快堆燃料组件内精细功率分布的计算

堆芯热通道因子是堆芯热工设计及安全分析的一项重要参数,确定热通道因子需用中子学计算给出较准确的燃料组件内元件棒功率分布。在三维六角形几何节块扩散理论基础上,使用多项式重构的方法计算节块内中子通量密度分布和功率密度分布。针对快堆六角形燃料组件的特点,用小六角形积分的方法计算组件内元件棒功率,得到组件内各元件棒功率分布。在NAS程序基础上,编制了元件棒功率分布计算模块NAS PIN。通过与蒙特卡罗程序的校验可发现,二者计算结果符合较好,计算精度可满足工程设计的需要。     

小型长寿命核能系统燃料物理性能的研究

本文在简要说明世界上小型长寿命核能系统研究现状的基础上,提出了使用钍-铀燃料和铅-铋冷却剂构造小型长寿命堆芯的设想,并为此进行了一系列燃料物理性能的研究.对于长寿命核能系统的堆芯物理设计,使反应性随燃耗变动最小非常重要,同时应该尽可能地提高堆芯的燃耗以满足长寿命运行的需求.本文使用MCNP和MCBurn程序详细计算分析了使用不同的初始驱动燃料、不同栅格、燃料成分和类型、富集度条件下,燃料栅元的燃耗反应性变化等性能,并对其进行了能谱、转换比、富集度变化等方面的分析,经过对比初步确定了使用钍-铀燃料构造长寿命堆芯的物理条件,并以此为起点构造出一个堆芯,计算给出了反应性空泡系数等安全参数.     

CERMET-SNRE堆芯物理计算分析

核火箭发动机功率高、寿命长、比冲大,在执行深空探测和星际航行任务时具有不可替代的优势。小型化是核火箭发动机的一个重要趋势,基于此提出了一种使用钨基金属陶瓷燃料的小型核火箭发动机（CERMET-SNRE）堆芯方案,并采用蒙特卡罗程序（MCNP）进行了精确建模,计算了相关物理参数。计算分析结果表明：CERMET-SNRE堆芯能谱硬,燃耗浅,后备反应性足够,功率分布合理,控制鼓与安全棒价值足够,发射掉落事故下有效增殖因数小于0.98,堆芯方案合理,满足设计要求。     

径向步进倒料行波堆的数值研究

行波堆是一种可实现自持增殖-燃耗的新概念快堆,它可直接使用天然铀、贫铀、钍等可转换核材料,实现非常高的燃料利用率。基于行波堆的原理,提出了具有现实应用价值的径向步进倒料行波堆的概念,并将其与典型钠冷快堆的设计相结合,采用数值方法对由外而内的径向步进行波堆二维渐近稳态特性进行了研究。计算结果表明:渐近keff随倒料循环周期近似抛物线分布,而渐近燃耗随倒料循环周期线性增长,满足临界条件的倒料循环周期中最大燃耗可达38%;堆芯功率峰随着倒料循环周期的增长,从燃料卸出区(堆芯中心)向燃料导入区(堆芯外围)移动,功率峰值逐渐降低,在高燃耗情况下,靠近堆芯中心的轴向功率分布呈M形。     

Application of candle burnup strategy for future nuclear energy utilization

The CANDLE burnup strategy is a new reactor burnup concept, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed along the core axis from bottom to top (or from top to bottom) of the core and without any change in their shapes. Therefore, any burnup control mechanisms are not required, and reactor characteristics do not change along burnup. The reactor is simple and safe. If this burnup scheme is applied to some neutron rich fast reactors, either natural or depleted uranium can be utilized as fresh fuel after second core and the burnup of discharged fuel is about 40%. It means about 40% of natural or depleted uranium can be utilized without either enrichment or reprocessing.

