Feasibility study of applying the passive safety system concept to fusion–fission hybrid reactor

The fusion–fission hybrid reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc., with the fusion neutron source striking the subcritical blanket. The passive safety system consists of passive residual heat removal system, passive safety injection system and automatic depressurization system was adopted into the fusion–fission hybrid reactor in this paper. Modeling and nodalization of primary loop, partial secondary loop and passive core cooling system for the fusion–fission hybrid reactor using relap5 were conducted and small break LOCA on cold leg was analyzed. The results of key transient parameters indicated that the actuation of passive safety system could mitigate the accidental consequence of the 4-inch cold leg small break LOCA on cold leg in the early time effectively. It is feasible to apply the passive safety system concept to fusion–fission hybrid reactor. The minimum collapsed liquid level had great increase if doubling the volume of CMTs to increase its coolant injection and had no increase if doubling the volume of ACCs.     

Thermal hydraulic phenomena related to small break LOCAs in AP1000

Since the TMI accident in 1979, a lot of attention in the nuclear engineering field has been drawn to the small break LOCA issue, around which plenty of work has been done both experimentally and theoretically. Subsequent reactor designs have also been greatly influenced.As a Generation III + reactor that received Final Design Approval by U.S. NRC, AP1000 employs a series of passive safety systems to improve its safety. However, the thermal hydraulic phenomena related to small break LOCAs in AP1000 have not been fully understood and further studies are still required.This paper investigated the available literature and information on thermal hydraulic phenomena that occur during small break LOCAs in AP1000, which included the critical flow, natural circulation, counter-current flow limiting, entrainment, reactor vessel level swell, direct contact condensation and thermal stratification. In particular, the physical phenomena, theoretical and experimental research conducted in the past few decades, and prediction models as well as their comparison and evaluation for the thermal hydraulic phenomena related to the small break LOCAs in AP1000 were concluded.     

直接注入管线失水事故非能动安全系统运行特性研究

小型模块式反应堆ACP100采用了非能动安全和模块化设计技术,可用于地区集中供暖、海水淡化和核动力商船等多个方面。其中,非能动安全设计主要包括非能动应急堆芯冷却系统、非能动余热排出系统等非能动安全系统和自动卸压等专设措施。针对ACP100非能动安全设计技术特点,在中国核动力研究设计院非能动安全系统综合性能缩比试验装置上开展了大量失水事故系统特性试验研究,根据试验数据分析,获得了非能动安全系统在直接注入管线发生破口后系统的综合响应特性,掌握了系统间的相互影响规律,并初步评估其对堆芯的冷却效果。     

小型堆破口失水事故初步研究

为验证中国广核集团小型堆方案设计,尤其是其中非能动安全注入系统的初步设计,基于RELAP/SCDAPSIM程序,建立了小型堆的一、二回路系统和非能动安全注入系统模型,模拟计算了冷管段0.04 m等效直径破口、冷管段0.2 m等效直径破口、直接注入管道双端断裂、自动卸压系统误启动等LOCA工况。计算结果表明,一回路可实现有效的冷却和降压,堆芯不会过热,验证了其非能动安全注入系统的设计合理性和反应堆系统的安全性。     

LOCA事故下安全注射系统可靠性的GO-FLOW法分析

反应堆安全注射系统是包含复杂操作时序的动态系统,本文研究了应用GO-FLOW方法对其进行可靠性分析,导出了能将GO-FLOW用于含两种失效模式的可修部件状态概率计算的可靠性参数等效模型,并验证了模型的正确性。给出了实际算例,结果表明,GO-FLOW方法是对含时序问题的动态系统进行可靠性分析的有效工具,本文导出的可靠性参数等效模型提高了GO-FLOW对多状态问题的分析能力。     

中国氦冷固态实验包层模块In-box LOCA事故分析研究

中国氦冷固态实验包层模块(CN HCCB TBM)将在ITER 2号窗口进行测试,在测试期间,聚变中子和TBM内部材料发生核反应,产生氚和其他放射性物质。考虑到ITER的运行和工作人员与公众的安全,在进入ITER测试之前需要进行事故安全分析。本文应用MELOCR对HCCB TBM及其氦冷系统(HCS)进行建模,开展了TBM增殖区冷却板流道破口事故(In-box LOCA)安全研究,并对泄压罐体积,破口面积,隔离阀关闭延迟时间等关键参数进行敏感性分析。结果表明:在保守假设流道全破裂的工况下,box压力超过其压力限值4 MPa,而单根流道和5根流道破裂的工况下,box均未超过其压力限值;安装泄压罐和改变隔离阀关闭延迟时间能够有效的控制box压力。     

