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.     

100～500keV能区~9Be(d,n)~(10)B反应中子产额及角分布测量

金属铍具有较稳定的理化性质,同时也易与入射的轻核粒子发生反应产生中子,在小型中子发生器的研制中具有其他靶材所不具有的优势。9Be(d,n)10 B反应的中子产额和中子角分布数据,尤其是低能区的数据,对小型中子发生器的设计非常重要。本文在中国原子能科学研究院的600kV高压倍加器上测量了100～500keV能区9Be(d,n)10B反应的中子产额和中子角分布。     

厚铍靶9Be(d,n)反应中子产额测量

能量在3MeV以下厚靶D-Be反应的中子产额实验数据至关重要,但较为缺乏。本工作在北京大学4.5MV静电加速器上对氘束轰击厚铍靶的中子产额进行测量。对入射氘核能量在0.35~2MeV之间的若干能量点用长中子计数管进行了0°方向中子产额、中子角分布及中子总产额的测量。与已有的测量结果和经验公式进行了比较,并拟合出氘束轰击厚铍靶中子总产额的经验公式。     

New Evaluation of Neutron Nuclear Data for Zinc Isotopes

New evaluation of neutron-induced nuclear data for five stable isotopes of zinc (mass numbers A = 64, 66, 67, 68, and 70) was consistently carried out in the incident neutron energy range from 10^?5 eV up to 20MeV. In the low energy region up to about 100keV, the resonance parameters were evaluated by taking account of the available measured data. In the fast neutron region, the comprehensive calculations with nuclear reaction models, in which compound, preequilibrium, and direct processes are taken into account, were performed to estimate cross sections for various reactions and double differential cross sections of emitted neutrons and γ-rays. The comparisons of the evaluated cross sections with the experimental data and existing evaluated nuclear data libraries are made and show a good agreement with the measurements.     

Thermal neutron cross sections and diffusion parameters of Plexiglass

The neutron total cross section of Plexiglass has been measured for energies between 10⁻³ and 10³ eV by the transmission method with pulsed-neutron time-of-flight techniques. A calculation based on a synthetic scattering function shows a very good agreement with the measured values over the entire energy range. This model has been used to evaluate other quantities of interest in moderator design problems, including energy-transfer kernels and thermal neutron diffusion parameters. These experimental and theoretical results are compared with available data for Plexiglass.     

Neutron energy dependence of delayed neutron yields and its assessments

Incident neutron energy dependence of delayed neutron yields of uranium and plutonium isotopes is investigated. A summation calculation of decay and fission yield data is employed, and the energy dependence of the latter part is considered in a phenomenological way. Our calculation systematically reproduces the energy dependence of delayed neutron yields by introducing an energy dependence of the most probable charge and the odd–even e?ect. The calculated fission yields are assessed by comparison with JENDL/FPY-2011, delayed neutron activities, and decay heats. Although the fission yields in this work are optimized to delayed neutron yields, the calculated decay heats are in good agreement with the experimental data. Comparison of the fission yields calculated in this work and JENDL/FPY-2011 gave an important insight for the evaluation of the next Japanese evaluated nuclear data library (JENDL) .     

Prompt fission neutron multiplicity and spectrum evaluations for n+238U reaction

The prompt fission neutron multiplicity and spectra for n+²³⁸U reaction are calculated using an improved Los Alamos model which includes the linear relation between the average prompt gamma ray energy and the prompt neutron multiplicity and also the average fission fragment kinetic energy dependence on the incident neutron energy. The coefficients describing the quadratic variation of the fission fragment kinetic energy versus the incident energy are obtained by extrapolation of the data and procedure used for n+²³⁵U reaction. The inverse process compound nucleus cross-section of the fissioning nucleus is calculated using the coupled channel method. In the incident energy range where only the first fission chance is involved the comparison of present spectrum evaluation with spectrum calculation using multi-modal model is made too. The calculated prompt neutron multiplicity and spectra of ²³⁸U neutron induced fission are in good agreement with the experimental data for the entire incident energy range required in evaluations, proving the validity of the used procedure.     

基于金活化片的单球谱仪研制及能谱测量

中国先进研究堆(CARR)H-8水平孔道是提供中子的实验孔道,可以提供稳定的辐射场,对于不同的中子实验,其所需的中子能谱谱形不同,准确测量中子能谱具有重要意义。为测量H-8水平孔道中子能谱,研制一种以金活化片为热中子探测器的被动式单球中子能谱仪,使用MCNP程序对10^-11～15 MeV能区的中子能量响应进行计算,并分析能量响应的合理性。在CARR堆导管大厅对单球谱仪进行测试实验,使用高纯锗探测器测量各金活化片活度,使用UGA(unfolding based on genetic algorithm)解谱程序对实验数据进行解谱计算。结果表明,导管大厅出射中子能量在10^-9～10^-6 MeV范围内,单球中子谱仪可以较为精确的给出中子能谱数据,适用于CARR堆H-8水平孔道中子能谱测量研究。     