In the ideal nuclear energy utilization system, the radioactive toxicity in the environment should remain or decrease after the utilization. This requirement is very severe and difficult to be satisfied. It may take too much time for its realization. The CANDLE burnup may substitute this period. Though it is a once-through fuel cycle, the discharged fuel burnup is about ten times of the present value for light water reactors. The space necessary for final disposal can be drastically reduced. However, in order to realize such a high burnup of discharged fuels some innovative technologies should be developed. Either new material standing still for such a high burnup or intermediate recladding will be required. Especially new fuel development will take a lot of time. For the time being a small reactor with CANDLE burnup may be a good option for nuclear power generation. Even this kind of reactor requires some innovative technologies and a long period for their developments. For the first stage of CANDLE burnup the prismatic fuel high-temperature gas cooled reactor is preferable. Since the design of this reactor fits to the CANDLE burnup very well, only a little time is required for its research and development.     

轻水堆燃料组件计算程序包TPFAP

TPFAP是一个同时适用于PWR和BWR的穿透几率法燃料组件燃耗计算程序包。它首先利用碰撞几率方法在库能群结构下完成三区或四区圆环几何的栅元输运计算。载钆燃料棒或硼棒可燃毒物栅元的有效吸收截面由微燃耗程序CMB产生,两维穿透几率法组件计算是在(x,y)几何下进行。基模计算用来考虑中子泄漏修正。根据反应率等效,计算组件等效扩散参数。在每一燃料棒和可燃毒物棒进行燃耗计算,TPFAP给出每一燃耗步的组件和栅元少群截面、功率分布,提供核设计和安全分析所需参数。     

DRAGON&DONJON程序在MSR中堆芯燃耗计算的适用性

DRAGONDONJON组件-堆芯"两步法"程序通过合理简化,理论可适用于任何堆芯与工况。使用蒙特卡罗方法 RMC(Reactor Monte Carlo code)、MCNP(Monte Carlo Neutron Particle transport code)程序验证DRADON程序是否能够承担快/热谱型熔盐堆(Molten Salt Reactor,MSR)焚烧TRU、Th U燃料燃耗计算。选出熔盐增殖堆(Molten Salt Breeder Reactor,MSBR)与熔盐锕系元素再循环和嬗变堆(Molten Salt Advanced Reactor Transmuter,MOSART)堆型进行比较,同时分别利用RMC程序验证DRAGON程序组件燃耗计算的准确性,利用MCNP程序验证DRAGON程序组件均匀化方法以及DONJON程序截面调用和程序全堆扩散的准确性。结果表明,组件燃耗计算中,TRU和Th U燃料满足燃耗计算要求;堆芯临界计算中,快/热谱堆芯计算误差均小于0.001。证明DRADON程序可以胜任快、热谱型MSR焚烧TRU、Th U燃料的物理计算任务。     

Development and validation of the code COUPLE3.0 for the coupled analysis of neutron transport and burnup in ADS

The analysis of the fuel depletion behavior is critical for maintaining the safety of accelerator-driven subcritical systems(ADSs). The code COUPLE2.0 coupling 3-D neutron transport and point burnup calculation was developed by Tsinghua University. A Monte Carlo method is used for the neutron transport analysis, and the burnup calculation is based on a deterministic method. The code can be used for the analysis of targets coupled with a reactor in ADSs. In response to additional ADS analysis requirements at the Institute of Modern Physics at the Chinese Academy of Sciences, the COUPLE3.0 version was developed to include the new functions of(1) a module for the calculation of proton irradiation for the analysis of cumulative behavior using the residual radionuclide operating history,(2) a fixed-flux radiation module for hazard assessment and analysis of the burnable poison, and(3) a module for multi-kernel parallel calculation, which improves the radionuclide replacement for the burnup analysis to balance the precision level and computational efficiency of the program. This paper introduces thevalidation of the COUPLE3.0 code using a fast reactor benchmark and ADS benchmark calculations. Moreover,the proton irradiation module was verified by a comparison with the analytic method of calculating the210 Po accumulation results. The results demonstrate that COUPLE3.0 is suitable for the analysis of neutron transport and the burnup of nuclides for ADSs.     

