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.     

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.     

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。     

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.     

Adapting and applying SCDAP/RELAP5 to CANDU in-vessel retention studies

In CANDU reactors, the cool moderator surrounding the calandria tubes provides a potential heat sink following an accident initiator if the emergency coolant injection fails. However, in scenarios when a subsequent loss of all heat sinks occurs, the fuel channels fail and ultimately, the entire reactor core collapses and relocates into the bottom of calandria vessel (CV), which is externally cooled by shield-tank water. Previous studies using MAAP4-CANDU and ISAAC computer codes were found to investigate the long-term coolability of the CV in the late phase of core degradation in course of a severe accident. SCDAP/RELAP5 was applied in a previous work of the authors to the study of the in-vessel retention issue using the COUPLE models with user-defined slumping inside the 2D COUPLE mesh. This option allows for thermal and mechanical analyses of the reactor lower head avoiding the necessity to calculate the preceding course of core degradation during the accident. The former analyses used an equivalent spherically shaped CV while, for the present paper, calculations are performed with COUPLE routines modified to properly use the option for a horizontal pipe in plane geometry. The paper describes the modifications and the application of the resulted SCDAP/RELAPSIM/MOD3.4 code version to the study of the coolability of a CV starting with a dry debris bed. The vessel rupture time is compared to the ISAAC calculated value for a LOCA with loss of all heat sinks and no recovery actions. Parametric studies are performed in order to quantify the effect of several identified sources of uncertainty: boundary conditions of the vessel above debris, gap heat transfer coefficient and metallic fraction of zirconium inside the debris.     

Core coolability of an ATR by heavy water moderator in situations beyond design basis accidents

In the case of a loss-of-coolant accident (LOCA) with coincident loss of emergency coolant injection (LOECI), core cooling is generally very severe. However, as the ATR plant has heavy water at about 60°C in the core, decay heat can be removed by the heavy water cooling system. Separate-effects tests relating to heavy water cooling were conducted with each setup. The important thermal hydraulics was radiation heat transfer, ballooning of a pressure tube, contact conductance between the pressure tube and a calandria tube and critical heat flux of the calandria tube. Constants and correlations obtained by the tests were incorporated into several codes to assess the core cooling. Long term core cooling capability with the heavy water cooling system was assessed. The core was cooled without melting under the postulated events due to inherent characteristics of the ATR.     

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.     

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.     

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.     

严重事故下正常余热排出系统向堆芯注水策略分析

选取导致堆芯熔化频率最高的始发严重事故--直接注入（DVI）管线断裂事故,以及典型高压熔堆事故--丧失主给水始发事故(LOFW),利用MAAP4程序,分析反应堆堆芯热工水力行为,并对正常余热排出系统（RNS）堆芯注水策略的有效性与负面效应进行评估。分析结果表明,在DVI管线断裂事故和LOFW严重事故序列中,利用RNS进行堆芯注水可有效终止堆芯熔化进程,维持堆芯长期冷却。但堆芯再淹没会产生更多的氢气,存在增加安全壳氢气燃烧风险的可能性。此外通过分析利用严重事故管理导则中辅助计算文件给出的堆芯最小流量实施堆芯注水策略,讨论注水流量对堆芯冷却的影响,结果表明,在实施堆芯注水策略时,建议在系统允许的情况下采用更高的流速进行堆芯冷却。     

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。      

严重事故管理导则入口条件研究

严重事故管理导则的入口是从电厂应急运行规程(EOP)向严重事故管理导则(SAMG)转换的条件,也是严重事故缓解行动的重要依据。本文选取失去四级电源导致的典型高压熔堆序列以及大破口失水事故(LLOCA)导致的典型低压熔堆序列,根据严重事故堆芯剧烈氧化机理,得出了燃料温度、氢气产生速率及产氢量、入口集管过冷度以及慢化剂液位的关系。结果表明入口集管过冷度小于0且持续十几分钟,同时慢化剂系统的状态指示慢化剂液位低于6 900mm,可以作为严重事故管理的入口条件。最后,阐述了目前电厂EOP向SAMG转换的机制,并提出了改进的意见。     

