Measurement of neutron capture cross sections of Pd-107 at J-PARC/MLF/ANNRI

Measurements of the neutron capture cross sections of ¹⁰⁷Pd were carried out at the Materials and Life Science Experimental Facility (MLF) of the Japan Proton Accelerator Research Complex (J-PARC). Gamma-rays were detected with an NaI(Tl) spectrometer of the Accurate Neutron–Nucleus Reaction Measurement Instrument (ANNRI). The neutron capture cross sections were determined by the time-of-flight method in the neutron energy region from the thermal to keV energies.     

Measurements of fast—neutron capture cross sections for ^159Tb and ^169Tm

The neutron capture cross sections for ^159Tb and ^169Tm relative to the ^197Au (n,γ)^198Au reaction are measured at neutron energies of 0.57,1.10 and 1.60 MeV by using the activation method.The activities of the products are measured with a high resolution HPGe detector gamma-ray spectrometer.The errors of the present work are 5-6% for Tb,6-7% for Tm.The recommended data in energy region of 0.4-3.0MeV are given as compared with other data published previously.     

Estimation of Covariance Matrices for Nuclear Data of 237Np, 241Am and 243Am

Covariance matrices were estimated for the fission and capture cross sections and the numbers of neutrons per fission of ²³⁷Np, ²⁴¹Am and ²⁴³Am given in JENDL-3.3. GMA and KALMAN codes were applied to estimate them for the fission and capture cross sections, respectively. In the low energy region, the errors of resonance parameters were given. The covariance matrices for the numbers of prompt neutrons per fission (V_p ) were evaluated by assuming a linear equation. For the delayed neutrons (v_d ), only their standard deviations were estimated. The results were compiled in the ENDF-6 format and merged with JENDL-3.3.     

Measurement of fast-neutron capture cross sections for 75As

1 INTRODUCTIONThe cross sections of the "As(n,7)"As reaction are boortant in evaluating theradiation damage of the material. EXPerimeats[1-6] and evaluations have been performedto deterAnne the "As(n, 7)"As reaction cross section, but there are large discrepanciesamong them especially in the MeV neutron energy region. Therefore, new eXPerAnent isneeded.2 EXPERIMENT DETAILSThe measuremenis were perfOrmed at the 4.5 MV Van de Grand accelerator of theInstitute Of Heavy Ion Phy…     

Radiochemical Determination of Neutron Capture Cross Sections of 241Am

The thermal neutron capture cross sections and the neutron capture resonance integrals of ²⁴¹Am leading to the production of the isomer ²⁴²Am and the ground-state ^242gAm were measured radiochemically by the Cd-ratio technique with neutron flux monitors of Co/Al and Au/Al alloy. Highly-purified ²⁴¹Am targets were irradiated in an aluminum capsule by using JMTR. The neutron fluxes and their epithermal neutron fractions were determined by measuring γ-rays of ⁶⁰Co and ¹⁹⁸Au. The yields of ^242mAm and ^242gAm were decided by analyzing growth and decay curves of the α-ray activity ratios ²⁴²Cm/²⁴¹Am. The resultant thermal neutron capture cross sections are 85.7 ± 6.3 b and 768 ± 58 b for ^242mAm and ^242gAm, and the resonance integrals 114±7 b and 1,694±146 b, respectively. The differences between the present results and the evaluated values by Mughabghab are 38–59%. The isomeric ratios, g/(m+g), of 0.90±0.09 for thermal neutrons and 0.94±0.11 for epithermal neutrons are, however, almost consistent with evaluated values.     

Measurement of the neutron-induced total cross sections of natPb from 0.3 eV to 20 MeV on the Back-n at CSNS

The neutron-induced total cross sections of natural lead have been measured in a wide energy range(0.3 eV-20 MeV)on the back-streaming white neutron beamline(Back-n)at the China Spallation Neutron Source.Neutron energy was determined by the neutron total cross-section spectrometer using the time-of-flight technique.A fast multi-cell fission chamber was used as the neutron detector,and a 10-mm-thick high-purity natural lead sample was employed for the neutron transmission measurements.The on-beam background was determined using Co,In,Ag,and Cd filters.The excitation function of natPb(n,tot)reaction below 20 MeV was calculated using the TALYS-1.96 nuclear-reaction modeling program.The present results were compared with previous results,the evaluated data available in the five major evaluated nuclear data libraries(i.e.,ENDF/B-Ⅷ.0,JEFF-3.3,JENDL-5,CENDL-3.2,and BROND-3.1),and the theoretical calculation curve.Good agreement was found between the new results and those of previous experiments and with the theoretical curves in the corresponding region.This measurement obtained the neutron total cross section of natural lead with good accuracy over a wide energy range and added experimental data in the resonance energy range.This provides more reliable experimental data for nuclear engineering design and nuclear data evaluation of lead.     