三门核电厂稳压器安全阀误开启事故研究

稳压器安全阀用于核电站一回路系统和设备的超压保护,如果发生故障卡开,将造成冷却剂丧失事故(LOCA)。本文使用机理性分析程序对三门核电厂1号机组进行建模,并对稳压器安全阀误开启导致的LOCA事故进行模拟分析,研究在稳压器水位较高的情况下,非能动安全设施对LOCA事故的响应情况。之后,为验证三门核电站对类似三哩岛事故的应对能力,假设丧失给水叠加稳压器安全阀卡开事故并进行相应事故分析。通过以上两个事故的分析表明,三门核电厂的非能动安全设计能够应对稳压器安全阀故障造成的LOCA事故,保证对一回路补水,不会造成非常严重的事故后果。     

美国小型轻水堆计划

本文介绍了美国小型轻水堆的特点和研究发展计划。     

AC600非能动安全系统首期实验研究

综述性地介绍了“八五”期间陆续完成的ＡＣ６００非能动安全壳冷却系统风洞实验研究、ＡＣ６００全压堆芯补水实验研究和ＡＣ６００二次则非通动应急堆芯余热排出系统实验研究等三项前期原则性实验研究概况和主要结果。实验研究表明：三大非能动安全系统的设计是合理、可行的,基本能满足其赋与功能要求。实验研究发现了在设计中应值得高度重视的一些热工水力现象,如“水锤现象”。实验研究所获得的实验数据,已用于设计改进和下一     

AC600非能动安全壳冷却系统的概率安全分析

ＡＣ６００是我国改进型压水堆核电站，本文对其在概念设计阶段的非能动专设安全设施中的安全壳冷却系统进行了概率安全分析（ＰＳＡ）。文中采用故障树技术，定旧计算出了系统的不可用度及置信区间，主要部件故障对不同度的贡献和各组成单元的重要度等，并将计算结果与国内外现有压水堆核电站进行了比较，经比较得出ＡＣ６００采用非能动安全冷却系统，将能明显提高核电站的安全性，可靠性和经济性，由于它是一种新的设计，因此围绕     

Best Estimate BWR Transient Analysis with TRACG Assessment using Data from BWR Startup Tests and LOCA Integral Tests

TRACG is a new version of the best estimate BWR transient analysis code, which utilizes a multi-dimensional two-fluid model for the thermal hydraulics and a three-dimensional neutron kinetics model. A three-dimensional neutronics, a fully implicit integration scheme and models for advanced BWR components have been implemented in the code upon TRAC-BF1.

Assessment of TRACG has been performed in this study for the predictive capability of plant transients, which include thermal-hydraulic and neutronic interactions, as affected by responses of the plant control system. Simulations were presented for BWR representative transient tests, which were done as part of a series of BWR5 startup tests. As for the capability to predict thermal hydraulics during the design basis LOCAs, simulations were presented for the LOCA integral tests conducted in the ROSA-III at JAERI and the Hitachi TBL, which had been used for assessment of the TRAC former version.

Consequently, (1)the space-dependent power flow transitions in a BWR were confirmed by TRACG simulations in which the module coupled with neutronics and thermal hydraulics during transients has been newly introduced, and (2) the characteristic thermal-hydraulic phenomena including multi-channel effects during the design basis LOCAs were confirmed, as well as the TRAC former version, by TRACG simulations on which the influence due to a fully implicit integration scheme has not extended. Capability of TRACG to predict BWR transients ranging from simple plant operational transients to design basis LOCAs was successfully demonstrated.     

船用堆小破口失水事故处置的影响因素分析

针对船用堆小破口失水事故处置复杂的特点,利用运行安全分析平台对事故进行了仿真研究,探讨了补水系统、危急冷却系统、二回路设备等对事故处置过程和后果的影响,为运行人员的处理和操作提供了参考,有助于失水事故应急处置规程的制定。     

200MW核供热站方案设计

200MW核供热示范站反应堆设计中采用了一系列先进技术,如自然循环、一体化布置、自稳压、双层壳结构、控制棒水力驱动系统和非能动式安全系统等,使得供热站更安全、可靠、结构简单、易于建造和维修。本文简要介绍了该站的安全原则、主要设计考虑、总体方案和主要设计特点等。     

