中子与58,60Ni反应的理论计算

根据中子与天然Ni及其同位素反应的总截面、去弹截面和弹性散射角分布的实验数据，得到中子的光学模型势参量。应用得到的光学模型势参量，根据光学模型、统一的Hauser-Feshbach和激子模型理论以及扭曲波玻恩近似理论，系统计算和分析了中子与^58,60Ni反应的非弹散射角分布和双微分截面，理论结果与实验很好地一致。     

多探测器快中子飞行时间谱仪

描述了一台串列加速器HI-13上的多探测器快中子飞行时间谱仪,并与国际上同类谱仪进行了比较。本谱仪主要用于能量大于8 MeV的快中子散射实验、次级中子双微分截面及带电粒子引起的出射中子能谱的测量。简要介绍了谱仪各主要部分(包括零信号拾取筒、氘气体靶、探测器、电子学等)的结构和特性及其在快中子实验中的应用。     

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.     

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.     

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.     

Neutron damage calculations in Cu,Nb and Au to 32 MeV: Application to sputtering and deuteron-breakup neutron sources

Primary recoil distributions and specific damage energies have been computed for high energy deuteron-breakup neutrons in Cu, Nb and Au. The calculations are based on theoretical neutron cross sections and consider in particular a d-Be spectrum broadly peaked at 15 MeV with some neutrons above 30 MeV. The theoretical results are similar to corresponding calculations for monoenergetic 15-MeV neutrons and are in good agreement with range measurements of (n, 2n) recoils generated by high energy d-Be neutrons in Nb and Au. The calculations are also consistent with recent d-Be neutron sputtering experiments in Nb and Au and demonstrate the usefulness of deuteron-breakup neutron sources for simulating fusion neutron effects.     

Theoretical calculation and analysis of the p+Co reaction

All cross sections, elastic and inelastic scattering angular distributions, energy spectra, and double differential cross sections of neutron, proton, deuteron, triton, helium and alpha particle emission for the p+⁵⁹Co reaction have been calculated and analyzed at incident energies from threshold to 200 MeV. The optical model, the intra-nuclear cascade model, direct, pre-equilibrium and equilibrium reaction theories are used. It is found that the theoretical calculated results are in good agreement with experimental data.     

Calculation and analysis of n+Zr reactions in the En ⩽ 250 MeV energy range

All of reaction cross sections, angular distributions, energy spectra, γ-ray production cross sections, and the double differential cross section for neutron, proton, deuteron, triton, helium and alpha emission are calculated and analyzed for n+^{90,91,92,94,96,nat}Zr at incident neutron energies from 0.1 to 250 MeV. The optical model, intranuclear cascade model, the unified Hauser–Feshbach theory and the exciton model which included the improved Iwamoto–Harada model are used. Theoretical calculated results are compared with existing experimental data and other evaluated data from ENDF/B-VI.8, ENDF/B-VII.0 and JENDL-3.3. The optical model potential parameters are obtained according to the experimental data of total, nonelastic cross sections and elastic scattering angular distributions.     

Measurements of neutron total and capture cross sections of 241Am with ANNRI at J-PARC

Neutron total and capture cross sections of ²⁴¹Am have been measured with a new data acquisition system and a new neutron transmission measurement system installed in Accurate Neutron Nucleus Reaction measurement Instrument at Materials and Life Science Experimental Facility of Japan Proton Accelerator Research Complex. The neutron total cross sections of ²⁴¹Am were determined by using a neutron time-of-flight (TOF) method in the neutron energy region from 4 meV to 2 eV. The thermal total cross section of ²⁴¹Am was derived with an uncertainty of 2.9%. A pulse-height weighting technique was applied to determine neutron capture yields of ²⁴¹Am. The neutron capture cross sections were determined by the TOF method in the neutron energy region from the thermal to 100 eV, and the thermal capture cross section was obtained with an uncertainty of 4.1%. The evaluation data of JENDL-4.0 and JEFF-3.2 were compared with the present results.     

A new technique for spectral interference correction on pin-by-pin BWR core analysis

A new correction technique to capture the spectral interference effect on collapsed cross sections, which focuses on application to the pin-by-pin boiling water reactor (BWR) core analysis, is proposed. The spectral interference effect, which is caused by adjacent loadings of different types of fuel assemblies, has relationship with variations of neutron leakage in each pin-cell from the viewpoint of neutron balance. Variation of neutron leakage affects neutron spectrum and thus the neutron leakage is considered to be important to correct coarse-group cross sections used in core calculations. We focus on the neutron leakage in each pin-cell and use it as a correction index (i.e., a leakage index (LI)), which is defined as the volume-averaged neutron leakage in a pin-cell. By utilizing the leakage index, we represent the variations of coarse-group cross sections as the linear combination of LIs. In order to verify and discuss the applicability of the present correction technique, two-dimensional benchmark calculations considering typical characteristics of BWR cores are carried out. From the calculation results, the present correction technique well reproduces the reference coarse-group cross sections and improves the calculation accuracies.     