Moderator Analysis of Wolsong Units 2/3/4 for the 35% Reactor Inlet Header Break with a Loss of Emergency Core Cooling Injection

Three-dimensional numerical calculations have been performed for a transient moderator circulation inside the CANDU (Canada Deuterium Uranium) calandria vessel of Wolsong Units 2/3/4. The porous media approach was applied for the core region containing 380 calandria tubes. An anisotropic hydraulic resistance model for the porous media has been developed based on the empirical pressure loss correlations. The selected event was the 35% RIH (Reactor Inlet Header) break with a loss of ECC (Emergency Core Cooling) injection, which has been known to give the largest heat load to the moderator among all the DBA's (Design Basis Accidents). The calculation has been successfully done until 1,200 s after the break, when most of the considerable heat transfer procedure has been completed. During this LOCA (Loss of Coolant Accident) transient, the local subcoolings in the vicinity of any PT/CT (Pressure Tube/Calandria Tube) contact does not drop below the experimentally derived subcooling threshold of 30°C. Because the minimum subcoolings reach only a few degrees to the threshold temperature during the initial 20-40 s, future work on the CANDU moderator circulation needs to be aimed at determining whether this small subcooling margin covers the uncertainty of the moderator analysis.     

压水堆严重事故下封头热斑计算

压水堆堆芯熔化事故情况下,下封头热斑会造成压力容器局部过热,导致临界热流密度发生。利用FLUENT软件对堆芯熔化事故时的下封头热斑进行计算,从流动和换热角度预测热斑导致的下封头薄弱环节。计算结果表明:堆芯熔化事故时,压力容器下封头存在两处最薄弱的位置,分别为下封头正下方正对外部冷却水位置和氧化壳与压力容器交界处。特别是在氧化壳与压力容器交界处,由于多种原因导致临界热流密度发生,使得该处熔化严重。通过设置延伸小管和附加冷却水可延迟压力容器壁面熔穿的时间。     

Simulation of Targets Feeding Pipe Rupture in Wendelstein 7-X Facility Using RELAP5 and COCOSYS Codes

Wendelstein nuclear fusion device W7-X is a stellarator type experimental device, developed by Max Planck Institute of plasma physics. Rupture of one of the 40?mm inner diameter coolant pipes providing water for the divertor targets during the “baking” regime of the facility operation is considered to be the most severe accident in terms of the plasma vessel pressurization. “Baking” regime is the regime of the facility operation during which plasma vessel structures are heated to the temperature acceptable for the plasma ignition in the vessel. This paper presents the model of W7-X cooling system (pumps, valves, pipes, hydro-accumulators, and heat exchangers), developed using thermal–hydraulic state-of-the-art RELAP5 Mod3.3 code, and model of plasma vessel, developed by employing the lumped-parameter code COCOSYS. Using both models the numerical simulation of processes in W7-X cooling system and plasma vessel has been performed. The results of simulation showed, that the automatic valve closure time 1?s is the most acceptable (no water hammer effect occurs) and selected area of the burst disk is sufficient to prevent pressure in the plasma vessel.     

On the thermal analysis of pressure tube/calandria tube contact in CANDU reactors

Under abnormal conditions contact between a pressure tube and the surrounding calandria tube in the core of a CANDU reactor may take place. The resulting temperature field may adversely affect the hydrogen diffusion characteristics in the pressure tube material. This paper is concerned with the thermal aspects of contacting pressure and calandria tubes. A critical review of existing thermal interfacial conductance correlations and their applicability to this problem was carried out. Experiments were also carried out to obtain detailed temperature distribution in the walls of typical pressure and calandria tubes in contact under simulated operating conditions. The thermal fields in both tubes were obtained as functions of the contact pressure and system temperatures. The results showed that the heat flow within the contact area is essentially one-dimensional. The data was used to calculate the interfacial thermal conductance as a function of contact pressure. The results were compared with available interfacial conductance correlations and an assessment of their applicability was accordingly made.     

