成核密度模型对弧形表面CHF的影响

反应堆发生严重事故时,必须及时对反应堆压力容器（RPV）下封头进行外部冷却以降低下封头损毁可能性,事故期间下封头具有很高的热流分布,在实施外部冷却时可能出现由于过冷沸腾导致的气泡聚集而产生换热恶化从而烧毁。本研究利用ANSYS Fluent软件进行RPV外部冷却的临界热流密度（CHF）数值计算,并通过实验对比发现Basu Warrier和Dhir研究的成核密度模型可以很好地应用于球形表面CHF计算。通过对比球形和椭球形下封头CHF,认为椭球形下封头的CHF特性与球形结构完全不同,并不能用球形结构的实验和计算结果去推测椭球形结构的数值和变化规律。      

IVR-ERVC下压水堆压力容器下封头传热及应力/应变分析

以某1000?MW压水堆为例,利用二维极坐标热模型分析RPV壁面与双层堆芯熔池和外部冷却水堆腔之间的传热,计算下封头壁面瞬态二维温度场分布和烧蚀情况,同时通过有限元分析程序计算下封头壁面的各瞬态温度场和烧蚀引起的热应力/应变情况,分析压水堆RPV下封头在压力容器内熔融物滞留-压力容器外冷却（IVR-ERVC）下的结构完整性。计算结果表明:①芯熔融坍塌后200?s下封头壁面开始熔融,最薄厚度直线下降;3000?s后熔融区沿下封头内壁呈一片柳叶形状分布;②下封头内表面的吸热热流大于外表面的散热热流,在两层熔池界面处内外表面热流密度达到最大值;③RPV下封头热应力在0~400?s时集中于下封头内壁面;在400 s后,下封头内壁面热应力逐渐减小,形变量逐渐增大,下封头完整性可以得到保证;④2000?s以后,RPV下封头烧蚀损伤处内外壁面均产生应力集中,下封头烧蚀处内外壁应力值均大于许用应力,在2000?s后有可能发生断裂,在烧蚀损伤边缘处可能出现破口。      

压力容器外部冷却技术研究进展及建议

2011年福岛核事故再次让人们意识到核泄漏的严重性,压力容器外部冷却技术作为压水堆严重事故下堆内熔融物滞留的关键技术,再次成为国际核能领域的研究热点。压力容器外部冷却技术通过非能动方式向反应堆腔室充水,使压力容器半球形封头淹没并进行沸腾冷却,从而保持压力容器的完整性,避免核泄漏的发生。本文总结了国际压力容器外部冷却技术的研究进展,详细分析了试验研究的三个阶段,简要介绍了ERVC的应用情况,指出了压力容器外部冷却技术性能提高的三种途径。     

真实表面材料及其老化效应对反应堆压力容器ERVC-CHF影响的试验研究

通过反应堆压力容器外部冷却（ERVC）实现熔融物堆内滞留（IVR）技术是核电厂严重事故缓解的重要措施之一。在本文的研究中,建立了二维切片式、全尺寸的试验台架FIRM,开展严重事故条件下反应堆压力容器ERVC-临界热流密度（CHF）试验研究。试验采用去离子水作为试验工质,获得了反应堆压力容器下封头ERVC过程的CHF限值。研究了真实表面材料对CHF的影响及其影响机理,讨论了在去离子水下表面材料SA508 Gr3. Cl.1钢的老化效应。本试验研究对于认识反应堆压力容器IVR-ERVC条件下的CHF行为、提高反应堆压力容器安全性有重要意义。     

RPV下封头熔融池换热特性数值模拟研究

熔融物堆内滞留（In-vessel Retention,IVR）指的是在核电厂严重事故发生后,通过在压力容器和保温层间隙注入冷却水防止压力容器熔穿失效。本文基于COMSOL Multiphysics软件建立了一个流-热-固耦合计算模型,对IVR技术作用下的反应堆压力容器（Reactor Pressure Vessel,RPV）下封头双层熔融池的演变过程进行了仿真研究。当前模型计算结果表明：在稳态分层的状态下,与氧化物层接触的下封头未发生明显的熔化,与金属层接触的下封头会发生明显的熔化,但在被冷却条件下依然可以保持压力容器的完整性。     