Burnup analysis and in-core fuel management study of the 3 MW TRIGA MARK II research reactor

The principal objective of this study is to formulate an effective optimal fuel management strategy for the TRIGA MARK II research reactor at AERE, Savar. The core management study has been performed by utilizing four basic types of information calculated for the reactor: criticality, power peaking, neutron flux and burnup calculation. This paper presents the results of the burnup calculations for TRIGA LEU fuel elements. The fuel element burnup for approximately 20 years of operation was calculated using the TRIGAP compute code. The calculation is performed in one-dimensional radial geometry in TRIGAP. Inter-comparison of TRIGAP results with other two calculations performed by MVP-BURN and MCNP4C-ORIGEN2.1 show very good agreement. Reshuffling at 20,000 MWh step provides the highest core lifetime of the reactor, which is 64,500 MWh. Besides, the study gives valuable insight into the behaviour of the reactor and will ensure better utilization and operation of the reactor in future.     

确定论数值反应堆程序NECP-X的开发及应用

数值反应堆是基于大规模并行计算平台,利用先进的物理模型和数值模拟算法,采用精细化建模,从而精确模拟反应堆在正常运行与事故工况中发生的各类物理现象的模拟技术。西安交通大学NECP团队基于自研的多群和连续能量数据库,提出了全局 局部耦合输运计算方法、大规模并行的2D/1D耦合输运方法等,开发了基于确定论方法的数值反应堆物理程序NECP X,并在此基础上实现了物理 热工 燃料性能分析的多物理耦合模拟计算。基于该程序及其耦合系统,在商用大型压水堆、研究堆和实验堆中进行了验证应用。数值结果表明,NECP X程序及其耦合系统可准确预测反应堆在运行过程中的关键安全参数随时间的演变情况,如有效增殖因数、功率、温度、应力、间隙宽度等,可为商用大型压水堆、研究堆和研究堆的设计及安全分析提供可靠的工具。     

空间离散对弥散颗粒燃料燃耗计算的影响

单栅元燃耗计算是全堆芯燃耗计算的基础，栅元空间离散对燃耗计算的结果有显著影响。弥散颗粒燃料由于双重非均匀性的存在，空间离散的情况更为复杂。本文基于ALPHA组件程序，分析了颗粒在平源区上归类的宏观离散方案与颗粒内部细分燃耗区的微观离散方案对弥散颗粒燃料燃耗计算的影响。算例包括无毒物的UC颗粒单栅元，含Gd₂O₃层的QUADRISO颗粒单栅元和含UC颗粒与Gd₂O₃毒物颗粒的双颗粒单栅元。数值结果表明，无毒物栅元宏观需分3圈以上，含Gd₂O₃栅元宏观需分5圈以上；无毒物算例微观不需要分圈，含Gd₂O₃层的QUADRISO颗粒需在微观燃料区细分2圈，双颗粒问题的Gd₂O₃毒物颗粒微观需分12～15圈。     

Small PWR Using Coated Particle Fuel of Thorium and Plutonium

An innovative concept of PFPWR50 for district heating has been studied, which is a small PWR of 50MWt capability using coated particle fuels with conventional zircaloy cladding. This concept takes advantages of fuel integrity against fission products release of coated particle fuels and a high reliability of PWR technology based on the long history of a successful operation. We have investigated burnup characteristics of fuel rods, assemblies, and reactor cores by the calculation code SRAC95 in order to establish a core concept of long life without on-site refueling. The loading pattern of assemblies with various concentrations of burnable poison is optimized to obtain a flat excess reactivity during the core life in order to eliminate a soluble boron control system. The core life of a cycle is about 8.9 equivalent full power years. And we have also studied the applicability of SiC/SiC composite cladding in place of zircaloy cladding, which is now under development for gas cooled fast reactor fuels. It could be applicable to high burnup fuel rods for a long term operation. From the calculation results, it is found out that the burnup characteristics do not change significantly with SiC cladding and contribute to elongate the core life to 9.2 equivalent full power years.     

