP6 calculation presented a good agreement compared to the experimental data. The proper ACE files generation by FRENDY was confirmed for the HTTR criticality calculations. [ABSTRACT FROM AUTHOR]

• A new module designed to generate TSL data is developed in NECP-Atlas. • The module can calculate TSL data for solid materials of any crystal structure. • A technique based on the anisotropic displacement parameters method is implemented. A new module designed to generate thermal scattering law (TSL) data is developed in the nuclear data processing code NECP-Atlas, which is currently under development at Xi'an Jiaotong University. The module can calculate TSL data for solid materials of any crystal structure and liquid materials in the free gas or the Egelstaff and Schofield approximations. To acquire a precise description of the neutron thermalization, the module eliminates the cubic approximation and atom site approximation used in the conventional method. In addition, a technique based on the anisotropic displacement parameters method is implemented to consider the anisotropy of interatomic forces and the correlation of forces from different directions. Comparison with the LEAPR and the Debye–Waller matrix method used in ENDF/B-VIII.0 enabled verification of the TSL data generated by the module. The numerical results revealed the good performance of the module. [ABSTRACT FROM AUTHOR]

• A new linearization workflow for the generation of point-wise cross sections is developed. • The Lagrangian interpolation method is applied to accelerate the linearization. • The precision of the linearized point-wise cross sections is verified in microscopical and macroscopical aspects. In the evaluations, there are various types of nuclear data with different formats, and they should be firstly processed into the libraries with specific formats by the nuclear data processing codes so that the processed libraries can be used by the neutronics codes. In the nuclear data processing, one of the most important procedures is to get the point-wise cross sections (PW XSs) by the linearization. In general, the conventional linearization workflow contains the interval-halving technique and a series of linearization criteria. The precision of the PW XSs and the processing efficiency are controlled by the linearization criteria including the fractional tolerance, the maximum fractional tolerance and the resonance-integral tolerance. The resonance-integral tolerance is used to avoid generating a large number of energy points within the high energy range (HER). It is acceptable for the many applications such as the thermal spectrum reactors. However, for some fast spectrum applications, the PW XSs with higher precision within the HER are important. Therefore, the fractional linearization criterion should be used only (strict criteria) for the conventional linearization workflow to reach the higher precision but the processing efficiency is greatly decreased. To improve the processing efficiency and keep the high precision of PW XSs simultaneously, a new linearization workflow based on the Lagrangian interpolation is proposed and implemented into the nuclear data processing code NECP-Atlas which is developed at Xi'an Jiaotong University. Numerical results show that the precision of PW XSs based on the new linearization workflow agree well with that based on the conventional linearization workflow with strict criteria and the processing efficiency is significantly improved. [ABSTRACT FROM AUTHOR]

It has been proved that resonance elastic scattering kernel (RESK) has significant impact on the energy and angle distributions of the secondary neutrons, and consequently affects the eigenvalues and fuel temperature coefficients. Therefore, it is necessary for the nuclear data processing codes to generate the accurate data considering RESK for the neutronics analysis. In this paper, a processing method is proposed to generate the S (α , β , T) tables efficiently for Monte Carlo codes to account for the RESK. A multi-point linearization method is developed to generate the incident energy grids of S (α , β , T) tables, which can minimize the data size of S (α , β , T) tables without losing accuracy. The equally probable cosines in S (α , β , T) tables are obtained by solving the sets of nonlinear equations which are established by Legendre polynomials and the differential moments of RESK. The generation procedure of the S (α , β , T) tables is more practical by using the proposed calculation method of equally probable cosines. Based on the proposed processing method, a new module is developed in the nuclear data processing code called NECP-Atlas. The ACE-format libraries with the S (α , β , T) tables are generated by NECP-Atlas and supplied to the OpenMC code to perform the verifications. The numerical results show that the impact on eigenvalues caused by the RESK can be reflected correctly by the Monte Carlo calculations using the S (α , β , T) tables. Meanwhile, the results based on the S (α , β , T) tables match well with those based on the Doppler broadening rejection correction approach. • A method for generating S (α , β , T) tables considering resonance elastic scattering kernel is proposed. • The processing method is a highly efficient method and the accuracy of S (α , β , T) tables is high. • The ACE-format libraries are generated by NECP-Atlas and several benchmarks are calculated. [ABSTRACT FROM AUTHOR]