基于先进程序+保守评价模型的300MW压水堆核电站大破口失水事故分析

大破口失水事故(LBLOCA)是决定核电站运行功率的设计基准事故之一,本文利用最佳估算系统分析程序RELAP5/MOD3,通过修改其相关模型或关系式,结合有关分离效应与整体效应试验数据验证,形成满足10CFR50附录K中保守评价模型要求的LOCA分析工具——先进程序+保守评价模型程序及分析方法。在此工具与方法开发基础上,对300MW压水堆核电站进行了一回路冷管段双端剪切断裂LBLOCA计算分析,计算的包壳峰值温度(PCT)与应急堆芯冷却系统(ECCS)验收准则及相应最终安全分析报告对比表明：应用该工具与分析方法,可望获得进一步的PCT裕量。     

WCOBRA／TRAC最佳估算方法在AP600大破口失水事故分析中的应用

AP600属于简易的先进压水堆设计，采用非能动安全系统，该系统起到了现行反应堆（McIntyre和Beck，1992）中能动的应急堆芯冷却系统（ECCS）的作用。为了验证AP600设计能够减缓假想大破口失水事故（LOCA）的后果，采用美国核管会最近批准的LOCA最佳估算方法（BELOCA），在标准安全分析报告中，对AP600大破口LOCA事故进行了分析。WCOBRA／TRAC程序针对AP600的特点进行建模，验证了匾柱型堆芯试验装置（CCrF）和上腔室试验装置（UPTF）的下降段注入试验的有效性，对AP600大LOCA事故工况下喷放和再淹没冷却传热的不确定性进行了再评估，保守的最小膜态沸腾温度用来定义喷放冷却的边界参数。由于采用了局部模型和总体模型，并采用了统计近似方法，再加上初始条件和边界条件的限制假设，BELOCA简化了对计算程序不确定性的定量计算。最终分析得到的95％包壳峰值温度（PCT95）为1186K，满足10CFR50．46的标准，并留有很大的裕量。本文因此得出结论：AP600的设计能够减缓假想大破口LOcA事故后果。     

田湾核电站长周期换料大破口失水事故分析

田湾核电站拟采用长周期换料策略,堆芯设计的改变需对设计基准事故进行重新分析。本文对反应堆入口主管道大破口失水事故进行了计算分析,在保守的初始输入及计算假设的基础上,通过对轴向功率分布及应急堆芯冷却系统的保守性分析,得出基于燃料包壳温度的最保守的计算工况,并进行了计算。计算结果表明,实施长周期策略后,大破口失水事故仍可满足验收准则的要求,堆芯设计具有足够的安全裕量。     

VVER机组LBLOCA始发严重事故工况下堆芯出口温度BEPU分析

最佳估算加不确定性（BEPU）方法目前广泛应用于核电厂设计基准事故（DBA）的分析。考虑到严重事故现象复杂及不确定性较大,BEPU方法在严重事故领域应用较少。堆芯出口温度（CET）是核电厂安全运行的重要监测参数,本文以VVER1000压水堆核电厂为研究对象,采用BEPU方法对大破口失水事故（LBLOCA）始发严重事故工况下包壳破裂对应的CET进行不确定性分析,并对输入参数进行敏感性分析。计算结果表明：气隙释放对应CET的单侧统计容忍限值（95/95）为430.85 ℃;CET对输入参数中的衰变热系数和包壳厚度较为敏感。     

大破口失水事故分析方法GSM的开发与应用

中广核确定论统计方法(GSM)是介于保守评价模型和最佳估算评价模型之间的失水事故(LOCA)分析方法。在该方法中,程序模型采用确定论现实方法(DRM)惩罚模型进行保守方法处理,对电厂模型采用保守假设,对电厂重要状态参数采用统计方法量化确定不确定性范围和分布,并对统计抽样计算得到的目标参数分别采用参数统计和非参数统计处理以得到包壳峰值温度的双95%值上限值。将该方法应用于CPR1000核电厂大破口LOCA分析,与传统DRM相比可挖掘约9%的LOCA裕量。     