压力容器-保温层流道变形条件下临界热流密度试验研究

针对压力容器外部冷却(ERVC)应用中的压力容器-保温层流道(RPV-保温层流道)变形问题,利用提高临界热通量影响因素(FIMR)的试验装置,在相同流量范围开展了变形条件下壁面临界热流密度(CHF)的试验研究,分析了流道变形和流量变化对压力容器(RPV)下封头壁面CHF的影响规律,获得了流道变形情况下ERVC的安全裕度。结果表明:随着RPV下封头角度升高,循环流量增加,下封头壁面CHF增大;与原型流道相比,变形流道下封头壁面CHF的变化幅度小于7%,流道变化的影响并不显著;变形流道中,下封头壁面安全裕量最小的位置与原型流道相同,其安全裕量略有提高。     

严重事故下大功率先进压水堆IVR-ERVC有效性分析

通过压力容器外部冷却(ERVC)以实现堆内熔融物滞留(IVR)作为反应堆严重事故缓解管理的一项重要举措一直以来广泛受到关注和研究。本文使用严重事故分析程序MELCOR,从瞬态角度对大型先进压水堆进行了IVR-ERVC相关研究。过程中重点关注了堆芯熔毁和重新定位,熔池形成、生长及其传热过程,并且对压力容器外部流动传热进行了分析。MELCOR计算所得下封头热流密度分布的瞬态结果与临界热流密度(CHF)比较和分析表明,1700 MWe大功率压水堆发生严重事故后在IVRERVC条件下能够保证压力容器的完整性,即,IVR-ERVC能够有效带出下封头熔融物的衰变热量,缓解严重事故后果。     

椭球形与球形下封头压力容器内熔融物滞留传热特性分析

严重事故下堆芯熔融物再分布于压力容器下封头,在衰变热作用下高温堆芯熔融物对压力容器壁面施加较大的热负荷,可能导致压力容器失效。针对压力容器内熔融物滞留下的传热过程,基于Fortran90语言开发了椭球形下封头压力容器内熔融物堆内滞留（IVR）分析程序IVRASA-ELLIP,计算具有椭球形下封头的压力容器在严重事故下稳态熔池的传热过程及IVR特性。利用IVRASA-ELLIP程序计算了VVER-1000压力容器内熔池的传热,分析具有椭球形下封头的压力容器各处的壁面热流密度、氧化物硬壳厚度和压力容器壁厚,并与运用IVRASA程序计算的AP1000稳态熔池传热结果进行对比分析。研究结果表明,在相同初始参数下椭球形下封头内的壁面热流密度较球形下封头内的小,与热流密度的变化趋势相对应,椭球形下封头内压力容器壁的消融量较球形下封头内的小,椭球形下封头内形成的氧化物硬壳厚度较球形下封头内的厚。     

压力容器下封头大变形模型的开发及在FOREVER实验中的应用分析

反应堆发生严重事故后，将堆芯熔融物滞留在压力容器内的策略（In-vessel Retention,IVR）是作为缓解严重事故的一项重要措施，该策略已成功应用于AP1000、华龙一号和CAP1400等先进压水堆的严重事故管理中。在实施IVR策略时，下封头受到高温熔融物的热负荷会发生变形，下封头的变形改变堆腔的冷却流道，这会直接影响压力容器外部冷却的排热能力和IVR策略的成功实施，有必要对下封头变形展开研究和应用。针对ISAA(Integrated Severe Accident Analysis)程序LHTCM(Lower Head Thermal Creep Module)模型简化薄膜应力模型十分简单和缺乏计算变形模块的问题，本文从机理出发，基于Timoshenko板壳理论、Nortron蠕变定律和大变形塑性理论开发了机理模型—下封头大变形模型，并将该模型集成到一体化严重事故分析程序ISAA中对FOREVER-EC2实验进行应用，预测失效时间与实验的误差仅为1.9%，预测底部伸长量与实验测量值较为符合，破口位置与实验一致。分析结果表明该模型能准确预测在堆芯熔化严重事故中下封头所受应力、...     

反应堆内熔融物冷却的三维数值模拟研究

目前国际上普遍采用堆芯熔融物压力容器内滞留（IVR）策略来缓解严重事故后果。本文基于日本应用能源研究所开发的核电厂事故分析程序SAMPSON,对其压力容器内熔融物冷却分析（DCA）模块进行改进,增加了熔池内金属和氧化物分层模型,开发了熔融物三维直角坐标网格与压力容器三维曲面坐标的交界面几何参数前处理程序,改进了压力容器外冷却的传热关系式。通过AP1000核电机组严重事故下的IVR对改进后的程序进行分析验证,并与实验结果进行对比。结果表明,改进后的SAMPSON程序可对核电厂严重事故下下封头内的熔融物冷却滞留开展有效的模拟分析。     

