先进压水堆熔融物堆内滞留参数不确定分析研究

压水堆核电厂在严重事故下将发生堆芯熔化事故而形成熔融池。形成熔融池的过程具有很大的不确定性,这影响到反应堆压力容器熔融物堆内滞留（IVR）策略的有效性。本工作以AP1000核电厂两层IVR模型为研究对象,对成功实施反应堆压力容器外部冷却（ERVC）的假想严重事故进行了熔融池参数不确定性分析,包括参数的敏感性分析和使用拉丁超立方抽样的概率分析。结果表明：衰变功率对IVR评价参数影响最大,应采取措施（如上堆腔注水）尽量延缓堆芯熔化的时间;熔融物中不锈钢的质量将对金属层参数造成较大影响,可考虑在压力容器内布置牺牲性材料来减小金属层的集热效应;氧化物层外压力容器失效的概率仅为1.2%,但金属层外压力容器失效的概率高达20%。本结果对今后IVR策略研究和设计具有一定的指导意义,同时也为压水堆核电厂安全评审提供理论支持。     

堆芯熔化严重事故下反应堆压力容器下封头高温蠕变分析

核电厂在发生堆芯熔化严重事故时,采用堆内熔融物滞留（IVR）策略将熔融物包容在反应堆压力容器（RPV）内是一项重要缓解措施。在IVR策略期间,RPV下封头在熔融物的极高温度载荷和力学载荷的共同作用下很有可能因过度蠕变变形而失效。因此,有必要对熔融物滞留条件下RPV下封头进行蠕变变形分析,以保证RPV结构完整性。该文在假定IVR条件下,采用有限元方法对RPV下封头进行热-结构耦合分析,通过计算得到容器壁的温度场和应力场,以及下封头的塑性和蠕变变形,并结合塑性和蠕变断裂判据对下封头进行失效分析。结果表明,考虑蠕变影响后,结构的变形将大大增加;严重事故下采取熔融物滞留策略期间,RPV下封头的主要失效模式为蠕变失效而非塑性失效;内压对蠕变变形量和蠕变失效时间有较大影响。该文为严重事故下RPV下封头的蠕变和失效研究提供了分析方法。      

DVI管线破裂始发严重事故的IVR分析

本文选取了直接注入管线破裂始发的严重事故,分析堆芯熔融物压力容器内保持（IVR）策略实施以后压力容器下腔室内堆芯碎片和压力容器下封头的响应、堆芯碎片与压力容器壁面的传热、压力容器外壁面与堆腔水之间的传热以及压力容器不同区域的热流密度。研究表明,该事故序列下未发生下封头蠕变失效,区域4有最早发生蠕变失效的可能性。     

基于SCDAP/RELAP5耦合堆腔注水的非能动压水堆熔融池冷却分析

采取堆腔注水策略冷却熔融池对缓解严重事故后果、降低安全壳的失效概率具有十分重要的作用。本文采用SCDAP/RELAP5程序,首先以韩国APR1400相关实验结果对堆腔外部注水自然对流冷却能力进行比对分析,然后建立了耦合堆腔注水措施的融熔池冷却的核电厂模型,以非能动压水堆为研究对象,针对冷段大破口失水事故(LBLOCA)始发严重事故序列,分析堆芯熔融进展过程中实施堆腔注水策略后融熔池的冷却特性及堆腔外部注水的自然循环能力。分析结果表明,LBLOCA下,当堆芯出口温度达到923K时,实施堆腔注水后能有效冷却下封头内的熔融池,从而保持压力容器的完整性。     

反应堆压力容器外水冷条件下贯穿件完整性分析

严重事敝下堆芯熔融物坍塌到反应堆压力容器(RPV)下封头时,可能造成贯穿件因高温熔融物热侵袭而失效,使压力容器丧失完整性,熔融物进入到反应堆堆腔中,导致熔融物堆内滞留(IVR)失效.在分析贯穿件脱落和熔融物流入贯穿件两种失效模式基础上,分别运用VTA程序和修正的整体凝固模型(MBF)计算贯穿件焊缝的熔化程度、热膨胀产生的摩擦力,估算贯穿件内熔融物流动的距离.结果表明,在成功实施反应堆压力容器外水冷(EVVC)措施条件下,300 MW压水堆核电厂压力容器的下封头不会因贯穿件失效而丧失完整性,堆芯熔融物小能通过贯穿件失效向堆腔迁移.     

