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.     

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.     

Calculation of cross sections for metastable state production in the (n, γ), (n,n′), (n, 2n) and (n, 3n) reactions of 93Nb

Toward the revision of ⁹³Nb data in the Japanese Evaluated Nuclear Data Library JENDL-4.0, we calculated cross sections for metastable state production (MSP) in the (n, γ), (n, n′), (n, 2n) and (n, 3n) reactions in the incident energy (E_n) range from 7 keV to 20 MeV. The cross sections were evaluated using nuclear reaction models such as the spherical optical model, the multi-step statistical model, preequilibrium models, and the distorted-wave Born approximation (DWBA). By adjusting parameters of nuclear level densities, we could obtain the MSP cross sections which are almost consistent with the experimental data.     

Evaluation of neutron nuclear data on xenon isotopes

Neutron nuclear data of Xe isotopes have been evaluated in the energy region, including the resolved resonance one, from 1 keV to 20 MeV by using the theoretical nuclear reaction models. The phenomenological optical model potential was employed to calculate the total cross section for natural Xe with the coupled-channels method. The cross sections for channels of capture, (n, 2n), (n, p) and (n, α) reactions were calculated and compared with available experimental results including recently measured data. The elastic scattering angular distributions and particle emission spectra were calculated, although there is no experimental information available. Reaction cross sections of evaluated libraries were considered for comparison with the calculated results. The presently calculated cross sections reproduce better the available experimental 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 of Natural Silver and Its Isotopes

Neutron nuclear data of natural silver and its isotopes (¹⁰⁷Ag and ¹⁰⁹Ag) have been evaluated in the energy range of 10^?5eV–20 MeV. Evaluated quantities are the total, elastic and inelastic scattering, capture, (n, Zn), (n, 3n), (n, p), (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 based on available experimental data and theoretical calculations. Multi-step Hauser-Feshbach calculation played an important role in the determination of the reaction cross sections. In the calculation, the precompound process was taken into account above 5MeV, in addition to the compound one. The evaluated data have been compiled into JENDL-3 in the ENDF-5 format.     

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.     

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.     

Evaluation of Neutron Cross Sections of Plutonium-241

The neutron cross sections of ²⁴¹Pu were evaluated in the energy range between 10^?5 eV and 15MeV, and are stored in the Japanese Evaluated Nuclear Data Library Version-1 (JENDL-1). In the energy range below 100eV, the evaluated data contained in ENDE/B-IV and the resonance parameters recommended in BNL-325 were tentatively adopted. The unresolved resonance parameters were determined between 100 eV and 21.5 keV so as to reproduce the experimental data of the fission and capture cross sections. Above 21.5 keV, the fission cross section was evaluated on the basis of the experimental data, most of which were reported as the ratio to the fission cross section of ²³⁵U and then were normalized by the fission cross section of ²³⁵U adopted in JENDL-1. The capture cross section was obtained from the experimental data of a in the energy range up to 250 keV. The capture cross section above 250 keV and the elastic and inelastic scattering, (n, 2n) and (n, 3n) reaction cross sections above 21.5 keV were obtained on the basis of the theoretical calculations. The calculated cross sections are connected smoothly with those obtained from the unresolved resonance parameters at 21.5 keV. This suggests the self-consistency of the present evaluation.     

Measurement of Gamma-Ray Production Cross Sections of Iron for Incident Neutron Energies between 6 and 33 MeV

Measurement of differential γ-ray production cross sections, i.e. (n, x γ) cross sections, of Fe was made for neutron energies from 6 to 33 MeV. Neutrons used in the experiment were white neutrons produced with (p, n) reactions by 35 MeV protons using a thick Be target. The neutron energy was analyzed by the time-of-flight method and bunched into 3 MeV wide energy bins, for each of which the spectrum of secondary γ-rays produced in an Fe sample was measured by a BGO scintillator at an angle of 144° to the neutron beam direction.

The obtained (n, xγ) cross sections agreed well with other data and the evaluated data file of ENDF/B-IV at neutron energies below 15 MeV where data were existing. The JENDL-3 file overestimated the γ-ray spectra at γ-ray energies of 3 to 7 MeV. The present work newly provided the data in the neutron energy range above 20 MeV. The GNASH calculation made by Young reproduced the measured data fairly well even at these higher energies.     

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.     

High Resolution Measurements of Double Differential (n, α) Cross Sections of 58Ni and natNi between 4.2 and 6.5 MeV Neutrons

Double differential (n, α) reaction cross sections of ⁵⁸Ni and ^natNi were measured for 4.2–6.5 MeV neutrons with energy resolution good enough to separate α-particles from the low-lying levels of residual nuclei by using a gridded ionization chamber. Angular distribution and excitation functions were derived for α₀, α₁ and α_i≥2 components (α-particles to the ground level, the 1st level and levels higher than the 2nd level, respectively). The experimental results were compared with these obtained from calculation based on Hauser-Feshbach model employing the optical potential and the level density parameters derived to reproduce the experimental values of total, (n,p) and (n,α)cross sections. The calculation showed fair agreement with the experimental data while it underestimated the (n,α)cross section above 6 MeV.     

