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 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 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.     

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.     

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.     

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.     

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 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.     

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.     

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.     

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 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.     

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.     

Application of BGO Scintillators to Absolute Measurement of Neutron Capture Cross Sections between 0.01 eV and 10 eV

Applying a total energy absorption γ-ray detector composed of 12 bricks (5 × 5 cm², 7.5 cm thick) of BGO scintillators, the absolute measurement of capture cross sections for Au and Sb has been made in an energy region between 0.01 and 10eV using the linac time-of-flight method. Incident thermal neutron flux was absolutely determined by using the BGO detection system with a Sm sample. To extend the neutron flux measurement from the thermal neutron region to higher neutron energies, the ¹⁰B(n, αγ) reaction was applied. Absolute capture yield for the relevant capture sample was obtained by the saturated capture yield at a large resonance of the sample.

Gold was selected to investigate the application of the BGO detection system to the absolute measurement of the capture cross sections, since the ¹⁹⁷Au(n, γ)¹⁹⁸ Au reaction cross section is a well known standard one. The result of the ¹⁹⁷Au(n, γ)¹⁹⁸ Au reaction cross section showed good agreement with the evaluated data in JENDL Dosimetry File and ENDF/B-VI. Then, the detection system was applied to the Sb(n, γ) cross section measurement. Antimony has a large scattering-to-capture cross section ratio comparing to that of gold. The result showed good agreement with the evaluated data in JENDL-3.2 and ENDF/B-VI.     

Measurements of Thermal Neutron Cross Section and Resonance Integral for 237Np(n, γ)238Np Reaction

Making use of a standard neutron spectrum field with a pure Maxwellian distribution, the thermal neutron cross section for the ²³⁷Np(n, γ)²³⁸Np reaction was measured at a neutron energy of 0.0253 eV by the activation method. The result is 158±3 b, which is obtained relative to the reference value of 98.65±0.09 b for the ¹⁹⁷Au(n, γ)¹⁹⁸Au reaction. Although the data in JENDL-3 is larger by about 15% than the present value, the recently revised data in JENDL-3.2 is close to the present. The ENDF/B-V, ENDF/B-VI, JEF-2 and Mughabghab's data are also larger by 7–15%. Old measurements are larger by 7–18% than the present data.

The resonance integral for the ²³⁷Np(n, γ)²³⁸Np reaction was also measured relative to the reference value of 1,550±28 b for the ¹⁹⁷Au(n, γ)¹⁹⁸Au reaction with a 1/E standard neutron spectrum field. By defining the Cd cut-off energy as 0.5 eV for the ²³⁷Np(n, γ)²³⁸Np reaction, the present resonance integral is 652 ± 24 b, which is in good agreement with the JENDL-3, -3.2, ENDF/B-V, -VI, JEF-2 and Mughabghab's data. However, most of the old experimental data are, in general, larger by 24–38% than the present measurement.     

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 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 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.