Re-evaluation of neutron displacement cross sections for silicon carbide by a Monte Carlo approach

Neutron displacement cross sections for SiC are re-evaluated by a Monte Carlo approach, with damage energies of primary recoils calculated by the stopping and range of ions in matter (SRIM) code. The validity of the Monte Carlo model is examined by the case of iron, and the results show good agreement with the reference values. Neutron displacement cross sections for SiC at energies up to 100 MeV are calculated, and averaged over the neutron spectra of a fusion DEMO reactor, the high flux test module of the International Fusion Materials Irradiation Facility, and typical fission test reactors. Gas production is also calculated for those neutron irradiation facilities. Finally, the suitability of the displacement cross sections is discussed. The results on comparison among neutron irradiation of different facilities by the current displacement cross sections are similar to those by results of the previous work. Moreover, since neutron displacement cross sections in this study are calculated with damage energies of primary recoils calculated by SRIM, neutron damage evaluated by our displacement cross sections is suitable for correlation with damage by heavy ions calculated by SRIM.     

Measurement and Analysis of Reactivity Worth of 241Am Sample in Water-Moderated Low-Enriched UO2 Fuel Lattices at TCA

The reactivity worths of 22.82 grams of ²⁴¹Am oxide sample were measured and theoretically analyzed in water-moderated UO2 fuel lattices in seven cores of the Tank-Type Critical Assembly (TCA) at the Japan Atomic Energy Agency for an integral test of ²⁴¹Am nuclear data. These cores provided a systematic variation in the neutron spectrum between the thermal and resonance energy regions. The sample reactivity worth was measured with an uncertainty of 2.1% or less. The theoretical analysis was performed using the JENDL-3.3 nuclear data by a Monte Carlo calculation method. Ratios of calculation to experiment (C/Es) of the reactivity worth were between 0.91 and 0.97, and showed no apparent dependence on the neutron spectrum. In addition, sensitivity analysis based on the deterministic calculation method was carried out to obtain the impact of changing the ²⁴¹Am capture cross section on the sample reactivity worth. The result of this analysis showed that the C/E could be significantly improved by almost uniformly increasing the ²⁴¹Am capture cross section of JENDL-3.3 by 25–30%.     

Neutron total cross section measurements of polyethylene using time-of-flight method at KURNS-LINAC

ABSTRACT

The neutron total cross sections of polyethylene have been measured in the energy region from 0.001 eV to 40 keV by the time-of-flight (TOF) method using the Kyoto University Institute for Integrated Radiation and Nuclear Science – Linear Accelerator (KURNS-LINAC). A ⁶Li detector and a gas electron multiplier (GEM) detector have been used as a neutron detector, and the polyethylene plates of 0.31 and 0.20 cm thickness were employed for the neutron transmission measurement.

The present results were compared with the previous results and the evaluated data in JENDL-4.0. In the energy region below 0.01 eV, the present results are in good agreement with the data measured by Herdade et al. (1973) and by Granada et al. (1987). On the other hand, the evaluated data in JENDL-4.0 are larger than all the measured data. In the energy region from 0.035 to 0.15 eV, the data measured by Granada et al. and the evaluated data in JENDL-4.0 are up to about 4 ~ 6% larger than the present results.     

Measurements of neutron total cross-sections and resonance parameters of erbium at the Pohang Neutron Facility

We measured neutron total cross-sections of natural erbium in the neutron energy region from 0.2 to 120 eV by using the neutron time-of-flight method at the Pohang Neutron Facility, which consists of an electron linear accelerator, a water-cooled tantalum target with a water moderator, and a 12-m-long time-of-flight path. A ⁶Li-ZnS(Ag) scintillator with a diameter of 12.5 cm and a thickness of 1.6 cm was used as a neutron detector, and a group of high-purity natural erbium metallic plates with various thickness was used for the neutron transmission measurements. The present measurement was compared with the existing experimental and the evaluated data. The resonance parameters of ¹⁶⁶Er, ¹⁶⁷Er, ¹⁶⁸Er, and ¹⁷⁰Er in the neutron energy region below 120 eV were extracted from the transmission by using the multilevel R-matrix SAMMY code and were compared with the evaluated data from ENDF/B VII.0 and other previous reported results.     

活化法测量197Au等核素的14MeV中子反应截面

采用活化法测量了197Au等核素的14MeV中子反应截面.当入射中子能量为14.75MeV时,197Au(n,2n)196Au反应截面测量结果为(2175±76)×10-31m2,并与其他测量结果和ENDF/B-6评价数据库数据进行了比较.对于结果的一些不确定度因素,采用MCNP程序进行了分析.     

