Severe accident development modeling and evaluation for CANDU

Romania as UE member got new challenges for its nuclear industry. Romania operates since 1996 a CANDU nuclear power reactor and since 2007 the second CANDU unit. In EU are operated mainly PWR reactors, so, ours have to meet UE standards. Safety analysis guidelines require to model nuclear reactors severe accidents.Starting from previous studies, a CANDU degraded core thermal hydraulic model was developed. The initiating event is a LOCA, with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperature inside a pressure tube reaches 1000 °C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator, eventually, begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which surrounds the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data.     

Analysis of severe core damage accident progression for the heavy water reactor

In this study,the severe accident progression analysis of generic Canadian deuterium uranium reactor 6 was preliminarily provided using an integrated severe accident analysis code.The selected accident sequences were multiple steam generator tube rupture and large break loss-of-coolant accidents because these led to severe core damage with an assumed unavailability for several critical safety systems.The progressions of severe accident included a set of failed safety systems normally operated at full power,and initiative events led to primary heat transport system inventory blow-down or boil off.The core heat-up and melting,steam generator response,fuel channel and calandria vessel failure were analyzed.The results showed that the progression of a severe core damage accident induced by steam generator tube rupture or large break loss-of-coolant accidents in a CANDU reactor was slow due to heat sinks in the calandria vessel and vault.     

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.     

An evaluation of the severe accident management strategies for CANDU-6 type plants

This paper provides an evaluation of the mitigation effects for the severe accident management strategies of the Wolsong plants which are typical CANDU-6 type reactors. The evaluation includes the effect of the following six mitigation strategies: (1) injection into the primary heat transport system (PHTS), (2) injection into the calandria vessel, (3) injection into the calandria vault, (4) reduction of the fission product release, (5) control of the reactor building condition, (6) reduction of the reactor building hydrogen. The tested scenario is a loss of coolant accident with a small out-of-core break, and the thermal hydraulic and severe accident phenomenological analyses were implemented by using the ISAAC computer program. The calculation results show that the most effective means for a primary decay heat removal is a low pressure safety injection, that for a calandria vessel integrity is an end-shield cooling injection, and that for a reactor building integrity is a pressure control via local air coolers. Besides the above, the usefulness of each safety component was evaluated in this analysis.     

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。     

Development of Debris Coolability Analysis Module in Severe Accident Analysis Code SAMPSON for IMPACT Project

Debris coolability in the lower plenum of the reactor pressure vessel is an important factor for the evaluation of in-vessel debris retention. The debris coolability analysis module has been developed to predict more mechanistically the safety margin of the present reactor vessels in a severe accident. The module calculates debris spreading and cooling through melting and solidification in combination with the temperature distribution of the vessel wall and it evaluates the wall failure. Debris spreading is solved by the explicit method on a quasi-three-dimensional scheme and debris coolability is solved on the basis of natural convection analysis with melting and solidification. The calculated results for spreading were compared with the results from a water spreading experiment on the floor and the results for coolability were compared with those from an n-octadecane melting experiment in the rectangular vessel. The comparisons showed the capability for predictions of the spearhead transportation in the debris spreading process and of the melting front transportation and time evolution of the fluid temperature in the melting process. The module provides a good tool for the prediction of the reactor pressure vessel safety margin in a severe accident through the analysis of debris spreading and coolability.     

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.     

