压水堆堆芯熔融物的运动及传热数值分析

针对HPR1000堆型堆芯熔融坍塌问题建立了精确的三维堆芯模型,使用时间推进方法通过求解熔融物的瞬态运动、传热微分方程,确定熔融物在堆芯中的瞬态位置和瞬时温度,以模拟堆芯升温及堆芯熔融进程。研究结果表明：停堆后约2 400 s开始出现熔融现象,熔融物在堆芯活性区域内下落且发生多重相变过程;在4 900 s后,熔融物在堆芯底部形成约1.5 m高的稳定熔池;由于外围组件与低温围栏装置换热,最外围的组件不会发生熔融。本文建立的堆芯熔融物运动与传热分析模型及相关计算结果,可为事故缓解和处理提供技术参考。     

LOCA事故后堆芯瞬态传热及熔融过程数值研究

获取某压水堆核电厂相关参数,建立堆芯及维护结构三维模型,采用大空间自然对流换热和相邻八棒辐射换热模型,求解二维瞬态导热微分方程,计算事故发生后堆芯温度发展及熔融过程。研究表明:随着事故进程的发展,堆芯水位降低,堆芯温度升高,堆芯最高温度点逐渐下移。在事故进程560 s后,控制棒开始熔融;1200 s后,不锈钢棒开始熔融;燃料芯块在2700 s后开始熔融,7000 s后,堆芯熔融份额超50%。大部分堆芯节点熔融时,围桶结构仍未熔融。熔融物直接掉落,向下封头内发生初始迁移。蒸汽对流换热和辐射换热均能影响燃料棒熔融时刻,且蒸汽对流换热占主导地位,蒸汽的影响不能被忽略。辐射换热具有展平堆芯温度的作用。     

严重事故下反应堆压力容器下封头耦合烧蚀传热分析

核电站严重事故发生后,反应堆压力容器（RPV）固壁在熔池作用下会发生烧蚀、减薄。开展RPV下封头耦合烧蚀传热分析对堆坑注水有效性论证和RPV剩余壁厚确认有重要的理论指导意义。本文以CPR1000反应堆压力容器为研究对象,在FLUENT 17.2平台下,基于动态网格方法和UDF二次开发,构建了综合考虑RPV固壁瞬态烧蚀与导热、RPV内壁热流密度再分布及RPV外壁过冷沸腾的全耦合计算模型,获取了9 000 s内的堆坑两相流场分布和RPV固壁烧蚀温度场,分析确定了最小剩余壁厚和发生位置。结果表明：使用动态网格捕捉壁面烧蚀的方法可行,本文全耦合计算模型在分析RPV固壁瞬态烧蚀过程方面有一定优势。     

严重事故下堆芯熔融物与混凝土的相互作用

当反应堆由于始发事件发展到压力容器熔融贯穿时,堆芯熔融物与混凝土相互作用（MCCI）可能引起安全壳晚期失效,包括地基熔穿及不可凝气体引起的安全壳超压失效。本文以600MW轻水堆核电厂为对象,选取全厂断电（SBO）叠加汽动辅助给水泵失效诱发的严重事故序列,应用MELCOR程序研究了该序列下发生MCCI的主要现象,着重关注了混凝土的消融速率及氢气的产生速率,为相应的严重事故管理提供支持。     

反应堆压力容器堆芯筒体快速断裂分析

反应堆压力容器的堆芯筒体受中子辐照最高,是辐照脆化敏感的关键部位。为防止堆芯筒体的快速断裂,在核电工程设计中有必要对该部位进行断裂力学分析,采用法国《压水堆核岛机械设备设计和建造准则》(RCC-M)规定的2种断裂力学分析方法对某核电工程的压力容器进行详细的快速断裂力学分析。分析结果表明,反应堆压力容器堆芯筒体在运行过程中不会发生快速断裂。     

压力容器下腔室冷却剂流动特性数值模拟研究

压力容器流场特性是反应堆热工水力设计的重要依据之一。论文采用三维数值模拟方法,建立了包括进口及环形下降段、下腔室及堆芯进口段、堆芯段的华龙一号反应堆压力容器下腔室分析模型,并采用多孔介质模拟堆芯段压降及流动,在网格数量级敏感性分析的基础上确定了最终网格模型,对运行工况下压力容器下腔室冷却剂的流动特性进行了研究。结果表明,下腔室出现逆时针漩涡流动,冷却剂在冲刷格架板后在下腔室底部汇集并向上流入堆芯;通过分析格架板的上、下表面压差发现大、小格架板所受水力冲击方向相反,载荷大小相近;对下堆芯板流水孔归一化流量分配进行了分析。通过求解附加标量浓度输运方程以标记并跟踪冷却剂的分布和交混,结果表明冷却剂随着流动发生逆时针横向交混,平均有43.7%的冷却剂份额会偏移至逆时针的相邻堆芯进口位置,表明交混特性较好。     