快堆非能动安全研究发展概况

近年来,在美国快堆研究计划中,提出了一种实现快堆安全目标的新概念,就是以“纵深防御”思想为基础的“非能动安全”(Passive Safety)概念,强调应用非能动的机理保护反应堆和公众的安全,而不是依靠增加能动的专设安全设施。本文扼要介绍有关快中子反应堆“非能动安全”研究的发展概况。     

应用CDAG方法进行秦山二期大破口LOCA严重事故堆芯损伤研究

应用MELCOR 1.8.5程序模拟了秦山二期无缓解措施的大破口LOCA严重事故序列,并利用西屋公司堆芯损伤评价导则(CDAG)对该事故早期堆芯损伤进行评价,得到了下封头失效前特定时刻的堆芯损伤状态和程度。初步分析结果表明,CDAG可以合理地评价秦山二期无缓解措施的大破口严重事故堆芯损伤状况和损伤程度,对进一步研究和验证CDAG的综合评价能力和适用性具有重要参考意义。     

1E级安全壳内用电动机鉴定的核安全审查

对压水堆核电厂1E级安全壳内电动机的鉴定过程和鉴定文件进行审查,要求对鉴定试验结果与标准法规的符合性以及与安全壳内环境要求的一致性做出评价。首先介绍1E级安全壳内用电动机的老化试验、设计基准事故试验等主要鉴定试验,对随后审查过程中遇到的典型问题进行分析,并将IEEE 334与RCC-E两套标准进行对比探讨。     

Development of TSACO and application to Chinese HCCB TBM cooling system

In this study, a thermal-hydraulic and safety analysis code (TSACO) for helium cooling system has been developed using Fortran 90 language, and the simulation has been performed for the cooling system of the Chinese helium cooled ceramic breeder test blanket module (CH HCCB TBM). The semi-implicit finite difference technique was adopted for the solution of the dynamic behavior of helium cooling system. Furthermore, a detailed illustration of the numerical solution for heat structures and critical model was presented. The code was verified by the comparison of RELAP5 code with the same initial condition, boundary condition, heat transfer and flow friction models. The TBM inlet/outlet temperatures and pressure drop were obtained and the results simulated by TSACO were shown in good agreement with those by RELAP5. Thereafter, the design basis accident in-vessel loss of coolant accident (LOCA), was investigated for the CH HCCB TBM cooling system. The critical flow model was also verified by comparing with RELAP5 code. The results indicated that the TBM can be cooled down effectively. The vacuum vessel (VV) pressure and the mass of helium spilled into the VV maintained below the design limits with a large margin.     

重力注硼系统压力响应特性实验研究

为了研究２００ＭＷ低温核供热堆重力注硼系统在不同初始条件下的压力响应性,建造了重力注硼模拟系统。并根据实际注硼系统的热工水力特性,给出了模拟相似准则。实验中,主要研究了冷态及热态条件下系统初始压力,汽液相管道阻力特性,汽液联通方式,堆芯罐与注硼罐上空腔体积比对两罐汽空间压力平衡时间和注硼响应时间的影响。     

基于自适应重要抽样法非能动系统功能故障概率评估

针对非能动系统功能故障概率评估,提出一种新的自适应重要抽样方法。这种方法先对失效域进行预抽样,然后拟合出失效域中样本分布的密度函数,以之作为重要抽样密度函数。以1000 MW非能动先进压水堆(AP1000)非能动余热排出系统为研究对象,考虑模型和输入参数的不确定性,将响应面法和自适应重要抽样法相结合,对其进行功能故障概率评估。结果表明:与传统的概率评估方法相比,自适应重要抽样法具有较高的计算效率,同时又能保证很高的计算精度。     

Study on the long-term passive cooling extension of AP1000 reactor

The AP1000 with high safety is a generation III pressurized water reactor(PWR),its significant feature is passive safety system.However,its passive cooling can only maintain for 72 h and requires additional support from inside or outside the plant.To solve this problem,this study utilized the WGOTHIC software to calculate and analyze the water inventory in the passive containment cooling water tank under different conditions.The results show that when the cooling water inventory is 6553.78 m3,the AP1000 nuclear power plants can achieve long-term,completely passive cooling without any inside or outside the plant.The same outcomes occur when 65-mm-thick containment wall increases the design pressure rating to 0.6 MPa at the cooling water inventory of 5673 m3.Also,the AP1000 shield building was accordingly improved.An ANSYS analysis of the structural stability of the shield building with a 6000 m3 cooling water inventory confirmed that the new design can meet the requirements of the seismic design and the safe residual heat removal requirements of a large-scale PWR.