Measurements of gamma-ray emission probabilities of 241, 243Am and 239Np

Gamma-ray emission probabilities of ^{241, 243}Am and ²³⁹Np have been precisely measured with gamma- and alpha-ray spectroscopic methods. The activities of the samples were determined by measuring alpha particles using a Si semiconductor detector. Gamma rays emitted from the samples were measured with a planar type High-Purity Germanium (HPGe) detector. An efficiency curve of the HPGe detector was derived with uncertainties from 0.7% to 2.5% by combining measured efficiencies and Monte Carlo simulation. The gamma-ray emission probabilities for the major gamma rays of these nuclides were determined with uncertainties less than 1.2%.     

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.     

Correction of the thermal neutron capture cross section of 241Am obtained by the Westcott convention

There is large discrepancy among the reported experimental data of the thermal neutron capture cross section of ²⁴¹Am, where the activation measurements provided larger cross sections than those in the time-of-flight ones. The Westcott convention has been widely used for the derivation of the thermal neutron capture cross section in the activation measurements. We have estimated that this large discrepancy is due to the existence of the resonances below the cadmium cut-off energy (E_Cd ～ 0.5 eV). By reviewing the Westcott convention, we developed the correction method taking account of the contribution of the resonances near or below E_Cd. The correction term was evaluated using the JENDL-4.0. Application of the present method successfully improved the existing discrepancy of the thermal capture cross section of ²⁴¹Am.     

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.     

Feedback on neutron capture cross sections of 238Pu and 241Am from analyses of measured isotopic compositions of irradiated LWR fuels and MOX core physics experiments

The atomic fractions of ²³⁸Pu and ²⁴¹Am in MOX fuels recycled in light water reactors are 1% to 2% and not significant compared with those of major Pu isotopes. On the other hand, recent evaluated nuclear data libraries, such as JENDL-4.0 and JEFF-3.2, give noticeably different thermal and epithermal neutron capture cross sections for ²³⁸Pu and ²⁴¹Am. The thermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am in JEFF-3.2 are 31% and 9% larger than those of JENDL-4.0, respectively. This paper shows the effect of the differences in the neutron cross sections on analysis results of two different integral experiments. The first is the isotopic compositions of ²³⁸Pu on UO₂ and MOX fuels irradiated in BWR and PWR, and the second is the critical experiments of the water moderated cores fully loaded with MOX fuels. The former was analyzed by using the continuous energy Monte Carlo burnup calculation code MVP-BURN and the latter by the continuous energy Monte Carlo calculation code MVP. The comparisons between the calculated and measured results indicate that the most likely thermal and epithermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am should be around at the middle between those of JEFF-3.2 and JENDL-4.0.     

一台测量快中子辐射俘获截面的大液体闪烁探测器

为了测量快中子辐射俘获截面,我们研制了一台测量γ射线的球形液体闪烁探测器,其直径为1m,容积680l。该探测器属4π几何类型,探测器效率达80％以上。为了降低本底,除了用10cm厚的铅和40cm厚的石蜡建成屏蔽室外,还采用符合方法,使本底计数小于100cps。探测器已经在2.5MeV脉冲质子静电加速器上测量了金、钽和铥等元素的辐射俘获截面。     

Measurements of keV-neutron capture cross sections and capture gamma-ray spectra of 105Pd

The neutron capture cross sections and capture gamma-ray spectra of ¹⁰⁵Pd were measured in the region from 15 to 100 keV and at 585 keV. A neutron time-of-flight method was utilised with an anti-Compton NaI(Tl) spectrometer and a 1.5-ns pulsed neutron source by the ⁷Li(p,n)⁷Be reaction. The capture yields were obtained by applying a pulse-height weighting technique to the observed net capture gamma-ray pulse-height spectra. The capture cross sections of ¹⁰⁵Pd were derived with errors less than 5%, using the standard capture cross sections of ¹⁹⁷Au. The evaluated capture cross sections of JENDL-4.0 and ENDF/B-VII.1 were compared with the present results. The evaluations of JENDL-4.0 and ENDF/B-VII.1 were larger than the present results by 3%–15% in the region from 15 to 100 keV and at 585 keV. The capture gamma-ray spectra of ¹⁰⁵Pd were also derived by unfolding the observed net capture gamma-ray pulse-height spectra. The multiplicities of capture gamma rays of ¹⁰⁵Pd were obtained from the capture gamma-ray spectra.     

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.     

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.     

Proton-induced activation cross section measurement for aluminum with proton energy range from 0.4 to 3 GeV at J-PARC

We have started an experimental program to measure activation cross sections systematically in the proton-induced spallation reaction in structural materials commonly used in high-intensity proton accelerator-based facilities, such as Japan Proton Accelerator Research Complex (J-PARC). As the first step of the program, aluminum (Al) was chosen to verify the adequacy of the measurement technique implemented in a J-PARC proton beam environment because data of Al have been relatively well studied both by experimental measurement and simulation. Activation cross sections of ⁷Be, ²²Na, and ²⁴Na in Al were measured at proton energy points from 0.4, 1.3, 2.2 to 3.0 GeV, which could be delivered smoothly from the synchrotron. The validity of experimental data has been verified by introducing an effective proton numbers determination procedure. We compared the measured data with existing experimental data, the evaluated data (JENDL-HE/2007), and the calculations with several intra-nuclear cascade models by the Particle and Heavy Ion Transport code System (PHITS) code. Although the experimental data agreed with JENDL-HE/2007, the calculations underestimated about 40%. This could come from the evaporation model (generalized evaporation model) being implemented in the PHITS code. We found that the calculations agreed with the experimental data by an upgraded PHITS code.     

中子核反应截面测量中的厚靶技术

在核反应截面测量实验中使用厚靶技术，对厚靶测量谱进行解谱，得到等效薄靶的结果。用该方法大大提高了事件的计数率，明显地减少了统计误差。     

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%.     

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.     

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.     

Neutron-capture cross-sections of 244Cm and 246Cm measured with an array of large germanium detectors in the ANNRI at J-PARC/MLF

The neutron neutron-capture cross cross-sections of ²⁴⁴Cm and ²⁴⁶Cm were measured by the time-of-flight method in the energy range of 1–300 300 eV with an array of large germanium detectors in the Accurate Neutron-Nucleus Reaction measurement InstrumentANNRI at Material and Life Science Experimental Facility (MLF) of the Japan Proton Accelerator Research ComplexJ-PARC/MLF. The ²⁴⁴Cm resonances at around 7.7 and 16.8 8 eV and the ²⁴⁶Cm resonances at around 4.3 and 15.3 3 eV were observed in the capture reactions for the first time. The uncertainties of the obtained cross cross-sections are 5.8% at the top of the first resonance of ²⁴⁴Cm and 6.6% at that of ²⁴⁶Cm. The rResonance analyses were performed for low-energy ones using the code SAMMY. The prompt γ-ray spectra of ²⁴⁴Cm and ²⁴⁶Cm were also obtained. Eight and five new prompt γ-ray emissions were observed in the ²⁴⁴Cm(n, γ) and ²⁴⁶Cm(n, γ) reactions, respectively.     

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.