Vulnerability of a partially flooded PWR reactor cavity to a steam explosion

When the hot molten core comes into contact with the water in the reactor cavity a steam explosion may occur. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and later, during the expansion of the water vapour, to production of missiles that may endanger surrounding structures.The purpose of the performed analysis is to provide an estimation of the expected pressure loadings on the typical PWR cavity structures during a steam explosion, and to make an assessment of the vulnerabilities of the typical PWR cavity structures to steam explosions. To achieve this, the fit-for-purpose steam explosion model is proposed, followed by comprehensive and reasonably conservative analyses of two typical ex-vessel steam explosion cases differing in the steam explosion energy conversion ratio. In particular, the vulnerability of the surrounding reinforced concrete walls to damage has been sought in both cases.     

Ex-vessel fuel-coolant interaction experiment in the DISCO facility in the LACOMECO project

In the frame of the LACOMECO (large scale experiments on core degradation, melt retention and containment behavior) project of the 7th European Framework Program, a test in the DISCO (dispersion of corium) facility was performed in order to analyze the phenomena which occur during an ex-vessel fuel–coolant interaction (FCI). The test is focused on the premixing phase of the FCI with no trigger used for explosion phase. The objectives of the test were to provide data concerning the dispersion of water and melt out of the pit, characterization of the debris and pressurization of the reactor compartments for scenarios, where the melt is ejected from the reactor pressure vessel (RPV) under pressure. The experiment was performed for a reactor pit geometry close to a French 900 MWe reactor configuration at a scale of 1:10. The corium melt was simulated by a melt of iron–alumina with a temperature of 2400 K. A containment pressure increase of 0.04 MPa was measured, the total pressure reached about 0.24 MPa. No spontaneous steam explosion was observed. About 16% of the initial melt (11.62 kg) remained in the RPV vessel, 60% remained in the cavity mainly as a compact crust. The fraction of the melt transported out of the pit was about 24%.     

基于TEXA-Ⅴ的AP1000堆外蒸汽爆炸分析

TEXAS-Ⅴ是一维三相瞬态蒸汽爆炸数学物理分析程序,本文采用TEXAS-Ⅴ程序对AP1000堆外蒸汽爆炸进行分析研究。结果表明：熔融物在粗混合阶段不断碎裂,并与冷却剂发生剧烈热量交换;AP1000堆外蒸汽爆炸的压力波随传播强度逐渐降低,压力波的传播会触发熔融物前沿后的熔融物碎裂产生更强的压力波,峰值可达70 MPa,且熔融物液柱具有合适的粗混合时间,较大的初始注入速度以及较大的注入直径能触发蒸汽爆炸产生更为强烈的压力波,具有更大的危险性。     

RPV侧边破口条件下堆外蒸汽爆炸计算分析

针对实际过程中更有可能发生的压力容器（RPV）侧边破口条件开展蒸汽爆炸计算分析。根据经济合作与发展组织（OECD）发布的现象识别与重要度排序表（PIRT）,选取堆外蒸汽爆炸敏感性分析参数,使用MC3D软件建立三维局部破口和二维环状破口几何模型,对影响计算结果的重要参数（破口尺寸、堆坑水位、破口位置、触发条件、液柱碎化和液滴碎化模型）开展RPV侧边破口条件下敏感性分析,获得最恶劣计算工况条件。敏感性分析结果表明,在大破口失水事故（LBLOCA）工况下,当堆坑处于满水位、RPV发生二维侧边环状破口、接触堆坑侧壁面时触发蒸汽爆炸、采用CONST模型和Classical模型时,堆坑侧壁面的压力载荷计算结果最为保守,对堆坑和安全壳完整性威胁最大。      

Numerical analysis of pressure load in a PWR cavity in an ex-vessel steam explosion

