基于离散纵标法的三维耦合燃耗与活化计算方法的发展

燃耗与活化分析在反应堆的燃耗管理与辐射屏蔽设计分析中起关键性的作用.基于一维、二维的输运程序的燃耗与活化分析方法难以解决复杂几何和强烈各向异性散射问题.本文通过耦合三维离散纵标(SN)方法粒子输运程序以及指数欧拉法活化计算程序,发展了快速精确的三维耦合燃耗与活化计算方法.该方法考虑了共振自屏效应动态修正,并采用重要核素...     

三维平几何各向异性散射问题的节块离散纵标输运计算

本文给出了三维（ｘ、ｙ、ｚ）几何各向异性散射问题的节块离散纵标输运计算模型，并研制了相应软件ＮＯＴＲＡＮ／３Ｄ，得到了满意的计算结果。     

角度自适应离散纵标输运计算方法

精确的屏蔽计算方法是核装置辐射屏蔽设计的重要基础,离散纵标法(SN)是主要的屏蔽计算方法之一。本文基于价值理论的目标导向与角度自适应相结合的方法,有效地减弱了角度的离散误差。求解输运共轭方程获得目标函数的重要性分布,采用局部角度离散误差与目标函数的重要性加权,产生后验误差估计,为角度自适应过程提供判断依据。角通量密度的映射采用多项式权重法和球谐函数拟合法。数值结果表明,对于具有直孔道或曲折孔道的屏蔽问题,在相同精度下离散角度数减少了1～2个数量级,极大地减少了计算量。角度自适应方法以较少的离散方向获得了准确的计算结果,有效地减弱了角度离散误差对屏蔽计算精度的影响。     

二维输运燃耗耦合程序开发与验证

为准确计算反应堆内燃耗问题,建立了基于二维离散纵标法及BATEMAN燃耗方法的输运燃耗耦合计算方法,并开发相应的计算程序。基于ENDF/B-Ⅶ评价库开发了175群中子和42群光子截面数据库MUSE-F1.0,采用OECD/NEA发布的MOX燃料快堆基准题对耦合计算方法及程序系统进行验证计算。结果表明,耦合计算程序结果与基准题吻合良好,误差在8%以内,初步验证了耦合计算程序在快堆嬗变工程应用中的可行性。     

三维离散纵标方法在反应堆精细屏蔽计算中的应用研究

随着计算能力的发展,三维离散纵标方法(SN)已逐渐为反应堆辐射屏蔽计算的主流方法之一。本文就三维SN方法应用于反应堆的精细化屏蔽建模与分析的关键问题进行了研究,主要包括精细化几何建模、堆芯固定源模型的创建和数据库截面参数敏感性分析等内容。在此基础上,本文以典型的压水反应堆为对象,构建了精细的三维SN计算分析模型,以压力容器快中子注量率为算例,完整实现了反应堆的精细化三维SN建模与分析,并将三维SN结果和蒙特卡罗方法的计算结果进行了比较分析。对比结果表明,精细化三维SN方法具有较高的计算精度,验证了三维SN方法在反应堆精细屏蔽计算问题中的有效性和正确性。     

二维中子输运问题的离散纵标短特征线方法研究

离散纵标法是求解中子输运方程的主要数值方法之一,空间变量离散及误差控制对保证输运计算精度至关重要。传统有限差分离散方法对于特定模型会产生非物理振荡问题,粗网精度不足使得低阶差分方法的应用具有局限性。本文研究了二维常数和线性短特征线方法,短特征线空间离散基于中子输运的特征线解,根据输运方程的空间矩守恒构造网格角通量密度完成输运方程求解。选取固定源和临界问题进行测试验证并分析了网格敏感性。数值结果表明,线性短特征线离散对网格敏感性较低,较常数短特征线和低阶差分方法具有更高的计算精度及效率。     

三维离散纵标方法在RPV快中子注量率计算中的初步应用

离散纵标(又称SN)方法是反应堆压力容器快中子注量率计算中最常用的方法之一.计算机的飞速发展和三维离散纵标方法的不断完善,使得三维离散纵标法应用于工程设计成为可能.本文以某压水堆为研究对象,采用基于三维全堆芯中子学程序SCIENCE的计算结果,使用三维离散纵标方法程序计算了压力容器内表面快中子注量率三维分布,并分别与采...     

