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。     

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。     

FRAPCON-2动力堆燃料棒稳态性能程序移植

一、程序移植目的1.模拟动力堆燃料元件单棒稳态运行性能供燃料元件单棒的设计和运行参考我国反应堆燃料性能程序研究起步较晚,移植美国NRC经过多年发展和校验的标准程序有利于我国动力堆燃料元件的设计、制造和运行方面的研究,对发展我国轻水堆燃料元件稳态性能程序有一定参考价值。     

动态工况下棒状燃料元件温度场的解析式

基于棒状燃料元件导热微分方程的一般形式,推导了反应堆动态工况下棒状燃料元件温度分布随时间变化的解析表达式.应用该表达式,进行了模拟计算.结果与实际工况的误差在可接受范围内,验证了该解析式的正确性.     

高燃耗下M5合金包壳燃料棒稳态和瞬态性能分析

修改并验证了分析程序FEMAXI-IVM,增加了程序的适用范围。对采用M5合金包壳的FA300-4高性能燃料组件中的燃料棒在稳态和瞬态运行工况下的燃料性能进行了分析。结果表明,此种燃料棒在稳态和瞬态工况下都能保持其完整性,能保证反应堆的安全运行。     

压水堆燃料辐照性能多物理耦合并行分析程序开发

核燃料元件是反应堆的核心部件,其性能影响反应堆的安全性与经济性,利用燃料元件性能分析程序开展燃料堆内稳态辐照性能分析对于燃料设计及安全评价具有重要意义。通过开发燃料温度分布、变形计算、裂变气体释放及内压等模型,结合燃料元件热工-力学多物理耦合计算分析耦合方案,基于先进并行计算方法构建了高性能并行化燃料性能分析程序Athena。利用典型商用压水堆核电站数据及同类程序计算结果进行了程序初步验证,结果表明Athena程序计算结果合理可靠。通过定义堆芯功率及热工水力边界条件,程序能够并行开展压水堆全堆芯燃料辐照性能分析,提高燃料辐照性能分析效率,是数值反应堆原型系统（CVR1.0）的重要组成。     

钠冷快堆燃料元件性能分析程序的开发与验证

为了评估钠冷快堆氧化物燃料元件稳态、瞬态和事故条件下的性能和行为演化,开发了钠冷快堆燃料元件性能分析程序FIBER。程序采用有限体积法实现燃料元件温度的计算,用有限元方法实现力学、裂变气体释放的计算,并通过时间步长控制模块控制程序的稳定运行。为验证程序的准确性,通过调研得到俄罗斯BN600反应堆辐照数据,与FIBER程序的裂变气体释放、柱状晶粒等计算结果进行对比分析。结果表明,FIBER程序对最大燃耗11.8at%、最大辐照损伤78 dpa的快堆燃料元件的辐照变形、柱状晶区、裂变气体释放性能评价是有效的。     

温度反馈阶跃反应性输入下的燃料元件温度场分析

对燃料元件的非稳态温度场进行分析计算.结合反应堆物理、堆芯元件传热和与温度耦合的物性参数,给出了物理数学模型.采用稳定的差分格式进行计算,获得了有温度反馈阶跃反应性输入条件下的棒状燃料元件温度分布和变化规律,计算结果的精度较高,对堆芯热工设计与运行安全分析有参考价值,特别对处于经常变工况的核动力反应堆更有现实意义.     

SONG—多功能栅格物理程序概述

面向下一代反应堆堆芯分析的多功能栅格物理程序(SONG)采用特征线方法(MOC)求解任意几何区域内的中子输运方程,可灵活模拟由棒状元件或板状元件组成的矩形栅格或六角栅格的两维燃料组件模型或两维堆芯模型。程序采用指数矩阵方法求解拓展的裂变产物链和全锕系嬗变链,以满足超长寿期、超深燃耗的钍铀燃料循环或铀钚燃料循环计算的功能需求。相关验证计算结果表明SONG程序满足预期要求。     

秦山核电厂燃料元件内辐照考验的分析和评价

为了验证秦山核电厂燃料元件的堆内性能,在重水试验堆开展了3×3-2小元件堆内综合辐照考验。本文就影响考验结果的若干技术问题和考验条件进行了仔细的分析,充分论证了该试验具有的实际意义。考验件在堆内经历了相当电厂堆稳态工况和一般事故工况的考验。 考验棒最大燃耗达34GWd/tU,棒最大表面热负荷达1.39MW/m~2。在整个考验过程中没有发生考验棒的破损。文章最后就考验结果在验证燃料元件性能及其在电厂堆内安全可靠运行方面进行了评价。     

基于MOOSE框架的五方程两相流分析程序开发

基于多物理场耦合框架MOOSE,采用五方程两相流模型开发了模块化程序ZEBRA,实现了高阶时间、空间离散格式两相流动传热问题的求解。采用Bartolomei开展的垂直圆管过冷沸腾实验对ZEBRA进行验证,在不同热流密度、质量流密度、压力工况下,将程序计算值与实验值进行了数值验证和计算分析。结果表明：ZEBRA中五方程模型预测值与实验值符合良好,沸腾起始点和空泡份额的预测合理,表明ZEBRA初步具备了处理两相流问题的能力。     

Preliminary study of parameter uncertainty influence on Pressurized Water Reactor core design

