~(56)Fe(n,p)~(56)Mn反应截面的测量

在E_n=12.8—18.2兆电子伏能区,用活化法测量了~(56)Fe(n,p)~(56)Mn反应截面,在E_n=14.63±0.20兆电子伏处做了绝对测量,结果为108.0±2.9毫靶。中子通量用伴随粒子法测定。对T(d,n)~4He反应的中子角分布也作了测定。~(56)Mn的放射性用φ10×7.6厘米的Nal(Tl)闪烁谱仪测量,~(56)Mn特征γ射线的探测效率用4πβ-γ符合法测定的~(56)Mn标准源刻度。本文还对用伴随粒子法测量中子通量作了较详细的叙述。测量结果同国外数据作了比较,并做了简短讨论。     

快中子64Zn(n,α)61Ni反应微分截面实验测量

由于64Zn(n,α)61Ni反应的剩余核是稳定的,不能用通常的活化法来测量,致使该反应截面实验数据缺乏.利用双屏栅电离室作为带电粒子探测器,在En=2.54,4.00,5.03,5.50与5.95MeV 5个能点,对64Zn(n,α)61Ni反应的微分截面进行了实验测量,并通过微分截面对角度的积分得到了反应截面.实验在北京大学4.5MV静电加速器上进行.2.54MeV的单能中子采用固体氚-钛靶T(p,n)3He反应产生,其余四种能量的准单能中子通过氘气体靶D(d,n)3He反应获得.绝对中子通量采用238U(n,f)反应来确定,实验过程中用BF3长中子计数器进行相对中子通量监测.测量结果与已有的实验与评价数据进行了比较.     

~(176)Hf(n,2n)~(175)Hf反应截面测量

采用相对测量技术,以活化法对13.4～14.8MeV范围内的176Hf(n,2n)175Hf反应截面进行了测量。样品固定在距离D-T中子源20cm处的圆环的不同位置上进行中子辐照,采用93Nb(n,2n)92Nbm作为监测反应,活化产物采用高纯锗探测器进行了测量,所得14MeV附近的176Hf(n,2n)175Hf反应截面实验值为(2100±85)mb,对实验结果与公开文献值和ENDF/B6.8评价库数据进行了比对。     

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.     

核反应截面协方差矩阵计算

一、引言裂变和聚变堆的设计者,在设计时需要许多准确的核反应截面数据。为准确估计设计中的误差,他们不仅要求知道反应截面的误差,还要求知道完整的协方差矩阵。因此,近年来评价核数据库要求给出反应截面的协方差矩阵。IAEA也多次举行专家会议讨论有关协方差矩阵问题。本文介绍了由反应截面测量中各参量的误差及其相关系数计算反应截面协方差矩阵的方法。并用此方法计算了~(59)Co(n,p)~(59)Fe反应截面的协方差矩阵。     

14.8 MeV中子诱发78Kr(n,2n)77Kr反应截面的测量

快中子诱发(n,2n)反应截面的测量在核反应机制研究和核技术应用等方面有着广泛的应用价值。本文在中国原子能科学研究院的高压倍加器上,基于活化法实验测量了78Kr(n,2n)77Kr在148 MeV能点的反应截面。样品靶为高纯78Kr气体样品,用十万分之一天平称重得到78Kr的质量,将两片高纯93Nb薄片分别固定在样品靶两侧以监测中子注量率。利用T(d,n)4He反应产生148 MeV中子,轰击距中子源约10 cm的样品靶大于4 h后,用准确刻度过效率的HPGe探测器测量活化产物 77Kr和92Nbm的活度。利用蒙特卡罗程序计算中子注量率修正、样品自吸收修正、样品几何修正等因子,得到了78Kr(n,2n)77Kr的反应截面,并将结果与文献值和评价数据库进行了比较。研究结果有助于提高78Kr(n, 2n)77Kr反应截面测量和评价的水平。     

