Neutron Total Cross Section for Iron around Resonance Valley near 24 keV

A neutron leakage spectrum from a nickel sphere surrounding a 14-MeV neutron source is measured and analyzed in order to verify the accuracy of nickel cross sections. Measurement is done by means of a time-of-flight technique in the range of 2–15 MeV using an NE213 scintillator, and compared with calculations carried out with MCNP, a continuous energy Monte Carlo transport code, using JENDL-3PR1 and ENDFIB-IV neutron libraries.

In spite of an overestimation of neutron flux near 13 MeV, the calculated result employing JENDL-3PR1 shows generally better agreement with the measured spectrum. In ENDF/B-IV usage, there is disagreement between measured and calculated spectra between 5 and 12 MeV. Problems in evaluated nuclear data for nickel are also described.     

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.     

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.     

Self-Shielding Factors for Neutron Capture Reactions of Uranium-238 and Thorium-232 in Energy Range of 1–35 keV

Self-shielding factors for the neutron capture reactions of ²³⁸U and ²³²Th were measured in the resonance energy region of 1–35 keV, using a neutron time-of-flight method with an electron linear accelerator. The self-shielding factors for arbitrary dilution cross sections were obtained from sets of neutron transmission ratios and self-indication ratios measured with several transmission samples of different thicknesses. The maximum experimental errors for ²³⁸U and ²³²Th were about 3 and about 7%, respectively.

The experimental results were compared with calculations based on JENDL-2, JENDL-3 and ENDF/B-IV. For ²³⁸U, an energy dependent structure was observed in the experimental self- shielding factors. The calculations based on JENDL-2 and ENDF/B-IV did not show this structure in the unresolved resonance region and were smaller than the experimental values from 4 to 6 keV. The calculation based on the resolved resonance parameters in JENDL-3 showed better agreement with the experiment from 4 to 6 keV, but discrepancies still remained in other energy ranges.

For ²³²Th, no remarkable discrepancy was observed in the unresolved resonance region, but JENDL-2 and JENDL-3 tended to give smaller values than the experimental self-shielding factors in the resolved resonance region.     

Geometrical Interpretation of Mutual Relatonship between Ordinary Coherence and Noise Power Contribution Ratio

Angular neutron fluxes leaking from the surface of lithium-oxide and graphite slab assemblies have been measured with irradiation of D-T neutrons. The spectrum measurement was performed using the time-of-flight technique with an NE213 scintillation detector. The thicknesses of the slabs were 0.6 to 5 mean free path for 14.8 MeV neutrons, and the measured leaking angles of the angular fluxes were 0.0°, 12.2°, 24.9°, 41.8° and 66.8°. The experimental results have been compared with the results calculated by the continuous energy Monte Carlo transport code MCNP, using the data in the JENDL-3PR1, ?3PR2, and ENDF/B-V nuclear data files. The comparisons between the experimental and calculated results show that the data of ⁷Li in JENDL-3PR2 is improved for the secondary emission spectra of the 4.63 MeV level and (n, 2n) reactions; the angular distributions of 3rd-and 4th-level inelastic reactions of C in the JENDLs are questionable. The thickness dependences for high energy neutrons also suggest that the total cross section of ⁷Li and the elastic cross sections of C are slightly inadequate.     

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.     

Cross Section Assessment by Measurement of Time Dependent Neutron Spectra from Lithium Assemblies

Time dependent neutron spectra from lithium assemblies were measured to assess the neutron cross sections of ⁷Li in ENDF/B-IV, which is important nuclide for the D-T fusion reactor blanket material. Pulsed neutrons produced by D-D or D-T reaction were used to measure leakage neutron spectra from cubical lithium assemblies as a function of time by the use of NE213 liquid scintillator. Calculations of time dependent neutron spectra were carried out by the Monte Carlo code SIMON, which was prepared for this study. The group constants used in these calculations were processed from ENDF/B-IV data. The calculated and the measured neutron spectra were compared for the following three; a stationary spectrum, spectra at each time interval and decay curves for specified energy groups. Discrepancies between the measured and the calculated neutron spectra were found in these comparisons. In order to assure the cause of these discrepancies, some calculations were carried out with recently measured cross sections of inelastic scattering which excite 0.478 and 4.63 MeV level of ⁷Li. It was concluded that some of the neutron cross section data of ⁷Li in ENDF/B-IV should be ameliorated.     