SCDAP/RELAP5与MELCOR程序对堆芯损伤过程预测的比较

SCDAP／RELAP5与MELCOR程序是目前得到广泛使用的两个严重事故分析程序．它们在模拟堆芯溶化及压力容器下封头失效过程中采用了基于不同理论的计算模型。本文利用两个程序分别对秦山二期核电厂发生假想的全厂断电事故下的堆芯损伤过程进行预测．并对比分析了这2个严重事故分析程序的优点及相应的计算结果．     

The evaluation of RCS depressurization to prevent core melting in pressure tube reactors (CANDU-type)

Pressure tube reactors, especially of the CANDU-type, have a low-pressure vessel calandria – under an internal pressure near atmospheric. The calandria vessel is immersed into the water contained inside a concrete structure – the calandria vault. In the case of accidents with the loss of normal core heat sinks, the moderator inside the calandria (heavy water) could become the ultimate heat sink. Accident analysis using a newly developed model (ASQR) strengthens the importance of the inside cooling of the fuel channels in order to prevent severe accidents. Even if implementing those methods related to moderator for eliminating the impairment of the outside cooling of fuel channels, these are not sufficient. The new model has been compared against the well-known in-reactor LOCA experiment – PBF – NRU.     

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.     

Cernavoda CANDU severe accident evaluation

The papers present the activities dedicated to Romania Cernavoda Nuclear Power Plant first CANDU Unit severe accident evaluation. This activity is part of more general PSA assessment activities. CANDU specific safety features are calandria moderator and calandria vault water capabilities to remove the residual heat in the case of severe accidents, when the conventional heat sinks are no more available. Severe accidents evaluation, that is a deterministic thermal hydraulic analysis, assesses the accidents progression and gives the milestones when important events take place. This kind of assessment is important to evaluate to recovery time for the reactor operators that can lead to the accident mitigation. The Cernavoda CANDU unit is modeled for the of all heat sinks accident and results compared with the AECL CANDU 600 assessment.     

An application of the new way to prevent core melting in pressure tube reactors (CANDU type)

The pressure tube reactors, especially CANDU type, have a calandria low pressure vessel (near to atmospheric pressure) immersed into a concrete vault filled with water. The accident analysis done by ELFIN-HTCELL code for the channel heat up and by fluid flow PHOENICS code as applied for moderator cooling system efficacy, showed that even the moderator cooling system operates, in some transients sequences where the normal heat sinks are lost, and the top core pressure tubes can reach burst conditions, which means that the fission product secondary retaining barrier gets destroyed, and yet the core can be cooled by water admission through the ruptured tubes from the emergency core cooling system (ECC), if it is available. Otherwise, if in many accident sequences the moderator cooling system remains the ultimate heat sink for the core fuel, and it is not available even from the accident start, a core melt appears. Taking into account the “natural” advantage offered by the presence of both pools in calandria and in the vault, separated by the calandria vessel, the introduction of density locks between them could be a safety passive design solution. When the temperature of moderator water gets higher the density lock cold-hot interface loss stability and thus the density locks get “open” fully permitting the admission of the cool water from the vault pool in calandria. Therefore, by natural circulation the decay heat is transferred via an air-cooling tower, and no mechanical moving parts are needed to open this circuit. Also, if the vault water is borated, it can be used to stop the nuclear reaction when the normal shutdown systems are not available and a positive reactivity coefficient appears, e.g. large loss of coolant accident (LOCA).     

Validation and application of the WABE code: Investigations of constitutive laws and 2D effects on debris coolability

The WABE-2D model aims at the problem of coolability of degraded core material during a severe accident in a light water reactor (LWR) and describes the transient boil-off and quenching behavior of debris beds. It is being developed in the frame of the KESS code system, which is considered to describe the processes of core heatup, melting, degradation and relocation to the lower plenum as well as the subsequent behavior. The models developed in this frame are being integrated in the German system code ATHLET-CD.An emphasis of the present contribution lies on multidimensional aspects of the cooling behavior. From multidimensional features a significant improvement of overall coolability is expected compared to what is concluded based on classical one-dimensional analyses. Such analyses – also mainly oriented at top cooling conditions – additionally miss the expected importance of interfacial drag which should support coolability in co-current flow situations due to bottom flooding. The latter situation plays a role in the multidimensional behavior expected under realistic conditions. Thus, a further emphasis in the present contribution lies on the constitutive drag laws and their effects in such configurations.Calculations comparing top and bottom flooding and the influence of interfacial friction are presented. An explanation for effects observed in related experiments at Forschungszentrum Karlsruhe is provided based on this influence. The significant increase of dryout heat flux with water inflow from below, driven by a lateral water column, is reproduced and understood. Enhanced cooling due to this and in general by lateral inflow is also demonstrated for reactor scenarios, considering particulate debris in the lower head of the reactor pressure vessel (RPV) of a LWR or in a deep water pool in the reactor cavity of a boiling water reactor (BWR). Cooling by steam flow through local dry zones can establish under lateral water supply to regions below and yield a further extension of coolability. Quenching of hot material is also analyzed. Finally, cases with loss of coolability, dry zone formation and melting are considered, especially in the perspective to analyze melt pool formation in the lower head of the RPV and the history of thermal interaction with the lower head wall. The latter will determine failure possibilities and modes of the RPV.     

