堆芯晚期注水对一回路压力的影响分析

分析典型的1000 MW级压水堆核电厂在高压严重事故序列下,堆芯晚期注水对压力容器失效时一回路压力的影响.分析结果表明,在开启1列稳压器卸压阀的情况下,稳压器波动管可能会在压力容器失效之前发生蠕变失效使一回路被动卸压,堆芯晚期注水不会造成一回路压力大幅增大,但波动管失效的时间和尺寸存在较大的不确定性.在开启2列或3列卸...     

压水堆一回路卸压措施影响因素研究

研究压水堆一回路管道小小破口失水事故叠加辅助给水失效导致的高压堆芯熔化严重事故进程，对比验证不同严重事故缓解措施入口温度条件下一回路卸压缓解途径的充分性和有效性，并确认较佳的一回路冷却系统（RCS）降压途径。结果显示，以低于650℃的温度作为降压缓解措施入口条件，可及时恢复可能的堆芯冷却能力。一、二回路卸压效果分析表明，考虑了长期衰变热移出注水流量和堆芯过冷度要求，较佳的卸压配置为初期打开一列稳压器卸压阀，同时迅速恢复辅助给水并开启蒸汽发生器卸压阀。      

AP1000小破口叠加重力注射失效严重事故分析

应用新版MELCOR程序,建立了AP1000一二回路、非能动安全系统及安全壳隔室的热工水力模型,并以热段小破口叠加重力注射系统失效事故为例,对该严重事故进程在压力容器内阶段进行模拟计算,对缓解措施的功能进行了分析和评价。结果表明：自动卸压系统（ADS1～4）的成功实施,可使来自堆芯补水箱和安注箱的冷却水快速有效地注入堆芯,在冷却水完全耗尽前,堆芯始终处于淹没的状态。ADS4爆破阀开启后,使回路压力快速与安全壳压力平衡;非能动安全壳冷却系统对抵御严重事故下由于衰变热和非冷凝气体带来的缓慢升温升压是行之有效的措施;点火器在氢气浓度较低时点火,缓解了安全壳大空间发生全局燃爆而引发安全壳超压失效的风险,但连续点火燃烧会引起局部隔室温升远超出设计温度而危及后备缓解设施的存活。     

基于氢气风险分析的ADS手动卸压策略研究

先进非能动压水堆设计采用自动卸压系统(ADS)对一回路进行卸压,严重事故下主控室可手动开启ADS,缓解高压熔堆风险。然而ADS的设计特点可能导致氢气在局部隔间积聚,带来局部氢气风险。本文基于氢气负面效应考虑,对利用ADS进行一回路卸压的策略进行研究,为严重事故管理提供技术支持。选取全厂断电始发的典型高压熔堆严重事故序列,利用一体化事故分析程序,评估手动开启第1~4级ADS、手动开启第1~3级ADS、手动开启第4级ADS 3种方案的卸压效果,并分析一回路卸压对安全壳局部隔间的氢气负面影响。研究结果表明,3种卸压方案均能有效降低一回路压力。但在氢气点火器不可用时,开启第1~3级ADS以及开启第1~4级ADS卸压会引起内置换料水箱隔间氢气浓度迅速增加,可能导致局部氢气燃爆。因此,基于氢气风险考虑,建议在实施严重事故管理导则一回路卸压策略时优先考虑采用第4级ADS进行一回路卸压。     

