Evaluation of neutron nuclear data on mercury isotopes

Neutron nuclear data on 10 isotopes of mercury have been evaluated for the next version of Japanese evaluated nuclear data library (JENDL) general-purpose file in the energy region from 10^?5 eV to 20 MeV. Resolved resonance parameters (RRPs) for ^{200, 202}Hg were supplemented with the data which had not been considered in the previous library. Unresolved resonance parameters (URPs) were obtained by fitting to the total and capture cross sections calculated from nuclear models. A statistical model code was applied to evaluate cross sections above the resolved resonance region. Coupled-channel optical model parameters were employed for the interaction between neutrons and nuclei except ²⁰³Hg. Compound, pre-equilibrium, and direct-reaction processes were considered for cross-section calculation in the high energy region. The present results reproduce experimental data very well. The evaluated data are compiled into evaluated nuclear data file (ENDF) formatted data files.     

Generation of nuclear data using Gaussian process regression

ABSTRACT

A new approach for generating nuclear data from experimental cross-section data is presented based on Gaussian process regression. This paper focuses on the generation of nuclear data for proton-induced nuclide production cross-sections with a nickel target. Our results provide reasonable regression curves and corresponding uncertainties and demonstrate that this approach is e?ective for generating nuclear data. Additionally, our results indicate that this approach can be applied in experimental design to reduce the uncertainty of generated nuclear data.     

Sensitivity of charged particle activation analysis for long-lived radioactive nuclide determination

ABSTRACT

Charged Particle Activation Analysis (CPAA) utilizing an 8-MeV proton beam has been studied for determination of 35 long-lived radioactive nuclides. We accumulated the reaction cross section and nuclear decay data by referring to nuclear database supplied by National Nuclear Data Center in Brookhaven National Laboratory. We also calculated the reaction cross sections by using statistical model code ALICE. By using the nuclear data, we have derived determination sensitivity of the radioactive nuclides relative to unit weight and specific radioactivity. The result indicates that several hardly measurable nuclides with long half-lives such as ¹³⁵Cs, ²⁴⁴Pu, ¹²⁹I, ¹²⁶Sn, ⁹³Mo, ¹⁰⁷Pd, ²³⁶U, ²⁴⁸Cm, and ²³⁷Np have high sensitivity. It may be concluded that CPAA can be applied to determination of several long-lived nuclei and will provide a quick and non-destructive analysis method.     

Extension of a nuclear reaction calculation code CCONE toward higher incident energies — multiple preequilibrium emission,and spectrum in laboratory system

A nuclear reaction calculation code CCONE, which was developed for nuclear data evaluation for JENDL/AC-2008 and JENDL-4, has been upgraded to improve the prediction accuracy for calculated cross sections at nucleon incident energies higher than 20 MeV. Multiple particle emission, in which nucleons and complex particles up to α-particle are involved, from pre-equilibrium reaction process was implemented based on the sequential-decay calculations for all produced exciton states within the framework of the two-component exciton model. The effect of velocity-change of particle-emitting nuclei on the multiple emission in preequilibrium and compound processes, which was not included in the previous evaluations, was taken into account to obtain spectra in the laboratory system using an average velocity approximation for each composite/compound nucleus. Calculated nucleon emission spectra at nucleon incident energies from 20 to 200 MeV were compared with experimental and evaluated data for the proton- and neutron-induced reactions on ²⁷Al. The present results are in good agreement with experimental data. It was found that their predictions were better than those of JENDL/HE-2007 especially for low emission energies at high incident energies.     

Benchmark study of the recent version of the PHITS code

We performed a benchmark study for 58 cases (22 cases reported in this paper and 36 cases reported in online as supplementary materials of this paper) using the recent version (version 2.88) of the Particle and Heavy-Ion Transport code System (PHITS) in the following fields: (1) particle production cross-sections for nuclear reactions from 20 MeV to 1 GeV, (2) thick-target neutron yields and neutron shielding, (3) depth–dose distribution in water using ¹²C beam, and (4) electron and photon transportation over a wide-energy range from keV to GeV. Overall agreements were found to be sufficiently satisfactory; however, several discrepancies are observed, particularly in particle productions with energies below 100 MeV, neutron production for ⁷Li(p,n)⁷Be, and photonuclear reactions. To overcome these inaccuracies and to further improve the code, it will be necessary to incorporate a high-energy version of the evaluated nuclear data library JENDL-4.0/HE and the photonuclear data file JENDL-PD in the PHITS package.     

