一体化压水堆动态特性的仿真研究

研究了一体化压水堆堆芯模型的简化问题．用集总参数法建立了堆芯数学模型,包括中子动力学模型．反应性反馈模型和堆芯热工水力模型。得到了堆芯模型的增量式的传递函数方程组,并通过仿真对一体化压水堆和传统的分散布置压水堆的动态特性进行了比较,验证了模型的是合理性。     

直接三维中子输运特征线法并行计算软件的实现

作为中子输运问题的一种重要确定论方法,特征线法（MOC）具有强几何适应性、计算流程简洁、易于大规模并行的优点。ANT-MOC是自主开发的中国数值反应堆1.0（CVR1.0）中的三维特征线法中子输运计算程序,主要用于压水堆、快堆的堆芯输运计算。ANT-MOC支持基于构造实体几何（CSG）的复杂几何建模、高效的用户输入方式、面向矩形/六边形网格的射线追踪算法,以及基于轨迹链分解的并行算法和负载平衡策略。在国产超算上,ANT-MOC可以扩展到约10万处理器核,并行效率在50%以上。针对压水堆、快堆计算问题进行验证和参数敏感性分析,结果表明ANT-MOC计算结果具有较好的稳定性和准确度。     

压水堆核电站堆芯集中参数模型的微机仿真

阐述了PWR核电站堆芯的模型化问题,提适用于微机仿真的核电站堆芯的物理数学模型,将核电站堆芯分为三大块分别建立模型,中子动力学模块,反应性反馈模块,堆芯热力学模块,建立系统传递函数,运用MATLA仿真,得到良好结果。     

压水堆重反射层堆芯核热耦合高精度计算分析研究

重反射层的应用可提高反应堆中子经济性,其结构和中子吸收特性均与压水堆常规围板/反射层差异较大,因此对核设计程序的计算分析能力提出了新的要求。为分析重反射层建模方案对堆芯中子学计算结果的影响,使用先进中子学程序SCAP N和确定论堆芯高保真模拟程序NECP X对压水堆重反射层问题进行了高保真模拟,分析了5种反射层建模方案下计算结果的差异,并将高精度计算结果与商用核设计程序系统进行了对比。数值结果表明,重反射层水洞内冷却剂温度变化对计算结果影响较小;相较精确建模方案,重反射层铁水打混建模方案造成的反应性计算偏差在±30 pcm以内、组件相对功率分布计算偏差在±2%以内。     

基于PAnySimu的PWR核电站堆芯建模与仿真研究

基于PAnySimu仿真支撑系统对PWR核电站一回路堆芯部分进行建模与仿真分析.通过研究分析岭澳二期3/4号机组堆芯实际结构,将其分为功率计算、堆芯传递计算、控制棒引起的反应性、反应性反馈、毒物计算五个模型.在此基础上,分析堆芯中子通量,考虑控制棒位置、燃料和慢化剂温度、氙和钐中毒、硼浓度等因素对中子通量的影响.利用P...     

Implementation of control rod movement and boron injection options by using control variables in RELAP5/PARCS V2.7 coupled code

In present neutron kinetics codes, control rods banks do not have the possibility of dynamic movement during the simulation of a transient; besides it is necessary to send the boron concentration from the thermal-hydraulic code to the neutronic code to account for changes in cross-sections due to boron dilution. For instance, control rod movements are pre-programmed with simple instructions introduced before the beginning of the calculation. Hence, control rod positions are not related to the core characteristics and the control systems at any time of the simulation. This work presents the changes introduced in RELAP5/PARCS v2.7 codes to achieve that control rods and the boron injection become more dynamic and realistic components in such kind of simulators. Furthermore, in order to test the modifications introduced in both codes, it has been analyzed a boron injection transient in a typical PWR Nuclear Power Plant. The thermal-hydraulic model includes all the primary loop components of a PWR, the core fuel assemblies modeled with PIPE components, pumps, steam generators, pressurizer, etc. The neutronic representation of the reactor has been made in a one-to-one basis fuel channel model for the whole core.     