重水堆核电厂典型严重事故氢气风险分析

核电厂严重事故下的氢气控制一直是核电厂关注的热点问题之一。本文采用重水堆一体化事故分析程序建立了主热传输系统(PHTS)模型、排管容器及端屏蔽系统、堆腔以及安全壳模型。分别选取代表高压熔堆和低压熔堆的全厂断电及出口集管大破口失水事故始发严重事故序列,从堆芯氧化产氢以及系统热工水力行为出发,对重水堆产氢特性及点火器的消氢效果进行了研究。分析表明:严重事故下随着堆芯冷却恶化,排管容器内发生锆水反应而产生氢气,排管容器和堆腔内的水对氢气产生有较长时间的抑止作用,随着排管容器和堆腔内水的逐渐烧干,排管容器蠕变失效,熔融堆芯落入堆腔发生堆芯熔融物与混凝土的相互作用而产生大量氢气。当氢气点火器失效时,安全壳隔间内氢气体积份额持续增加,存在燃爆风险;点火器开启时,隔间中的氢气混合气体在较低浓度下点燃,氢气燃烧模式处于慢速燃烧区。     

福岛第一核电厂2号机组严重事故进程模拟分析

本文应用MELCOR程序,通过建立全厂详细的模型,对福岛第一核电厂2号机组在地震发生后4天(96h)内的严重事故进程进行了模拟分析并与电厂实测数据进行了比较。基于文中假设的模拟计算得到的趋势与电厂现有实测数据较为一致,分析结果表明:假设TORUS隔间内海水淹没一半时,作为新增的外部热阱与RCIC系统耦合工作,可有效地将堆芯衰变热排出,并延缓了安全壳压力上升。96h内安全壳压力未达到过滤排放系统开启值;RCIC系统在事故发生后近3天失效,此后4.6h操纵员通过开启主蒸汽泄压阀(SRV)对反应堆进行快速卸压,然而堆芯在消防水注入时接近完全裸露,继而发生强烈锆水反应;6h内产氢量达到近800kg。事故后期堆芯通道依然维持可冷却几何形状,最终操纵员通过开启第2组泄压阀对反应堆进行卸压,消防水泵得以有效向反应堆注入冷却水,堆芯重新淹没并冷却。     

稳压器卸压阀卸压效果研究

基于国际上模拟严重事故瞬态过程最详细的机理性程序SCDAP/RELAP5/MOD3.1,主要分析研究了核电站未紧急停堆的预期瞬变(ATWS)初因(失去主给水、失去厂外电和控制棒失控提升)叠加辅助给水失效导致的堆芯熔化严重事故进程,并验证阻止ATWS导致堆芯熔化进程的一次侧卸压缓解措施的充分性和有效性.计算分析结果显示,一列稳压器卸压阀不足以充分降低一回路压力,压力仍然停留在10MPa以上,存在很大高压熔堆的风险.增加一列卸压阀可把一回路压力降低到3MPa左右,安注系统得以投入,及时有效地阻止堆芯熔化进程,降低了高压熔堆风险.分析结果还显示高压安注系统的投入对一回路卸压具有重要影响.     

基于气动悬浮的高温氧化物熔体热物性测量

堆芯熔融物的热物性是研究反应堆严重事故进程及堆内堆外现象机理的重要基础参数。当堆芯熔化时，堆内温度达3 000 K,形成U-Zr-O-Fe多元混合物，而气动悬浮技术是优选的高温下测量堆芯熔融物基础热物性的技术。本文描述了一套基于气动悬浮和激光加热技术的密度、表面张力和黏度的测量装置，目前已实现高温氧化物密度的实验测量。装置采用收缩-扩张型锥形喷嘴悬浮球状样品，采用CO₂连续激光器加热并熔化样品，采用双色红外测温仪监测样品的温度并进行激光器功率反馈控制；采用高速相机记录样品轮廓的变化，并结合图像分析法计算样品的体积，最后得到被测材料在高温下的密度。实验测量得到2 750～3 200 K范围内氧化锆熔体的密度，其在熔点(2 988 K)处的密度为4.717 g/cm~3,温度系数为-7.202×10^-4 g/(cm~3·K)。     