Evaluation of neutron nuclear data on tantalum isotopes

Neutron nuclear data on four isotopes of tantalum have been evaluated for the next version of Japanese Evaluated Nuclear Data Library general-purpose file in the energy region from 10^?5 eV to 20 MeV. Unresolved resonance parameters were obtained by fitting to the total and capture cross sections calculated from nuclear models, while resolved resonance parameters were selected from experimental data. A statistical model code was applied to evaluate cross sections above the resolved resonance region. Compound, pre-equilibrium and direct-reaction processes were considered for cross-section calculation. Coupled-channel optical-model parameters were employed for the interaction between neutrons and nuclei. Giant-dipole and pygmy resonance parameters for E1 γ-ray transition from tantalum isotopes were determined so as to reproduce measured γ-ray spectrum for ¹⁸¹Ta. The present results reproduce experimental data very well. The evaluated data are compiled into Evaluated Nuclear Data File (ENDF) formatted data files.     

Covariances of Evaluated Nuclear Cross Section Data for Th, W and Mn

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 ²³²Th, ^{180,182,183,184,186}W and ⁵⁵Mn 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 ²³²Th, ^{180,182,183,184,186}W and ⁵⁵Mn nuclei are included into the ENDF/B-VII.1 library release.     

Evaluation of neutron nuclear data on iodine isotopes

Neutron nuclear data on six isotopes of iodine have been evaluated for the next version of Japanese Evaluated Nuclear Data Library (JENDL) general-purpose file in the energy region from 10^?5 eV to 20 MeV. Unresolved resonance parameters were obtained by fitting to the total and capture cross sections calculated from nuclear models, while resolved resonance parameters remain unchanged from JENDL-4.0. A statistical model code was applied to evaluate the cross sections above the resolved resonance region. Coupled-channel optical model parameters were employed for the interaction between neutrons and nuclei. Compound, pre-equilibrium and direct-reaction processes were considered for cross-section calculation in high-energy region. Giant-dipole resonance parameters for γ-ray transition from iodine isotopes were determined so as to reproduce measured γ-ray spectrum for ¹²⁷I. The present results reproduce experimental data very well. The evaluated data are compiled into ENDF-formatted data files.     

Evaluation of neutron cross sections for a complete set of Dy isotopes

Neutron cross sections for a complete set of Dy isotopes, ^{156,158,160,161,162,163,164}Dy, were evaluated in the incident energy range from 10⁻⁵ eV to 20 MeV. In the low energy region, including thermal and resolved resonances, our evaluations are based on the latest data published in the Atlas of Neutron Resonances. In the unresolved resonance region we performed additional evaluation by using the averages of the resolved resonances and adjusting them to the experimental data. In the fast neutron region, we used the nuclear reaction model code EMPIRE-2.19 with the model parameters adjusted to the experimental data. The results are compared with the available experimental data and with the existing nuclear data libraries, including ENDF/B-VI.8 and JEFF-3.1. The new evaluations are suitable for neutron transport calculations and they were adopted by the new US evaluated nuclear data library, ENDF/B-VII.0, released in December 2006.     

Nuclear data evaluation of Mn by the EMPIRE code with emphasis on the capture cross-section

Manganese is one of the constituents of alloys for structural components of fission and fusion devices and a well-known neutron dosimeter; however, existing ENDF-B/VII.0 ⁵⁵Mn evaluation was produced by Shibata (1989). This work is an attempt to re-evaluate neutron-induced cross-sections of ⁵⁵Mn using the latest release of the EMPIRE code. Sensitivity studies on the physical and fitting parameters are presented, with special emphasis on the capture and neutron inelastic cross-sections. A calculated nuclear data file in ENDF-6 format of the neutron interaction cross-sections is produced. It extends up to 150 MeV, which is of interest for fusion and accelerator driven system applications. This evaluation is compared with the ENDF/B-VII.0 evaluation and with a selection of experimental microscopic cross-sections. The evaluation is tested using integral data: the OKTAVIAN integral experiment on a manganese shell and an FNG experiment with manganese activation foils. Benchmark results provide needed feedback for the refinement of the physics parameters.     

