堆用蒙卡程序燃耗计算功能开发

堆用蒙卡程序(RMC)是由清华大学工程物理系REAL实验室自主开发的用于反应堆物理分析的中子输运蒙卡程序,本文主要介绍其燃耗计算功能的开发与验证。RMC的燃耗计算功能具有的特点:内部耦合ORIGEN,相比于外耦合方式,更加灵活和高效;使用基于能谱的单群截面统计方法,可在保证精度的前提下,显著提高计算效率;采取预估修正和中点近似等多种燃耗步策略,减小大燃耗步长时的计算误差。通过计算压水堆栅元、沸水堆组件、快堆等一系列基准题和算例,验证了RMC燃耗计算的正确性和速度优势。     

基于高阶CRAM的燃耗程序开发与验证

本文基于高阶切比雪夫有理近似方法（CRAM）研制了点燃耗程序ICRAM，并内耦合于蒙特卡罗输运程序OpenMC，形成了一套燃耗计算分析程序OPICE。与传统部分分式分解（PFD）形式的CRAM相比，高阶不完全局部分解（IPF）形式的CRAM具有数值稳定性好、计算精度高和步长包容性更好等特点，满足高保真燃耗计算发展的需求。为提高耦合计算精度，OPICE采用了预估-校正和子步法两种耦合策略，支持纯衰变、定通量和定功率3种计算模式。通过OECD/NEA压水堆栅元燃耗基准题和快堆燃耗基准题的验证，程序计算结果与实验值及各参考值吻合良好，初步验证了OPICE的正确性与有效性。     

基于最佳一致逼近多项式的燃耗计算方法研究

本文研究了一种基于最佳一致逼近多项式（MMPA）的燃耗计算方法求解燃耗方程。相比于切比雪夫有理近似方法（CRAM）和围道积分有理近似方法（QRAM）,MMPA方法只需一次矩阵求逆计算即可求解燃耗方程,且所有计算都是实数运算,具有数值稳定性好、求解效率高等优点。进一步研制了基于MMPA方法的点燃耗程序AMAC,并耦合蒙特卡罗输运程序OpenMC,采用衰变例题、固定辐照例题、OECD/NEA压水堆栅元燃耗基准题和沸水堆组件燃耗基准题进行验证,程序计算结果与实验值及各参考值吻合良好,初步验证了MMPA方法在理论和数值上的正确性和有效性。     

聚变-裂变混合堆程序开发及验证

针对聚变-裂变混合堆设计研究中原有燃耗计算程序MONK9A耗时长等问题,利用MCNP和SCALE5.1程序包中的Origen-s程序开发出1套可用于先进反应堆设计的燃耗耦合程序MOCouple-s.选取了压水堆燃耗基准题、ADS基准题对MOCouple-s程序进行了验证,结果表明,MOCouple-s程序关于反应性和核素成分的计算结果与实验测量结果和其他程序的计算结果符合良好,且在某些计算结果、参数设置、自动化执行等方面优于国内外类似程序.利用MOCouple-s程序对MONK9A程序在混合堆燃耗计算上的适用性进行了验证,结果差别不大,证明MONK9A程序用于混合堆初步研究设计得到的燃耗计算结果是可靠的.     

溶液堆物理计算程序FMCAHR燃耗功能及其验证

介绍了FMCAHR程序的燃耗计算模型及流程,并使用燃耗基准题和DRAGON程序对燃耗计算结果进行验证.验证结果表明,FMCAHR燃耗计算功能的准确性较高,适用于溶液堆的燃耗计算分析.     

燃耗截面基本库对输运燃耗耦合计算的影响分析

基于蒙特卡罗中子输运程序和ORIGEN2点燃耗程序的蒙特卡罗输运燃耗耦合计算方法应用广泛。但现有评价库中子连续截面的核素个数远小于燃耗计算涉及到的核素数量,即通过输运计算得到的燃耗截面不足以完全替代燃耗计算的基本库。采用经过栅元验证的蒙特卡罗燃耗程序MCBMPI,对最新的VERA燃耗计算基准题进行验证计算,对比分析不同的燃耗截面基本库对输运燃耗计算的影响。分析结果表明:1)在实际应用中尽量不要采用典型热中子截面库,会带来较大偏差;2)在燃耗计算核素替换较多的情况下,对该基准题而言,选取典型压水堆基本库还是典型快堆基本库,对结果影响不大,二者keff偏差在8‰以内,燃耗末期235U偏差在4‰以内,135Xe偏差在5‰左右;3)建议选取与研究对象能谱相近的基本库。     

基于CRAM的压水堆燃耗程序开发及验证

正基于CRAM方法开发了燃耗求解程序MCRAM,利用Cinder90燃耗数据库生成了3 400阶燃耗矩阵(图1),并耦合MCNP程序对重要的锕系核素和裂变产物核素的反应截面进行了修正。以OECD/NEA乏燃料成分基准数据库中的Takahama-3压水堆燃料组件为基准题,对MCRAM程序的计算结果进行验证,并与其他程序的计算结果进行比较。结果表明,MCRAM程序对重要裂变产物和主要锕系核素的计算结果相对误差小于5%,计算精度与ORIGIN2程序的相当(图2)。     