Calculation of Neutron-Induced Reaction Cross Sections of Manganese-55

Neutron-induced reaction cross sections of ⁵⁵Mn are calculated for the evaluated nuclear data libraries, ENDF/B-VI and JENDL-3. Simultaneously calculated are the inelastic scattering, (n,2n), (n,p), (n,α), (n,np), (n,nα) and (n,γ) reaction cross sections, the angular distributions of emitted neutrons and protons, and the energy distributions of emitted particles and γ-rays. A unified Hauser-Feshbach code is applied to calculate these quantities. Precompound and direct-reaction processes are taken into consideration, in addition to the compound process. The calculated results reproduce the experimental data very well. Forward-peaked angular distributions of continuous neutrons are obtained from the calculation, and found to be consistent with the measurements at 14MeV.     

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.     

Systematic investigation of spectroscopic factors from (d,p) reactions for deuteron nuclear data evaluation

Spectroscopic factors for the (d,p) reactions on ¹²C, ²⁷Al, ⁴⁰Ca, and ⁵⁸Ni for incident deuteron energies up to 100 MeV were extracted systematically by fitting theoretical distorted wave Born approximation calculations to the existing experimental data. The extracted spectroscopic factors show similar incident energy dependence for all the target nuclei. An empirical expression describing the energy dependence has been deduced and implemented in a computational code system developed specially for deuteron-induced reactions. The calculation with the code system reproduces the measured ²⁷Al(d,p)²⁸Al cross sections in the incident energy range from the threshold to 20 MeV and the measured double-differential cross sections for (d,xp) reactions on ¹²C and ⁵⁸Ni at 56 and 100 MeV fairly well.     

Measurements of Double-Differential Neutron Emission Cross Sections of 6Li, 7Li and 9Be for 11.5 MeV and 18.0 MeV Neutrons

Double-differential neutron emission cross sections (DDXs) of ⁶Li, ⁷Li and ⁹Be were measured for 18.0 MeV and 11.5 MeV incident neutrons produced by the T(d, n) and ¹⁵N(d, n) reactions respectively, using the Tohoku University Dynamitron time-of-flight (TOF) spectrometer. The data were obtained at 13 laboratory angles, and angular-differential cross sections (ADXs) of elastic and inelastic scattering neutrons were derived from the DDXs. For 11.5 MeV neutrons, we obtained the neutron emission spectra over the secondary neutron energies by newly employing the double TOF method as well as the conventional one. In the measurements at 18.0 MeV, we achieved better energy resolution than in our previous studies by using a neutron detector that has a larger solid angle and a thinner tritium target. The experimental results of DDXs and ADXs were compared with our previous results and the evaluated data given in JENDL-3.2, JENDL Fusion File and ENDF/B-VI. It is found that the JENDL data reproduce the experimental ones very well.     

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.     

p+~(107,109)Ag核反应数据的计算和分析

本工作通过理论计算的方法获得了一套适合入射能量为从阈值到200 MeV的p+107,109 Ag核反应全套微观数据。首先,使用光学模型理论进行调参计算,得到了一套适合入射能量为从阈值到340MeV的p+107,109 Ag核反应Becchetti-Greenlees光学势参数,这套参数与实验数据符合很好。其次,在这套光学势参数的基础上用扭曲波玻恩近似对入射能量从阈值到200MeV的p+107,109 Ag直接非弹性散射截面进行了计算。最后,使用核反应统计理论计算了入射能量从阈值到200MeV的p+107,109 Ag核反应各反应道的截面和出射粒子能谱,得到了该能区p+107,109 Ag核反应全套微观数据。将所有计算值与实验数据进行比较,结果表明,所得到的全套微观数据与实验数据符合很好。     

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.     

Theoretical calculation of neutron cross sections for 90,91,92,94,96Zr in the incident energy range between 200 keV and 20 MeV

Neutron cross sections of ^{90,91,92,94,96}Zr were calculated in the incident energy (E_n) range from 200 keV to 20 MeV for the revision of the 4th version of the Japanese Evaluated Nuclear Data Library (JENDL-4.0). The calculation was carried out by using conventional nuclear reaction models such as the spherical optical model, the distorted wave Born approximation, preequilibrium models, and the multi-step statistical model. Parameter values of these nuclear models were adjusted with the aid of experimental cross sections which were published after the JENDL-4.0 evaluation. Cross sections were computed for total, elastic and inelastic scattering, (n, γ), (n, 2n), (n, p), (n, α), (n, nα), and (n, x) = (n, d) + (n, np) reactions, and they were almost consistent with the experimental data. The cross sections were also estimated for the metastable states with the half-life larger than 1 sec. The obtained results well reproduced measured cross sections for the reactions ⁹⁰Zr(n, 2n)^89mZr, ⁹¹Zr(n, x)^90mY and ⁹¹Zr(n, nα)^87mSr.