COMMEN程序热膨胀模型的开发和验证

堆芯解体事故已成为钠冷快堆安全分析的主要关注点之一。COMMEN程序是中国原子能科学研究院开发的钠冷快堆堆芯解体事故分析程序。该程序耦合了二维的时空中子动力学模块,主要用于计算堆芯丧失原有几何形状之后的事故进程。为改进COMMEN程序,须在现有的中子学模块中添加热膨胀模型。改进后的COMMEN程序计算严重事故时考虑的反应性反馈更全面。为验证该模型,使用改进后的COMMEN程序对中国实验快堆(CEFR)进行建模计算,并将计算结果和SAS4 A程序进行对比。结果表明:添加了热膨胀模型后,COMMEN程序的计算结果得到了很大改善,其结果与SAS4 A符合的很好。     

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。     

钠冷快堆棒状燃料堆芯子通道分析程序开发及验证

热工水力分析软件的验证是安全审查重点关注的问题。为了实现不同设计软件间的对比验证,本工作开发出具有自主知识产权的钠冷快堆堆芯子通道分析程序SSCFR,进行中国实验快堆（CEFR）全堆芯稳态分析、子通道稳态分析及全堆芯瞬态分析,并将分析结果与CEFR运行和设计值进行对比。结果表明,SSCFR程序的计算结果与CEFR运行值及安全分析报告中的设计计算值符合较好,可用于钠冷快堆后续的软件对比验证及设计计算工作。     

MCCI过程模型开发及验证

概述了严重事故下堆芯熔融物与混凝土相互作用（MCCI）过程的机理性模型,并给出了大亚湾核电厂全厂断电及大破口叠加安注失效等典型初因事故导致的严重事故下的MCCI过程的计算分析结果,并与相同事故序列下的MELCOR计算结果进行对比。计算结果表明,所给出的严重事故下的MCCI过程模型正确合理,计算速度快,能满足在模拟机上应用的要求。     

钠冷快堆的金属燃料元件

使用铀-钚-锆金属合金燃料的钠冷快堆具有良好的固有安全性。采用小堆组合的模块化设计使这类金属燃料快堆电站具有很好的固有安全性、经济性、增殖性并可实现燃料的现场后处理。金属燃料的加工及后处理都采用高温冶金方法,因而制造方便,造成的放射性废物量少。金属型快堆燃料已重新受到世界上的重视。     

ATHLET程序的钠冷快堆应用扩展及其验证

系统分析程序是对钠冷快堆的冷却剂回路系统进行全局模拟、瞬态及事故安全分析的重要工具。本工作对德国核设施与反应堆安全机构（GRS）开发的轻水堆最佳估算系统程序ATHLET进行修改,增加了钠的物性公式和传热关系式,将其适用范围扩展到钠冷快堆。为验证修改过的ATHLET程序,对法国凤凰（Phenix）反应堆系统建模,并对其自然对流实验进行模拟,将计算结果与实验数据进行比较。结果显示,ATHLET程序的钠冷快堆应用扩展具有良好的适用性。     

钠冷快堆系统程序FR-Sdaso开发

为解决600 MW示范快堆（CFR600）事故分析和工况设计中的实际问题,自主开发了钠冷快堆系统程序FR-Sdaso,其建模范围包括堆芯、一回路、二回路、三回路、四回路和事故余热排出系统,主要物理模型包括点堆模型、单通道堆芯热工模型、多区钠池模型、四区蒸汽发生器模型等核岛设备或部件分析模型,汽轮机、凝汽器、给水加热器、除氧器等常规岛设备采用集总参数模型,泵、阀门、管道及控制体等采用通用模型。对程序进行了初步验证,结果表明,FR-Sdaso程序可用于分析全厂瞬态工况及超功率、失流、失热阱等典型事故过程。目前,FR-Sdaso程序已用于CFR600的设计和安全分析。     

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。     

钠冷快堆混合钠火程序开发

钠管道泄漏继而发生钠的燃烧为钠冷快堆特有的事故。在喷雾钠火模型和池式钠火模型基础上,将钠喷雾燃烧和池式燃烧进行了耦合,并针对钠冷快堆钠工艺间的结构特点,最终开发了混合钠火计算程序COMSFIRE。使用该程序计算了FAUNA喷雾钠火试验和CADARACHE池式钠火试验,并与试验结果和部分程序计算结果进行了对比。同时设计了混合燃烧算例,并使用该程序与CONTAIN-LMR程序进行了对比。通过计算结果的对比和分析,初步验证了程序的正确性。     

非能动停堆组件热工水力设计程序开发与验证

液体悬浮式非能动停堆组件主要用于对钠冷快堆发生无保护失流事故的缓解,作为国内钠冷快堆中首次研发使用的设备,开发适用于非能动停堆组件的热工水力设计程序是十分有必要的.本文通过分析非能动停堆组件工作原理及结构特点,建立非能动停堆组件的质量守恒方程、能量守恒方程、动量守恒方程和导热方程,利用交错网格思想进行网格划分,采用半隐...     

