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。     

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。     

Hydrogen management and overpressure protection of the containment for future boiling water reactors

New design and evaluation method for hydrogen management of containment atmosphere have been developed for application in the future boiling water reactor (BWR). These are intended as a part of consideration of severe accidents in the course of design so as to assure a high level of confidence that a large release of radioactivity to the environment that may result in unacceptable social consequences can reasonably be avoided. Emphasis on hydrogen management and protection against overpressure failure is based on the insights from probabilistic safety assessments (PSAs) that late phase overpressure (and associated leakage) and molten corium concrete reaction (MCCI) need attention to ensure that containment remains intact, in case energetic challenges to the containment such as DCH (direct containment heating) or FCI (fuel coolant interactions) are practically eliminated by design or resolved from risk standpoint of view. The authors studied the use of palladium-coated tantalum for hydrogen removal from containment atmosphere in order to avoid pressurization of the containment with small free volume by non-condensable gas and steam. Its effectiveness for ABWR (advanced boiling water reactor) containment was evaluated using laboratory test data. Although further experimental studies are necessary to confirm its effectiveness in real accident conditions, the design is a promising option and one that could be backfitted upon necessity to existing plants for which pressure retaining capability cannot be altered. Also new evaluation method for flammability control under severe accident conditions was developed. This method employes a realistic assessment of the amount of oxygen and hydrogen gases generated by radiolytic decomposition of water under severe accident conditions and their subsequent transport from water to containment atmosphere.     

Development and verification of molten corium–concrete interaction code

During a severe accident of Pressurized Water Reactor(PWR), the core materials was heated, melt located on the lower head of Reactor Pressure Vessel(RPV). With the temperature rise, the corium will melt through the lower head and discharge into the reactor cavity. Those corium will interact with the concrete and damage the integrity of the containment, so some coolability method should used to quench the corium. In order to investigate the progress of MCCI, a MCCI analysis code, that is MOCO, was developed. The MOCO includes the heat transfer behavior in axial and radial directions from the molten corium to the basemat and sidewall concrete, crust generation and growth, and coolability mechanisms reveal the concrete erosion and gas release, which are important for the interaction process. Cavity ablation depth, melt temperature, and gas release are the key parameters in the interaction research. The physical-chemistry reaction is also involved in MOCO code. In the present paper, the related MCCI experiment data were used to verify the models of the MOCO and the calculation results of MOCO were also compared with other MCCI analysis codes.     

Experimental simulation of corium dispersion phenomena in direct containment heating

In a direct containment heating (DCH) accident scenario, the degree of corium dispersion is one of the most significant factors responsible for the reactor containment heating and pressurization. To study the mechanisms of the corium dispersion phenomenon, a DCH separate effect test facility of 1:10 linear scale for Zion PWR geometry is constructed. Experiments are carried out with air-water and air-woods metal simulating steam and molten core materials. The physical process of corium dispersion is studied in detail through various instruments, as well as with flow visualization at several locations. The accident transient begins with the liquid jet discharge at the bottom of the reactor pressure vessel. Once the jet impinges on the cavity bottom floor, it immediately spreads out and moves rapidly to the cavity exit as a film flow. Part of the discharged liquid flows out of the cavity before gas blowdown, and the rest is subjected to the entrainment process due to the high speed gas stream. The liquid film and droplet flows from the reactor cavity will then experience subcompartment trapping and re-entrainment. Consequently, the dispersed liquid droplets that follow the gas stream are transported into the containment atmosphere, resulting in containment heating and pressurization in the prototypic condition. Comprehensive measurements are obtained in this study, including the liquid jet velocity, liquid film thickness and velocity transients in the test cavity, gas velocity and velocity profile in the cavity, droplet size distribution and entrainment rate, and the fraction of dispersed liquid in the containment building. These data are of great importance for better understanding of the corium dispersion mechanisms.     

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.     

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.     

An experimental study on quenching of a radially stratified heated porous bed

The quenching characteristics of a volumetrically-heated particulate bed composed of radially stratified sand layers were investigated experimentally in the POMECO facility. The sand bed simulates the corium particulate debris bed which is formed when the molten corium released from the vessel fragments in water and deposits on the cavity floor during a postulated severe accident in a light water reactor (LWR). The electrically-heated bed was quenched by water from a water column established over top of it, and later also with water coming from its bottom, which was circulating from the water overlayer through downcomers. A series of experiments were conducted to reveal the effects of the size of downcomers, and their locations in the bed, on the quenching characteristics of the radially stratified debris beds. The downcomers were found to significantly increase the bed quenching rate. To simulate the non-condensable gases generated during the MCCI, air and argon were injected from the bottom of the bed at different flow rates. The effects of gas flow rate and its properties on the quenching behaviour were observed. The results indicate that the non-condensable gas flows reduce the quenching rate significantly. The gas properties also affect the quenching characteristics.     