Implementation of control rod movement and boron injection options by using control variables in RELAP5/PARCS V2.7 coupled code

In present neutron kinetics codes, control rods banks do not have the possibility of dynamic movement during the simulation of a transient; besides it is necessary to send the boron concentration from the thermal-hydraulic code to the neutronic code to account for changes in cross-sections due to boron dilution. For instance, control rod movements are pre-programmed with simple instructions introduced before the beginning of the calculation. Hence, control rod positions are not related to the core characteristics and the control systems at any time of the simulation. This work presents the changes introduced in RELAP5/PARCS v2.7 codes to achieve that control rods and the boron injection become more dynamic and realistic components in such kind of simulators. Furthermore, in order to test the modifications introduced in both codes, it has been analyzed a boron injection transient in a typical PWR Nuclear Power Plant. The thermal-hydraulic model includes all the primary loop components of a PWR, the core fuel assemblies modeled with PIPE components, pumps, steam generators, pressurizer, etc. The neutronic representation of the reactor has been made in a one-to-one basis fuel channel model for the whole core.     

基于R2S方法的反应堆结构材料活化剂量计算研究

反应堆结构材料在堆芯中子辐照下由于中子活化反应而产生大量的放射性核素,其衰变光子是反应堆停堆检修、换料、退役过程中工作人员职业照射剂量的重要来源。本文基于严格两步法（R2S）,研究了反应堆结构材料栅元活化计算方法,并基于蒙卡粒子输运程序（MCNP）与点活化计算程序（ORIGEN）建立了反应堆结构材料活化剂量计算软件（MOCA）。通过开发功能接口与数据接口程序实现输运程序与活化计算程序的自动耦合,进而实现“中子输运-活化分析-剂量计算”全自动耦合分析。利用M5包壳活化计算模型、不锈钢活化计算模型和NUREG/CR-6115压水堆模型对MOCA进行基准验证,证明了MOCA的正确性与可靠性。      

Calculation of reactivity and power factors depending on the external source location

We used the neutron diffusion equation with external neutron sources, in cartesian geometry and the two groups of energy, to verify the influence of external neutron source locations in the calculation of reactivity and power factors. To this end, the Coarse Mesh Finite Difference (CMFD) method was applied to the adjoint flux calculation and to simplify reactivity calculation in PWR type reactor, using the output of the Nodal Expansion Method (NEM). Different locations on the two-dimensional plane, as well as different types of fuel elements in the reactor core were used in the present study.     

Development of an asymmetric multiple-position neutron source (AMPNS) method to monitor the criticality of a degraded reactor core

Tne analytical/experimental method has been developed to to monitor the subcritical reactivity and unfold the k_∞ distribution of a degraded reactor core. The method uses several fixed neutron detectors and a ²⁵²Cfneutron source placed sequentially in multiple positions in the core. Therefore, it is called the asymmetric multiple-position neutron source (AMPNS) method. The AMPNS method employs the nucleonic codes to analyze in two dimensions the neutron multiplication of a ²⁵²Cf neutron source. An optimization program, GPM, was utilized to unfold the k_∞ distribution of the degraded core, in which the desired performance measure minimizes the error between the calculated and the measured count rates of the degraded reactor core. The analytical/experimental approach is validated by performing experiments using the Penn. State Breazeale TRIGA reactor (PSBR). A significant result of this study has been to provide a means to plan the source and detector placements and assign core cells to the damaged TMI-2 core as well as to monitor the criticality during the recovery period.     

双环路压水堆非对称入口条件下物理-热工特性研究

双环路压水堆存在反应堆入口流量、温度不对称的非正常运行工况。本文建立了基于CFD方法的反应堆整体三维流场模型,并耦合中子动力学计算程序和RELAP5程序,对这种非对称入口条件下的反应堆物理-热工特性进行了数值模拟。结果表明：反应堆入口流量不对称会加剧堆芯入口流量分配的不均匀性,并进一步导致局部功率变化,对反应堆安全不利;在入口温度不对称的条件下,冷却剂在下腔室的混合非常不充分,并导致堆芯入口温度分布不均匀,引起局部功率变化较大,对反应堆安全不利。     

Design concept of a small nuclear reactor for large-diameter NTD-Si using a conventional PWR full-length fuel assembly

A design concept for a small nuclear reactor dedicated to large-diameter neutron transmutation doping silicon (NTD-Si) is proposed. Conventional PWR (Pressurized Water Reactor) full-length fuel assembly is used to assure stable and reliable supply of fuel. Criticality, neutron transportation, and core burn-up calculations are performed using the MVP/GMVP II code and MVP-BURN code. The calculation results show that the proposed reactor can be critical over 18 years, and excess reactivity can be suppressed by a combination of Gd₂O₃ burnable poison and soluble boron. Preliminary steady-state single-channel thermal hydraulic analysis showed that heat removal from core is possible under 1 atm operating pressure. Si ingots up to 30 cm in diameter can be irradiated in the reactor irradiation channels, and the uniform irradiation condition can be achieved for a large-diameter Si ingot.     