基于cosRMC的新一代PWR栅元和组件燃耗计算

新一代压水堆与现有压水堆的重要区别之一是燃料富集度不同,考虑到燃料制造、燃料燃耗等问题,目前压水堆的UO₂燃料富集度通常小于5%,MOX燃料中易裂变Pu含量通常小于6%。新一代压水堆的燃料富集度有可能超过现有标准,平均燃耗有望达到70 GW•d/tU,这对反应堆计算软件提出了新的要求。本文基于反应堆蒙特卡罗程序cosRMC对新一代压水堆栅元和组件基准进行了中子学分析,包括裂变反应率分布、中子通量密度分布及核子密度随燃耗的变化等,并对含Gd棒的组件燃耗计算进行了细致分析。计算结果表明,cosRMC的计算结果与国际上其他程序的计算结果符合较好。通过程序之间结果对比发现,随着燃耗的增加,不同程序计算的Pu含量差别变大。     

超临界水冷堆环形燃料组件核热耦合分析

在超临界水冷堆预概念设计中,组件设计是十分重要的,将影响堆芯性能。超临界水冷堆中水密度变化剧烈的特性要求必须进行核热耦合分析。从中子学及热工性能角度,使用三维核热耦合程序对环形燃料组件进行了优化设计。应用中子学计算程序FENNEL-N对环形燃料组件进行三维扩散计算,可得到组件内单棒功率分布,应用热工计算程序SUBSC对组件进行子通道分析。在计算过程中,分析了燃料棒间距及燃料棒与组件壁盒之间的间隙对组件性能的影响。计算结果显示,增大棒间距和棒壁间隙能提高组件k_inf,但会增大组件内功率峰因子;子通道受热不均匀性对组件热工性能影响较大,通过加入定位格架的方式能展平冷却剂出口温度,降低最大包壳温度。对环形燃料组件的安全分析表明,从中子学角度该组件是安全的。     

Burnup studies of spent fuels of varying types and enrichment

This paper describes the results of fuel burnup measurements, made over a period of several years on discharged fuel from nuclear power plant and research reactor. The measured and calculated burnup of different spent fuel types, viz.: Natural uranium CANDU fuel bundles; 10.5% enriched booster rods; 20% enriched MTR fuel elements have been presented. High-resolution gamma spectrometry, using ¹³⁷Cs and ¹³⁴Cs burnup monitors was employed in different reactors to estimate the amount of ²³⁵U depletion in the respective fuel. The experimental data was compared with those of calculations to optimize fuel-scheduling programme. The burnup data is useful for assessment of fuel performance in the core and resolving design issues related to long-term storage facilities. It has been observed that the gamma spectrometry is very effective in identifying exact position of individual booster bundles inside the discharged booster assemblies, which is useful in safeguard applications. It is concluded that the distribution of measured isotopic activity ratios of ¹³⁴Cs/¹³⁷Cs along the height of the spent fuel gives accurate estimate of the axial neutron flux profiles in the core. The activity ratio technique therefore provides a useful method to determine flux peaking factors at the particular locations of the fuel assemblies in the reactor.     

Analysis of fuel subassembly innerduct configurational effects on the core characteristics and power distribution of a sodium-cooled fast breeder reactor

The power distribution and core characteristics in various configurations of fuel subassemblies with an innerduct structure in the Japan sodium-cooled fast reactor were evaluated using a Monte Carlo code for neutron transport and burnup calculation. The correlation between the fraction of fuel subassemblies facing outward and the degree of power increase at the core center was observed regardless of the compositions. This indicated that the spatial fissile distribution caused by innerduct configurations was the major factor of the difference in the power distribution. A power increase was also found in an off-center region, and it tended to be greater than that at the core center because of the steep gradient of neutron flux intensity. The differences in the worth of control rods caused by the innerduct configurations were confirmed.     