Simulation of MCCI with the TOLBIAC-ICB code based on the phase segregation model

In the event of a severe accident in a pressurized water reactor, corium, a mixture of molten materials issued from the fuel, cladding and structural elements, appears in the reactor core. In some circumstances, corium is likely to melt through the reactor pressure vessel and spread over the concrete basemat of the reactor pit. Molten core concrete interaction (MCCI) then occurs. The main question that has to be addressed in this scenario is whether and when the corium will make its way through the basemat. For some years, CEA is developing a numerical code named TOLBIAC-ICB in order to simulate molten core concrete interaction in reactor case. The general approach used in this code is based on the phase segregation model developed by CEA. The solid phase is supposed to be located at the corium pool boundaries as a solid crust composed of refractory oxides, whereas the corium pool contains no solid. The interfacial temperature between the crust and the pool is the liquidus temperature calculated with the composition of the pool. The interaction between thermalhydraulics (mass and energy balances) and physico-chemistry (liquidus temperature, crust composition, chemical reaction) is modelled through a coupling between TOLBIAC-ICB and the GEMINI code for the determination of the physico-chemistry variables. The main purpose of this paper is to present the modelling used in TOLBIAC-ICB and some validation calculations using the data of experiments available in the literature.     

MELCOR2.2冷水入侵和熔融物喷出模型在MCCI计算中的应用

堆芯熔融物-混凝土相互作用（MCCI）产生的大量不凝气体导致安全壳压力快速升高，威胁安全壳的完整性。最新版本MELCOR2.2程序增加了冷水入侵和熔融物喷出模型以完善堆芯熔融物-混凝土相互作用现象的模拟。本文建立了大功率非能动压水堆堆腔几何模型，针对程序中添加的冷水入侵和熔融物喷出模型进行了应用分析。结果表明：在混凝土的消融过程中，冷水入侵和熔融物喷出模型的开启对熔融物和水接触面之间的硬质壳层有一定的影响；硬质壳层的反复形成和裂解加快了水的汽化，减弱了混凝土的消融速度；堆腔熔融物上部和侧部硬质壳层形成后，熔融物向下传热加强，混凝土的消融速度增加。新增模型能从细节上模拟堆芯熔融物-混凝土相互作用现象，为压水堆核电厂严重事故安全审评提供可靠技术保障。     

大亚湾核电站全厂断电诱发的严重事故过程研究

在大亚湾核电站严重事故计算分析结果的基础上,对全厂断电诱发的典型的严重事故序列及缓解对策进行了分析。结果表明,全厂断电事故发生后,大约1～2h堆芯上部会裸露,压力容器在5～7h后失效。在约100h安全壳超压失效,而堆坑地基在事故后8.7d会被熔蚀5.5m。结果还表明,堆坑注水措施可以防止堆坑地基熔穿并且减少事故中由于堆芯熔融物与混凝土反应产生的氢气。     

小型动力堆断电严重事故下熔融物分层对事故后果的影响研究

采用MELCOR程序,对小型动力堆全部电源丧失严重事故下,下封头失效后“堆坑”不同熔融物分层模型进行计算,并对熔融物分层对事故后果的影响进行了研究。结果表明：不同模型下熔融物的总厚度及其变化趋势基本一致;堆腔底板材料为混凝土时,堆芯熔融物的分层较为复杂,而金属材料相对简单。小型动力堆不会发生堆芯熔融物与混凝土相互作用(MCCI)。不同模型的计算对放射性后果基本无影响,但对“堆坑”熔穿进程有影响,强迫混合模型熔穿时间最快、机理计算模型熔穿时间最慢。从安全分析的角度,选择强迫混合模型较为保守。     

Estimation of ex-vessel steam explosion pressure loads

An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor vessel fails and the molten core pours into the water in the reactor cavity. 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 production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment.In this article, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which is being developed for the simulation of fuel–coolant interactions. A parametric study was performed varying the location of the melt release (central, right and left side melt pour), the cavity water subcooling, the primary system overpressure at vessel failure and the triggering time for explosion calculations. The main purpose of the study was to establish the influence of the varied parameters on the fuel–coolant interaction behaviour, to determine the most challenging cases and to estimate the expected pressure loadings on the cavity walls. For the most explosive central, right side and left side melt pour scenarios a detailed analysis of the explosion simulation results was performed. The study shows that for some ex-vessel steam explosion scenarios higher pressure loads are predicted than obtained in the OECD programme SERENA phase 1.     

In-vessel retention of corium at the Loviisa plant

In-vessel retention of corium has been approved to be part of the severe accident management strategy for IVO's Loviisa plant. The approach selected takes advantage of the unique features of the plant such as a low power density, a reactor pressure vessel (RPV) without penetrations at the bottom, and ice-condenser containment which ensures a flooded cavity in all risk significant sequences. The thermal analyses, which are supported by an experimental program, demonstrate that, in Loviisa, the molten corium on the lower head of the RPV is externally coolable with wide margins. This paper summarizes the approach, the thermal analyses and the plant modifications being implemented.     