大破口失水事故分析方法GSM的开发与应用

中广核确定论统计方法（GSM）是介于保守评价模型和最佳估算评价模型之间的失水事故（LOCA）分析方法。在该方法中，程序模型采用确定论现实方法（DRM）惩罚模型进行保守方法处理，对电厂模型采用保守假设，对电厂重要状态参数采用统计方法量化确定不确定性范围和分布，并对统计抽样计算得到的目标参数分别采用参数统计和非参数统计处理以得到包壳峰值温度的双95%值上限值。将该方法应用于CPR1000核电厂大破口LOCA分析，与传统DRM相比可挖掘约9%的LOCA裕量。     

基于非参数统计法的DRM分析方法改进研究

确定现实方法(DRM)基于响应面分析方法进行不确定性评价,并在现实模型中引入保守裕度,使得最终结果包络不确定性的可信度大于95%。本文在DRM的基础上,采用独立于不确定性参数的非参数统计法代替响应面分析方法,以峰值包壳温度为目标参数,对典型3环路压水堆核电厂的大破口失水事故(LB-LOCA)进行不确定性分析,结果表明:基于非参数统计法得到的满足95%/95%概率要求的峰值包壳温度被原有DRM的计算值所包络,通过该改进方法可以获得更大的安全裕量。     

压水堆大破口失水事故重要现象识别及数值计算不确定性量化分析研究

大破口失水事故是压水堆核电厂最重要的设计基准事故,对该事故的准确模拟可为提升反应堆功率提供重要支撑。本文采用最佳估算程序RELAP5对压水堆失水事故试验（LOFT）的实验工况FP-LP-2进行了模拟计算,并应用德国反应堆安全研究所（GRS）不确定性分析方法对计算结果进行不确定性量化和敏感性分析;给出了关键输出参数95%置信度的不确定性包络带,并分析了计算结果的不确定性变化趋势及原因。分析结果表明,对包壳峰值温度影响较大的重要现象包括堆芯衰变热、完整环路破口临界流喷放系数和燃料棒的热导率。本文研究确认了GRS方法的有效性,为改进现有核电站安全分析方法具有积极作用。      

铅铋-水直接接触沸水快堆系统子通道耦合热工分析

新概念铅铋-水直接接触沸水快堆（PBWFR）结构紧凑，具有可移动性，在海岛、偏远地区具有很强的应用价值。本文通过将铅铋合金冷却快堆子通道分析程序SUBAS和铅铋合金冷却快堆热工水力系统安全分析程序SACOL耦合，对PBWFR进行分析，重点分析了无保护超功率（UTOP）事故，得到了PBWFR堆芯子通道和系统热工水力特性。结果表明，SACOL程序与耦合程序计算结果的相对误差不超过4%，证明了单向耦合和分步计算的正确性和合理性。采用耦合计算能更加准确地描述事故后组件内各子通道的热工参数变化，弥补了单通道程序分析的不足。在UTOP事故分析中，随着功率上升，包壳温度会迅速升高，热通道内包壳温度最高会达到834 ℃，超过许用限值800 ℃而导致包壳失效。因此包壳温度需在事故开始时具有足够的安全裕量，才能保证事故后反应堆的长期安全运行。     

反应堆冷却剂泵水力特性对大破口失水事故的影响研究

以AP1000为研究对象,应用WCOBRA/TRAC程序对大破口失水事故进行模拟.主要分析4种不同的主泵特性曲线对系统压力、破口流量及包壳峰值温度的影响.研究结果表明,大破口失水事故下,由于主泵特性曲线的差异,导致喷放阶段及再淹没阶段的峰值包壳温度相差近150℃.通过合理优化或改进主泵特性可以为核电厂大破口失水事故带来更大的安全裕量.     

Dynamic simulation of accidental closure of intermediate heat exchanger isolation valve in a pool type LMFBR

In a pool type liquid metal cooled fast breeder reactor (LMFBR), core and other internals such as pumps, heat exchangers are immersed in a pool of sodium. Heat exchange from primary sodium circuit (pool) to secondary sodium circuit (loop) is through four intermediate heat exchangers (IHX) immersed in primary sodium pool. Each IHX is provided with a sleeve valve at its primary sodium inlet window for the purpose of isolating the shell side of IHX from the sodium pool. With such a provision, an inadvertent partial or complete closure of a sleeve valve of one of the IHX during normal operation of the reactor has been considered as a design basis event for the reactor. One dimensional transient thermal hydraulic models of the primary and secondary sodium circuits have been developed to study the thermal hydraulic consequences of such an event. The main areas of concern in the plant and the availability of safety parameters for the detection of the event have been evaluated.     