Development of System Code FR-Sdaso for Sodium cooled Fast Reactor

To solve actual problems in the accident analysis and working condition design of the 600 MW demenstration fast reactor (CFR600), the sodium-cooled fast reactor (SFR) system code FR-Sdaso was developed, which could be used to model the reactor core, primary system, secondary system, tertiary system, quadruple system and the decay heat removal system of the SFR. The physical models can be divided into three categories: The models for nuclear island equipment including point reactor model, single-channel core thermal model, multi-zone sodium pool model and four-zone steam generator model, etc., the lump parameter models for conventional island equipment, including turbine, condenser, feed water heater, deaerator, etc., and the general models for pump, valve, pipe and control volume. Preliminary V&V work for FR-Sdaso was conducted, and the results show that FR-Sdaso can be used to analyze the transient conditions of the whole plant and typical SFR accidents such as overpower, loss of flow, and loss of heat sink. FR-Sdaso was used in the design and safety analysis of the CFR600.     

Development of safety assessment methodology on fuel element failure propagation in SFR and its application to Monju

Probabilistic and deterministic safety assessments and experimental studies on local fault (LF) propagation in sodium-cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious fuel pin failures have been considered to be the most dominant initiators of LFs in these probabilistic assessments because of its high frequency of occurrence during reactor operation and possibility of subsequent pin-to-pin failure propagation. Four possible mechanisms of fuel element failure propagation from adventitious fuel pin failure (FEFPA) were identified from a state-of-the-art review of open papers. All the mechanisms for FEFPA analysis including thermal, mechanical and chemical propagation are modeled into a safety assessment code which is applicable to arbitrary SFRs by developing some needed but missing methods. Furthermore, an assessment on FEFPA of Japanese prototype fast breeder reactor (Monju) was performed using this methodology. It was clarified that FEFPA was highly unlikely and limited at most within one subassembly in Monju owing to its redundant and diverse detection and shutdown systems for FEFPA even assuming the propagation. These results also suggested future possibility of run-beyond-cladding-breach operation which would enhance the economic efficiency in Monju.     

SARAX程序系统在钠冷快堆瞬态分析中的应用

无保护事故下的瞬态分析是钠冷快堆安全分析的重要内容。基于OECD/NEA发布的MOX-3600和MET-1000基准题,本文利用SARAX程序系统对不同钠冷快堆进行了瞬态计算,分析了堆内各种反应性反馈效应,并计算了无保护失流(ULOF)事故和无保护超功率运行(UTOP)事故下燃料温度和冷却剂温度的变化。计算结果表明：SARAX程序系统在快堆瞬态分析中可给出合理的参数预测结果;ULOF事故对于钠冷快堆是更为严重的事故瞬态,会导致堆内的钠沸腾进而发生严重事故。     

A new empirical model for self-leveling behavior of cylindrical particle beds

ABSTRACT

During the material relocation phase of core disruptive accidents in sodium-cooled fast reactors, the rapid quenching and fragmentation of molten materials discharged from the reactor core into the lower plenum region can lead to the formation of debris beds. Coolant boiling may lead to leveling of the mound-shaped beds, which changes both the beds' coolability with decay heat in the fuel and the neutronic characteristics. In this study, a series of experiments using simulant materials were performed to develop an experimental database of self-leveling processes of particle beds in a cylindrical system. To simulate the coolant boiling in the beds in the experiments, a gas injection method was used to percolate nitrogen gas uniformly through the base of a bed with a conical-shaped mound. Time variations in bed height during the self-leveling process were measured for different particle sizes, densities and sphericities, and gas injection velocities. Using a dimensional analysis approach, a new model was proposed. This model correlates the experimental data on transient bed height with an empirical equation using a characteristic time for self-leveling development and an equilibrium bed height. The proposed model reasonably predicts the self-leveling development of particle beds.     

Development of Sodium-cooled Fast Reactor Sodium Combined Fire Code

Sodium fire caused by sodium pipe leakage is the specific accident for sodium-cooled fast reactor. Based on the sodium spray fire model and sodium pool fire model, sodium spray fire and sodium pool fire were coupled together. A sodium combined fire code COMSFIRE was finally developed based on the structure characteristic of sodium technology room in sodium-cooled fast reactor. FAUNA sodium spray fire experiment and CADARACHE sodium pool fire experiment were calculated with the developed COMSFIRE code, the results of which were compared with the experiments results and some other code results. A combined fire case was designed, and the results were compared with CONTAIN-LMR code. The correctness of the COMSFIRE code was primarily proved through the comparison and analysis.     

钠冷快堆瞬态热工水力及安全分析程序开发

钠冷快堆是第4代核反应堆的主力堆型,瞬态热工水力及安全特性是其设计研发和安全评审的重要工作,需要专用的分析工具。本文基于模块化建模思想,建立了钠冷快堆系统关键部件的热工水力模型和辅助模型,采用具有高稳定性和自动变步长能力的Gear算法,开发了钠冷快堆瞬态热工水力及安全分析软件THACS,并通过了国际基准题EBR-Ⅱ的有保护失流事故实验SHRT-17的初步验证。结果表明,THACS程序能较好模拟此实验的瞬态过程,具备钠冷快堆瞬态热工水力及安全分析的能力,可为我国钠冷快堆研发提供支持。