Ex-vessel steam explosion may happen as a result of melting core falling into the reactor cavity after failure of the reactor vessel and interaction with the coolant in the cavity pool. It can cause the formation of shock waves and production of missiles that may endanger surrounding structures. Ex-vessel steam explosion ener- getics is affected strongly by three dimensional （3D） structure geometry and initial conditions. Ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which is developed for simulating fuel-coolant interactions. The reactor cavity with a venting tunnel is modeled based on 3D cylin- drical coordinate. A study was performed with parameters of the location of molten drop release, break size, melting temperature, cavity water subcooling, triggering time and explosion position, so as to establish parame- ters＇ influence on the fuel-coolant interaction behavior, to determine the most challenging cases and to estimate the expected pressure loadings on the cavity walls. The most dangerous case shows the pressure loading is above the capacity of a typical reactor cavity wall.     

Evaluation of Containment Failure Probability by Ex-Vessel Steam Explosion in Japanese LWR Plants

The containment failure probability due to ex-vessel steam explosions was evaluated for Japanese BWR and PWR model plants. A stratified Monte Carlo technique (Latin Hypercube Sampling (LHS)) was applied for the evaluation of steam explosion loads, in which a steam explosion simulation code JASMINE was used as a physics model. The evaluation was made for three scenarios: a steam explosion in the pedestal area or in the suppression pool of a BWR model plant with a Mark-II containment, and in the reactor cavity of a PWR model plant. The scenario connecting the generation of steam explosion loads and the containment failure was assumed to be displacement of the reactor vessel and pipings, and failure at the penetration in the containment boundary. We evaluated the conditional containment failure probability (CCFP) based on the preconditions of failure of molten core retention within the reactor vessel, relocation of the core melt into the water pool without significant interference, and a strong triggering at the time of maximum premixed mass. The obtained mean and median values of the CCPF were 6.4x 10^?2 (mean) and 3.9x 10^?2 (median) for the BWR suppression pool case, 2.2x10^?3 (mean) and 2.8x10^?10 (median) for the BWR pedestal case, and 6.8X10^?2 (mean) and 1.4x10^?2 (median) for the PWR cavity case. The evaluation of CCFPs on the basis of core damage needs consideration of probabilities for the above-mentioned preconditions. Thus, the CCFPs per core damage should be lower than the values given above. The specific values of the probability were most dependent on the assumed range of melt flow rate and fragility curve that involved conservatism and uncertainty due to simplified scenarios and limited information.     

熔融物液柱碎化模型对先进压水堆堆外蒸汽爆炸计算的影响分析

在堆外蒸汽爆炸计算中,液柱碎化模型影响着熔融物液滴生成速率、液滴直径、液滴分布、液滴凝固和气泡比例等粗混合参数和现象,从而影响了蒸汽爆炸的冲击载荷。本文基于MC3D V3.8程序,采用不同的液柱碎化模型（CONST模型和KHF模型）对先进压水堆堆外蒸汽爆炸进行计算分析,探讨了CONST和KHF模型对蒸汽爆炸计算的影响。结果表明,两种模型计算的粗混合状态类似;在熔融物触底时刻,爆炸性准则几乎相同,此时触发爆炸得到的冲击载荷差别很小,表明该时刻触发爆炸时不同液柱碎化模型对爆炸冲击计算的影响较小;在本文所定义的工况下,先进压水堆堆坑墙体承受的最高压力约为20 MPa,最大冲量小于0.2 MPa•s。     

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.     

An evaluation of a direct cooling method for the ex-vessel corium stabilization

An evaluation of the ex-vessel core catcher system of a sample advanced light water reactor was presented. The core catcher was designed to cool down the molten corium through a combined injection of water and gas from the bottom of the molten corium, which could be effective in the reduction of rapid steam generation. By using the MELCOR code, a scenario analysis was performed for a representative severe accident scenario of the ALWR, that is, the 6-in. large break loss of coolant accident without safe injection. The spreading characteristics of ejected corium at vessel breach were asymptotically evaluated on the core catcher horizontal surface. The composition of the molten corium, the decay power level, and the sacrificial concrete ablation depth with time were obtained by a sacrificial concrete ablation analysis. The corium cooling history in the core catcher during the coolant injection was evaluated to calculate the temporal steam generation rate by considering an energy conservation equation. These were used as the major inputs for the temporal calculations of containment pressure which was performed by using the GASFLOW code. Several cases with change of water and gas injection rates were calculated. It was confirmed that the bottom water/gas injection system was an effective corium cooling method in the ex-vessel core catcher to suppress the quick release of steam.     