Investigation of absorption characteristics for thermal-load fluctuation using HTTR

A commercial very high-temperature gas-cooled reactor (VHTR) hydrogen cogeneration system named gas turbine high temperature reactor 300-cogeneration (GTHTR300C) is designed and developed in Japan Atomic Energy Agency (JAEA). Moreover, it has been planned that hydrogen production system and gas turbine system is connect to high-temperature engineering test reactor (HTTR). The reactor system is required to continue a stable and safety operation as well as a stable power supply in the case that thermal-load is fluctuated by the occurrence of abnormal event in the heat utilization system like as the hydrogen production facility. Then, it is necessary to confirm that the thermal-load fluctuation could be absorbed by the reactor system so as to continue the stable and safety operation. The thermal-load fluctuation absorption tests using the HTTR were planned to clarify the absorption characteristics of the HTGR system. However, it is difficult to clarify the phenomenon due to many kinds of fluctuation in nuclear thermal power in the reactor core. Moreover, the actual data regarding how the delay of the temperature response is effective for the reactor system had been gained quantitatively.

The thermal-load fluctuation absorption tests without nuclear heating were planned and conducted in JAEA to clarify the absorption characteristic of thermal-load fluctuation mainly by the reactor and by the intermediate heat exchanger (IHX). The absorption characteristics of thermal-load fluctuation can be revealed with sufficient temperature fluctuation. So the tests were conducted with the primary coolant temperature 120 °C which is the start-up temperature of the HTTR. As a result it was revealed that the reactor has the larger absorption capacity of thermal-load fluctuation than the expected one and that the IHX can be contributed to the absorption of the thermal-load fluctuation generated in the heat utilization system in the reactor system. It was confirmed from their results that the reactor and the IHX has effective absorption capacity of the thermal-load fluctuation generated in the heat utilization system. Moreover, the calculation with the safety evaluation code based on RELAP5/MOD3 was performed. It was confirmed that the calculated temperature for the reactor is almost same to the measured one with the new analysis model. On the other hand, it was confirmed that the calculated temperature for the IHX decreased faster than the measured one due to smaller absorption capacity in the calculation model than that in actual one. It can be considered that the calculation for the IHX produces the conservative result. It was summarized that the safety evaluation code can represent the thermal-load fluctuation absorption behavior conservatively.     

A thermal hydraulic calculation method of an intermediate heat exchanger of a loop type FBR

This paper describes a thermal-hydraulic calculation of an intermediate heat exchanger (IHX) with the computational fluid dynamics (CFD) code CFX. The motivation of this paper is to clarify a heat transfer degradation phenomenon in the IHX through three-dimensional calculation. The whole IHX of the “Monju” reactor is modeled with three parts, i.e., the primary side, the secondary side and the heat transfer region. Through a partial calculation using these models, the flows on the primary side and the secondary side are shown to be axisymmetric. Therefore, a sector model is adopted for the calculation model in the heat transfer region. The calculated temperatures in the IHX are compared with the measured results using the IHX in the “Monju” reactor. Good agreement is obtained for the predicted outlet temperatures and temperatures on the shell surface. As a result of the CFD calculation, it is evaluated that a heat transfer in the lower plenum on the secondary side is dominant under the low flow rate conditions. This fact contributes to lower the heat transfer coefficient in the IHX when the standard heat exchanger theory is applied to calculate the overall heat transfer coefficient between the primary and the secondary sides.     