压水堆核电厂全厂断电事故及其缓解措施

以1座典型的3环路压水堆核电厂为参考对象,分别研究了发生全厂断电事故时堆芯在低压和高压状态下的损坏进程。结果表明：在考虑稳压器波动管的蠕变失效时,虽避免了高压熔堆,但低压状态下堆芯损坏更为严重,且产生更多的氢气。分析了导致这一结果的原因,提出了在堆芯出口温度达923K时的严重事故缓解措施。计算结果表明：该缓解措施能有效地延缓堆芯损坏进程,为操纵员恢复交流电源以及采取其它缓解手段赢得更多时间。     

大功率非能动压水堆严重事故工况堆芯熔毁进程研究

应用MELCOR 2.1程序,建立了大功率非能动反应堆主要回路、非能动安全系统及安全壳的热工水力模型,并以热段小破口叠加ADS 1阀门失效和内置换料水箱失效触发严重事故为研究对象,对事故进程进行模拟,对堆芯熔毁进程进行了分析。分析结果表明:1)锆合金和不锈钢氧化释热功率在蒸汽充足的情况下高于燃料的衰变功率,将加速堆芯的恶化;2)约13.1%的不锈钢和27.1%的锆合金被氧化,共产生550.99kg氢气;3)堆芯构件的熔化主要依赖于材料自身的熔点和有无构件支撑,堆芯支撑板能够延缓熔融物跌落进入下封头的进程;4)熔池形成后若外部冷却的不足将很快导致下封头应力失效。     

Effect of water injection on hydrogen generation during severe accident in PWR

Effect of water injection on hydrogen generation during severe accident in a 1000 MWe pressurized water reactor was studied.The analyses were carried out with different water injection rates at different core damage stages.The core can be quenched and accident progression can be terminated by water injection at the time before cohesive core debris is formed at lower core region.Hydrogen generation rate decreases with water injection into the core at the peak core temperature of 1700 K,because the core is quenched and reflooded quickly.The water injection at the peak core temperature of 1900 K,the hydrogen generation rate increases at low injection rates of the water,as the core is quenched slowly and the core remains in uncovered condition at high temperatures for a longer time than the situation of high injection rate.At peak core temperature of 2100-2300 K,the Hydrogen generation rate increases by water injection because of the steam serving to the high temperature steam-starved core.Hydrogen generation rate increases significantly after water injection into the core at peak core temperature of 2500 K because of the steam serving to the relocating Zr-U-O mixture.Almost no hydrogen generation can be seen in base case after formation of the molten pool at the lower core region.However,hydrogen is generated if water is injected into the molten pool,because steam serves to the crust supporting the molten pool.Reactor coolant system (RCS) depressurization by opening power operated relief valves has important effect on hydrogen generation.Special attention should be paid to hydrogen generation enhancement caused by RCS depressurization.     

Response of gas-cooled fast breeder reactors to depressurization accidents

The gas-cooled fast breeder reactor (GCFR) under design by Gulf General Atomic is cooled with helium pressurized to 85 atm and has the reactor core, the steam generators and their associated steam turbine-driven helium circulators, and auxiliary core cooling loops all contained within a massive prestressed concrete reactor vessel (PCRV).The response of the GCFR to coolant depressurization accidents has been investigated and it has been shown that this class of accidents can be safely handled with considerable safety margin. Rapid depressurization is assumed to be caused by a seal failure in a large concrete plug closing one of the large PCRV cavities and the depressurization rate is controlled by a flow restrictor incorporated within the closure plug. Continued core cooling is provided by the main core cooling loops. The plant transient reponse following a depressurization accident has been calculated with a computer code developed at GGA. The results obtained indicate rather mild increases in peak clad temperature for a depressurization accident with the leak area defined by the flow restrictor.Additional cases investigating larger leak areas to explore safety margins indicate that the peak cladding temperature does not increase rapidly with increasing leak area. Secondary containment conditions in a depressurization accident have also been evaluated.     