Determination of 79Se and 135Cs in Spent Nuclear Fuel for Inventory Estimation of High-Level Radioactive Wastes

⁷⁹Se and ¹³⁵Cs are long-lived fission products and are found in high-level radioactive waste (HLW). The estimation of their inventories in HLW is essential for the safety assessment of geological disposal, owing to their mobility in the strata. In this study, the amounts of ⁷⁹Se and ¹³⁵Cs in a spent nuclear fuel solution were measured. About 5 g of irradiated UO₂ fuel discharged from a commercial Japanese pressurized water reactor (PWR) with a burn-up of 44.9 GWd/t was sampled and dissolved with 50mL of 4M nitric acid in a hot cell for 2 h. After Se and Cs were chemically separated, the amounts of ⁷⁹Se and ¹³⁵Cs in the spent nuclear fuel solution were measured by inductively coupled plasma quadrupole mass spectrometry (ICP-QMS). The amounts of ⁷⁹Se and 135Cs were 5:2 ± 1:5 and 447 ± 40 g/MTU, respectively. The results presented in this study, which are the first postirradiation experimental data in Japan, showed good agreement with those obtained by the ORIGEN2 code using the data library of JENDL-3.3.     

Evaluation of neutron nuclear data on tellurium isotopes

Neutron nuclear data on 13 Te isotopes have been evaluated for the next release of Japanese Evaluated Nuclear Data Library (JENDL) general purpose file in the energy region from 10^{? 5} eV to 20 MeV. Thermal capture cross sections of ^{120, 121m, 127m, 129m, 131m, 132}Te were determined from the latest measurements or from a simplified formula, although the resolved resonance parameters, which describe the low-energy behavior of cross sections, remain unchanged from JENDL-4.0 for ^{122, 123, 124, 125, 126, 128, 130}Te. A statistical model code was applied to evaluate the cross sections above the resolved resonance region. Coupled-channel optical model parameters were employed for the interaction between neutrons and nuclei. Compound, pre-equilibrium, and direct-reaction processes were considered for cross-section calculation in high-energy region. The present results reproduce experimental data very well, and are found to be much better than the JENDL-4.0 data. The evaluated data are compiled into evaluated nuclear data file (ENDF)-formatted data files.     

The influence of Xe-135m on steady-state xenon worth in thermal molten salt reactors

There has been a resurgence of interest in fuel-in-salt Molten Salt Reactors (MSR); a number of governments and private companies are currently undertaking efforts to develop and commercialize MSR technology. Recent nuclear models used in the TENDL nuclear data library have estimated the cross section of the metastable state of ¹³⁵Xe, ^135mXe, to have a much larger cross section than the ground state of ¹³⁵Xe. Thermal MSRs with continual online noble gas stripping of the fuel salt can operate in a regime where ^135mXe makes up a notable fraction of the xenon worth, necessitating the implementation of these new cross-sections in the neutronic analysis of these advanced reactor types. To estimate the effect of ^135mXe on reactor operation, a simplified mathematical model was produced with one neutron energy group and ^135mXe cross section data from the TENDL-2015 nuclear data library. ²³⁵U and ²³³U systems were investigated. It was found that the steady-state xenon reactivity worth was considerably higher for some modes of operation when ^135mXe was included in the xenon worth calculations. Based on available literature, it was found that proposed MSR concepts may operate in the modes of operation where ^135mXe has a notable impact on steady-state xenon worth. This work highlights the need to include ^135mXe in MSR models and the importance of acquiring evaluated cross-sections for ^135mXe.     

Comparison of Level Density Models for the <Superscript>60,61,62,64</Superscript>Ni(p,n) Reactions of Structural Fusion Material Nickel from Threshold to 30 MeV

The knowledge of level density for reaction cross-section calculations are needed for various applications such as fission and fusion reactor design, accelerator driven sub-critical systems, nuclear medicine, neutron capture and astrophysics. In this study, the excitation functions for (p, n) reactions from reaction threshold to 30 MeV proton incident energy on ⁶⁰Ni, ⁶¹Ni, ⁶²Ni and ⁶⁴Ni isotopes were calculated using TALYS 1.6 nuclear code involving the level density models. This is of importance to the validation of nuclear model approaches with increased predictive power. There are several models of level density that can be used to predict cross-section. In this work, the (p, n) cross-sections would be calculated using three different model of level density, such as constant temperature model, back-shifted fermi gas model and generalized superfluid model on ^60,61,62,64Ni reactions. The (p, n) reaction cross-section calculations for ^60,61,62,64Ni target nuclei were compared with each other and the experimental nuclear reaction data obtained from EXFOR database.     