Dynamic response analysis of tube array in calandria to rotational vault motion

An efficient procedure has been presented for dynamic response analysis of horizontal tube array in partially filled calandria including hydrodynamic interaction effects (S.P. Joshi, A. Goyal, Dynamic response analysis of tube array in partially filled calandria, Earthquake Eng. Struct. Dyn. 28 (1999) 287–309). The procedure is general enough to consider transfer of energy between the fluid-coupled tubes and effects of moderator sloshing on the magnitude and the distribution of hydrodynamic forces. It can also simulate the added damping effects due to hydrodynamic interaction. In this paper, the procedure, originally developed for translational excitations of the calandria vault, is extended to include the rotational excitations of the vault. The procedure uses semi-analytical approach for the evaluation of hydrodynamic terms that leads to considerable economy in the computation.     

基于热胀冷缩原理的TACR压力管与排管间非能动热开关设计

提出了一种可应用于钍基先进CANDU型反应堆(TACR:Thorium-based Advanced CANDU Reactor)压力管与排管间的非能动热开关设计方案.该方案应用金属的热胀冷缩性质,通过热胀冷缩部件推动开关滑块移动来控制压力管与排管间的传热介质种类,以改变压力管与排管之间的热阻.该方案在满足TACR正常运行工况下对压力管和排管间高热阻要求的同时,能够在事故工况下降低二者之间的热阻导出余热.由于利用了金属热胀冷缩性质作为推动力,并利用改变传热介质种类来改变热阻,因此,高度的可靠性和有效性是该方案设计的特点.     

A high-temperature creep model for Zr-2.5 wt% Nb pressure tubes

For a postulated loss-of-coolant accident in a CANDU reactor, in which the primary cooling circuit fails to remove the heat generated in the core, the temperature of the pressure tubes could rise very quickly. Since any deformation of the pressure tubes would control how the core heat is transferred to the surrounding moderator, which is a large heat sink, the accurate prediction of this transient deformation is essential. The majority of the pressure tubes in CANDU reactors are cold-worked Zr-2.5 wt% Nb and creep equations for this material have been developed from uniaxial creep tests. These creep equations were successful in predicting the creep strain in constant-stress uniaxial tests in which the temperature was ramped at rates ranging from 1° C/s to 50° C/s. They also successfully predicted the ballooning of internally pressurized sections of pressure tube that were heated at about 5° C/s.     

Simulation of the small modular reactor severe accident scenario response to SBO using MELCOR code

Advanced small modular reactors (SMRs) use different design in the systems, structures, components from large reactors for achieving a high level of safety and reliability. In present work, the SMRs severe accident caused by the station blackout (SBO) was modeled and analyzed using MELCOR code, and the simulation of the accident scenario response to SBO was conducted. Based on the steady state calculation, which agrees well with designed values, we introduced the SBO accident for transient calculation. First, the case of the SBO accident without the passive core cooling system (PXS) was calculated. The progression and scenario in the reactor pressure vessel (RPV) and the containment were simulated and analyzed, including the transient response, cooling capacity and thermal-hydraulic characteristics and so on. The station black-out transient in the SMR can be simulated accurately, and the main failure model in the accident process can be concluded. Then three other cases of the SBO accident with different passive safety systems (core makeup tank (CMT), accumulator (ACC), passive residual heat removal system heat exchanger (PRHR HX), automatic depressurization system (ADS)) of the PXS were calculated respectively, and the results for different passive safety systems were compared. The passive core cooling system can not only provide water to the primary coolant system, but also take away the reactor decay residual heat. So in a station black-out transient, we can get more time for restoring AC power, and effectively prevent the accidents such as Fukushima.