Evaluation of intentional depressurization strategy in Chinese 600 MWe PWR NPP

Intentional depressurization is one of the effective strategies in preventing high-pressure melt ejection (HPME) and direct containment heating (DCH), which is most feasible for the operating nuclear power plants (NPPs) in China. In order to evaluate this strategy of a Chinese 600 MWe PWR NPP, the plant model is built using SCDAP/RELAP5 code. ATWS, SBO, SGTR and SLOCA are selected as the base cases for analysis of intentional depressurization. The results show that opening safety valves of pressurizer manually when the core exit temperature exceeds 922 K can reduce the RCS pressure effectively and prevent the occurrence of HPME and DCH. Several uncertainties such as the operability of safety valves, ex-vessel failure and the transitory rise of RCS pressure are also analyzed subsequently. The results show that the opening of the safety valves can be initiated normally and that opening three safety valves is a more favorable strategy in the event of possible failure of one or more of the safety valves; the probability of ex-vessel failure is reduced after intentional depressurization is implemented; the transitory rising of reactor coolant system (RCS) pressure when the molten core materials relocate to the lower head of reactor pressure vessel (RPV) will not influence the effect of depressurization.     

非能动堆芯冷却系统LOCA下冷却能力分析

本文基于机理性分析程序建立了包括反应堆一回路冷却剂系统、专设安全设施及相关二次侧管道系统的先进压水堆分析模型,对典型的小破口失水事故和大破口失水事故开展了全面分析。针对不同破口尺寸、破口位置的失水事故,分析了非能动堆芯冷却系统（PXS）中非能动余热排出系统（PRHRS）、堆芯补水箱（CMT）、安注箱（ACC）、自动卸压系统（ADS）和安全壳内置换料水箱（IRWST）等关键系统的堆芯注水能力和冷却效果。研究表明,虽然破口尺寸、破口位置会影响事故进程发展,但所有事故过程中燃料包壳表面峰值温度不超过1 477 K,且反应堆堆芯处于有效淹没状态。PXS能有效排出堆芯衰变热,将反应堆引导到安全停堆状态,防止事故向严重事故发展。     

Consequences of delayed air ingress following a depressurization accident in a high temperature reactor

Air ingress is a specific event in a high temperature reactor (HTR). The potential threat posed by air ingress lies in the chemical reaction of oxygen with hot graphite at a temperature above 500 °C leading to reaction heat and graphite corrosion. In order to assess the consequence of air ingress into the reactor, it is postulated that breaks are present above and below the reactor core and that unobstructed ingress of air through them is possible. It is obvious that the air ingress incident has to be preceded by a depressurization accident. For this hypothetical scenario the maximum possible air flow rate through the core resulting solely from the pressure losses in the core is estimated as a function of the break cross-sections exposed above and below the core.In this paper, the thermal behavior of an HTR with prismatic fuel (operating inlet/outlet temperatures 450/850 °C) during air ingress accident conditions has been investigated. In particular, maximum temperatures and burn-off of the fuel and bottom graphite reflector for various air flow rates for the postulated hypothetical scenario have been analyzed. It also indicates the limiting time at which the graphite layer of fuel will be completely burnt-off and the fuel compacts exposed. In addition, the consequences of delayed air ingress after a core heat up following depressurization have been investigated.This paper, thus, throws light on the safety aspects of the new generation HTRs with prismatic fuels (e.g. NGNP/ANTARES) conceived for power generation and process heat application.     

基于混合自适应遗传退火算法的AHTR堆芯优化研究

板状先进高温堆（AHTR）的预设计采用均一富集度的燃料组件,导致功率峰因子（PPF）过大,总PPF高达2.09,一定程度制约了反应堆的安全性与经济性。文章采用富集度分区法对其进行改进优化,为了加快堆芯燃料最优化布置的搜索速度,设计了一种自适应的混合智能算法,该算法整个优化过程均基于一个用MATLAB语言编辑的程序自动完成,优化后的径向功率峰因子降低至1.122,相比原设计降低25.02%。温度场模拟结果表明,优化方案温度分布更均匀,峰值温度从1030 K降低至1010 K,有效地提高了堆芯的安全裕量。      