20 MeV以下快中子与~(56)Fe反应的理论计算及评价

正我国国防核装置的研发始终把Fe列为16个最重要的核素中仅有4个结构材料核之一。现今,国际合作评价库CIELO则把它列为最重要的6个核素之一(~1 H、~(16) O、~(56) Fe、~(235,238) U、~(239) Pu),特别是56Fe的非弹性散射截面~(56)Fe(n,inl)在核装置中子输运计算中扮演着重要的角色。本工作首先基于详细的实验数据分析,并利用反应道截面之间的关系,给出了20 MeV以下n+~(56)Fe非弹反应道建立非弹截面的评价推荐值;其次,为获得20 MeV以下n+~(56)Fe各反应道的全能区的反应数据,本工作开展了重要反应道截面     

14MeV中子反应截面的综合评价

文章叙述了14MeV中子反应截面综合评价的基本思想及数学处理方法。对~(56)Fe(n,p)~(56)Mn,~(27)Al(n,α)~(24)Na,~(63)Cu(n,2n)~(62)Cu,~(27)Al(n,p)~(27)Mg和~(41)K(n,p)~(41)Ar等五种反应截面进行综合评价,得到了较满意的结果。评价值分别为109.7,114.2,533.5,71.8和48.0mb,误差(1σ)分别为0.88%,1.1%,1.5%,2.6%和3.2%。     

~(58)Ni全套中子核数据的评价(英文)

评价了~(58)Ni在10~(-5)eV到20MeV能区的全套中子数据。数据包括全截面、弹性、去弹、总非弹,九条分立能级和连续态的非弹及(n,2n),(n,3n),(n,n′α) (n,αn′),(n,n′p) (n,pn′),(n,p),(n,d),(n,t),(n,α)和(n,γ)反应截面,还包括次级中子角分布、双微分截面(DDCS),γ产生数据和共振参数。评价依据直到1995年测量的实验数据和用UNF程序的理论计算。评价数据以ENDF/B-6格式入中国评价核数据库第三版(CENDL-3)。     

天然镍全套中子核数据的评价(英文)

评价了天然镍在10~(-5)eV～20MeV能区的全套中子数据。数据包括全截面、弹性、去弹、总非弹、33条分立能级和连续态的非弹及(n,2n),(n,3n),(n,n′α) (n,αn′),(n,n′p) (n,pn′),(n,p),(n,d),(n,t),(n,α),(n,γ)和(n,2p)反应截面,还包括有关反应的次级中子角分布、双微分截面(DDCS),γ产生数据和共振参数。评价依据直到1996年测量的实验数据和用UNF程序的理论计算。评价数据以ENDF/B-6格式给出并纳入中国评价核数据库第三版(CENDL-3)。     

235U快中子核反应截面协方差评价

本文基于最小二乘不确定度传递方法,建立²³⁵U中子裂变核反应截面模型依赖型与非模型依赖型协方差评价体系。通过针对实验测量较丰富的中子反应总截面、辐射俘获、(n,2n)等核反应实验数据不确定度源项分析,为协方差评价提供实验基础,并给出对应核反应截面的非模型依赖型协方差评价数据。通过开展快中子能区²³⁵U核反应理论模型参数灵敏度计算与分析,导出实验测量缺乏的核反应截面模型依赖型协方差评价数据。经上述系统评价,所得协方差数据与核反应截面中心值研究过程自洽、物理合理,并按国际标准ENDF-6格式输出,便于核工程用户使用。     

Response of multi-step compound pre-equilibrium reaction cross sections for the (p,n) reactions to forms of optical model parameters