A Pebble Bed Reactor cross section methodology

A method is presented for the evaluation of microscopic cross sections for the Pebble Bed Reactor (PBR) neutron diffusion computational models during convergence to an equilibrium (asymptotic) fuel cycle. This method considers the isotopics within a core spectral zone and the leakages from such a zone as they arise during reactor operation. The randomness of the spatial distribution of fuel grains within the fuel pebbles and that of the fuel and moderator pebbles within the core, the double heterogeneity of the fuel, and the indeterminate burnup of the spectral zones all pose a unique challenge for the computation of the local microscopic cross sections. As prior knowledge of the equilibrium composition and leakage is not available, it is necessary to repeatedly re-compute the group constants with updated zone information. A method is presented to account for local spectral zone composition and leakage effects without resorting to frequent spectrum code calls. Fine group data are pre-computed for a range of isotopic states. Microscopic cross sections and zone nuclide number densities are used to construct fine group macroscopic cross sections, which, together with fission spectra, flux modulation factors, and zone buckling, are used in the solution of the slowing down balance to generate a new or updated spectrum. The microscopic cross-sections are then re-collapsed with the new spectrum for the local spectral zone. This technique is named the Spectral History Correction (SHC) method. It is found that this method accurately recalculates local broad group microscopic cross sections. Significant improvement in the core eigenvalue, flux, and power peaking factor is observed when the local cross sections are corrected for the effects of the spectral zone composition and leakage in two-dimensional PBR test problems.     

基于OpenMC的多群截面库制作及有效性验证

OpenMC是麻省理工大学计算反应堆物理组开发的开源蒙特卡罗程序,能够方便地制作适用于特定堆芯中子能谱分布的多群反应截面及高阶勒让德散射截面以用于离散坐标输运程序ANISN的计算。本文基于ENDF/B-Ⅶ.1和CENDL-3.1评价数据库,利用OpenMC计算制作了ANSIN格式的多群截面并通过基准题的计算验证计算结果的准确性。通过截面转换程序的编写,将OpenMC给出的堆芯各阶勒让德散射分量,堆芯中子能谱分布,散射、吸收反应率以及裂变中子产生速率等信息转换为ANISN程序可读取的截面库格式。采用制作的截面库利用ANINS计算有效中子增殖因子及堆芯中子通量分布。结果表明,ANISN确定论的计算结果与OpenMC给出的蒙特卡罗计算结果相吻合,验证了这种方法可有效地为ANISN提供截面数据,将来可推广应用于二维、三维确定论中子输运计算。     

Theoretical spectrum-threshold detector data correlation in EBR-II

An experimental and computer program to further examine the neutron environment in the Experimental Breeder Reactor-II (EBR-II) has been completed. Monte Carlo and S₄ Transport methods were used to determine the neutron spectrum at various positions in the EBR-II core and blanket regions. Response functions for the threshold detectors ⁵⁸Ni (n, p) ⁵⁸Co and ⁵⁴Fe (n, p) ⁵⁴Mn were determined for each position and the corresponding predicted induced activities are compared with experimental results. Based on combinations of calculated neutron spectra, experimental detector responses, and cross section end points an empirical differential cross section was determined for the ⁴⁶Ti (n, p) ⁴⁶Sc threshold reaction. Spectrum averaged cross sections for the three threshold reactions which have been determined at various positions in this facility suggest that significant errors in fast neutron fluences will result if the usual fission spectrum averaged cross sections are used.     

氘核中子反应截面的法捷耶夫方程理论计算

中子引起的轻核反应是核数据研究的重要内容。当前我国核数据库中氘核中子反应截面的计算结果局限于采用s 波可分离势,且入射能量在20 MeV以下。需要发展三体核反应的法捷耶夫方程理论方法,采用超出s 波的核子 核子相互作用,从而对更高能量范围内氘核全套中子反应截面做出准确的描述。本文介绍了利用法捷耶夫方程计算n+d三核子反应体系的弹性散射微分截面、破裂反应、破裂反应出射中子和质子的双微分截面的理论框架及数值计算结果,同时计算了弹性散射总截面和破裂反应总截面的激发函数。计算结果与实验数据及CENDL 32、ENDF/B Ⅷ.0、JENDL 5、JEFF 33等数据库中的评价数据符合较好。     

Measurement and Analysis of Cadmium Ratios of Activation Foils in Thermal Critical Assembly (KUCA B 3/8″P36EU-NU-EU)

