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 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.     

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.     

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 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.     

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 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.     

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.     

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.     

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.     

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.     

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 of praseodymium-141 and -143

For the next version of JENDL general purpose file, neutron nuclear data on ^141,143Pr are evaluated considering cross sections and spectra provided from experiments in the fast neutron energy region up to 20 MeV. Total and elastic cross sections are derived from optical model. Pre-equilibrium and direct-reaction processes are taken into account in addition to the compound process. The statistical model is applied to calculate the cross sections above the resolved resonance region. For unstable ¹⁴³Pr nuclide having less experimental data, we use the same parameters as ¹⁴¹Pr for mass-dependent and -independent terms in the optical potential. The resolved resonance parameter remains unchanged from JENDL-4.0. The present evaluations give consistent results with experimental data, reflecting newly measured one. The evaluated data are compiled in ENDF-6 format for the next version of JENDL.     

Neutron Excess Generation by Fusion Neutron Source for Self-Consistency of Nuclear Energy System

The present day fission energy technology faces with the problem of transmutation of dangerous radionuclides that requires neutron excess generation. Nuclear energy system based on fission reactors needs fuel breeding and, therefore, suffers from lack of neutron excess to apply large-scale transmutation option including elimination of fission products. Fusion neutron source (FNS) was proposed to improve neutron balance in the nuclear energy system. Energy associated with the performance of FNS should be small enough to keep the position of neutron excess generator, thus, leaving the role of dominant energy producers to fission reactors. The present paper deals with development of general methodology to estimate the effect of neutron excess generation by FNS on the performance of nuclear energy system as a whole. Multiplication of fusion neutrons in both non-fissionable and fissionable multipliers was considered. Based on the present methodology it was concluded that neutron self-consistency with respect to fuel breeding and transmutation of fission products can be attained with small fraction of energy associated with innovated fusion facilities.