Core cooling in pressurized-water reactor during water injection

In this paper,the reactor core cooling and its melt progression terminating is evaluated,and the initiation criterion for reactor cavity flooding during water injection is determined.The core cooling in pressurized-water reactor of severe accident is simulated with the thermal hydraulic and severe accident code of SCDAP/RELAP5.The results show that the core melt progression is terminated by water injection,before the core debris has formed at bottom of core,and the initiation of reactor cavity flooding is indicated by the core exit temperature.     

Verification for Flow Analysis Capability in the Model of Three-Dimensional Natural Convection with Simultaneous Spreading,Melting and Solidification for the Debris Coolability Analysis Module in the Severe Accident Analysis Code ‘SAMPSON’, (I)

The debris coolability analysis module in the severe accident analysis code ‘SAMPSON’ has been enhanced to predict more mechanistically the safety margin of present reactor pressure vessels in a severe accident. The module calculates debris three-dimensional natural convection with simultaneous spreading, melting and solidification using the ‘debris spreading-cooling model’ in combination with the temperature distribution of the vessel wall and it evaluates the wall failure. Debris spreading is solved by the free surface calculation method in which the height function is applied. The model makes possible a multiplex heat and mass transfer analysis with flow spearhead and melt front transportation for a single-phase flow analysis code through the resetting of two types of mesh attributions and re-arrangement of the pressure matrix at each time step. The results calculated with the present model are compared with the results from a water spreading experiment. The comparisons verify the model capability for predictions of debris flow in the spreading process. The module provides a good tool for prediction of the reactor safety margin in a severe accident through the three-dimensional natural convection analysis of debris with simultaneous spreading, melting and solidification.     

Experimental investigation of coolability behaviour of irregularly shaped particulate debris bed

In case of a severe nuclear reactor accident, the core can melt and form a particulate debris bed in the lower plenum of the reactor pressure vessel (RPV). Due to the decay heat, the particle bed, if not cooled properly, can cause failure of the RPV. In order to avoid further propagation of the accident, complete coolability of the debris bed is necessary. For that, understanding of various phenomena taking place during the quenching is important. In the frame of the reactor safety research, fundamental experiments on the coolability of debris beds are carried out at IKE with the test facility “DEBRIS”. In the present paper, the boiling and dry-out experimental results on a particle bed with irregularly shaped particles mixed with stainless steel balls have been reported. The pressure drops and dry-out heat fluxes of the irregular-particle bed are very similar to those for the single-sized 3 mm spheres bed, despite the fact that the irregular-particle bed is composed of particles with equivalent diameters ranging from 2 to 10 mm. Under top-flooding conditions, the pressure gradients are all smaller than the hydrostatic pressure gradient of water, indicating an important role of the counter-current interfacial drag force. For bottom-flooding with a liquid inflow velocity higher than about 2.7 mm/s, the pressure gradient generally increases consistently with the vapour velocity and the fluid-particle drag becomes important. The system pressures (1 and 3 bar) have negligible effects on qualitative behaviour of the pressure gradients. The coolability of debris beds is mainly limited by the counter-current flooding limit (CCFL) even under bottom-flooding conditions with low flow rates. The system pressure and the flow rate are found to have a distinct effect on the dry-out heat flux.Different classical models have been used to predict the pressure drop characteristics and the dry-out heat flux (DHF). Comparisons are made among the models and experimental results for DHF and pressure drop characteristics. Considering the overall trend in prediction of DHF and two-phase pressure drop, it was observed that none of the models could provide accurate predictions for both DHF and pressure drop under top- and bottom-flooding conditions. This implies that developments of more accurate models are needed including the effects of non-uniform particle sizes and the multidimensional nature of particulate debris beds, which are not reflected so far in these models.     