Objectives in developing the European pressurized water reactor (EPR)

In 1991, major German utilities and Electricité de France agreed to develop, together with Siemens and Framatome, the nuclear island for the next generation of nuclear power plants. This nuclear island design is based on German and French experience in the construction and operation of pressurized water reactors. The major step in the evolutionary European pressurized water reactor (EPR) design is the systematic inclusion of events beyond classic design events. The mitigation of core melt accidents by special means, such as reinforcement of the containment function, primary loop depressurization, hydrogen reduction and ultimate heat removal, is part of the design, in addition to a number of features which increase the reliability of plant operation and accident control. Nevertheless, utilities and designers are aware of the economic challenge facing the EPR by the need to compete with other nuclear power plant designs and fossil plants for electricity generation.     

Meeting fuel temperature limits in an HTR-Module Reactor during depressurized core heat-up

The thermal behaviour of an HTR-Module Reactor is discussed for the design basis event of core heat-up after fast depressurization taking into account the most unfavourable initial state and uncertainties of input data. The reactor is designed to retain its fission products inside the fuel coatings even if all active components for decay heat removal and reactivity control should fail. To meet this goal maximum fuel temperatures during core heat-up should not exceed the technological limit of 1620°C, for which the integrity of the fuel coatings has been proven experimentally.Two-dimensional thermal-hydraulic calculations show that the maximum fuel temperature during core heat-up is expected to be 1472°C taking into account nominal full power operation as an initial state, a sudden depressurization in the beginning of the event, and nominal input data. The most unfavourable initial state is the steady state operation close to the scram set points, i.e. 105% power and increased cold and hot gas temperatures. Accounting for this leads to a maximum fuel temperature of 1522°C. Relevant uncertainties of input data are those of decay heat production, power distribution and core thermal conductivity and specific heat capacity. Their individual standard deviations can be combined to an integral uncertainty margin of ±86 K which covers two standard deviations. Hence the maximum fuel temperature taking into account unfavourable initial state and uncertainties is 1608°C.     

Analysis on hydrogen risk mitigation in severe accidents for Pressurized Heavy Water Reactor

Hydrogen source term and hydrogen mitigation under severe accidents is evaluated for most nuclear power plants (NPPs) after Fukushima Daiichi accident. Two units of Pressurized Heavy Water Reactor (PHWR) are under operating in China, and hydrogen risk control should be evaluated in detail for the existing design. The distinguish feature of PHWR, compared with PWR, is the horizontal reactor core surrounded by moderator in calandria vessel (CV), which may influence the hydrogen source term. Based on integral system analysis code of PHWR, the plant model including primary heat transfer system (PHTS), calandria, end shield system, reactor cavity and containment has been developed. Two severe accident sequences have been selected to study hydrogen generation characteristic and the effectiveness of hydrogen mitigation with igniters. The one is Station Blackout (SBO) which represents high-pressure core melt accident, and the other is Large Break Loss of Coolant Accident (LLOCA) at reactor outlet header (ROH) which represents low-pressure core melt accident. Results show that under severe accident sequences, core oxidation of zirconium–steam reaction will produce hydrogen with deterioration of core cooling and the water in CV and reactor cavity can inhibits hydrogen generation for a relatively long time. However, as the water dries out, creep failure happens on CV. As a result, molten core falls into cavity and molten core concrete interaction (MCCI) occurs, releasing a large mass of hydrogen. When hydrogen igniters fail, volume fraction of hydrogen in the containment is more than 15% while equivalent amount of hydrogen generate from a 100% fuel clad-coolant reaction. As a result, hydrogen risk lies in the deflagration–detonation transition area. When igniters start at the beginning of large hydrogen generation, hydrogen mixtures ignite at low concentration in the compartments and the combustion mode locates at the edge of flammable area. However, the power supply to igniters should be ensured.