Two-phase natural circulation flow of air and water in a reactor cavity model under an external vessel cooling during a severe accident

As a part of a study on a two-phase natural circulation flow between the outer reactor vessel and the insulation material in the reactor cavity under an external reactor vessel cooling of the Advanced Power Reactor (APR) 1400, a Hydraulic Evaluation of Reactor cooling Mechanism by External Self-induced flow-HALF scale (HERMES-HALF) experiment has been performed by using the non-heating method of an air injection. This large-scale experiment uses a half-height and half-sector model of the APR1400. This experiment has been analyzed to verify and evaluate the experimental results by using the RELAP5/MOD3 computer code. The RELAP5/MOD3 results have shown that the water circulation mass flow rate is very similar to the experimental results of the HERMES-HALF, in general. Increases in the water inlet area and the water level in the reactor cavity lead to an increase in the water circulation mass flow rate. The effects of an air injection mass flow rate and the water outlet area on the water circulation mass flow rate are dependent on the water inlet area size. As the water outlet moves to a lower position, the water circulation mass flow rate increases slowly.     

Debris interactions in reactor vessel lower plena during a severe accident II. Integral analysis

The integral physico-numerical model for the reactor vessel lower head response has been exercised for the TMI-2 accident and possible severe accident scenarios in PWR and BWR designs. The proposed inherent cooling mechanism of the reactor material creep and subsequent water ingression implemented in this predictive model provides a consistent representation of how the debris was finally cooled in the TMI-2 accident and how the reactor lower head integrity was maintained during the course of the incident. It should be recalled that in order for this strain to occur, the vessel lower head had to achieve temperatures in excess of 1000 °C. This is certainly in agreement with the temperatures determined by metallographic examinations during the TMI-2 Vessel Inspection Program. The integral model was also applied to typical PWR and BWR lower plena with and without structures under pressurized conditions spanning the first relocation of core material to the reactor vessel failure due to creep without recovery actions. The design application results are presented with particular attention being focused on water ingression into the debris bed through the gap formed between the debris and the vessel wall. As an illustration of the accident management application, the lower plenum with structures was recovered after an extensive amount of creep had damaged the vessel wall. The computed lower head temperatures were found to be significantly lower (by more than 300 K in this particular example) with recovery relative to the case without recovery. This clearly demonstrates the potential for in-vessel cooling of the reactor vessel without a need to externally submerge the lower head should such a severe accident occur as core melting and relocation.     

External cooling of a reactor vessel under severe accident conditions

The TMI-2 accident demonstrated that a significant quantity of molten core debris could drain into the lower plenum during a severe accident. For such conditions, the Individual Plant Examinations (IPEs) and severe accident management evaluations, consider the possibility that water could not be injected to the RCS. However, depending on the plant specific configuration and the accident sequence, water may be accumulated within the containment sufficient to submerge the lower head and part of the reactor vessel cylinder. This could provide external cooling of the RPV to prevent failure of the lower head and discharge of core debris into the containment.This paper evaluates the heat removal capabilities for external cooling of an insulated RPV in terms of (a) the water inflow through the insulation, (b) the two-phase heat removal in the gap between the insulation and the vessel and (c) the flow of steam through the insulation. These results show no significant limitation to heat removal from the bottom of the reactor vessel other than thermal conduction through the reactor vessel wall. Hence, external cooling is a possible means of preventing core debris from failing the reactor, which if successful, would eliminate the considerations of ex-vessel steam explosions, debris coolability, etc. and their uncertainties. Therefore, external cooling should be a major consideration in accident management evaluations and decision-making for current plants, as well as a possible design consideration for future plants.     

小型压水堆压力容器内部三维流场计算

In the reactor safety analysis process, it is important to obtain an accurate flow field inside the pressure vessel. Taking the small pressurized water reactor as the research object, the computational fluid dynamics (CFD) method was used to calculate and analyze the internal flow field of the reactor pressure vessel, and the fuel assembly flow distribution and the lower head mixing characteristics were obtained. The results show that the maximum flow distribution coefficient of the fuel assembly is 1.032, the minimum value is 0.934, and the overall flow distribution is characterized by “large in the middle and small in the edge” under the high-speed symmetrical inlet condition of the two pumps. The flow vortex of the lower head is enhanced, and the uneven distribution of the flow distribution of the fuel assembly is increased, under the high-speed asymmetric inlet condition of the pump. The minimum mixing factor of the coolant flow at the core inlet was calculated to be 0.022 due to the insufficient mixing characteristics of the lower head.     

