New Evaluation of Neutron Nuclear Data for Zinc Isotopes

New evaluation of neutron-induced nuclear data for five stable isotopes of zinc (mass numbers A = 64, 66, 67, 68, and 70) was consistently carried out in the incident neutron energy range from 10^?5 eV up to 20MeV. In the low energy region up to about 100keV, the resonance parameters were evaluated by taking account of the available measured data. In the fast neutron region, the comprehensive calculations with nuclear reaction models, in which compound, preequilibrium, and direct processes are taken into account, were performed to estimate cross sections for various reactions and double differential cross sections of emitted neutrons and γ-rays. The comparisons of the evaluated cross sections with the experimental data and existing evaluated nuclear data libraries are made and show a good agreement with the measurements.     

Evaluation of Neutron Nuclear Data for Mercury

Neutron nuclear data of stable mercury isotopes (¹⁹⁶Hg, ¹⁹⁸Hg, ¹⁹⁹Hg, ²⁰⁰Hg, ²⁰¹Hg, ²⁰²Hg and ²⁰⁴Hg) have been evaluated in the energy range of 10^?5eV–20MeV. Evaluated quantities are the total, elastic and inelastic scattering, capture, (n, 2n), (n, 3n), (n, p), (n, d), (n, α), (n, np), (n, nα) reaction and γ-ray production cross sections, the resonance parameters, and the angular and energy distributions of emitted neutrons and γ-rays. The evaluation is mainly based on nuclear reaction model calculations. Statistical-model calculation played a significant role in the determination of the reaction cross sections. The evaluated data have been compiled in the ENDF-6 format, and are used for the design study of a mercury target system proposed at the Neutron Science Research Center, Japan Atomic Energy Research Institute.     

Neutron Nuclear Data Evaluation of Cesium Isotopes for JENDL-4.0

For the development of JENDL-4.0, neutron nuclear data for fission product nuclides, ^{133,134,135,136,137}Cs, were revised in the incident neutron energy range from 1 eV to 20MeV by using a coupled-channels optical model (OM), and nuclear reaction models. The OM potential parameters were determined for stable ¹³³Cs to reproduce the experimental data of total and elastic scattering cross sections and angular distributions of elastically scattered neutrons. The present results reasonably reproduce measured data for (n; 2n), (n; p), (n; α), and capture reactions on ¹³³Cs. Important differences between the present results and JENDL-3.3 are found for the capture cross sections of ^134,137Cs. The cross section obtained for ¹³⁷Cs was smaller than that in JENDL-3.3. This result makes the transmutation of medium-lived ¹³⁷Cs increasingly difficult. The production probabilities of metastable states for ^134,138Cs via capture reactions on ^133,137Cs are compared with experimental values. The present result for ^134m Cs production is marginally consistent with measured data. However, a large discrepancy is recognized for ^138m Cs production. The γ-ray emission data were evaluated with available measurements, and newly compiled in JENDL-4.0. Maxwellian-averaged capture cross sections were calculated in the energy range from 1 to 103 keV, and are compared with other derived data.     

Evaluation of Neutron Nuclear Data of Natural Nickel and Its Isotopes

Neutron nuclear data of natural nickel and its isotopes have been evaluated. Evaluated are the total, elastic and inelastic scattering, capture, (n, 2n), (n, 3n), (n, p), (n, α),(n, n'p) and (n, n'α) reaction cross sections, the resonance parameters, the angular and energy distributions of secondary neutrons in the energy range of 10^?5 eV～20 MeV. The evaluation has been made on the basis of recently measured data with the aid of the spherical optical model and statistical model. The results of the benchmark tests of JENDL-1 have been also taken into consideration. Special care has been taken on the background cross sections in the resonance region, the remaining resonance structure in the unresolved resonance region up to a few MeV, and grouping of the inelastic scattering levels in the natural nickel file. The problems left for future work are also discussed. The results of the present evaluation were adopted in JENDL-2.     