模块式小型堆非能动堆腔注水冷却策略研究

在模块式小型堆MELCOR分析模型的基础上,对典型严重事故序列进行计算分析,得到压力容器下腔室内堆芯熔融池特征参数,并使用自主研发的CISER程序对模块式小型堆堆腔注水冷却效果进行研究。通过影响参数的敏感性及保守性分析证明模块式小型堆堆腔注水冷却措施可行且有适当的安全裕量。     

IRIS反应堆严重事故下堆内自然循环及下封头失效分析

使用REALP5/SCDAP分析了IRIS堆汽轮机停机和部分失流事故导致的严重事故进程及缓解措施。分析结果表明IRIS堆内水装量大,使得堆芯较长时间处于淹没状态,事故发生后近7个小时堆芯开始裸露,10小时后堆芯开始损坏。对于不卸压不安注的情况,压力容器会完全干涸,堆芯和蒸汽发生器之间形成蒸汽自然循环流动,堆芯温度缓慢升高,低熔点的控制棒金属首先熔化落入下腔室并加热下封头,使得下封头底部区域发生蠕变断裂失效。在不卸压的情况下一个上充泵的安注流量就能够缓解事故。     

严重事故IVR下反应堆压力容器耦合传热数值模拟分析

堆芯熔化严重事故下保证反应堆压力容器（RPV）完整性非常重要,高温蠕变失效是堆芯熔化严重事故下反应堆压力容器的主要失效模式。在进行严重事故堆芯熔化物堆内包容（IVR）下RPV结构完整性分析中,RPV内外壁和沿高度方向的温度分布以及剩余壁厚是结构分析的重要输入。本文采用CFD分析方法对RPV堆内熔融物、RPV壁以及外部气液两相流动换热进行热-固-流耦合分析,获得耦合情况下的温度场、流场、各相份额分布以及RPV的剩余壁厚,为RPV在严重事故IVR下的结构完整性分析提供依据。     

假想堆芯熔化严重事故下反应堆压力容器完整性的研究进展与建议

堆芯熔化严重事故下保证反应堆压力容器完整性非常重要,高温蠕变失效是堆芯熔化严重事故下反应堆压力容器的主要失效模式。本文介绍了近年来在假想堆芯熔化严重事故下国内外反应堆压力容器高温蠕变行为的研究进展及现状,着重阐述了在材料高温蠕变试验、缩比模型试验和数值模拟等方面取得的成果,以及国内在RPV结构完整性高温蠕变行为研究方面的最新成果,指出了目前研究中存在的问题并提出开展多轴拉伸试验、三维耦合效应的温度场分析和缩比模型试验等研究方向。     

船用堆严重事故下熔融物堆内滞留能力分析

堆芯熔化的严重事故,可能导致船用堆下封头失效、熔融物进入堆坑,危害人员及船体安全。本文采用严重事故一体化程序MAAP4,以船用堆全船断电事故为研究对象,针对低压安全注射系统投入时机、低压安全注射水流量,研究下腔室熔池形成后,投入低压安全注射系统对熔融物堆内滞留的作用。结果表明:在下腔室熔池形成后1576?s时,投入两台安全注射泵仍能有效阻止压力容器失效,实现熔融物堆内滞留;在下腔室熔池形成2646?s后,投入低压安全注射系统不能阻止压力容器失效。      

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。     

Core performance of new concept passive-safety reactor “kamado” - safety, burn-up and uranium resource problem -

New concept of a passive-safety reactor “KAMADO” has a negligible possibility of core melting and flexibility of total reactor power. The reactor core of KAMADO consists of fuel elements of graphite blocks, which have UO₂ fuel rods and cooling water holes. These fuel elements are located in a reactor water pool of atmospheric pressure (1 atm) and low temperature (< 60°C). In case of LOCA, decay heat from fuel rods is removed by conduction heat transfer to the reactor water pool. Since the cooling water does not contact a fuel rod directly, core design has much flexibility without considering dry-out limitation and Minimum Critical Power Ratio (MCPR). Additionally an effective use of spent fuel is expected.     