中国第一台高性能白光中子源——CSNS反角白光中子源及其应用

基于中国散裂中子源（CSNS）建设的我国第一台高性能白光中子源--反角白光中子源（Back-n）是国际上综合性能最好的白光中子源之一,能区范围覆盖meV～百MeV,飞行时间测量分辨率可在全能区达到1%以内,中子注量率国际领先。自2018年3月建成以来,Back-n已开展了一系列的核数据测量实验、探测器标定实验、中子辐射效应实验和中子照相研究,科研产出效率非常高,实验数据质量达到了研究要求,为我国多领域的科学研究和应用研究提供了一个强大的平台。本文对该白光中子源的性能特点、已投入运行和规划中的核数据测量实验谱仪进行了综述,并指出了主要应用方向。     

先进裂变核能的关键核数据测量和CSNS白光中子源

在设计加速器驱动的次临界系统（ADS）、核废料嬗变装置及钍基熔盐堆时亟需一些关键核数据,当前核数据库受实验条件或中子能区的限制,存在核数据精度不高甚至少部分核素数据缺失的情况。本文综述了国内外相关的核数据研究和相应的白光中子源情况。基于中国散裂中子源(CSNS)的反角通道白光中子源实验终端的中子束流具有非常宽的能谱（0.01 eV~200 MeV）和很好的时间特性。模拟得到距靶80 m处的实验终端的中子注量率为9.3×10⁶cm^-2•s^-1,1 eV ~ 1 MeV能量间隔内的中子数占总中子数的53%;同时,加速器运行在双束团模式或单束团模式,时间分辨率均在0.3%～0.9%之间,适合开展核数据测量。     

冷中子三轴谱仪的整体配置和主要部件的参数设计

从冷中子导管系统的实际情况出发,以程序模拟为主要手段,通过计算和分析得到最重要的设计参数：可用中子波长范围、样品处束流强度和能量分辨率。在这些参数的基础上,结合实际给出了谱仪的整体配置以及主要部件的部分参数,如能量和动量转移范围、单色器和分析器的起飞角范围和嵌镶角、样品台散射角、准直器的准直度及本底和剂量要求等。     

(α, n) Neutron Spectra on Thick Light Targets

Atomic Energy - Accurate data on the neutron yield and data on the energy spectrum of neutron radiation are needed in order to develop methods for the safe handling of fresh nuclear fuel,...     

Fission Product Yields from Fission Spectrum n+Pu for ENDF/B-VII.1

We describe a new cumulated fission product yield (FPY) evaluation for fission spectrum neutrons on plutonium that updates the ENDF/B-VI evaluation by England and Rider, for the forthcoming ENDF/B-VII.1 database release.¹ We focus on FPs that are needed for high accuracy burnup assessments; that is, for inferring the number of fissions in a neutron environment. Los Alamos conducted an experiment in the 1970s in the Bigten fast critical assembly to determine fission product yields as part of the Interlaboratory Reaction Rate (ILRR) collaboration, and this has defined the Laboratory's fission standard to this day. Our evaluation includes use of the LANL-ILRR measurements (not previously available to evaluators) as well as other Laboratory FPY measurements published in the literature, especially the high-accuracy mass spectrometry data from Maeck and others. Because the measurement database for some of the FPs is small — especially for ⁹⁹Mo — we use a meta-analysis that incorporates insights from other accurately-measured benchmark FP data, using R-value ratio measurements. The meta-analysis supports the FP measurements from the LANL-ILRR experiment. Differences between our new evaluations and ENDF/B-VI are small for some FPs (less than 1-2%-relative for ⁹⁵Zr, ¹⁴⁰Ba, ¹⁴⁴Ce), but are larger for ⁹⁹Mo (4%-relative) and ¹⁴⁷Nd (5%-relative, at 1.5 MeV) respectively. We present evidence for an incident neutron energy dependence to the ¹⁴⁷Nd fission product yield that accounts for observed differences in the FPY at a few-hundred keV average energy in fast reactors versus measurements made at higher average neutron energies in Los Alamos' fast critical assemblies. Accounting for such FPY neutron energy dependencies is important if one wants to reach a goal of determining the number of fissions to accuracies of 1-2%. An evaluation of the energy-dependence of fission product yields is given for all A values based on systematical trends in the measured data, with a focus on the energy dependence over the fast neutron energy range from 0.2-2 MeV. Based on these trends, we present an evaluation of the FPY data at 0.5 and 2.0 MeV average incident neutron energies. This new set of ENDF/B-VII data will enable users to linearly interpolate between the pooled FPY data at ∼0.5 MeV and our new data at 2 MeV to obtain FPYs at other energies.     