Measurements of absolute electron capture cross sections in He2+-He and Ne8+-O2,N2,CH4 collisions

The total absolute cross sections of single-and double-electron capture(SEC and DEC)in the collisions of He2+with He and Ne8+with O2,N2,and CH4 were studied in the energy ranges 3.5-50 keV/u and 2.8-40 keV/u,respectively.Through a deep analysis of the experimental systematic uncertainties in the measurement procedure and data evaluation,the error in the experimental results of the SEC cross sections is less than 9％.Within the uncertainties,the present results of the He2+-He collision show good consistency with previous measurements,validating the experimental system and paving the way for precise measurements of EC cross sections for a variety of ions and neutral gases.The present measurements allow for a test of EC theory and provide crucial EC cross section data for the establishment of plasma models in fusion research and astrophysical X-ray studies.     

Reconstruction method of homogenized cross sections

The reconstruction method of homogenized cross sections in the direct response matrix method has been developed. In this reconstruction method, homogenized cross sections, which take into consideration the influences of neighboring fuel assemblies, can be reconstructed with the response relationship of incoming neutron partial currents and neutron production rates. Calculations for heterogeneous multi fuel assembly systems were done to verify the developed method. The thermal energy group fuel assembly cell-averaged homogenized cross sections reconstructed by this method agreed with those evaluated by the direct calculation of the whole system using the Monte Carlo method within 0.2%. The effect using the reconstructed fuel assembly cell-averaged homogenized cross sections in a conventional core analysis code using cross sections homogenized in a fuel assembly cell was also investigated. The results obtained showed that the analysis accuracy of k-infinity can be improved by using the cross sections reconstructed by the method. Because almost no influences on the analysis accuracy could be found related to the divided numbers of the surfaces and the angles, and the response relationship with neutron production rates of fuel rods or a fuel assembly cell-averaged neutron production rate, this reconstruction method is applicable to a conventional core analysis code using homogenized cross sections in a fuel assembly cell.     

Measurement of Effective Capture Cross Section of Americium-243 for Thermal Neutrons

The effective capture cross section of ²⁴³Am for thermal neutrons was measured with an activation method. A sample of ²⁴³Am was irradiated for 10 hrs at Kyoto University Reactor, KUR. After the irradiation, the sample was cooled for one month. In the cooling time, ^244mAm and ^244gAm produced by the irradiation decayed out to ²⁴⁴Cm. The α rays emitted from ²⁴³Am and ²⁴⁴Cm were measured with a silicon surface barrier detector. The neutron flux at the irradiation position was monitored using Au/Al and Co/Al wires. The effective capture cross section was deduced as 174.5±5.3b from the α-ray counts and the neutron flux. The quantity r√T/T₀ in Westcott's convention was 0.037±0.004. The present result was compared with the effective capture cross sections from the JENDL-3.3 and the Mughabghab evaluations.     

~(115)In中子非弹性散射截面的实验测量及蒙特卡罗修正

~(115)In是一种重要的活化材料,准确测量它的中子非弹性散射截面数据对中子注量监测具有重要意义。在四川大学原子核科学技术研究所2.5 MV静电质子加速器上,利用核反应D(d,n)~3He产生的单能中子,以~(197)Au作为标准,采用活化法测量了2.95 Me V、3.94 Me V、5.24 Me V能点的~(115)In中子非弹性散射截面。用Monte Carlo程序MCNPX(Monte Carlo N-Particle eXtended)对靶头材料、冷却水层和样品的包层材料等引起的多次散射效应及注量率衰减效应等进行了修正计算,得到最终结果与Loevestam的计算值符合较好,并且实验中可通过减小靶管、靶底衬、水层及样品的包层材料等厚度来减小多次散射效应和自屏蔽效应的影响。     

Thermal-Neutron Capture Cross Section and Resonance Integral of Americium-241

The thermal-neutron capture cross section (σ_0;g) and the resonance integral (I _0,g) leading to the ground state of 242Am were measured by an activation method for neutron capture by²⁴¹ Am. A method with gadolinium, which was similar to the cadmium difference method, was used to measure the cross section σ_0;g with attention to resonances of²⁴¹ Am. Americium chloride samples containing²⁴¹ Am radioisotope were irradiated for 68 h in the long-irradiation plug of the Kyoto University Research Reactor, KUR. Wires of Co/Al and Au/Al alloys were used as monitors to determine thermal-neutron fluxes and epithermal Westcott's indexes at the irradiation positions. An α-ray spectrometer was used to measure the activity ratios of²⁴² Cm to²⁴¹ Am. On the basis of Westcott's convention, the σ_0;g and I _0,g values were determined as 628 ± 22 b and 3:5 ± 0:3 kb, respectively.