压力管式超临界水堆极限事故下“无堆芯熔化”概念评估

在自主研发的事故分析程序SCTRAN的基础上,开发并验证了二维导热模型和辐射换热模型,并将改进后的SCTRAN应用于加拿大压力管式超临界水堆在失水事故（LOCA）叠加丧失紧急堆芯冷却系统（LOECC）事故中的堆芯安全评估,并对燃料棒到慢化剂之间的传热效率以及关键的影响因素进行了评估。计算结果表明,在LOCA叠加LOECC工况下,燃料棒到燃料通道的辐射换热和燃料棒到蒸汽的自然对流换热能够有效导出反应堆的衰变余热,最高功率的燃料组件内、外圈燃料棒的最高包壳温度分别为1278℃和1192℃,均低于不锈钢包壳的熔化温度,因此整个事故过程中不会发生堆芯熔化。      

Analysis of decay heat removal from RBMK-1500 reactor in decommissioning phase by natural circulation of water and air

The main problem in nuclear energy is providing of safety at all stages of lifetime of nuclear installations in conditions of normal operation, accidents and at shutdown. Ignalina NPP, located in Lithuania, is one of the latest with RBMK reactors at highest capacity. Ignalina NPP has two units, both are closed for decommissioning now (in 2004 and 2009). Both units are equipped with RBMK-1500 reactors, the thermal power output is 4200 MW, the electrical power capacity is 1500 MW for each. In RBMK-1500 reactor the fuel assemblies remain for long time inside reactor core after the final shutdown. The paper discusses possibility of heat removal from the RBMK-1500 core at shutdown condition by natural circulation of water (1) and air (2) inside the fuel channels. In first case the decay heat from fuel assemblies is removed due to natural circulation of water and the piping above reactor core should be cooled by means of ventilation in the drum separator compartments. To warrant free access of air in to fuel channels (in the second case) the reactor cooling system should be completely dry out and the pressure headers and the steam discharge valves in steam lines should be opened. If mentioned conditions will be fulfilled, the reactor core will be cooled by natural circulation of water or air and fuel rods remain intact.     

Prediction, analysis and solution of the flow inversion phenomenon in a typical MTR-reactor with upward core cooling

Research reactors of power greater than 20 MW are usually designed to be cooled with upward coolant flow direction inside the reactor core. This is mainly to prevent flow inversion problems following a pump coast down. However, in some designs and under certain operating conditions, flow inversion phenomenon is predicted. In the present work, the best-estimate Material Testing Reactors Thermal-Hydraulic Analysis program (MTRTHA) is used to simulate a typical MTR reactor behavior with upward cooling. The MTRTHA model consists of five interactively coupled submodels for: (a) coolant, (b) fuel plate, (c) chimney, lower plenum, suction box and cold leg, (d) flap valve and (e) natural circulation flow. The model divides the active core into a specified axial regions and the fuel plate into a specified radial zones, then a nodal calculation is performed for both average and hot channels with a chopped cosine shaped heat generation flux. The reactor simulation under loss of off-site power is performed for two cases namely: two-flap valves open and one flap-valve fails to open. The simulation is performed under a hypothetical case of loss of off-site power. Unfortunately, the flow inversion phenomenon is predicted under certain decay heat and/or pool temperature values below the design values. In most cases, the flow inversion phenomenon is accompanied by boiling which is undesirable phenomenon in this type of reactors as it could affect the fuel-clad integrity. The model results for the flow inversion phenomenon prediction are analyzed and a solution of the problem is suggested.     

A simplified refueling scheme for nuclear heating reactor

The nuclear heating reactor is a clean energy resource with high reliability and high economic benefits. Its basic design characteristics are different from those of big reactors. Some features, such as in-pressure vessel spent fuel storage, long refueling period, etc., provide the possibility to simplify the refueling system for economic purposes. After being stored in a pressure vessel for about 15 years, the spent fuel assemblies, with very low radioactivity and decay heat capacity, may be removed from the reactor pressure vessel to a storage pool by a simplified system including a shielded flask, reactor building crane, and some auxiliary tools. It is demonstrated that this ‘dry-method’ refueling scheme is safe and reliable.     

