Transmission probability method based on triangle meshes for solving unstructured geometry neutron transport problem

In the advanced reactor, the fuel assembly or core with unstructured geometry is frequently used and for calculating its fuel assembly, the transmission probability method (TPM) has been used widely. However, the rectangle or hexagon meshes are mainly used in the TPM codes for the normal core structure. The triangle meshes are most useful for expressing the complicated unstructured geometry. Even though finite element method and Monte Carlo method is very good at solving unstructured geometry problem, they are very time consuming. So we developed the TPM code based on the triangle meshes. The TPM code based on the triangle meshes was applied to the hybrid fuel geometry, and compared with the results of the MCNP code and other codes. The results of comparison were consistent with each other. The TPM with triangle meshes would thus be expected to be able to apply to the two-dimensional arbitrary fuel assembly.     

Two-dimensional nodal transport method for triangular geometry

The advanced nodal method for solving the multi-group neutron transport equation in two-dimensional triangular geometry is developed. To apply the transverse integration procedure, an arbitrary triangular node is transformed into a regular triangular node using coordinate transformation. The angular distributions of intra-node neutron fluxes and its transverse-leakage are represented by the S_N quadrature set. The spatial distributions of neutron flux and source in the regular triangle are given approximately by an orthogonal quadratic polynomial, and the spatial expansion of transverse-leakage is approximated by a second-order polynomial. To establish a stable and efficient iterative scheme, the improved nodal-equivalent finite difference algorithm is used. The results for several benchmark problems demonstrate the higher capability of the method to yield the accurate results in significantly smaller computing times than those required by the standard finite difference method and the finite element spherical-harmonics method.     

Burnup Calculation Sensitivity and Uncertainty Analysis Method of Fission Yield and Decay Half-life

The calculation model of sensitivity coefficient for decay half-life and fission product yield in burnup calculation was derived based on generalized perturbation theory, which considered the interaction between nuclear concentration and neutron flux. A code was developed to calculate sensitivity and uncertainty of effective neutron multiplication factors and nuclide concentration caused by nuclear data. Covariance matrix of fission yield for a simplified burnup library was generated based on standard deviation data of independent fission yield in evaluated nuclear data library to improve the accuracy of uncertainty quantification. Uncertainties induced by decay half-life and fission yield on infinite neutron multiplication factors and nuclide concentration for TMI-1 pin-cell in the UAM burnup benchmark were quantified based on ENDF/B-Ⅶ.1. The numerical results show that the uncertainty of infinite neutron multiplication factors induced by decay half-lives and fission yields is low, while the uncertainty of concentration of some fission product nuclide is high.     

圆筒形薄壳热屈曲临界温升一般表达式构建

随着航天和深海工程中薄壳结构的广泛应用,圆筒形薄壳结构热屈曲问题越来越引起广泛关注.首先给出了圆筒形金属薄壳热屈曲临界温升的一般计算表达式,然后基于有限元方法建立了该薄壳结构热屈曲分析的数值模型,计算结果验证了热屈曲临界温升计算式的准确性,同时进行了多参数对比分析.构建的两种热屈曲临界温升计算方法,将对今后高温环境下圆筒形薄壳结构的科学设计具有重要参考作用.     

绿色表面活性剂喷雾降尘试验研究

为提高城市喷雾降尘效率,通过对喷雾降尘用表面活性剂进行筛选,旨在寻求更为优越的性能。采用实验研究方法,测定了APG、AES、1 307、CAO-30、BS-12和APG-ET表面活性剂的表面张力和润湿性,发现阴离子表面活性剂和非离子表面活性剂的性能明显优于两性离子表面活性剂,以烷基糖苷（APG）为主,与筛选出的另外两种性能优良的表面活性剂异构十三醇聚氧乙烯醚（1 307）和烷基糖苷酒石酸酯盐（APG-ET）进行复配,测定复配表面活性剂的表面张力和润湿性。结果表明：APG、1307、APG-ET三元复配体系的性能最佳,在m（APG）∶m(1307)∶m（APG-ET）=6∶2∶2条件下,表面活性剂溶液质量浓度为0.1%时的表面张力可达23.13 mN?m^-1、粉尘的沉降时间为10 s;通过降尘试验研究了复配表面活性剂在不同浓度下对细颗粒物的去除能力,当复配表面活性剂质量浓度为0.1%时,对PM₁₀、PM_2.5和PM_1.0的去除率分别达到86.2%、84.5%、76.5%。降尘用表面活性剂的最优配方为m（APG）∶m(1307)∶m（APG-ET）=6∶2∶2,喷雾质量浓度为0.1%,该研究为提高城市喷雾降尘效果提供了参考。     