一阶中子输运方程的离散纵标有限元方法研究

利用离散纵标(SN)方法离散SN一阶多群中子输运方程,建立了基于剖分单元的有限元离散与非结构网格扫描方法相结合的求解模型。针对给定的三角形单元应用Galerkin变分,获得线性方程组,通过引入定解条件,求解线性代数方程组,获得该三角形单元所有节点上的角通量,然后对其他三角形单元进行扫描,从而解出所有节点处中子通量密度。根据上述理论模型,编制了相应的计算程序FEGT,对一系列例题进行校验的数值结果表明,该程序的计算精度满足要求。     

离散纵标屏蔽计算空间自适应方法研究

离散纵标法是目前国际上通用的核装置辐射屏蔽计算的主要方法之一。空间变量离散误差对离散纵标屏蔽计算的精度至关重要。实际屏蔽计算中存在较强的非均匀性,传统的空间离散方法不能高效地在全局达到较高精度。本文采用线性间断有限元方法进行空间离散,由基于两网格的误差估计和基于残差的误差估计驱动空间网格自适应细化;基于树状结构的六面体网格,对传统的输运扫描进行改进,在粗-细网映射中保持零阶、一阶空间矩守恒。数值结果表明：线性间断有限元方法具有较好的射线传递特性和空间收敛性;自适应算法能较好地对通量密度间断、通量密度梯度大、光学厚度大的区域进行网格细化。在达到相同计算精度时,自适应方法所需的网格数较均匀细分的方法减少约1个数量级,能有效提高屏蔽计算效率。     

三维中子-光子离散纵标输运程序CTDOS开发与验证

基于离散纵标方法开发了三维输运屏蔽程序CTDOS测试版。该程序可实现三维几何系统内中子-光子输运过程k_eff和固定源计算问题的求解。采用CTDOS程序对Takeda基准题第一模型、Kobayashi基准题第一模型进行了计算,k_eff计算结果与基准题及TORT计算结果的最大误差小于0.04%,群通量密度计算结果与基准题及TORT计算结果的平均误差约为3%。结果初步证明：CTDOS开发过程及功能是正确的,计算精度较高。     

基于离散纵标法与蒙特卡罗方法的三维耦合程序开发

辐射屏蔽设计是核装置工程设计的核心内容之一。单一的离散纵标法(比如SN)或蒙特卡罗方法(MC)在大型核装置屏蔽计算分析方面均存在一定限制。为了满足大型复杂核装置精确辐射屏蔽计算要求,本文实现了三维SN-MC耦合方法,并发展了相应的三维耦合程序系统。该程序结合了SN方法解决深穿透问题的优势和MC方法模拟复杂几何的长处,克服两种方法的缺点,为保证屏蔽系统优化设计的质量提供有力的技术支持。采用接口程序和MC自定义源抽样程序将SN计算得到的粒子角注量率转换为MC计算所需的源粒子信息,为下一步MC计算提供源项,实现三维SN-MC耦合输运计算。采用MC、SN、SN-MC耦合三种方法对直角坐标系和圆柱坐标系下的测试例题进行了计算比较分析。计算结果吻合良好,初步证明了所开发的三维SN-MC耦合程序的正确性。     

基于蒙特卡罗方法与指数欧拉法耦合的反应堆三维燃耗计算程序的研发

燃耗计算在反应堆设计、分析研究中起着重要作用.一维、二维耦合燃耗程序因其几何限制难以满足先进反应堆精细设计分析的要求.本文研发了基于蒙特卡罗方法与指数欧拉法耦合的三维燃耗程序.程序采用编写耦合MCNP与FISPACT的接口程序的方法,结合了MCNP处理复杂几何能力强,FISPACT计算核素全面、能谱多样的特点,实现了考...     