Assessment of Neutron Cross Sections for Titanium with Neutron Spectrum in Titanium Pile

The measurement of angular neutron spectrum in a quasi-spherical pile of Ti was carried out by the linac time-of-flight method for the assessment of neutron cross sections for Ti in the energy range from a few keV to a few MeV. The measured spectrum in the pile is generally in good agreement with the calculated one from ENDF/B-IV (MAT = 1,286 for Ti) except in the energy range from about 60 keV to a few 100 keV, where the calculation gives considerably lower neutron flux than the measurement.

In order to investigate the cause of this discrepancy between the measured and calculated spectra, the total cross sections for Ti were measured by the transmission method. The results give larger values of total cross sections for Ti by about 30% than ENDF/B-IV below 200 keV, and smaller values by about 10% above 200 keV. These results were ensured at 55 and 147 keV by the measurement using a Si-filtered neutron beam. The calculation based on the measured cross sections shows better agreement with the measured spectrum than that based on ENDF/B-IV. The discrepancy is still observed around 100 keV.

The sensitivity analysis shows the importance of cross sections above 1 MeV and elastic cross sections in the resonance energy region to solve the disagreement between the measured spectrum and the calculated one.     

Benchmark experiment of large-angle scattering reaction cross section of iron at 14 MeV using two shadow bars – Comparison of experimental results with ENDF/B-VIII –

ABSTRACT

It is important to perform neutron transport simulations with accurate nuclear data in the neutronics design of a fusion reactor. However, absolute values of large-angle scattering cross sections vary among nuclear data libraries even for well-examined nuclide of iron. Benchmark experiments focusing on large-angle scattering cross sections were thus performed to confirm the correctness of nuclear data libraries. The series benchmark experiments were performed at a DT neutron source facility, OKTAVIAN of Osaka University, Japan, by the unique experimental system established by the authors’ group, which can extract only the contribution of large-angle scattering reactions. This system consists of two shadow bars, target plate (iron), and neutron detector (niobium). Two types of shadow bars were used and four irradiations were conducted for one experiment, so that contribution of room-return neutrons was effectively removed and only large-angle scattering neutrons were extracted from the measured four Nb reaction rates. The obtained experimental results were compared with calculations for five nuclear data libraries including JENDL-4.0, JEFF.-3.3, FENDL-3.1, ENDF/B- VII, and recently released ENDF/B-VIII. It was found from the comparison that ENDF/B-VIII showed the best result, though ENDF/B-VII showed overestimation and others are in large underestimation at 14 MeV.     

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.     

Measurements and Analyses of Reactivity Effect of Fission Product Nuclides in Epithermal Energy Range

Experimental data usable for evaluating cross sections of main fission product elements (Rh, Cs, Nd, Sm, Eu and Gd) in the epithermal energy range were measured. A cadmium-covered vessel containing a pure water or an aqueous solution of a fission product element was inserted at the center of TCA (Tank-type Critical Assembly) core. Reactivity effects were obtained by the difference in the critical water levels between a pure water and an aqueous solution in the vessel. The measured reactivity was more than 1 φ and it was greater than the experimental uncertainties. Since the adjoint thermal flux below the cadmium-cutoff energy are largely depressed in the vessel, the reactivity effects in epithermal energy range could be measured. The analyses for the experiments were performed using the SRAC code system and neutron transport calculation code TWOTRAN. The exact Perturbation theory was applied to calculate the reactivity effects of fission product elements. The calculated reactivity effects using JENDL-3.2 and ENDF/B-IV cross-section libraries were compared against the measured values. The analyses using JENDL-3.2 gave reasonable results for these measurements.     

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.     

Measurement of keV-Neutron Capture Cross Section of 96Zr

The neutron capture cross section of ⁹⁶Zr at incident neutron energies from 15 to 100 keV has been measured by the time-of-flight method. Capture γ-rays were detected with an anti-Compton NaI(Tl) spectrometer, and the pulse-height weighting technique was applied to derive the neutron capture cross section. The present measurement provided the capture cross section as a function of incident neutron energy in the keV region. The results were compared with previous measurements and cross section data in the evaluated nuclear data libraries, JENDL-4.0, JENDL-3.3, ENDF/B-VII.0, and ENDF/B-VI.8. The present results revealed considerable underestimation of the evaluated cross sections in the high-energy region of 35–100 keV.     