小型可运输长寿命铅铋冷却快堆堆芯设计研究

为满足偏远地区供电需求,提出了一种小型可运输长寿命铅铋冷却快堆（STLFR）堆芯设计方案,额定热功率为20 MW,在不换料条件下可运行18 EFPY（有效满功率年）。为减小堆芯体积,堆芯采用蜂窝煤型燃料组件,内设若干冷却剂管道,管外为燃料,实现了较高的堆芯燃料体积占比。为展平堆芯径向功率分布,将堆芯燃料区沿径向划分为三区,分别采用不同的冷却剂管道尺寸。为降低堆芯高度,设计使用含高富集度⁶Li的液态锂作为吸收体的液态吸收体控制系统。为降低初始剩余反应性,在堆芯控制组件与安全组件中布置两组固定式可替换吸收体,分别在堆芯燃耗1/3和2/3寿期时替换为固定式反射体。提出的堆芯设计方案在整个运行寿期内满足热工设计限值,控制系统和安全系统能独立满足堆芯控制和停堆要求。采用准静态反应性平衡方法对5种典型无保护事故工况进行分析,初步证明了堆芯具有固有安全特性。     

全陶瓷微密封燃料有效多群截面计算方法研究

全陶瓷微密封（FCM）燃料是一种弥散颗粒燃料。由于弥散颗粒燃料存在双重非均匀性,传统的确定论方法及蒙特卡罗方法皆难以处理这种双重非均匀效应以获得有效多群截面。本文基于超细群方法建立FCM燃料的有效多群截面计算方法。为描述燃料棒内TRISO颗粒的非均匀性,在共振能量段,通过采用超细群方法求解包含TRISO颗粒的一维球模型得到超细群缺陷因子,通过超细群缺陷因子修正所有核素的超细群截面即可将颗粒和基质均匀化。由于TRISO颗粒在热能区也存在较强的自屏效应,在热能区,利用穿透概率及碰撞概率等价得到多群缺陷因子,通过多群缺陷因子修正所有核素的多群截面将燃料和基质均匀化。均匀化后的FCM燃料组件即可视为普通压水堆燃料组件进行共振计算。利用丹可夫修正因子等价得到FCM燃料组件各燃料棒的等效一维棒模型,对一维棒模型求解超细群慢化方程从而得到共振能量段的有效自屏截面。数值结果表明,该方法能有效处理FCM燃料的双重非均匀性,得到精确的有效自屏截面。     

表面活性剂与离子液体对β-葡萄糖苷酶的协同影响

为考察表面活性剂和离子液体1-乙基-3-甲基咪唑磷酸二乙酯盐（[EMIM]DEP）对类芽孢杆菌sp. LLZ1 β-葡萄糖苷酶活性的影响,在酶活测定体系中加入一定浓度的表面活性剂和[EMIM]DEP。结果表明：添加5%的[EMIM]DEP使β-葡萄糖苷酶的活性增强了12.00%,进一步添加0.1%鼠李糖脂、Span20、PEG4000和Tween80分别使酶活增强了21.85%、12.07%、8.57%和5.25%,而Triton X-100和SDS分别使酶活降低了4.59%和10.63%。动力学曲线和动力学参数表明随着表面活性剂和5%[EMIM]DEP对β-葡萄糖苷酶活性的增强,米氏常数K_m随之减小。圆二色谱（CD）分析表明分别经0.1%鼠李糖脂、Span20、PEG4000和Tween80处理后,β-葡萄糖苷酶的α-螺旋分别增加1.00%、0.78%、0.72%和0.80%,添加SDS导致α-螺旋减少5.72%。荧光光谱表明同时添加表面活性剂和5%[EMIM]DEP改变了β-葡萄糖苷酶的最大发射波长。差示扫描量热法(DSC)表明0.1%鼠李糖脂和5%[EMIM]DEP提高了β-葡萄糖苷酶的中点温度和平均展开焓。使用0.1%鼠李糖脂协同5%[EMIM]DEP水解纤维二糖,转化率提高了21.93%。     

小型铅铋冷却快堆瞬态安全分析

本文基于多通道热工模型与功率计算模型,在快堆分析程序SARAX的基础上开发了可用于分析小型铅铋冷却快堆在无保护超功率事故、无保护失流事故及无保护失热阱事故发生时瞬态安全特性的计算功能,并利用该程序计算了在不同事故情况下,堆芯反应性、功率以及热工参数随时间的变化,分析评价了堆芯的中子学和热工水力学性能。结果表明所设计的堆芯在发生事故时具有固有安全特性。     

A hybrid method to generate few-group cross sections for fast reactor analysis

The accuracy of fast reactor core calculation is usually determined by the accuracy of self-shielded few-group cross sections. To further improve the accuracy of cross section generation, a hybrid method is proposed. In the hybrid method, the Monte-Carlo method is used to deal with the resonance effect in both the resolved and unresolved resonance range. The self-shielded ultrafine-group total, fission and elastic scattering cross sections are tallied by the Monte-Carlo method. The scattering transfer matrices are then generated in a synthesis way by using the tallied elastic scattering cross sections and a problem-independent elastic scattering function. The angular flux moments for the group condensation are calculated in an explicit deterministic way. Several tests are done to verify the hybrid method. The results show that the hybrid method avoids the disadvantages of both the traditional deterministic method and the pure Monte-Carlo method. It is a more accurate method to generate the few-group cross sections for fast reactor cores.