SADHANA facility for simulation of natural convection in the SGDHR system of PFBR

The Prototype Fast Breeder Reactor (PFBR) is a 500 MWe sodium cooled pool type fast reactor being constructed at Kalpakkam, India. PFBR has all the reactor components immersed in the pool of sodium and the fission heat generated in the core, is removed by the sodium circulating in the pool. During normal operation this fission heat is transferred by primary sodium to secondary sodium, which in turn transfers the heat to water in the steam generator for producing steam. The removal of the decay heat generated in the reactor core after the reactor shutdown is also very important to maintain the structural integrity of reactor core components. PFBR employs two independent systems namely, Operational Grade Decay Heat Removal system (OGDHRS) and Safety Grade Decay Heat Removal System (SGDHRS) for decay heat removal. SGDHR system is a passive system working on natural convection to ensure the core coolability even under station blackout condition. It is very important to study the thermal hydraulic behavior of Safety Grade Decay Heat Removal system of PFBR to ensure its reliable operation. A scaled down model of the circuit, named SADHANA has been modeled, designed, constructed and commissioned for demonstration and evaluation of these systems. The facility has completed around 2000 h of high temperature operation. The performance of the experimental system is satisfactory and it meets all the design requirements. At 550 °C sodium pool temperature in test vessel the secondary sodium loop generated a sodium flow of 6.7 m³/h. These experiments have revealed the adequacy and capability of SGDHR system to remove the decay heat from the fast breeder reactor core after its shutdown.     

快中子通量实验堆失流事故三维数值模拟

为验证计算流体动力学（CFD）方法在钠冷快堆失流事故模拟计算中的可靠性和可行性,针对快中子通量实验堆（FFTF）,建立了包含冷池、热池、堆芯在内的全三维模型,其中堆芯组件简化为多孔介质模型,堆芯保留了盒间特征,各类隔板简化为无厚度面。失流事故下主要参数计算结果与实验数据的对比表明,CFD方法能有效捕捉冷池、热池以及盒间复杂的流动换热现象,堆芯最热组件的位置在瞬态过程发生了变化,热管段出口温度与实验值符合良好,装有温度测点的组件出口温度模拟值较实验值低。CFD方法仍需针对组件盒间进行相应的模型开发和验证,此外还需进行大量全堆级别的实验验证,以保证计算结果的合理性。     

Beyond design basis accidents in spent-fuel pools — Generic Issue 82

Nuclear reactor plants include storage facilities for the wet storage of spent-fuel assemblies. The safety function of the spent-fuel pool (SFP) and storage racks is to cool the spent-fuel assemblies and maintain them in a subcritical array during all credible storage conditions and to provide safe means of loading the assemblies into shipping casks.Generic Issue 82 (GI-82) relates to the concern that for a postulated accident sequence that results in the loss of water from a light-water reactor (LWR) spent-fuel storage pool, a Zircaloy cladding fire could occur and propagate to older stored fuel. This issue was identified during hearings concerning SFP reracking amendments in the late 1970s when licensees were starting to use high-density storage racks. High-density racks are used to accommodate the storage of spent fuel in SFPs at reactor sites until such time as the Department of Energy (DOE) repository is available and spent fuel can be removed from the reactor sites. Maintaining a low-density storage configuration for recently discharged spent fuel would reduce the Zircaloy cladding fire probability by an order of magnitude, but at a greater cost for additional onsite storage space.The accident sequences that could result in water loss from the SFP, including beyond design basis earthquakes, various types of seal failures and dropped shipping casks, and the Zircaloy cladding fire issues have been studied by the NRC staff. The results of these studies are provided in NUREG-1353, “Regulatory Analysis for the Resolution of Generic Issue 82, Beyond Design Basis Accidents in Spent-Fuel Pools”. Although these studies conclude that most of the spent-fuel pool risk is derived from beyond design basis earthquakes, this risk is not greater than the risk from core damage accidents due to these beyond design basis earthquakes. Therefore, reducing the risk from spent-fuel pools due to events beyond the safe shutdown earthquake would still leave a comparable risk due to core damage accidents. The risk due to beyond design basis accidents in spent-fuel pools, while not negligible, is sufficiently low that the added cost involved with further risk reduction is not warranted.