Development of a control rod depletion methodology for the Westinghouse NEXUS system

The NEXUS project is an effort to merge and modernize the methods employed in Westinghouse PWR and BWR steady-state reactor physics codes. The NEXUS system relies on a once-through nodal cross-section generation methodology with an innovative and efficient technique for pin power recovery. The pin power methodology overcomes a well-known limitation of existing methodologies, namely the incapacity to properly account for heterogeneity changes due to the depletion environment. The so-called control rod history problem where control rods are repeatedly inserted and withdrawn during core depletion is a good example of such a case. In addition to the control rod history impact on pin power distributions, the insertion of control rods during extended periods leads to significant control rod depletion that affects the reactivity worth of the control rods which in turn can have a significant impact on pin powers. The importance of accurately predicting pin powers, combined with the need to adequately estimate the reactivity worth and nuclear end of life of control rods in BWRs and in generation III+ PWRs, has motivated the development of a novel control rod depletion model. This methodology and its numerical qualification, initially for PWR application only, is the topic of this paper. The focus is on describing the salient features of the model and on illustrating its performance by means of numerical experiments. It is shown that together with the NEXUS pin power recovery model, the control rod depletion methodology accurately predicts the reactivity feedback from repeated control rod insertions in a PWR core.     

An experimental investigation of accumulation and transmutation behavior of americium in the MOX fuel irradiated in a fast reactor

Americium isotopes generated in the MOX fuel irradiated in the experimental fast reactor JOYO were analyzed by applying a sophisticated radiochemical technique. Americium was isolated from the irradiated MOX fuel by a combined method of anion-exchange chromatography and oxidation of Am. The isotopic ratios of americium and its content were determined by thermal ionization mass spectroscopy and α-spectrometry, respectively. The americium isotopic ratio was similar for all the specimens, but was significantly different from that of PWR-MOX. On the basis of present analytical results, the accumulation and transmutation behavior of americium nuclides in a fast reactor is discussed from the viewpoints of neutron spectrum dependence and the isomeric ratio of the ²⁴¹Am capture reaction. The estimated isomeric ratio is about 87%, which is close to the latest evaluated value. A rapid estimation method of Am content by using the ²⁴⁰Pu to ²³⁹Pu ratio was adopted and proved to be valid for the spent fuel irradiated in the fast reactor.     

基于OpenFOAM的中子输运动力学求解器ntkFoam研究

由于中子输运模拟的复杂性及其与其他物理过程耦合的困难性，全堆芯精细中子输运-热工水力多物理计算是核工程领域的难点。本文基于有限体积C++开源软件OpenFOAM，采用有限体积法建立稳态和瞬态中子输运动力学方程数值求解模型，开发了中子输运动力学求解器ntkFoam。通过对多个基准问题进行模拟验证表明，本文建立的ntkFoam求解器能准确模拟中子输运动力学问题，并能很好地适应于不同维度及复杂几何条件；可实现中子输运、传热传质的精细耦合，为基于中子输运计算的全堆芯多物理模拟提供了一些精确耦合的思路与方法。      

Influence of the external neutron sources in the criticality prediction using 1/M curve

The influence of external neutron sources in the process to obtain the criticality condition is estimated. To reach this objective, the three-dimensional neutron diffusion equation in two groups of energy is solved, for a subcritical PWR reactor core with external neutron sources. The results are compared with the solution of the corresponding problem without external neutron sources, that is an eigenvalue problem. The method developed for this purposes it makes use of both the nodal method (for calculation of the neutron flux) and the finite differences method (for calculation of the adjoint flux). A coarse mesh finite difference method was developed for the adjoint flux calculation, which uses the output of the nodal expansion method. The results regarding the influence of the external neutron source presence for attaining criticality have shown that far from criticality it is necessary to calculate the reactivity values of the system.     