严重事故条件下安全壳响应模拟研究

压水堆核电厂发生严重事故期间,从主系统释放的蒸汽、氢气以及下封头失效后进入安全壳的堆芯熔融物均对安全壳的完整性构成威胁。以国内典型二代加压水堆为研究对象,采用MAAP程序进行安全壳响应分析。选取了两种典型的严重事故序列：热管段中破口叠加设备冷却水失效和再循环高压安注失效,堆芯因冷却不足升温熔化导致压力容器失效,熔融物与混凝土发生反应（MCCI）,安全壳超压失效;冷管段大破口叠加再循环失效,安全壳内蒸汽不断聚集,发生超压失效。通过对两种事故工况的分析,证实了再循环高压安注、安全壳喷淋这两种缓解措施对保证安全壳完整性的重要作用。     

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。     

Effect of melt water interaction configuration on the process of steam explosion

Steam explosion experiments are performed at various modes of melt water interaction configuration using prototypic corium melt. The tests are performed to simulate both melt water interaction in a partially flooded cavity and melt water interaction in a cavity with submerged reactor. The tests are performed using zirconia and corium melts. The behavior of melt jet fragmentation during the flight in the air and fragmentation and mixing of melt jet in water is investigated by a high-speed video visualization and by comparison of debris size distribution and morphology of debris. Strength of steam explosion is estimated by measuring dynamic pressure and dynamic force.     

Ex-vessel transport and mixing of a deborated slug in a PWR primary geometry

The transport and mixing of a slug of deborated water in a lowered loop PWR is modeled by partitioning the volumes of the primary system according to chemical rector theory. Piping is modeled as plug flow volumes while the steam generator outlet plenum and the reactor coolant pumps are modeled as backmixed volumes. This simple approach provides a good representation of the transport and mixing phenomena outside the reactor vessel. The proposed methodology can be used to generate initial and boundary conditions for separate effects tests and CFD computations for the reactor vessel complex geometry. The decoupling of the ex-vessel primary system greatly enhances the resolution of boron dilution transient issue.     

Experiments on the interactions of molten ZrO2 with water using TROI facility

Korea Atomic Energy Research Institute (KAERI) launched an intermediate scale steam explosion experiment named ‘Test for Real cOrium Interaction with water (TROI)’ using reactor material. The objective of the program is to investigate whether the corium would lead to energetic steam explosion when interacted with cold water at a low pressure. The melt/water interaction is made in a multi-dimensional test section located in a pressure vessel. The inductive skull melting, which is basically a direct inductive heating of an electrically conducting melt, is implemented for the melting and delivery of corium. In the first series of tests using several kg of ZrO₂ where the melt/water interaction is made in a heated water pool at 30–95 °C, either a quenching or a spontaneous steam explosion was observed. The spontaneous explosion observed in the present ZrO₂ melt/water experiments clearly indicates that the physical properties of the UO₂/ZrO₂ mixture have a strong effect on the energetics of steam explosion.     

出口集管LLOCA始发严重事故分析

采用一体化分析程序建立了包括热传输系统、慢化剂系统、端屏蔽系统、蒸汽发生器二次侧系统的重水堆核电厂的严重事故分析模型。并选取出口集管发生双端剪切断裂的大破口失水事故（LLOCA）,同时叠加低压安注失效,辅助给水强制关闭的严重事故序列进行热工水力分析。由于主热传输系统环路隔离阀的关闭,使得两个环路的热工水力响应过程不同。最终由于低压安注的失效,慢化剂系统逐渐被加热,最终导致堆芯熔化、排管容器蠕变失效。在LLOCA事故序列中叠加向排管容器中注水的缓解措施,可以终止事故进程,使堆芯保持安全、稳定的状态。     