Sub-channel analysis can improve the accuracy of reactor core thermal design. However, the important initial parameters contain various uncertainties during reactor operation. In this work, the Sub-channel Analysis Code of Supercritical reactor (SACOS) code, which is also applicable for Pressurized Water Reactor (PWR), was used to study the coolant flow characteristic and fuel rod heat transfer characteristic of 1/8 assembly which has the maximum linear power density in 300 MWe PWR core firstly. Then the Wilks' method and Response Surface Method (RSM) were utilized to determine the influence of sub-channel input parameters uncertainties on the highest temperature of reactor core fuel rod and Minimum Departure from Nucleate Boiling Ratio (MDNBR). The results show that in the most conservative conditions, the maximum temperature of the fuel rod and MDNBR were 2167.4 °C and 1.08, respectively. Considering the uncertainties of assembly inlet flow rate, inlet coolant temperature and system pressure, the 95% probability values (with 95% confidence) of fuel rod maximum and MDNBR calculated using response surface methodology were 2144.0 °C and 1.6, while they were 2137 °C and 1.74 calculated by Wilks' approach. Results show that the uncertainty analysis methods can provide larger reactor design criteria margin to improve the economy of reactor. Furthermore, the code was developed to have the capacity to perform the uncertainty study of sub-channel calculation.     

复杂几何燃料组件的参数计算

利用加拿大蒙特利尔大学研制的DRAGON程序对反应堆复杂几何组件进行参数计算，并通过压水堆柱状元件基准问题、MTR型反应堆板状元件基准问题和其他不同几何形状的燃料组件进行校核计算。结果表明：DRAGON程序可用于多种复杂几何燃料组件参数的计算，且具有良好的计算精度。      

RANNS Code Analysis on the Local Mechanical Conditions of Cladding of High Burnup Fuel Rods under PCMI in RIA-Simulated Experiments in NSRR

The RANNS code analyzes behavior of a single fuel rod in reactivity-initiated accident (RIA) conditions. The code has two types of mechanical model; one-dimensional deformation model for each axial segment length of rod, and newly-developed two-dimensional local deformation model for one pellet length. Analyses were performed on the RIA-simulated experiments in the Nuclear Safety Research Reactor (NSRR), OI-10 with high burnup PWR rods, and results of cladding deformation were compared between calculations by the two models and PIE data. The pre-accident, or End-of-Life conditions of the rod were predicted by the fuel performance code FEAMXI-6. In the calculations by the two-dimensional model of RANNS, the plastic strain increases at the cladding ridges during PCMI were compared with those in between the ridges and with the PIE data, and effect of stress variance induced by local non-uniformity of strain on the crack growth was discussed.     

The development of a transient version of the HOTROD computer code

The full length fuel rod steady state performance code HOTROD has been developed to predict fuel performance during a transient. This paper explains the theory used to calculate the transient fuel temperatures using the Crank-Nicolson method. Transient HOTROD predictions of SGHWR and PWR axial clad temperatures during a loss-of-coolant accident are given and are compared with those predicted by other codes. The code is being further developed to model Zircaloy clad ballooning and the creep equations coupled to the heat transfer equation derived in this paper.     

TP2008核数据库验证

TP2008是新研制的用于TPFAP程序的69群核数据库,本文利用IAEA压水堆棒状燃料组件基准问题和零功率临界实验结果对TP2008核数据库进行了验证分析。结果表明,燃料组件无限增殖因数k_∞与机构TUR的符合相对好;棒状燃料组件相对功率分布计算结果与参考程序的符合较好。零功率临界实验的堆芯有效增殖因数k_eff的相对偏差大部分在-0.5%以内,符合较好。     

数值反应堆堆芯通道级三维热工水力程序CorTAF开发及初步验证

堆芯是核动力系统的核心部件,其完整性是反应堆安全运行的重要前提。传统核反应堆堆芯热工水力分析方法无法满足未来先进核动力系统的高精度模拟需求。本文依托开源CFD平台OpenFOAM,针对压水堆堆芯棒束结构特点建立了冷却剂流动换热模型、燃料棒导热模型和耦合换热模型,开发了一套基于有限体积法的压水堆全堆芯通道级热工水力特性分析程序CorTAF。选取GE3×3、Weiss和PNL2×6燃料组件流动换热实验开展模型验证,计算结果与实验数据基本符合,表明该程序适用于棒束燃料组件内冷却剂流动换热特性预测。本工作对压水堆堆芯安全分析工具开发具有参考和借鉴意义。     

FRED fuel behaviour code: Main models and analysis of Halden IFA-503.2 tests

The FRED fuel rod code is being developed for thermal and mechanical simulation of fast breeder reactor (FBR) and light-water reactor (LWR) fuel behaviour under base-irradiation and accident conditions. The current version of the code calculates temperature distribution in fuel rods, stress-strain condition of cladding, fuel deformation, fuel-cladding gap conductance, and fuel rod inner pressure. The code was previously evaluated in the frame of two OECD mixed plutonium-uranium oxide (MOX) fuel performance benchmarks and then integrated into PSI's FAST code system to provide the fuel rod temperatures necessary for the neutron kinetics and thermal-hydraulic modules in transient calculations. This paper briefly overviews basic models and material property database of the FRED code used to assess the fuel behaviour under steady-state conditions. In addition, the code was used to simulate the IFA-503.2 tests, performed at the Halden reactor for two PWR and twelve VVER fuel samples under base-irradiation conditions. This paper presents the results of this simulation for two cases using a code-to-data comparison of fuel centreline temperatures, internal gas pressures, and fuel elongations. This comparison has demonstrated that the code adequately describes the important physical mechanisms of the uranium oxide (UOX) fuel rod thermal performance under steady-state conditions. Future activity should be concentrated on improving the model and extending the validation range, especially to the MOX fuel steady-state and transient behaviour.