Concept of a core cooling system and experiments performed

One of the problems which must be solved in severe accidents is the melt-concrete interaction which does occur when the core debris penetrates the lower pressure vessel head and contacts the basement. To prevent these accident consequences, a core catcher concept is proposed to be integrated into a new pressurized-water reactor design. The core catcher achieves coolability by spreading and fragmentation of the ex-vessel core melt based on the process of water inlet from the bottom.In order to justify the dominant process during flooding of the melt from the bottom, prototypic experiments with thermite melts in laboratory scale have been carried out. In these experiments flooding and early coolability of the melt is demonstrated. To obtain more detailed information on the important process of water penetration into the melt, a simulant experiment has been conducted using a transparent plastic melt with the typical viscosity behaviour of an oxidic corium melt and a temperature allowing evaporation of water. In every experiment the melt is flooded, and complete freezing in the form of a porous layer occurs within a few minutes only.     

压水堆核电站全厂断电事故及辅助给水系统的缓解能力研究

以美国surry核电站为参考对象,采用最佳估算程序SCDAP/RELAP5/MOD3.4,建立了一个典型的三环路压水堆核电站严重事故计算模型,对全厂断电(SBO)事故的物理现象及堆芯熔化进程进行了详细分析,并研究了全厂断电事故发生后辅助给水(AFW)分别持续1800s和3600s对事故的缓解效果.计算结果显示,辅助给水能有效地延缓堆芯熔化进程,大大推迟反应堆压力容器的失效时间,为操纵员恢复交流电源以及实施其它缓解措施赢得更多的时间.     

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。     

Debris interactions in reactor vessel lower plena during a severe accident I. Predictive model

An integral predictive physico-numerical model has been developed to understand and interpret debris interactions in the reactor vessel plenum such as those which took place in the TMI-2 accident. The model represents the extent of debris jet disintegration by a jet-water entrainment model which can result in two types of debris configurations. One is particulated debris which eventually quenches in the water as a result of the entrainment process. The remainder of the debris penetrates to the bottom of the lower plenum and collects as a continuous layer. Each is treated as a separate region and has governing principles for its behavior. The potential for creating gap (contact) resistance and boiling heat removal is considered for heat transfer between the debris bed, the reactor vessel and steel structures and, most importantly, the vessel-to-crust gap water. The proposed in-vessel cooling mechanism due to material creep and water ingression into the expanding gap between the core debris and the vessel wall was found to explain the non-failure of the TMI-2 vessel in the course of the accident. The particulate debris bed is a mixture of metal and oxide, which is distributed as individual spherical particles of sizes determined at the time of entrainment. Energy is received from the continuum bed below by radiation and convection. The continuum debris bed is described by the crust behavior with the heat flux to the crust given by the natural convection correlations relating the Nusselt and Rayleigh numbers for the central region of debris. Using these governing principles, the rate laws for heat and mass transfer are formulated for each type of debris condition in the lower plenum. With the integration of the individual rates, the formation, growth and possible shrinkage of these regions are calculated. The potential reactor vessel breach is accounted for by considering the combined thermal and mechanical response of the vessel wall. The two-step failure model allows the vessel to fail at two different locations and at two different times.