熔融物堆芯冷却滞留特性研究

以模块式小型堆ACP100为分析对象,建立MELCOR程序严重事故分析模型,分析了堆芯衰变热依次经过吊篮、压力容器壁面然后进入堆腔注水系统(CIS)的传热行为。采用燃料棒失效模型评价燃料组件坍塌行为,并通过ANSYS程序蠕变断裂模型评价堆芯下板失效行为。分析结果表明,严重事故后堆芯中心燃料组件坍塌形成堆芯熔融池,堆芯周围燃料组件保持完整结构状态,堆芯下板支撑堆芯熔融池和未坍塌的燃料组件且未发生蠕变断裂失效;CIS冷却压力容器外壁面并导出堆芯衰变热,最终实现熔融物堆芯滞留,避免下封头内形成熔融池。     

熔融堆芯过程中混凝土热分解融化计算

研究发展了一多区瞬态混凝土分解融化模型,模型由热影响区、干区、无气体区及混凝土熔渣区构成。应用热平衡积分方法计算混凝土的分解速率和分解深度。研究结果与国外实验结果进行了比较,两者符合良好。     

熔融物堆芯冷却滞留特性研究

The severe accident analysis model of the small modular reactor ACP100 is built using MELCOR code, and the core heat removed process through the barrel and wall of reactor pressure vessel (RPV) is analyzed by the cavity injection system (CIS). The collapse behavior of the fuel assemblies is estimated by the fuel rod degradation model, and the failure behavior of the lower core plate is estimated by ANSYS program. The results show that the fuel assemblies in the core center melt and collapse to form the core melting pool, while the structure of the fuel assemblies surrounding the core melting pool remains intact, and the core lower plate supports the core melting pool and un-collapsed fuel assemblies all the time, and no creep rupture phenomenon occurs; the core heat can be removed by CIS and the debris in-vessel retention successfully avoids the formation of molten pool in the lower head.     

压水堆核电厂严重事故下堆芯熔融物的冷却研究

堆芯熔融物的冷却和捕集在严重事故后长期的进程对安全壳完整性有很重要的影响,本文综述了核电厂特别是先进核电厂在堆芯熔融物冷却和保持方面的设计,并进行简要分析比较.     

Modelling of heat and mass transfer processes during core melt discharge from a reactor pressure vessel

The objective of this paper is to study the heat and mass trasnfer processes related to core melt discharge from a reactor vessel in a light water reactor severe accident. The phenomenology modelled includes the convection in, and heat transfer from, the melt pool in contact with the vessel lower head wall, the fluid dynamics and heat transfer of the melt flow in the growing discharge hole and multi-dimensional heat conduction in the ablating lower head wall. A research programme is underway at the Royal Institute of Technology (Kungliga Tekniska Högskolan, KTH) to (1) identify the dominant heat and mass transfer processes determining the characteristics of the lower head ablation process: (2) develop and validate efficient analytical/computational models for these processes; (3) apply models to assess the character of the melt discharge process in a reactor-scale situation; (4) determine the sensitivity of the melt discharge to structural differences and variations in the in-vessel melt progression scenarios. The paper also presents a comparison with recent results of vessel hole ablation experiments conducted at KTH with a melt simulant.     

Does vessel rupture contribute to core melt probability?

In current risk studies it has been assumed that the probability of failure of a nuclear pressure vessel due to fast fracture is of the order of 10⁻⁷ per year. Calculations underlying the Marshall Report have essentially confirmed this figure. It appears, however, that the number of cracks in the vessel which has been assumed in these calculations and, therefore, the failure probability of the vessel is too low, possibly by orders of magnitude. As a consequence, gross vessel rupture could no longer be excluded from contributing to the PWR core melt probability and to large consequence accidents.     

Controlling and cooling core melts outside the pressure vessel

For future reactors, the control and cooling of ex-vessel corium melts is under consideration to increase the passive safety features even for very unlikely severe accidents. In this context, different research activities are studying ex-vessel corium behaviour and control, including the implementation of a core cooling device outside the reactor pressure vessel in order to prevent basement erosion and to maintain the integrity of the containment. This paper describes current research on key phenomena which must be understood and quantified to be finally controlled by the cooling device. These are the release of corium melt from the pressure vessel, the temporary retention of the melt in the reactor cavity until melt through of the gate, spreading of the melt on a large surface, and finally the cooling and solidification of the melt by direct water contact. The experiments use high temperature melts which are similar to corium melts. Where necessary, models are developed to transfer the results to reactor scale.     