钠冷快堆二回路系统热工水力瞬态分析程序SELTAC的开发与应用

针对钠冷快堆二回路系统的具体结构和运行特点，对中间热交换器、直流蒸汽发生器、钠缓冲罐以及泵、管道等设备和部件建立模型，采用FORTRAN语言自主编制了二回路系统热工水力瞬态分析程序SELTAC。利用中国实验快堆的停堆试验数据对所编制程序进行了初步验证。结果表明，程序计算值与试验值趋势一致，最大相对偏差不超过4.34%，吻合程度较好。将验证后的程序与一回路系统程序耦合，分析了某600 MW钠冷快堆在主热传输系统保持排热能力时的紧急停堆工况，得到了二回路系统的瞬态特性，为大型商用快堆电站的设计提供了参考。     

Assessment of flow induced vibration in a sodium-sodium heat exchanger

The 500 MWe Prototype Fast Breeder Reactor (PFBR) is under construction at Kalpakkam. It is a liquid metal sodium cooled pool type fast reactor with all primary components located inside a sodium pool. The heat produced due to fission in the core is transported by primary sodium to the secondary sodium in a sodium to sodium Intermediate Heat Exchanger (IHX), which in turn is transferred to water in the steam generator. PFBR IHX is a shell and tube type heat exchanger with primary sodium on shell side and secondary sodium in the tube side. Since IHX is one of the critical components placed inside the radioactive primary sodium, trouble-free operation of the IHX is very much essential for power plant availability. To validate the design and the adequacy of the support system provided for the IHX, flow induced vibration (FIV) experiments were carried out in a water test loop on a 60° sector model. This paper discusses the flow induced vibration measurements carried out in 60° sector model of IHX, the modeling criteria, the results and conclusion.     

400 MW低温供热堆功率调节系统仿真研究

随着日益增长的居民供暖需求，以及对环保的重视，核能供热以其显著减排、供热量大、安全性高的优点，对保护环境、减少污染、缓解燃煤需求等具有积极意义。通过以400 MW低温供热堆一回路中间热交换器为仿真边界，依回路建立各部件的数学模型，基于Matlab/Simulink软件平台建立上述模型的仿真模型。通过设置功率阶跃适应负荷变化，研究低温供热堆控制系统调节能力及一回路负荷跟踪能力。仿真结果表明：低温供热堆一回路功率调节系统跟随负荷变化调节性能良好，控制系统对反应性扰动的响应良好，对于以后设计低温供热堆的运行方式，可考虑负荷运行。     

倾斜条件对海上小型堆LOCA事故的影响研究

基于热工水力系统分析程序RELAP/SCDAPSIM，建立了倾斜条件下海上小型堆一、二回路系统模型和安全注入系统模型，模拟计算了不同横向和纵向倾斜角度下压力容器上接管发生双端剪切破口事故工况。计算结果表明，事故发生后，系统主要热工水力参数受纵向倾斜影响较小，受横向倾斜影响较为显著，且存在陡边效应；发生较大角度的横向倾斜时，一回路冷却剂在重力的作用下重新分布，导致堆芯水位显著降低，燃料包壳峰值温度相较于非倾斜条件下升高约520℃。      

压水堆核电站大破口失水事故分析

压水堆核电站安全分析报告是核安全监管部门对其进行安全审查的重要文件，大破口失水事故是核电站运行的设计基准事故，是安全分析报告中的重要内容。本文使用RELAP5/MOD3.2进行压水堆冷管段大破口失水事故的计算，对比发现一回路冷管段发生双端断裂大破口时燃料元件包壳温度峰值（PCT）最高，且长时间维持在较高温度，此条件下反应堆最危险。计算结果表明，事故发生后，一回路压力迅速下降，堆芯冷却剂的流动性变差，导致堆芯裸露，燃料包壳温度又重新回升。通过安注系统和辅助给水系统等一系列动作，能保证燃料元件包壳温度不超过1204 ℃的限值。     

中间换热器的传热和阻力特性

中间换热器在高温气冷堆氦气透平间接循环发电系统中是耦合高温气冷堆和氦气透平的关键部件,承担着将高温气冷堆中高温氦气的能量传递到氦气透平回路的任务.中间换热器给氦气透平的设计和运行维护带来方便,但它的传热与阻力性能不可避免地影响循环效率,因此,中间换热器的设计和选型需综合考虑传热效率、压力损失、材料性能和紧凑度等因素.本文介绍了印制板式换热器(PCHE)的主要特点,分析了它在间接循环系统中应用的可行性,重点研究了该中间换热器的传热和流动阻力特性,以及影响PCHE换热效率和压力损失的主要因素.在此基础上,提出了优化中间换热器传热和阻力特性的途径和方法.     

Thermal power calibrations of the IPR-R1 TRIGA reactor by the calorimetric and the heat balance methods

Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculated as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor.