Double-Differential Neutron Emission Cross Sections of 6Li and 7Li at Incident Neutron Energies of 4.2, 5.4, 6.0 and 14.2 MeV

Energy-angle double-differential neutron emission cross sections of lithium isotopes were measured at incident neutron energies of 4.2, 5.4 and 14.2 MeV for ⁶Li and of 5.4, 6.0 and 14.2 MeV for ⁷Li using a time-of-flight spectrometer. Care was taken in background subtraction and in data correction for sample-size effects. Detailed comparison of the present results was made with the evaluated data in JENDL-3PR1. A spectrum fitting method was used to extract the ^6,7Li(n, n'x)α and (n, 2n) reaction cross sections. Neutrons emitted from the (n, 2n) reactions were well described by the conventional evaporation model. A simple calculation with a final-state Coulomb interaction was effectively applied for the ^6,7Li(n, n'x)α reactions. Angle-integrated cross sections of the ⁷Li(n, n't)α reaction were in good agreement with the JENDL-3PR1 data except the data measured at 6.0MeV. The angular distributions of elastically and inelastically scattered neutrons were successfully analyzed with the coupled-channel method at the incident neutron energy of 14.2 MeV.     

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.     

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 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-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 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.     

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.     

Measurement of Double-Differential Neutron Emission Spectra from Uranium-238

We have performed the measurement of neutron emission spectra from ²³⁸U using a time-of-flight technique, and deduced the following data; (1) the prompt fission neutron spectra for 2 MeV incident neutrons at two emission angles of 90° and 135°, (2) the double-differential neutron emission cross sections at the incident energies of 1.2, 2.0, 4.2, 6.1 and 14.1 MeV. The emission spectra and the cross sections for scattering process were also deduced by subtracting the fission neutrons from the experimental spectra. The experimental results were compared with other experiments and the evaluations of JENDL-3 and ENDF/B-IV.

From the fission spectrum data ranging from 2 to 12 MeV, we have derived the best fit parameters for the Maxwellian and Watt type distribution functions. The experimental spectra are described with the Maxwellian spectrum with temperature of 1.24–1.26 MeV and are softer than both evaluations.

The spectra and cross sections for inelastic-scattering showed substantial disagreement with the evaluations concerning the discrete levels between 0.5 and 1.2 MeV, and continuum neutrons due to evaporation and pre-equilibrium processes. The secondary neutron angular distributions at 14 MeV incident energy were reproduced fairly well with the systematics.     

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 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.     

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.     

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 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.     

Measurement of Leakage Neutron Spectra from a Spherical Pile of Zirconium Irradiated with 14MeV Neutrons and Validation of Its Nuclear Data

In order to make a benchmark validation of the nuclear data for Zr, the leakage neutron spectrum from a Zr sphere of a 61-cm diameter was measured between 0.1 and 16MeV using a time-of-flight technique with a 14MeV neutron source facility, OKTAVIAN. The result was compared with the calculation using the Monte Carlo code MCNP-4A. To investigate the spectrum dependence on the individual neutron reactions, test calculations were carried out with the MCNP-4A code using the JENDL-3.2-based libraries, in which partial cross section values were reduced from the original values. From the comparison between the measured and the calculated spectra, it was found that each of the results could predict well the experiment in general. However, in detail, both ENDF/B-VI and EFF-2.4 gave considerable overestimation above 1 MeV. The JENDL-3.2 predicts the spectrum almost satisfactorily except below 0.8 MeV and around 10 MeV. The discrepancy found in JENDL-3.2 calculation is considered due to the cross section values of the (n, 2n) reaction and its secondary energy distributions (SED). The modified JENDL-3.2 library with the reduced (n, 2n) reaction values and the lower SED below 1 MeV reproduced the experiment with better agreement over the whole energy range.     

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 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.     

Measurement and analysis of reaction rates in a large spherical depleted uranium pile surrounding a 14-MeV neutron source

Measurement was made of the reaction rate distributions for ²³⁸U(n, f), ²³⁵U(n, f), ²³⁸U(n, γ), ²⁷Al(n, α) and ⁵⁸Ni(n, p) in a large depleted uranium (DU) pile. The pile consisted of DU blocks forming a spherical shell of 45.72 cm radius and 40.64 cm thick. 14-MeV neutrons were generated at the center. Fast neutron leakage spectrum was also measured by an NE-213 spectrometer. In order to assess the ²³⁸U neutron cross sections of JENDL-2, the experiment was analyzed using the Monte-Carlo transport code MCNP with continuous energy cross sections. The agreement between the calculation and the experiment is generally unsatisfactory. The ratios of calculation to experiment of low energy reactions decreased with the thickness of the DU layer. The analysis of the Livermore pulsed sphere experiment for the small DU sphere revealed underestimation of leakage neutron spectrum around 10 MeV. The ²³⁸U cross sections of JENDL-2 should be improved for 14-MeV neutronics calculation.     

Production of high-purity medical radio isotope 64Cu with accelerator-based neutrons generated with 9 and 12 MeV deuterons

We conducted a feasibility study for producing a high-purity medical radioisotope ⁶⁴Cu from natural zinc with accelerator-based neutrons. ⁶⁴Cu isotopes were produced via the ⁶⁴Zn(n,p) reaction. The accelerator-based neutrons were generated via the C(d,n) reaction using low-energy deuterons of 9 and 12 MeV on a 1-mm-thick carbon target. First, the production purity was estimated using the evaluated nuclear data library JENDL-4.0 and our previously measured thick target neutron yield. We found that even when natural zinc was used as the starting material, significantly high-purity ⁶⁴Cu could be obtained. Next, irradiation experiments for producing ⁶⁴Cu using natural zinc were conducted at Kyushu University Tandem Laboratory, with the amounts of ⁶⁴Cu isotopes and other gamma-emission nuclides measured by a high-purity germanium detector. As a result, high-purity ⁶⁴Cu isotopes of 1.11(49) × 10⁰ and 3.70 (17) × 10⁰ Bq/g/μC were produced with incident deuteron energies of 9 and 12 MeV, respectively.     

Measurement and Analysis of Neutron Spectra in Lithium Fluoride and Polytetrafluoroethylene Piles

For the assessment of neutron cross section data for fluorine, angular neutron spectra in the lithium fluoride (LiF) and polytetrafluoroethylene ((CF₂)n) piles were measured in the energy range from a few keV to a few MeV by the time-of-flight method with an electron linac, and the results were compared with those calculated by using nuclear data from JENDL-2 and ENDF/B-IV. Spatial distributions of neutron and X-ray fluxes were also measured in the test piles by the activation method, and the influence of photoneutrons generated in the sample material on the neutron spectrum in each pile was estimated. As a result, it was found that their influence on the neutron spectrum shape below 1 MeV was not so large as was necessary to be taken into account for the present assessment.

The calculated spectra using the JENDL-2 data and the ENDF/B-IV data show generally good agreement with those measured in both piles. However, both calculations underestimate the neutron fluxes around several 100 keV, and overestimate those below 100 keV, when they are normalized in the energy range of 10 keV～1 MeV. Large discrepancies are found between the shapes of the measured and calculated spectra around the resonances of fluorine cross section below 100 keV. The present measurements and analyses suggest that the reevaluations of the inelastic and elastic scattering cross sections below 1MeV and the resonance cross sections below 100 keV are necessary to reduce the observed discrepancies.