Melt quenching and coolability by water injection from below: Co-injection of water and non-condensable gas

The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The goal of our work is to provide the fundamental understanding needed for melt–water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability. The work considers the ex-vessel coolability phenomena in two stages. The first stage is the melt quenching process and is being addressed by Argonne National Lab and University of Wisconsin in modified test facilities. Given a quenched melt in the form of solidified debris, the second stage is to characterize the long-term debris cooling process and is being addressed by Korean Maritime University via test and analyses. In this paper, experiments on melt quenching by the injection of water from below are addressed. The test section represented one-dimensional flow-channel simulation of the bottom injection of water into a core melt in the reactor cavity. The melt simulant was molten lead or a lead alloy (Pb–Bi). For the experimental conditions employed (i.e., melt depth and water flow rates), it was found that: (1) the volumetric heat removal rate increased with increasing water mass flow rate and (2) the non-condensable gas mixed with the injected water had no impairing effect on the overall heat removal rate. Implications of these current experimental findings for ALWR ex-vessel coolability are discussed.     

大型先进压水堆IVR策略评估和裂变产物分布计算

大型先进压水堆通过堆内熔融物滞留（IVR）策略来缓解严重事故后果以降低安全壳失效风险。其中堆腔注水系统（CIS）被引入来实现IVR。本文使用严重事故分析软件计算大型先进压水堆在冷管段双端断裂事故下的事故进程、热工水力行为、堆芯退化过程和下封头熔融池传热行为,评估能动CIS的事故缓解能力。计算结果表明,事故后72 h,下封头外表面热流密度始终低于临界热流密度（CHF）,表明IVR策略有效。此外,计算分析了惰性气体、非挥发性和挥发性裂变产物的释放和迁移行为。计算发现,IVR下更多的放射性裂变产物分布在主系统内,壁面核素再悬浮形成气溶胶的行为被消除,安全壳壁面上沉积的核素被大量冷凝水冲刷进入底部水池。总体来说,IVR策略能更好地管理放射性核素分布,减小放射性泄漏威胁。     

Simulation of the physical effects of the interaction of molten fuel with the coolant

Conclusions The method proposed makes it possible to obtain computational estimates of the intensity of a steam explosion inside a reactor vessel and in the space below the reactor inside the melt trap. The computational investigations of the intensity of a steam explosion inside a VVéR vessel in the most likely scenario of a serious accident with efflux of melt into the bottom pressurized chamber show that under certain conditions a high pressure capable of destroying separate structural elements can develop. The mass of the interacting melt, the initial temperature, the fragmentation time, and the final size of the fragments, as well as the type of contact realized, have the greatest effect on the intensity of the steam explosion. Local steam explosions in pipes of the melt trap have a relatively low intensity and cannot have a large effect on the construction in the space below the reactor and on the containment envelope. Deceased. State Science Center of the Russian Federation — Physics and Power Engineering Institute. Translated from Atomnaya énergiya, Vol. 80, No. 1, pp. 3–10, January, 1996.     

Safety design of Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR)

In Pb–Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb–Bi coolant above the core, and Pb–Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb–Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits.     

Ex-vessel releases: Aerosol source terms in reactor accidents

An ex-vessel aerosol and fission-product source term may arise from various events occurring in the containment building of a nuclear reactor. The research into the source terms associated with three of these events is reviewed. These source terms are from steam explosions, pressurized melt ejection, and melt/concrete interaction.The least is known about the steam explosion source term. Analyses indicate that its magnitude is likely lower than that assumed in the Reactor Safety Study (WASH-1400), but no conclusive experimental data are as yet available.The aerosol and fission-product source term from pressurized ejection of melt is an issue only recently addressed. Experimental evidence has allowed estimates to be made of the magnitude of this source term.The source term from melt/concrete interaction has been long recognized and has the largest data base. Experimental programs have addressed this source term for several years. A mechanistic model of material release has been developed and is discussed.