Secondary particle yield and energy release data from intranuclear-cascade-evaporation model calculations of high energy (20–1100 MeV) neutron interaction with elements of shielding and biological importance

Secondary particle (neutron, proton, pion and heavy ion) yields and energy release data for 20, 50, 100, 300, 500, 800 and 1100 MeV neutron collisions with H, C, N, O, Al, Si, Ca, Fe and Pb have been calculated using the intranuclear-cascade-evaporation model. The low-energy limit is discussed and compared with the ENDF/B-IV neutron data. The mean elastic recoil energies have also been estimated by means of the Ranft formula for angular distribution. The multigroup response functions (kerma factors and production cross-sections) have been obtained and applied calculating the energy deposition and particle yields in the deep-penetrated graphite and iron spheres. The results for energy deposition of the ANISN neutron transport code using a high energy cross-section library and the derived kermas are about 10% overestimated relative to the thick-target results from the high energy particle transport code HETC. In the neutron transport discrete ordinates calculation the secondary production is underestimated when comparing with the HETC code.     

Ep≤200 MeV能区p+58Ni反应的计算与分析

以现有质子诱发⁵⁸Ni的各种核反应截面、能谱、双微分截面、弹性散射角分布等实验数据为基础，利用自行研制的大型核模型计算程序MEND计算质子能量在200MeV能区内⁵⁸Ni(p,x)反应的截面、能谱、角分布和n、p、α、d、t、³He6种出射轻粒子的双微分截面。MEND程序的理论框架基于球形光学模型、核子的核内级联发射模型、以激子模型为基础的预平衡发射理论、蒸发模型和带宽度涨落修正的Hauser Feshbach统计理论。光学模型中的势参数由APMN程序通过符合p+⁵⁸Ni反应的去弹截面和弹性散射角分布获得。出射粒子的双微分截面则利用MEND程序输出的能谱再通过Kalbach系统学公式计算。将计算结果与实验数据及ENDF/B6评价库进行了比较，计算结果与实验数据基本一致，与ENDF/B6相比，增加了³He的计算，且将能区上推至200MeV。      

快堆群常数连接程序LEK

LEK核数据处理程序是实现从ENDF/B-IV格式的评价核数据库中,用计算机自动取出快堆群常数程序KQCS的输入数据的程序。LEK取数既准确又节省大量的人力,改变了KQCS程序需手工输入上万个核数据的繁烦、落后局面,使中国核数据中心(CNDC)用计算机计算大量核素的快中子反应堆多群常数成为可能,为检验中国评价核数据库(CENDL)创造了条件。     

Detail analysis of fusion neutronics benchmark experiment for iron

In order to specify the best nuclear data on iron, the fusion neutronics benchmark experiment on iron at Japan Atomic Energy Agency (JAEA)/Fusion Neutronics Source (FNS) was analyzed in detail with MCNP-4C and the latest nuclear data libraries, JENDL-3.3, FENDL-2.1, JEFF-3.1 and ENDF/B-VII.0. As a result, totally the calculation result with ENDF/B-VII.0 agreed with the measurement best, except that it underestimated the measured neutron flux above 10 MeV with the depth. It was noted that the calculation result with JENDL-3.3 overestimated the measured neutrons below a few keV. Through the DORT calculations based on the iron data in ENDF/B-VII.0, it was found out that the first inelastic scattering cross-section data of ⁵⁷Fe in JENDL-3.3 caused the overestimation.     

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.     

Calculation and evaluations for n+Cu reactions

All cross sections of neutron induced reactions, angular distributions, energy spectra and double differential cross sections are consistently calculated and analyzed for n+^63,65,nat.Cu reactions at incident neutron energies below 200 MeV based on the nuclear theoretical models. The optical model, preequilibrium and equilibrium reaction theories, the distorted wave Born approximation theory are used. Theoretical calculated results are compared with existing experimental data and the evaluated results in ENDF/B-VII and JENDL-3 libraries. The optical model potential parameters are obtained according to the experimental data of total, nonelastic scattering cross sections and elastic scattering angular distributions.     

A validation of copper, nickel and iron data in endf/b-vi.2, jef-2.2 and jendl-3.2 through a series of br-1 critical core analysis

This report discusses the lessons learned from a validation of copper, nickel and iron data in ENDF/B-VI.2, JEF-2.2 and JENDL-3.2 through an analysis of 97.6 at. % ²³⁹PU metal fueled BR-1 fast reactor benchmarks. The specifications for the benchmarks in this study were taken from the ICSBEP handbook. The Monte Carlo code MCNP4B and continuous-energy cross-section libraries processed from the above three evaluated nuclear data files were used to perform the critical benchmark calculations. There is reasonable agreement between the predicted k_eff-values and the benchmarks within the uncertainties of measurements and Monte Carlo calculations, with the exception of the JEF-2.2 and JENDL-3.2 results for the copper-reflected core. The JEF-2.2 and JENDL-3.2 data give higher k_eff-values than those of ENDF/B-VI.2 for all the three reflector materials.