Evaluation of neutron nuclear data on platinum isotopes

Neutron nuclear data on 10 isotopes of platinum have been evaluated for the next version of Japanese Evaluated Nuclear Data Library general-purpose file in the energy region from 10^?5 eV to 20 MeV. Resolved resonance parameters of naturally occurring isotopes were taken from a compilation work, while unresolved resonance parameters were obtained by fitting to the total and capture cross sections calculated from nuclear models. A statistical model code was applied to evaluate cross sections above the resolved resonance region. Compound, pre-equilibrium and direct-reaction processes were considered for cross-section calculation. Coupled-channel optical model parameters were employed for the interaction between neutrons and nuclei. Giant-dipole and pygmy resonance parameters for E1 γ-ray transition from platinum isotopes were determined so as to reproduce measured γ-ray spectrum. The present results reproduce experimental data very well. The evaluated data are compiled into Evaluated Nuclear Data File formatted data files.     

Evaluation of neutron nuclear data for Technetium-99

Neutron nuclear data of ⁹⁹Tc was evaluated, considering cross-sections and spectra provided from recent experiments. The evaluation was made in the incident neutron energy range from 1 keV to 20 MeV, using the optical model and nuclear reaction models. The optical model calculation based on the coupled-channels method was performed for the interaction of neutrons with ⁹⁹Tc, and potential parameters appropriately chosen reasonably explain the measured data of total cross-section. The cross-section of inelastic scattering, capture, (n, 2n), (n, p), (n, α) and (n, n′α) reactions, and γ-ray emission spectra were calculated on the basis of statistical model with preequilibrium and direct components, and they were compared with available experimental data. It is found that the presently evaluated cross-sections and γ-ray emission spectra well reproduce those experimental values and that there is a large discrepancy among the present result and evaluated data for neutron emission spectra. The obtained capture cross-section increases at the energies below 1 MeV, relative to that in JENDL-4.0. This makes the transmutation efficiency of ⁹⁹Tc into stable ¹⁰⁰Ru by accelerator driven system enhanced. The production cross-section of ⁹⁹Mo important for the medical use of nuclear diagnostics reduces by 5–30% at the energies above 12 MeV, compared with JENDL-4.0.     

Estimate of (n,p) Cross Section for Unstable Nuclide 60Co by Multi-Step Hauser-Feshbach Model with Pre-Equilibrium Correction

Abstract

The cross section of the (n, p) reaction for unstable target nuclide ⁶⁰Co was estimated by a multi-step Hauser-Feshbach model code with pre-equilibrium correction. Parameters used in the model calculations were determined from the fitting of the cross sections to existing experimental data of (n, p), (n, α) and (n, 2n) cross sections and particle emission spectra for the adjacent stable nuclide ⁵⁹Co. The present result for ⁶⁰Co(n, p)⁶⁰Fe was compared with the values estimated from semi-empirical formulae around 14 MeV, The prediction accuracy of the present calculation is considered to satisfy the requirement for the fusion reactor applications. The theoretical model calculations with a reliable parameter set are recommended rather than the semi-empirical formulae for the important activation cross sections of exotic target nuclei including unstable ones.     

Excitation functions of (p,x) reactions on natural nickel up to 40 MeV

Production cross-sections of the residual radionuclides for the ^natNi(p,x)^{55,56,57,58 m + g}Co, ^56,57Ni nuclear reactions were measured up to 40 MeV by using a stacked-foil activation technique at the MC-50 cyclotron of the Korea Institute of Radiological and Medical Sciences. The results were compared with the reported experimental data as well as the theoretical calculations based on the TALYS and the ALICE-IPPE codes. The present results are in general good agreement with the other experimental data and calculated results. The integral yields for thick target were also deduced from the measured excitation functions of the produced radionuclides. The present experimental results will play an important role in enrichment of the literature data base for proton-induced reactions on natural nickel leading to various applications.     

Calculation of neutron cross-sections on copper-63 and -65

Toward the development of the next version of Japanese Evaluated Nuclear Data Library (JENDL) general-purpose file, we calculate neutron cross-sections on ^{63, 65}Cu from 50 keV to 20 MeV, which is the incident energy range above the resolved resonance region in JENDL-4.0. A dispersive optical model potential is adopted with a coupled-channel method for interaction between neutron and ^{63, 65}Cu. Direct, pre-equilibrium, and compound processes are taken into account in the calculation. All cross-sections, differential and double-differential cross-sections are consistently calculated with a single set of model parameters. The calculation results reproduce the measured data very well. In addition, disagreement between the calculated and experimental values seen in an integral test for the ⁶³Cu(n, α)⁶⁰Co reaction is improved by using the cross-section data obtained from the present work.