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。      

Corium crust strength measurements

Corium strength is of interest in the context of a severe reactor accident in which molten core material melts through the reactor vessel and collects on the containment basemat. Some accident management strategies involve pouring water over the melt to solidify it and halt corium/concrete interactions. The effectiveness of this method could be influenced by the strength of the corium crust at the interface between the melt and coolant. A strong, coherent crust anchored to the containment walls could allow the yet-molten corium to fall away from the crust as it erodes the basemat, thereby thermally decoupling the melt from the coolant and sharply reducing the cooling rate. This paper presents a diverse collection of measurements of the mechanical strength of corium. The data is based on load tests of corium samples in three different contexts: (1) small blocks cut from the debris of the large-scale MACE experiments, (2) 30 cm-diameter, 75 kg ingots produced by SSWICS quench tests, and (3) high temperature crusts loaded during large-scale corium/concrete interaction (CCI) tests. In every case the corium consisted of varying proportions of UO₂, ZrO₂, and the constituents of concrete to represent a LWR melt at different stages of a molten core/concrete interaction. The collection of data was used to assess the strength and stability of an anchored, plant-scale crust. The results indicate that such a crust is likely to be too weak to support itself above the melt. It is therefore improbable that an anchored crust configuration could persist and the melt become thermally decoupled from the water layer to restrict cooling and prolong an attack of the reactor cavity concrete.     

Direct containment heating integral effects tests in geometries of European nuclear power plants

The DISCO test facility at Forschungszentrum Karlsruhe (FZK) has been used to perform experiments to investigate direct containment heating (DCH) effects during a severe accident in European nuclear power plants, comprising the EPR, the French 1300 MWe plant P’4, the VVER-1000 and the German Konvoi plant. A high-temperature iron–alumina melt is ejected by steam into scaled models of the respective reactor cavities and the containment vessel. Both heat transfer from dispersed melt and combustion of hydrogen lead to containment pressurization. The main experimental findings are presented and critical parameters are identified.The consequences of DCH are limited in reactors with no direct pathway between the cavity and the containment dome (closed pit). The situation is more severe for reactors which do have a direct pathway between the cavity and the containment (open pit). The experiments showed that substantial fractions of corium may be dispersed into the containment in such cases, if the pressure in the reactor coolant system is elevated at the time of RPV failure. Primary system pressures of 1 or 2 MPa are sufficient to lead to full scale DCH effects. Combustion of the hydrogen produced by oxidation as well as the hydrogen initially present appears to be the crucial phenomenon for containment pressurization.     

Analysis on containment depressurization under severe accidents for a Chinese 1000 MWe NPP

Containment depressurization has been implemented for many nuclear power plants (NPPs) to mitigate the risk of containment overpressurization induced by steam and gases released in LOCA accidents or generated in molten core concrete interaction (MCCI) during severe accidents. Two accident sequences of large break loss of coolant accident (LB-LOCA) and station blackout (SBO) are selected to evaluate the effectiveness of the containment venting strategy for a Chinese 1000 MWe NPP, including the containment pressure behaviors, which are analyzed with the integral safety analyses code for the selected sequences. Different open/close pressures for the venting system are also investigated to evaluate CsI mass fraction released to the environment for different cases with filtered venting or without filtered venting. The analytical results show that when the containment sprays can't be initiated, the depressurization strategy by using the Containment Filtered Venting System (CFVS) can prevent the containment failure and reduce the amount of CsI released to the environment, and if CFVS is closed at higher pressure, the operation interval is smaller and the radioactive released to the environment is less, and if CFVS open pressure is increased, the radioactive released to the environment can be delayed. Considering the risk of high pressure core melt sequence, RCS depressurization makes the CFVS to be initiated 7 h earlier than the base case to initiate the containment venting due to more coolant flowing into the containment.     

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%.     

Three-dimensional behaviors of the hydrogen and steam in the APR1400 containment during a hypothetical loss of feed water accident

During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by an active reaction of the fuel-cladding and the steam in the reactor pressure vessel and released with the steam into the containment. In order to mitigate hydrogen hazards which could possibly occur in the NPP containment, a hydrogen mitigation system (HMS) is usually adopted. The design of the next generation NPP (APR1400) developed in Korea specifies that 26 passive autocatalytic recombiners and 10 igniters should be installed in the containment for a hydrogen mitigation. In this study, an analysis of the hydrogen and steam behavior during a total loss of feed water (LOFW) accident in the APR1400 containment has been conducted by using the computational fluid dynamics (CFD) code GASFLOW. During the accident, a huge amount of hot water, steam, and hydrogen is released into the in-containment refueling water storage tank (IRWST). The current design of the APR1400 includes flap-type openings at the IRWST vents which operate depending on the pressure difference between the inside and outside of the IRWST. It was found from this study that the flaps strongly affect the flow structure of the steam and hydrogen in the containment. The possibilities of a flame acceleration and a transition from deflagration to detonation (DDT) were evaluated by using the Sigma–Lambda criteria. Numerical results indicate that the DDT possibility was heavily reduced in the IRWST compartment by the effects of the flaps during the LOFW accident.