Evaluation of Neutron Nuclear Data for Sodium-23

Neutron nuclear data of ²³Na have been evaluated in the neutron energy region up to 20 MeV. Evaluated are the elastic and inelastic scattering, capture, (n, 2n), (n, p), (n, α), (n, np), (n, nα) reaction and γ-ray production cross sections, and the angular and energy distributions of neutrons and γ-rays. The evaluation is mainly based on nuclear model calculations. The pre-equilibrium and direct-reaction processes were taken into account in addition to the compound process. The evaluated data have been compiled into the latest version of JENDL, JENDL-3.3.     

Calculation of Neutron Nuclear Data on Molybdenum Isotopes for JENDL-4

Neutron nuclear data on ⁹²,⁹⁴,⁹⁵,⁹⁶,⁹⁷,⁹⁸,⁹⁹,¹⁰⁰ Mo have been calculated for the evaluated nuclear data library JENDL-4. Simultaneously calculated are the total, elastic and inelastic scattering, (n, p), (n, d), (n, t), (n, ³He), (n, α), (n, np), (n, nd), (n; nα), (n, 2n), (n, 3n) reaction cross sections, the angular distributions of emitted particles, and the energy distributions of emitted particles and γrays. The statistical model was applied to calculate these quantities. Coupled-channel optical model parameters were used for neutrons. Preequilibrium and direct-reaction processes were taken into account in addition to the compound process. The present calculations are almost consistent with available experimental data. The calculated results are compiled into JENDL-4.     

Evaluation of Neutron Nuclear Data on Bromine and Krypton Isotopes for JENDL-4.0

Neutron nuclear data on ^79,81Br and ^{78,80,82,83,84,85,86}Kr have been evaluated for the evaluated nuclear data library JENDL-4.0 in the energy region from 10^?5 eV to 20 MeV. The resolved resonance parameters were adjusted so as to reproduce recommended or measured thermal capture cross sections for some isotopes. The statistical model was applied to calculate the cross sections above the resolved resonance region. In the calculations, coupled-channel optical model parameters were used for neutrons. Preequilibrium and direct-reaction processes were taken into account in addition to the compound process. The present evaluation is consistent with available experimental data. The evaluated results were compiled into JENDL-4.0.     

Calculation of Neutron Nuclear Data on Rubidium and Strontium Isotopes for JENDL-4

Neutron nuclear data on ^85,86,87Rb and ^{84,86,87,88,89,90}Sr have been calculated for the evaluated nuclear data library JENDL-4 in the energy region from 10 keV to 20MeV. Simultaneously calculated are the total, elastic, and inelastic scattering, (n,γ), (n, p), (n, d), (n, t), (n,³He), (n,α), (n, np), (n, nd), (n, nα), (n, 2n), (n, 3n) reaction cross sections, angular distributions of emitted particles, and energy distributions of emitted particles and γ-rays. The statistical model was applied to calculate these quantities. Coupledchannel optical model parameters were used for neutrons. Preequilibrium and direct-reaction processes were taken into account in addition to the compound process. The present calculations are consistent with available experimental data. The calculated results are compiled into JENDL-4.     

Evaluation of Neutron Nuclear Data for Uranium-233

Neutron nuclear data of ²³³U have been evaluated in the energy range from 10-⁵ eV to 20 MeV. Evaluated quantities are the total, fission, capture, elastic and inelastic scattering, (n,2n) and (n,3n) reaction cross sections, and the average numbers of prompt and delayed neutrons emitted per fission. The thermal and resonance cross sections have been evaluated on the basis of the measured data. The resolved resonance parameters are given up to 100 eV and the unresolved resonance parameters between 100 eV and 30keV. The total and fission cross sections have been evaluated in the higher energy region on the basis of the recently measured data, while the theoretical calculation with the optical, statistical and evaporation models has been used for evaluation of the other cross sections. The presently adopted optical potential parameters have reproduced well the experimental total cross section in the entire energy range as well as the measured data of the s-wave strength function. The structure observed in the vp values below 1 MeV is reproduced by the semi-empirical formula based on the fission fragment kinematics. The presently evaluated fission cross section is considerably lower than that of ENDF/B-IV between 10 and 50keV. This low fission cross section is expected to resolve the ^Keff discrepancy pointed out from the benchmark tests in ²³³U critical assemblies.     