反中微子监测反应堆运行的数值模拟

利用反应堆出射反中微子计数监测反应堆运行,是国际上新兴的防扩散监测技术,已经过了实验检验。为研究该方法监测反应堆的能力,我们通过在现有MCNP5和MCORGS数值模拟软件中增加了蒙特卡罗方法模拟出射反中微子数目、能量和方向等信息的功能,开发出了用于模拟探测反应堆运行时出射反中微子的数值模拟软件。利用该软件我们研究了反应堆燃耗与出射反中微子计数关系、不同燃耗下铀和钚材料同位素比与出射反中微子计数关系、不同反应堆运行和换料条件下出射中子随反应堆运行时间的变化规律等问题。数值模拟结果表明,反应堆出射反中微子计数可以提供与反应堆运行情况相关的信息。     

An assessment of acoustic emission for nuclear pressure vessel monitoring

Recent research has greatly improved our understanding of the basic mechanisms of deformation and fracture that generate detectable acoustic emission signals in structural steels. A critical review of the application of acoustic emission (AE) to the fabrication, proof testing and in-service monitoring of nuclear pressure vessels is presented in the light of this improved understanding. The detectability of deformation and fracture processes in pressure vessel steels is discussed, and recommendations made for improving source location accuracy and the development of quantitative source assessment techniques.

Published data suggest that AE can make an important contribution to weld fabrication monitoring, and to the detection of defects in lower toughness materials during vessel proof testing. In high toughness materials, however, the signals generated during ductile crack growth may frequently be too weak for reliable detection. The feasibility of AE for continuous monitoring has not yet been adequately demonstrated because of high background noise levels and uncertainty about AE signal strengths from the defect growth processes that occur in service. In-service leak detection by AE shows considerable promise.

It is recommended that further tests are carried out with realistic defects, and under realistic conditions of loading (including thermal shock and fatigue) and of environment.     

Small specimen predictions of fracture toughness for nuclear pressure vessel steels

The use of fracture mechanics in the fracture-safe design and continued safe operation of nuclear reactor pressure vessels has provided an incentive for the development of small specimens for obtaining pertinent fracture toughness data. Small specimens are required for economic reasons when a large number of heats are involved and for space limitation reasons such as in surveillance programs. Several approaches to obtaining fracture toughness from small specimens by either direct measurements or indirect correlations and calculations are reviewed, and their merits and limitations are discussed. Emphasis is placed on techniques which have been developed to determine static and dynamic fracture toughness from surveillance-type specimens. Recently developed techniques for obtaining J-initiation values from a single test specimen and methods for estimating lower and upper shelf fracture toughness from tensile properties are also presented.     

Nondestructive Magnetic Adaptive Testing of nuclear reactor pressure vessel steel degradation

Inspection of neutron-irradiation-generated degradation of nuclear reactor pressure vessel steel (RPVS) is a very important task. In ferromagnetic materials, such as RPVS, the structural degradation is connected with a change of their magnetic properties. In this work, applicability of a novel magnetic nondestructive method (Magnetic Adaptive Testing, MAT), based on systematic measurement and evaluation of minor magnetic hysteresis loops, is shown for inspection of neutron irradiation embrittlement in RPVS. Three series of samples, made of JRQ, 15CH2MFA and 10ChMFT type steels were measured by MAT. The samples were irradiated by E > 1 MeV energy neutrons with total neutron fluence of 1.58 × 10¹⁹–11.9 × 10¹⁹ n/cm². Regular correlation was found between the optimally chosen MAT degradation functions and the neutron fluence in all three types of the materials. Shift of the ductile–brittle transition temperature, ΔDBTT, independently determined as a function of the neutron fluence for the 15CH2MFA material, was also evaluated. A sensitive, linear correlation was found between the ΔDBTT and values of the relevant MAT degradation function. Based on these results, MAT is shown to be a promising (at least) complimentary tool of the destructive tests within the surveillance programs, which are presently used for inspection of neutron-irradiation-generated embrittlement of RPVS.     

Magnetic properties of a highly neutron-irradiated nuclear reactor pressure vessel steel

We report results of minor B–H loop measurements on a highly neutron-irradiated A533B-type reactor pressure vessel steel. A minor-loop coefficient, which is a sensitive indicator of internal stress, changes with neutron fluence, but depends on relative orientation to the rolling direction in the low fluence regime. At a higher fluence of ～10 × 10²³ m^?2, on the other hand, an anomalous increase of the coefficient was detected irrespective of the orientation. The results were interpreted as due to competing irradiation mechanisms of the formation of Cu-rich precipitates, recovery process, and the formation of late-blooming Mn–Ni–Si-rich clusters.