In furtherance to improving agreement between calculated and experimental nuclear data, the nuclear reaction code GAMME was used to calculate the multistep compound(MSC) nucleus double differential cross sections(DDCs) for proton-induced neutron emission reactions using the Feshbach-Kerman-Koonin(FKK) formalism. The cross sections were obtained for reactor structural materials involving ~(52)Cr(p, n)~(52)Mn,~(56)Fe(p,n)~(56)Co, and ~(60)Ni(p, n)~(60)Cu reactions at 22.2 MeV incident energy using the zero-range reaction mechanism. Effective residual interaction strength was 28 MeV, and different optical potential parameters were used for the entrance and exit channels of the proton-neutron interactions. The calculated DDCs were fitted to experimental data at the same backward angle of 150°, where the MSC processes dominate. The calculated and experimental data agree well in the region of pre-equilibrium(MSC) reaction dominance against a weaker fit at the lower emission energies. We attribute underestimations to contributions from the other reaction channels and disagreement at higher outgoing energies to reactions to collectively excited states. Contrary to the FKK multi-step direct calculations, contributions from the higher stages to the DDCs are significant. Different sets of parameters resulted in varying levels of agreement of calculated and experimental data for the considered nuclei.     

Uncertainty quantification in (α,n) neutron source calculations for an oxide matrix

We present a methodology to propagate nuclear data covariance information in neutron source calculations from (α,n) reactions. The approach is applied to estimate the uncertainty in the neutron generation rates for uranium oxide fuel types due to uncertainties on 1) ^17,18 O(α,n) reaction cross-sections and 2) uranium and oxygen stopping power cross sections.The procedure to generate reaction cross section covariance information is based on the Bayesian fitting method implemented in the R-matrix SAMMY code. The evaluation methodology uses the Reich-Moore approximation to fit the ^17,18 O(α,n) reaction cross-sections in order to derive a set of resonance parameters and a related covariance matrix that is then used to calculate the energy-dependent cross section covariance matrix. The stopping power cross sections and related covariance information for uranium and oxygen were obtained by the fit of stopping power data in the α-energy range of 1 keV up to 12 MeV.Cross section perturbation factors based on the covariance information relative to the evaluated ^17,18 O(α,n) reaction cross sections, as well as uranium and oxygen stopping power cross sections, were used to generate a varied set of nuclear data libraries used in SOURCES4C and ORIGEN for inventory and source term calculations. The set of randomly perturbed output (α,n) source responses, provide the mean values and standard deviations of the calculated responses reflecting the uncertainties in nuclear data used in the calculations. The results and related uncertainties are compared with experiment thick target (α,n) yields for uranium oxide.     

Covariances of Evaluated Nuclear Cross Section Data for Th, W and Mn

The EMPIRE code system is a versatile package for nuclear model calculations that is often used for nuclear data evaluation. Its capabilities include random sampling of model parameters, which can be utilised to generate a full covariance matrix of all scattering cross sections, including cross-reaction correlations. The EMPIRE system was used to prepare the prior covariance matrices of reaction cross sections of ²³²Th, ^{180,182,183,184,186}W and ⁵⁵Mn nuclei for incident neutron energies up to 60 MeV. The obtained modelling prior was fed to the GANDR system, which is a package for a global assessment of nuclear data, based on the Generalised Least-Squares method. By introducing experimental data from the EXFOR database into GANDR, the constrained covariance matrices and cross section adjustment functions were obtained. Applying the correction functions on the cross sections and formatting the covariance matrices, the final evaluations in ENDF-6 format including covariances were derived. In the resonance energy range, separate analyses were performed to determine the resonance parameters with their respective covariances. The data files thus obtained were then subjected to detailed testing and validation. Described evaluations with covariances of ²³²Th, ^{180,182,183,184,186}W and ⁵⁵Mn nuclei are included into the ENDF/B-VII.1 library release.     

A new approach for cross section evaluations in the high energy region

A new approach for a cross section evaluation above the resonance region is introduced and described. It employs the KALMAN code and the Monte-Carlo tuning program called the TalysTuner to automatically adjust the model parameters. As a nuclear reaction model code, TALYS is used. The cross sections for ²³⁵U are calculated up to 150 MeV using the new method and compared with the available experimental data and other evaluations.     