Japanese Evaluated Nuclear Data Library Version 3 Revision-2: JENDL-3.2

The revision work of JENDL-3 has been made by considering feedback information of various benchmark tests. The main revised quantities are the resonance parameters, capture and inelastic scattering cross sections, and fission spectra of main actinide nuclides, the total and inelastic scattering cross sections of structural materials, the resonance parameters the capture and inelastic scattering cross sections of fission products, and the γr-ray production data. The revised data were released as JENDL-3.2 in June 1994. The preliminary benchmark tests indicate that JENDL-3.2 predicts various reactor characteristics more successfully than the previous version of JENDL-3.1.     

Neutron Nuclear Data Evaluation for Thorium-232

An evaluation was made on the neutron cross sections, resonance parameters and average neutron yield in fission for ²³²Th in the energy range from thermal energy to 20 MeV. The fission and capture cross sections were evaluated on the basis of the experimental data by converting the relative ratio data into cross section values by making use of recent evaluations for reference cross sections. The total cross section was determined from experimental data in the region from 24 keV to 15 MeV and then extrapolated to lower and higher energies by using the optical model whose parameters had been adjusted as so to reproduce the measured data. The elastic and inelastic scattering, (n, 2n) and (n, 3n) reaction cross sections were calculated by means of the statistical model combined with the optical model. A set of resonance parameters were recommended in the energy range below 3.5 keV and average resonance parameters were deduced in the unresolved resonance region. A value of 7.40 b was chosen for the capture cross section at 0.025 eV, and the picket-fence negative-energy levels were introduced so as to reproduce the non-l/v behavior of the capture cross section in the epithermal region.

The results were incorporated in the Japanese Evaluated Nuclear Data Library, Version 2 (JENDL-2). Comparison was made between the present and other evaluations such as ENDF/B-V and possible reasons for the discrepancy were discussed.     

Neutron Resonance Parameters of Silver-107 and Silver-109

Neutron transmission measurements were carried out on the separated isotopes of silver using the time-of-flight facility at the Japan Atomic Energy Research Institute electron linear accelerator. Neutrons were detected with the ⁶Li-glass detectors at 56 and 191 m. The samples used were metallic powder enriched to 98.2% for ¹⁰⁷Ag and 99.3% for ¹⁰⁹Ag. Transmission data were analyzed with the multi-level Breit-Wigner formula incorporated in a least squares fitting program. Resonance energies and neutron widths were determined for the large number of resolved resonances in the neutron energy region of 400 eV~7 keV. The s-wave strength functions and average level spacings were obtained to be; S₀= (0.43±0.05) × 10^?4, D₀ = 20±2 eV for ¹⁰⁷Ag and S₀= (0.45 ± 0.05) × 10^?4, D₀ = 20 ± 2eV for ¹⁰⁹Ag.     

Evaluation of Neutron Nuclear Data for Z>88 Minor Nuclides

Neutron nuclear data for 15 minor nuclides (Z>88) have been evaluated in the energy range of 10^?5 eV–20 MeV. Since only few experimental data are available, the present evaluation was mainly based on the systematics of the data from neighboring nuclides and also optical and statistical model calculations. The evaluations have been carried out for neutron cross sections of total, elastic scattering, inelastic scattering, (n, 2n), (n, 3n), (n, 4n), fission and capture reactions. In addition, angular and energy distributions of the emitted neutrons and average number of the emitted neutrons per fission were also evaluated. The results were compiled in the ENDF/B-V format and stored in the JENDL-3.     