裂变产额数据对铅基快堆RBEC-M燃耗计算影响研究

第四代核能系统是一种具有更好安全性、经济竞争力、核废物减少,以及防止核扩散的先进核能系统,代表了先进核能系统的发展趋势和技术前沿。铅基快堆是第四代核能系统中重要堆型之一。目前国际上通用的反应堆程序,比如MCNP+ORIGEN、RMC或者Serpent,很多研究主要针对压水堆,国际上也有研究发现针对铅基快堆基准题RBEC-M,确定论方法和蒙卡方法计算结果有较大偏差。本文深入研究了蒙卡程序使用的裂变产额对计算结果的影响。首先对反应堆蒙特卡罗程序RMC自带和燃耗库中的部分核素的裂变产额数据进行了更新,采用国际上著名RBEC-M基准题和OECD/NEA发布的快堆Pu循环燃耗基准题进行了验证分析,计算得到了裂变份额数据对快堆燃耗计算的影响。计算结果表明:更新后的裂变产额数据对系统的有效增殖因子和主要重核的质量变化影响较小,但对部分裂变产物的质量变化影响较大,部分核素偏差超过86%。对于快堆Pu循环燃耗基准题,长寿命高放废物~(133)Cs和~(129)I的计算结果偏差分别可达22.4%和47.8%,这将对长寿命高放废物的嬗变效率和核燃料循环有重要影响。     

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。     

确定论方法计算气冷快堆燃耗的程序系统开发

为了快速准确地求解气冷快堆的燃耗问题,开发了一套确定论方法程序系统。基于通用的燃耗过程计算框架,选取了输运程序WIMS和扩散程序CITATION,编写了相关的数据转换、截面管理、燃耗插值、自动控制程序,实现了WIMS和CITATION程序的耦合,并对一个气冷快堆模型燃耗过程进行了模拟。结果表明:开发的确定论方法程序系统是可行的,具备分析气冷快堆燃耗过程中重要物理特性的功能。     

MUSE-F1.0输运库的开发与基准验证

运用NJOY99程序,以微观评价库ENDF/B-Ⅶ.0为基础,开发了适用于快堆研究设计的175群中子、42群光子的多群截面数据库MUSE-F1.0。采用权重谱thermal--1/e--fast reactor-fission+fusion及勒让德P6近似。采用ANISN程序,从临界计算及屏蔽计算两方面对该库进行了较全面的检验;屏蔽检验涉及裂变堆、聚变堆、加速器等装置屏蔽材料所常用的相关核素截面数据的检验。检验结果表明：MUSE-F1.0在临界计算及屏蔽计算方面具有较高的精确度和较强的适用性,可满足快堆设计研究方面的应用要求。     

Analysis of a multigroup stylized CANDU half-core benchmark

An 8-group cross section library is provided to augment a previously published 2-group 3D stylized half-core Canadian deuterium uranium (CANDU) reactor benchmark problem. Reference eigenvalues and selected pin and bundle fission rates are also included. This benchmark is intended to provide computational reactor physicists and methods developers with a stylized model problem in more than two energy groups that is realistic with respect to the underlying physics. In addition to transport theory code verification, the 8-group energy structure provides reactor physicist with an ideal problem for examining cross section homogenization and collapsing effects in a full-core environment. To this end, additional 2-, 4- and 47-group full-core Monte Carlo benchmark solutions are compared to analyze homogenization-free transport approximations incurred as a result of energy group condensation.     

Development of a 2-D 2-group neutron noise simulator for hexagonal geometries

In this paper, the development of a neutron noise simulator for hexagonal-structured reactor cores using both the forward and the adjoint methods is reported. The spatial discretisation of both 2-D 2-group static and dynamic equations is based on a developed box-scheme finite difference method for hexagonal mesh boxes. Using the power iteration method for the static calculations, the 2-group neutron flux and its adjoint with the corresponding eigenvalues are obtained by the developed static simulator. The results are then benchmarked against the well-known CITATION computer code. The dynamic calculations are performed in the frequency domain which leads to discarding of the time discretisation. Then, the developed 2-D 2-group neutron noise simulator calculates both the discretised forward and the adjoint reactor transfer function between a point source and its induced neutron noise, by assuming the neutron noise source as an “absorber of variable strength” type. The neutron noise induced by a “vibrating absorber” type of noise source may also be modeled using the calculated transfer function. The viability of the simulator is verified for different benchmark cases.     

JENDL-4.0 Benchmarking for Fission Reactor Applications

Benchmark testing for the newly developed Japanese evaluated nuclear data library JENDL-4.0 is carried out by using a huge amount of integral data. Benchmark calculations are performed with a continuous-energy Monte Carlo code and with the deterministic procedure, which has been developed for fast reactor analyses in Japan. Through the present benchmark testing using a wide range of benchmark data, significant improvement in the performance of JENDL-4.0 for fission reactor applications is clearly demonstrated in comparison with the former library JENDL-3.3. Much more accurate and reliable prediction for neutronic parameters for both thermal and fast reactors becomes possible by using the library JENDL-4.0.     