小型压水堆压力容器内部三维流场计算

反应堆安全分析过程中,获得反应堆压力容器内部准确的流场至关重要。以小型压水堆为研究对象,运用计算流体力学(CFD)方法对反应堆压力容器内部流场进行计算分析,获得燃料组件流量分配和下封头混合特性。结果表明:两泵高速对称入口条件下,燃料组件流量分配系数最大值为1.032,最小值为0.934,且流量整体分布呈现"中间大、边缘小"的特点;一泵高速非对称入口条件下,下封头流动漩涡增强,燃料组件流量分配的不均性增大;下封头混合特性计算得到堆芯入口冷却剂流量混合因子最小值为0.022,下封头冷却剂混合能力不足。     

压水堆核电厂严重事故下堆腔注水措施研究

针对百万千瓦级压水堆核电厂,采用一体化严重事故分析工具,对一回路冷段大破口冷却剂丧失（LB-LOCA）始发严重事故下,采取堆腔注水（ERVC）缓解措施的事故进程进行模拟,对该措施缓解堆芯熔化进程、保持压力容器完整性的有效性进行分析验证,并对影响该措施的因素进行研究。分析结果表明,在充足的水源条件下,保证一定的注水速率和水位高度,LB-LOCA始发严重事故下采取堆腔注水的缓解措施可为下封头提供有效的冷却,保持压力容器的完整性。     

Late-phase melt conditions affecting the potential for in-vessel retention in high power reactors

If cooling is inadequate during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 accident. In such a case, concerns about containment failure and associated risks can be eliminated if it is possible to ensure that the lower head remains intact so that relocated core materials are retained within the vessel. Accordingly, in-vessel retention (IVR) of core melt as a key severe accident management strategy has been adopted by some operating nuclear power plants and planned for some advanced light water reactors. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements can provide sufficient heat removal to assure IVR for high power reactors (i.e., reactors with power levels up to 1500 MWe). Consequently, a joint United States/Korean International Nuclear Energy Research Initiative (I-NERI) has been launched to develop recommendations to improve the margin of success for in-vessel retention in high power reactors. This program is initially focussed on the Korean Advanced Power Reactor—1400 MWe (APR1400) design. However, recommendations will be developed that can be applied to a wide range of existing and advanced reactor designs. The recommendations will focus on modifications to enhance ERVC and modifications to enhance in-vessel debris coolability. In this paper, late-phase melt conditions affecting the potential for IVR of core melt in the APR1400 were established as a basis for developing the I-NERI recommendations. The selection of ‘bounding’ reactor accidents, simulation of those accidents using the SCDAP/RELAP5-3D^© code, and resulting late-phase melt conditions are presented. Results from this effort indicate that bounding late-phase melt conditions could include large melt masses (>120,000 kg) relocating at high temperatures (3400 K). Estimated lower head heat fluxes associated with this melt could exceed the maximum critical heat flux, indicating additional measures such as the use of a core catcher and/or modifications to enhance external reactor vessel cooling may be necessary to ensure in-vessel retention of core melt.     

增强IVR有效性的堆内注水策略研究

熔融物堆内滞留（IVR）是一项核电厂重要的严重事故管理措施,通过将熔融物滞留在压力容器内,以保证压力容器完整性,并防止某些可能危及安全壳完整性的堆外现象。对于高功率和熔池中金属量相对不足的反应堆,若下封头形成3层熔池结构,则其顶部薄金属层导致的聚焦效应可能对压力容器完整性带来更大的威胁。本文考虑通过破口倒灌及其他工程措施实现严重事故下熔池顶部水冷却,建立熔池传热模型,分析顶部注水的带热能力,建立事件树,分析顶部注水措施的成功概率及IVR的有效性。结果表明,通过压力容器内外同时水冷熔融物,能显著增强IVR措施的有效性。     

针对REPEC加热实验的RELAP5程序模拟与分析

为研究压力容器外部流道的冷却能力及流动传热过程,在反应堆压力容器外部冷却(REPEC, Reactor Pressure vessel External Cooling)实验台架前期加热实验的基础上,采用RELAP5程序对实验工况进行模拟和对比。模拟结果与实验数据一致性较好。随加热热流、进出口面积的增加,系统内自然循环流量也增加;入口欠热度对自然循环流量的影响不是很明显;近饱和沸腾条件下,系统出现明显的两相不稳定流动。