基于白光中子源的169Tm辐射俘获截面测量和共振参数分析

中子辐射俘获截面及共振参数在核工程设计、核天体物理等研究领域中有重要的应用价值。在中国散裂中子源（CSNS）反角白光中子束线（Back n）上,使用C6D6测量系统开展了169Tm辐射俘获反应测量。通过脉冲高度权重技术、共振吸收法和饱和归一法得到169Tm辐射俘获反应的产额。利用SAMMY程序拟合169Tm的产额数据,得到169Tm在1~100 eV能量区间的共振能量、中子宽度、辐射俘获宽度等共振参数。使用实验测得的共振参数和Reich Moore近似计算了169Tm在1~100 eV能量区间的辐射俘获截面。实验测量结果与ENDF/B Ⅷ.0数据库的推荐值总体符合较好,部分共振参数和截面存在一定的差异。产生这些差异的原因与Back n的源中子能谱结构、能量分辨率、实验本底的精度有关。     

用金活化法测定加速器中子源中子注量及MCNP4C修正

在四川大学720所2.5MeV静电质子加速器上,由核反应7Li(p,n)7Be,T(p,n)3He产生中子,对中国工程物理研究院研制的新型中子探测器进行效率刻度实验中,需要知道探测器位置处的中子绝对注量,为此我们测量了0.165、0.352、0.576、1.400MeV四个能点的中子注量。测量方法采用的是金活化法,在实验测量中,由靶头材料、冷却水层和样品的包层材料等引起的多次散射效应及中子在样品中的自屏蔽效应等均对实验结果产生影响。这些因素在实验中不可避免,也难以通过实验方法扣除,因此用Monte Carlo程序MCNP4C对上述效应进行了修正计算。     

Evaluation Methods of Resonance Absorption for System with Pellet Surrounded by Fuel Solution

Abstract

The reaction cross sections of ²⁷Al(n, p)²⁷Mg, ²⁷Al(n, a)²⁴Na, ⁵⁶Fe(n, p)⁵⁶Mn, ⁹⁰Zr(n, 2n)^89m+gZr and ⁹³Nb(n, 2n)^92mNb have been measured by the activation method in an energy range of 13.3–14.9 MeV using the intense D-T neutron source, FNS. Absolute flux was determined by the associated α-particle counting method incorporated with neutron spectra obtained from both a Monte Carlo calculation and a time-of-flight measurement. Corrections were extensively performed not only for the neutron flux determination, but also for the low energy neutron contribution to the reaction rates. The present data were compared with comprehensive evaluations as well as recent experimental data. The measured cross sections of ²⁷Al(n, a)²⁴Na, ⁵⁶Fe(n, p)⁵⁶Mn and ⁹⁰Zr(n, 2n)^89m+gZr are generally in good agreement within experimental errors with the values in both the JENDL Dosimetry File and IRDF-90. It is also shown that there are the overestimation of the cross sections for ⁹³Nb(n, 2n)^92mNb in the JENDL Dosimetry File, and the over- estimation and underestimation of the cross section for ²⁷Al(n, p)²⁷Mg in the JENDL Dosimetry File and IRDF-90, respectively.     

CENDL-3.2与ENDF/B-Ⅷ.0的56Fe评价截面对屏蔽计算影响研究

CENDL-3.2评价库对56Fe非弹性散射截面进行了更新,为了验证其与ENDF/B-Ⅷ.0评价库中截面以及屏蔽计算能力的差异,通过NJOY2016程序对56Fe共振重造后的非弹性散射、总截面等微观截面进行了比较;并制作了多群截面,在56Fe非弹性散射能量范围对以56Fe为主要核素的3个系列屏蔽基准题ILL-Fe、OKTAVIAN-Fe、IPPE-Fe进行了屏蔽计算性能的比较。结果表明,CENDL-3.2评价库的非弹性散射截面在4~12 MeV能量范围内低于ENDF/B-Ⅷ.0评价库的结果;多群截面基准题验证表明,CENDL-3.2评价库计算结果与实验值总体符合较好;对于OKTAVIAN-Fe基准题,在0.1~1 MeV能量范围内两评价库计算结果吻合较好。此外,所有基准题验证结果都有共同的现象,即在56Fe非弹性散射截面占主要贡献的1~10 MeV能量范围内,CENDL-3.2的计算结果比ENDF/B-Ⅷ.0的计算结果偏高。