Abstract

To examine the applicability of foil activation technique for the estimation of neutron spectrum in a thermal reactor, Cd ratios of 8 activation foils (Au, Th, Dy, In, Mn, W, D.U. and E. U.) were measured in the void at the core center of KUCA B3/8″P36 EU-NU-EU assembly. The Cd ratios were analyzed with SRAC code system using 107 group cross sections based on ENDF/B4. To make the correction for polyethylene plates facing to the void to the calculated spectrum with 2-dimensional (r-z) diffusion model, softening factor calculated with 1-dimensional infinite slab model was introduced. This model gave almost same neutron spectrum as that without this correction. For the model which distributes atoms of Al sheath and support cylinder homogeneously into simulating materials, and using pointwise (fine group) cross sections for Au, Th, W and D.U., the calculated values except for W and D.U. almost agreed with the experimental ones. For W and D.U. C/E values were–1.1. Since Cd ratios are sensitive to the change in neutron spectrum except for D.U., this method is useful to judge the appropriateness of calculated neutron spectrum.     

Some New Applications of Neutron Slowing Down Time Spectrometry

A lead neutron slowing-down-time spectrometer LESP (LEad Standard Pile or LEad neutron SPectrometer) with a number of novel features was devised. It was applied to measurements on the effective sensitivity of counters and on the effective absorption and total cross sections of reactor materials. The basic principle of this method centers around the correlation that exists between the average neutron energy and the neutron slowing-down-time after fast neutrons are pulsed into a block of heavy medium with small neutron absorption such as a large lead assembly. This is analogous to the well established time-of-flight spectrometry.

The proposed method of spectrometry should be suitable for measurement of such effects as the resonance self and mutual shielding and geometrical heterogeneity. As trial experiment, the effective efficiency of a ²³⁵U fission chamber was measured and found to correspond fairly well with BNL-325 data. As another example, blocks of heavy resonance material such as natural uranium, antimony and tungsten were placed in the LESP for producing the standard neutron spectrum fields of these materials, and the neutron spectra and effective cross sections measured on these materials were compared with calculation.

It is concluded that these new applications of the method are quite practical for measurements of such properties as the effective efficiency of neutron counters and group averaged cross sections.     

An evaluation of neutron and gamma-ray-production cross-section data for lead

A survey was made of the available information on neutron and gamma-ray-production cross-section measurements of lead. From these and from relevant nuclear-structure information on the Pb isotopes, we prepared recommended neutron cross-section data sets for lead covering the neutron energy range from 0.00001 eV to 20.0 MeV. The cross sections are derived from experimental results available to February 1972 and from calculations based on optical-model, DWBA, and Hauser-Feshbach theories. Comparisons which show good agreement between theoretical and experimental values are displayed in a number of graphs. Also presented graphically are smoothed total cross sections, Legendre coefficients for angular distributions, and a representative energy distribution of gamma rays from resonance capture.     

The (p,n) reaction in lithium and the ground state of Be6

Using a time-of-flight method the neutron spectra in the Li⁶ + + p and Li⁷ + p reactions have been investigated at a proton energy of 9 Mev. Neutron groups have been found in the (p, n) reaction corresponding to the ground state in Be⁶ and the three lowest states of Be⁷ as well as a continuous neutron spectrum at lower energy, due to more complicated reactions. The observation of the neutron group for the Li⁶(p, n)Be⁶ reaction is the first experimental indication of the existence of the Be⁶ nucleus. The energy of the Li⁶(p, n)Be⁶ reaction is 5.2 Mev, the width of the ground state in Be⁶ is T < 0.3 Mev. The differential cross sections for neutron formation have been measured at 0, 15, 30, 60 and 120 °.     

Application of correction technique using leakage index combined with SPH or discontinuity factors for energy collapsing on pin-by-pin BWR core analysis

A correction technique to capture the spectral interference effect on collapsed cross sections is combined with the superhomogenization (SPH) factor or the discontinuity factor (DF) and is applied to the pin-by-pin core analysis for boiling water reactors (BWRs). The spectral interference effect has relationship with variations of neutron leakage in each pin-cell from the viewpoint of neutron balance. In order to correct collapsed cross sections, a new correction technique, in which the neutron leakage in each pin-cell is used as a correction index, was proposed in the previous study. By this correction technique, the reference coarse group cross sections are well reproduced and the calculation accuracies are improved. However, the reference fine group calculation results could not be reproduced since the correction technique cannot reduce energy collapsing errors. Thus, we combine the correction technique with the SPH factor or the DF to reduce energy collapsing errors. In order to verify and discuss the applicability of the correction technique with the SPH factor or the DF, two-dimensional benchmark calculations considering typical characteristics of BWR cores are carried out. The correction technique with the DF more accurately reproduces the reference fine group calculation results than that with the SPH factor.