Examination of pin-by-pin fission rate distribution in large geometry evaluated by the Monte-Carlo method

Accuracy and reliability of pin-by-pin fission rate distribution in large geometries calculated by the multi-group Monte-Carlo method is examined through comparison with a deterministic transport code based on the method of characteristics. Various calculation geometries from a single assembly to a PWR full core are used for comparison of pin-by-pin fission rate distribution. An integral parameter, i.e., k-effective, can be accurately calculated by the Monte-Carlo method with a practical number of neutron histories (10⁶–10⁷) regardless of the size of the calculation geometry. On the other hand, comparison with the deterministic calculation shows that the estimated statistical errors for pin-by-pin fission rate distribution obtained by a Monte-Carlo calculation are somewhat underestimated in a large geometry, e.g., a PWR full core, under the present calculation conditions. Such underestimation of the statistical uncertainty of a local parameter should be carefully considered when the Monte-Carlo method is used as a reference tool for verification of a deterministic code, especially in large geometries.     

Numerical solution of the chain model of fission product nuclides. Effect on the reactivity of the reactor

Runge–Kutta algorithm of fourth order devised for the numerical solution of ordinary differential equations of fuel isotopic composition and fission products accumulation systems is described. Several hundreds fission product isotopes are included in partially depleted fuel as a result of direct yield from nuclear fission reaction, radioactive decay of fission products and/or neutron capture. The algorithm of the suggested method was coded and compared with both the analytical method and the international ORIGEN2 code system. The accuracy and speed of the calculations as obtained by implementing the method in a PCs with Linux system are found to be acceptable and faster than their resemblance. The effect of the calculated fission product concentrations on the reactivity of the reactor has been included.     

三维六角形组件压水堆堆芯燃料管理计算及程序系统研究

介绍所研制的WWER型压水堆堆芯燃料管理计算程序系统TPFAP-H／CSIM-H，六角形组件均匀化计算程序TPFAP-H是在压水堆正方形组件程序TPFAP的基础上，采用穿透概率法与响应矩阵方法相结合计算六角形组件内中子能谱分布，并考虑六角形栅元特点改造开发而成的CSIM-H是以先进六角形节块扩散程序为基础．参照SIMULATE程序功能而研制的物理-热工水力耦合的三维六角形节块PWR堆芯燃料管理程序两者通过接口程序LINK连接起来，可以考虑燃耗，功率、慢化剂密度变化．控制棒、氙等参数的多种反馈效应对IAEA的WWER-1000型Kalinin核电厂基准问题的校算的结果表明，临界硼浓度、功率和燃耗分布等结果与国际各研究机构的结果吻合良好，偏差均在工程要求之内。     

Prediction of MgO volume fraction in an ADS fresh fuel for the scenario code CLASS

Subcritical reactors, also called Accelerator Driven Systems (ADS), are specifically studied for their capacity in transmuting Minor Actinides (MA). Nuclear fuel cycle scenarios involving MA transmutation in ADS are widely researched. The nuclear fuel cycle simulation tool code CLASS (Core Library for Advanced Scenarios Simulations) is dedicated to the inventory evolution calculation induced by a complex nuclear fleet. For managing reactors, the code CLASS includes physic models. Loading models aim to provide the fuel composition at beginning of cycle according to the stocks isotopic composition and the reactors requirements. A cross section predictor aims to provide mean cross sections needed for solving Bateman equations. Physic models are built from reactors calculation set ahead of the scenario calculation. An ADS standard composition at BOC is a mixture of plutonium and MA oxide. The high number of fissile isotopes present in the subcritical core leads to an issue for building an ADS fuel loading model. A high number of isotopic vector at BOC is needed to get an exhaustive simulation set. Also, ADS initial reactivity is adjusted with an inert matrix which induces an additional degree of freedom. The building of an ADS fuel loading model for CLASS requires two steps. For any heavy nuclide composition at beginning of cycle, the core reactivity must be imposed at a subcritical level. Also, the reactivity coefficient evolution should be maintained during the irradiation. In this work, the MgO volume fraction is adjusted to reach the first requirement. The methodology based on a set of reactor simulations and neural network utilization to predict the MgO volume fraction needed to reach a wanted k_eff for any initial composition is presented. Also, a complete neutronic study is done that highlight the effect on MgO on neutronic parameters. Reactor simulations are done with the transport code MCNP6 (Monte Carlo N particle transport code). The ADS geometry is based on the EFIT (European Facility for Industrial-Scale Transmutation) concept. The simulation set is composed of more than 8000 randomized runs from which a neural network has been built. The resulting MgO prediction method allows reaching a k_eff at 0.96 and the distribution standard deviation is around 200 pcm.