A comparative study of integral parameters for TRX and BAPL benchmark lattices of thermal reactors using the cross-section library generated from JENDL and JEF evaluated data files

The objective of this paper is to present the results of comparative study of integral parameters for TRX and BAPL benchmark lattices of thermal reactors. The nuclear data processing code NJOY'99 was deployed for the generation of the 69-group cross-section library from the basic evaluated nuclear data files JENDL-3.2 and JEF-2.2. TRX and BAPL benchmark lattices were modeled with optimized inputs, which were suggested in the final report of the WIMS Library Update Project Stage-I. The inputs were the results of a detailed parametric study of the WIMS input options and also optimized for accuracy. The integral parameters (such as k_eff, ρ²⁸, δ²⁵, δ²⁸, C¹) of five uranium-fuel thermal assemblies: TRX-1 and TRX-2 and BAPL-1, BAPL-2, and BAPL-3 were calculated with the help of WIMSD-5B code based on the generated 69-group cross-section library. The calculated results are compared with those of experiments and it is found that the obtained results between the two libraries are in good agreement with each other. Besides, the calculated integral parameters are also well consistent with the measured values, which reflect the validation of the generated 69-group cross-section library and this library thus obtained is necessary to meet up the nuclear data for neutronics calculation of TRIGA Mark-II research reactor at AERE, Savar, Dhaka, Bangladesh.     

铅基快堆氧化物燃料性能分析程序FUTURE开发与初步验证

为了对铅基快堆氧化物燃料元件稳态工况下的服役性能和行为演化进行模拟计算,本文基于串行的半隐式耦合求解方法开发了铅基快堆氧化物燃料性能分析程序FUTURE。程序采用两步分析法实现了铅基快堆氧化物燃料棒全域热力分析与局部行为模型的多物理场耦合计算。通过各计算模块与模型算例、基准公式和现有程序的对比分析,对FUTURE程序进行了各分离效应的初步验证。结果表明,FUTURE程序能准确模拟铅基快堆稳态工况条件下氧化物燃料元件内部的温度演化、结构变形、应力分布和相互作用,并实现对燃料重构、氧和钚元素的迁移、裂变气体释放和服役期内液态铅铋腐蚀等内容的计算模拟。     

Validation of CENDL and JEFF evaluated nuclear data files for TRIGA calculations through the analysis of integral parameters of TRX and BAPL benchmark lattices of thermal reactors

The aim of this paper is to present the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through the analysis of the integral parameters of TRX and BAPL benchmark lattices of thermal reactors for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. In this process, the 69-group cross-section library for lattice code WIMS was generated using the basic evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 with the help of nuclear data processing code NJOY99.0. Integral measurements on the thermal reactor lattices TRX-1, TRX-2, BAPL-UO₂-1, BAPL-UO₂-2 and BAPL-UO₂-3 served as standard benchmarks for testing nuclear data files and have also been selected for this analysis. The integral parameters of the said lattices were calculated using the lattice transport code WIMSD-5B based on the generated 69-group cross-section library. The calculated integral parameters were compared to the measured values as well as the results of Monte Carlo Code MCNP. It was found that in most cases, the values of integral parameters show a good agreement with the experiment and MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through benchmarking the integral parameters of TRX and BAPL lattices and can also be essential to implement further neutronic analysis of TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh.     

Development and Verification of an Efficient Spatial Neutron Kinetics Method for Reactivity-Initiated Event Analyses

A space/time nodal diffusion code based on the nodal expansion method (NEM), EPISODE, was developed in order to evaluate transient neutron behavior in light water reactor cores. The present code employs the improved quasi-static (IQS) method for spatial neutron kinetics, and neutron flux distribution is numerically obtained by solving the neutron diffusion equation with the nonlinear iteration scheme to achieve fast computation. A predictor-corrector (PC) method developed in the present study enabled to apply a coarse time mesh to the transient spatial neutron calculation than that applicable in the conventional IQS model, which improved computational efficiency further. Its computational advantage was demonstrated by applying to the numerical benchmark problems that simulate reactivity-initiated events, showing reduction of computational times up to a factor of three than the conventional IQS. The thermohydraulics model was also incorporated in EPISODE, and the capability of realistic reactivity event analyses was verified using the SPERT-III/E-Core experimental data.     

池式钠冷快堆系统分析程序开发

针对池式钠冷快堆的特点,在对快堆系统的水力模型、热工模型和中子动力学模型进行详细分类和建模的基础上,利用FORTRAN95语言开发了可用于池式钠冷快堆事故分析的系统分析程序（FASYS程序）。以中国实验快堆为计算对象对FASYS程序模型进行了初步验证,所获得的结果和试验值与其他系统程序计算值符合良好,证明了所开发的系统分析程序的正确性。