Measurements of neutron induced activation of concrete at 64.5 MeV

The present paper presents the measurement of neutron induced activations on concrete using the 64.5 MeV quasimonoenergetic neutrons produced at the intense ⁷Li(p, n) neutron source at Cyclotron and Radioisotope Center, Tohoku Univeristy (CYRIC). The data were corrected for the effect of continuous neutrons in the source. The neutron energy, neutron yields and the spectrum of continuous neutrons were confirmed with the neutron time-of-flight method and the neutron activation measurement of the ²⁰⁹Bi(n, Xn) reactions having various threshold energy values. The nuclides produced by thermalized source neutrons are negligible. New data were obtained for concrete activation.     

MCNP simulation of neutron energy spectra for a TN-32 dry shielded container

This paper presents a numerical analysis of neutron energy spectra for a TN-32 spent fuel dry storage cask using Monte Carlo simulation. The analysis results were compared with experimental measurements to determine the suitability of using such codes for neutron flux calculations in soft-spectrum neutron environments. Complete spent fuel compositions were generated using Scale 4.4a. Variations in source definition and geometry determined that geometric and source simplifications in the computational model have negligible effect on final neutron energy distribution. Variations between experimental and computed spectra at energies above 1 MeV and below 100 keV demonstrated the shortfalls of various detection instruments used to collect the experimental neutron energy spectra data principally because these instruments were calibrated based on high neutron energy spectra. The MCNP calculations were generally in agree with the experimental data, but predicted that the detectors would over-respond to the neutron spectra around a spent fuel dry shielded container. Computed neutron energy spectra were always conservative when compared to experimental spectra.     

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.     

Measurement of neutron capture cross-section of the Ga(n, γ)Ga reaction at 0.0536 eV energy

The neutron capture cross-section for the ⁷¹Ga(n,  γ)⁷²Ga reaction at 0.0536 eV energy was measured using activation technique based on TRIGA Mark-II research reactor. The ¹⁹⁷Au(n, γ)¹⁹⁸Au monitor reaction was used to determine the effective neutron flux. Neutron absorption and γ-ray attenuation in gallium oxide pellet were corrected in determination of cross-section. The cross-section for the above reaction at 0.0536 eV amounts to 2.75 ± 0.14 b. As far as we know there are no experimental data available at our investigated energy. So far we are the first, who carried out experiment with 0.0536 eV neutrons for cross-section measurement. The present result is larger than that of JENDL-3.3, but consistent within the uncertainty range. The value of ENDF/B-VII is higher than this work. The result of this work will be useful to observe energy dependence of neutron capture cross-sections.     

Speed-up of cross-section collapsing with weight-window for subcritical burnup calculation

In the author’s group, a fusion–fission (FF) hybrid energy system has been analyzed using our own burnup calculation system consisting of Monte Carlo transport code MCNP-4C and point burnup code ORIGEN2.1. Since the neutron energy spectrum changes along with progress of burnup in a subcritical system, it is necessary to update one-group cross-section library in each burnup step. The one-group cross-sections are normally updated by collapsing the evaluated nuclear data such as JENDL and ENDF using a neutron flux calculated by an appropriate transport code such as MCNP. The collapsed cross-sections are handed over to ORIGEN, and the reaction rates for burnup of elements are thereafter estimated accurately.As well known, MCNP generates track-length (TL) data in the neutron transport calculation, which are base data to estimate the neutron flux. We thus use the track-length data directly instead of the calculated neutron flux, in order to evaluate the reaction rate as accurately as possible. However, the number of TLs becomes extremely large and thus it takes a longer computation time. We therefore reduce the number of TLs used in the cross-section collapsing process as far as the accuracy is conserved. However, in some energy region the number of TLs is inversely too small to conserve the original cross-section accuracy of the evaluated nuclear data files, because the number of TL data per unit energy is smaller than that of the nuclear data.In the present study, the weight-window (WW) technique of MCNP was applied to our burnup calculation system in order to control the number of TLs in such an energy region artificially and to complete the collapsing process accurately in the whole energy region. As a result, the variance of the calculated neutron flux thus deteriorates slightly, but the number of TLs could be successfully adjusted to conserve the accuracy of the nuclear data file in the whole energy region. And the accurate reaction rate estimation for burnup with MCNP was finally realized and simultaneously the computation time could be saved reasonably.     

用于反应堆内相对中子通量密度在线测量的闪烁体光纤探测系统研制

为实现反应堆不同空间和能量的相对中子通量密度在线监测,本文研究开发了一套新型的用于狭小空间且位置灵敏的闪烁体中子探测系统。该套系统由5种探头、5路光子计数器、1台计算机及相应的软件组成。5种探头的主要构成物质分别为~6 LiF+ZnS(Ag)、~(232) ThO_2+ZnS(Ag)、~(238) UO_2+ZnS(Ag)、~9Be+ZnS(Ag)以及BGO晶体,故可测量不同能量的相对中子通量密度。其中,掺有~6 LiF的探头用于热中子的测量,BGO探头用于γ测量,其余3种探头用于快中子的测量。利用该系统进行了启明星1#装置内热中子及快中子的相对通量密度分布测量,并将测量结果与利用蒙特卡罗方法得到的理论分布结果进行了比较。考虑到理论设置参数与实际实验参数的差别,可认为测量结果是可信的。