基于蒙特卡罗方法的三维燃耗计算研究

采用通过编写连接MCNP程序和ORIGEN2程序的接口处理程序的方法进行快中子系统的燃耗计算。由MCNP、ORIGEN2、接口处理程序和截面文件组成的软件系统可用于燃料或堆芯非均匀布置快中子系统的燃料同位素成分和燃耗反应性损失计算,在燃耗反应性损失计算中采用了伪裂变产物的方法。介绍程序系统的研制情况,并给出用该软件系统计算中国实验快堆首炉堆芯和OECD/NEAMOX燃料快堆基准题的燃耗计算结果。     

A linear multiple balance method with high order accuracy for discrete ordinates neutron transport equations

A linear multiple balance method (LMB) is developed to provide more accurate and positive solutions for the discrete ordinates neutron transport equations. In this multiple balance approach, one mesh cell is divided into two subcells with quadratic approximation for the angular flux distribution. Four multiple balance equations are used to relate center angular flux with average angular flux by Simpson's rule. From the analysis of spatial truncation error, the accuracy of the linear multiple balance scheme is O(Δ⁴) whereas that of diamond differencing is O(Δ²). The positivity of the method is also stronger than that of diamond differencing. To accelerate the linear multiple balance method, we also describe an additive angular dependent rebalance factor scheme which combines a modified boundary projection acceleration scheme and the angular dependent rebalance factor acceleration scheme. It is demonstrated, via Fourier analysis of a simple model problem as well as numerical calculations, that the additive angular dependent rebalance factor acceleration scheme is unconditionally stable with spectral radius <0.2069 c (c being the scattering ratio). The numerical results tested so far on slab-geometry discrete ordinates transport problems show that the solution method of linear multiple balance with additive angular rebalance acceleration is effective and sufficiently efficient.     

Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises

The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor–corrector approach as the integration method to determine the fuel composition at a desired burnup step. Validation of BUCAL1 was done by code vs. code comparison. Results of two different kinds of codes are employed. The first one is CASMO-4, a deterministic multi-group two-dimensional transport code. The second kind is MCODE and MOCUP, a link MCNP–ORIGEN codes. These codes use different burnup algorithms to solve the depletion equations system. Eigenvalue and isotope concentrations were compared for two PWR uranium and thorium benchmark exercises at cold (300 K) and hot (900 K) conditions, respectively. The eigenvalue comparison between BUCAL1 and the aforementioned two kinds of codes shows a good prediction of the systems’ k-inf values during the entire burnup history, and the maximum difference is within 2%. The differences between the BUCAL1 isotope concentrations and the predictions of CASMO-4, MCODE and MOCUP are generally better, and only for a few sets of isotopes these differences exceed 10%.     

The infinite medium Green's function method for multigroup discrete ordinates transport problems in multi-layered slab geometry

Analytic solutions of the multigroup discrete ordinates transport equation with linearly anisotropic scattering and arbitrarily distributed source for multi-layered slab problems are obtained by using the infinite medium Green's function (IMGF) and Placzek's lemma. In this approach, the infinite medium Green's function is derived analytically by using the spectral analysis for the multigroup discrete ordinates transport equation and its transposed equation, and this infinite medium solution is related to the finite medium solution by Placzek's lemma. The resulting equation leads to an exact relation that represents the outgoing angular fluxes in terms of the incoming angular fluxes and the interior inhomogeneous source for each slab. For heterogeneous problems having multi-layered slabs, the slabs are coupled through the interface angular fluxes. Since all derivations are performed analytically, the method gives exact solution with no truncation error. After the interface angular fluxes are calculated by using an iterative method, the continuous spatial distribution of the angular flux (i.e. analytic solution) in each slab is given straightforwardly in terms of the IMGF and the boundary angular fluxes.     

Validation of IRBURN calculation code system through burnup benchmark analysis

Assessment of the reactor fuel composition during the irradiation time, fuel management and criticality safety analysis require the utilization of a validated burnup calculation code system. In this work a newly developed burnup calculation code system, IRBURN, is introduced for the estimation and analysis of the fuel burnup in LWR reactors. IRBURN provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP4C as well as the versatile code for calculating the buildup and decay of nuclides in nuclear materials, ORIGEN2.1, along with other data processing and linking subroutines. This code has the capability of using different depletion calculation schemes.