A function to provide neutron spectrum produced from the 9Be + p reaction with protons of energy below 12 MeV

A function to give the total neutron production cross section, angular distribution, and energy spectrum via the ⁹Be + p reaction has been created by fitting experimental data to characterize compact neutron sources with thick Be targets bombarded by protons with energy below 12 MeV. To examine the suitability of the function, calculations of the angle-dependent neutron energy spectra produced in thick Be targets with 4- and 12-MeV protons using the function were compared with corresponding experiments and calculations using the nuclear data libraries of ENDF/B-VII.0 and JENDL4.0/HE. The function was in better agreement with the experiments than the calculations using the libraries except for at backward angles. The ¹¹⁵In(n,n’)^115mIn reaction rates calculated using GEANT4 with source neutrons given by both the function and ENDF/B-VII.0 were compared with that measured at the RIKEN Accelerator-Driven Compact Neutron Source to evaluate the neutron spectrum above 1 MeV. The function slightly overestimated the measurement by 14% and the calculation with ENDF/B-VII.0 underestimated by 35%. It was concluded that the function can be applied in compact neutron source designs.     

Significance of Covariance Matrices in Nuclear Data Evaluation

Tritium decontamination using ultra violet (UV) lamp and laser was performed. Simulated co-deposited layer on tungsten substrate was deposited by C₂H₂ or C₂D₂ glow discharge. The co-deposited layer was irradiated to UV lights from a xenon excimer lamp (172 nm) or ArF excimer laser (193 nm) and the in-situ decontamination behavior was evaluated by a mass spectrometer. After the UV irradiation, the hydrogen concentration in the co-deposited layer was evaluated by elastic recoil detection analysis (ERDA) and the depth profile was analyzed by secondary ion mass spectrometry (SIMS).

For the co-deposited layer formed by C₂D₂ glow discharge, it was found that M/e 3 (HD) gas was released mainly during the UV lamp irradiation while both M/e 3 (HD) and M/e 4 (D₂) gases were detected during the UV laser irradiation. Though the co-deposited layer was not removed by UV lamp irradiation, almost all the co-deposited layer was removed by UV laser irradiation within 1 min. The ratio of hydrogen against carbon in the co-deposited layer was estimated to be 0.53 by ERDA and the number of photon needed for removing 1 fim thick co-deposited layer was calculated to be 3.7×10¹⁸ cm^-2 for the UV laser by SIMS measurement. It is concluded that C-H (C-D) bond on the co-deposited layer were dissociated by irradiation of UV lamp while the co-deposited layer itself was removed by the UV laser irradiation.     

Evaluation of neutron nuclear data for gallium

Neutron nuclear data on ^69,71Ga have been evaluated for the next version of the Japanese Evaluated Nuclear Data Library (JENDL) general purpose file in the energy region from eV to 20 MeV. The resolved resonance parameters at negative energies were adjusted so as to reproduce measured thermal capture cross sections. The statistical model was applied to calculate the cross sections above the resolved resonance region. In the calculations, coupled-channel optical model parameters were used for neutrons. Pre-equilibrium and direct-reaction processes were taken into account in addition to the compound process. The present evaluation is consistent with available experimental data. The evaluated data are compiled into an ENDF-formatted data file.     

Measurements and Calculations of Angular Flux Spectra Emitted from Lithium and Graphite Slabs with D-T Neutron Source

Angular dependent flux spectra from slab assemblies (lithium and graphite) were measured to test nuclear data and calculational methods for D-T fusion reactor neutronics. The collimated 14 MeV neutron source could be applied by the use of an associated particle method and the neutron spectra from 14 to 2 MeV were observed with TOF technique. The measured spectral pattern was dependent on the anisotropy of secondary neutrons emitted from both the elastic and the non-elastic scattering for 14 MeV neutrons. As for the numerical calculations, one-dimensional discrete ordinates transport codes (ANISN and NITRAN) were used. The multigroup cross sections processed with SPTG4Z from ENDF/B-IV were used as common nuclear data base. The problems of calculational methods and nuclear data were discussed in comparison with the experimental data and it was clarified that sufficient nuclear data of angular dependent cross sections for the non-elastic scattering have not been available in ENDF/B-IV and that the anisotropy of the scattering could not be calculated with ANISN which utilized the scattering kernel generated by incorrect treatment of scattering kinematics in the processing code. However, good agreement between the measurements and calculations was obtained by the use of NITRAN system with the appropriate processing codes of inelastic scattering anisotropies. It was shown that the NITRAN system was useful for anisotropic neutron transport calculations.