复杂燃料成分的多核素共振干涉计算方法研究

共振干涉现象广泛存在于反应堆系统中，是影响共振计算精度的重要因素之一。当前提出的干涉因子方法，其计算效率难以适用于燃耗过程中的复杂燃料成分。基于改进的伪核素理论与超细群慢化方程求解程序，提出了一种针对实际压水堆燃耗过程的快速共振干涉计算方法。对于燃耗过程中的复杂燃料成分，在均匀问题和压水堆栅元几何下进行了共振自屏分析。结果表明，该方法的计算精度与严格的超细群计算及蒙特卡罗方法相当，效率上优于干涉因子方法，适用于压水堆燃耗过程中的快速共振计算。     

空间锂冷概念快堆堆芯中子学特性研究

本文研究了一种空间锂冷概念快堆的堆芯中子学特性。反应堆燃料采用氮化铀,冷却剂采用⁷Li液态金属,主要结构材料采用W-25%Re。反应堆的控制靠反射层内的控制鼓来实现。建立了程序的计算模型,通过计算和分析,给出了堆芯的主要尺寸和物理参数,计算了堆芯的控制鼓价值、燃耗和功率分布。分析了堆芯中Re的谱移吸收特性和满功率运行7 a不需换料的性能,谱移吸收特性能确保反应堆在发射失败浸在水或湿沙中时处于次临界状态。     

Benchmark Problem Suite for Reactor Physics Study of LWR Next Generation Fuels

This paper proposes a benchmark problem suite for studying the physics of next-generation fuels of light water reactors. The target discharge burnup of the next-generation fuel was set to 70GWd/t considering the increasing trend in discharge burnup of light water reactor fuels. The UO₂ and MOX fuels are included in the benchmark specifications. The benchmark problem consists of three different geometries: fuel pin cell, PWR fuel assembly and BWR fuel assembly. In the pin cell problem, detailed nuclear characteristics such as burnup dependence of nuclide-wise reactivity were included in the required calculation results to facilitate the study of reactor physics. In the assembly benchmark problems, important parameters for in-core fuel management such as local peaking factors and reactivity coefficients were included in the required results. The benchmark problems provide comprehensive test problems for next-generation light water reactor fuels with extended high burnup. Furthermore, since the pin cell, the PWR assembly and the BWR assembly problems are independent, analyses of the entire benchmark suite is not necessary: e.g., the set of pin cell and PWR fuel assembly problems will be suitable for those in charge of PWR in-core fuel management, and the set of pin cell and BWR fuel assembly problems for those in charge of BWR in-core fuel management.     

Optimization layer by layer networks for in-core fuel management optimization computations in PWRs

The optimization layer by layer (OLL) learning algorithm is applied to prediction of assembly-wise power and burnup distribution, the critical soluble boron concentration, and the pin power peaking factor (PPPF) with core burnup in the pressurized water reactor (PWR) of the Korean Nuclear Unit (KNU) 11. It is shown that the OLL trained neural networks can predict core depletion characteristics as accurately as the high-precision modern nodal method codes and that the OLL trained neural networks can compute core depletion characteristics about 40 times faster than the modern nodal method code. The OLL networks are then utilized for determining the optimum fuel assembly (FA) loading pattern (LP) of the equilibrium cycle KNU 11 PWR core by a simulated annealing (SA) scheme. By demonstrating that the FA LP optimization by the SA scheme can be carried out within 10 to 15 min thanks to the speedy neutronics evaluation of the OLL networks, it is proposed that the OLL networks can make a satisfactory substitute for core evaluation codes based on modern nodal methods in in-core fuel management optimization computations where neutronics analysis for a large number of trial loading patterns has to be carried out.