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.     

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.     

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.     

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).     

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。     

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.     

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.     

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.     

CANDU堆核电厂全厂断电始发严重事故进程研究

采用一体化分析程序建立了适用于CANDU堆核电厂的严重事故分析模型。该模型主要包括热传输系统、慢化剂系统、端屏蔽系统、蒸汽发生器二次侧系统等。针对全厂断电始发的严重事故进行了相应的热工水力现象分析,得知慢化剂系统和端屏蔽系统内的大量水使事故进程大幅推迟。同时,对重要时间进程与ISAAC2.0程序结果进行了初步比对,两者的结果基本吻合。分析结果可为开展重水堆严重事故现象及缓解措施研究提供技术参考。     

Generic CANDU 6 plant severe accident analysis employing SCAPSIM/RELAP5 code

The purpose of the present study is to assess the capability of SCDAPSIM/RELAP5 to perform the deterministic analysis for postulated severe accidents for CANDU plant and to gain information for potential improvements in code modelling. SCDAPSIM/RELAP5 is a widespread and detailed computer code for severe accident analysis that can be adapted to benchmark the CANDU dedicated tools, MAAP4–CANDU and ISAAC. Simulations of station blackout (SBO) and large loss-of-coolant accident (LOCA) scenarios, which, through further system failures, may eventually lead to severe core damage (SCD) accident in a CANDU 6, are presented. The paper provides details concerning the methodology and nodalization used, and interprets the results obtained. Comparisons of the SCDAPSIM/RELAP5 simulations with the MAAP4–CANDU code reported results are presented. Also, some insights are given on possible reasons for the discrepancies between the SCDAPSIM/RELAP5 and MAAP4–CANDU code predictions.     

Methodology for developing channel disassembly criteria under severe accident conditions for PHWRs

This paper presents a methodology to develop a model for disassembly of the coolant channels in Pressurized Heavy Water Reactors under severe accident conditions. This model gives criteria to decide when under severe accident condition coolant channels will rupture due to deterioration in material properties at high temperatures and increase in load due to creep sag of channels above it and hence get disassembled. Presently available severe accident codes use simplistic and optimistic criteria based on a predefined temperature to predict failure of fuel channels and an explicit criterion for disassembly of the channel is not covered. The coolant channel disassembly model developed in this paper is based on modeling the sag and pile up of channels. A uniform temperature along the length of the channel is assumed. The disassembly of the channel is assumed when the total strain at any location exceeds the failure strain for a given temperature. A 3D failure surface which is a plot of time to failure, temperature of the calandria tube and load on the calandria tubes (on account of no of channels piled up) is developed. This failure surface can be used as an input to severe accident codes to predict the progress of the core disassembly. A set of failure surfaces is recommended to be used if metal–water reaction on the outer surface is to be accounted for loss in ductility due to metal water reaction. The temperature transient of the calandria tube for a severe accident obtained from system thermal hydraulic codes can be mapped onto the failure surface. The time at which the mapped transient crosses the failure surface gives the time at which the calandria tube is disassembled. This disassembly model is an engineered model which is much more realistic as compared to the current temperature based conservative model for predicting severe accident progression.     

Approach to accident management in RBMK-1500

In order to ensure the safe operation of the nuclear power plants accident management programs are being developed around the world. These accident management programs cover the whole spectrum of accidents, including severe accidents. A lot of work is done to investigate the severe accident phenomena and implement severe accident management in NPPs with vessel-type reactors, while less attention is paid to channel-type reactors CANDU and RBMK.Ignalina NPP with RBMK-1500 reactor has implemented symptom based emergency operation procedures, which cover management of accidents until the core damage and do not extend to core damage region. In order to ensure coverage of the whole spectrum of accidents and meet the requirements of IAEA the severe accident management guidelines have to be developed.This paper presents the basic principles and approach to management of beyond design basis accidents at Ignalina NPP. In general, this approach could be applied to NPPs with RBMK-1000 reactors that are available in Russia, but the design differences should be taken into account.     

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.     

Development of severe accident management guidance (SAMG) for the Canadian CANDU 6 nuclear power plants

This paper discusses the severe accident management guidance (SAMG) development process undertaken for the Canadian CANDU 6 nuclear power plants (NPPs); the customization process of the generic CANDU SAMG for the Point Lepreau NPP is presented. Examples of severe accident management (SAM) guidelines related to containment pressure control are included in this paper. This paper also provides an overview summary of the severe accident analysis program at Atomic Energy of Canada Limited (AECL) that complements the SAM guidelines development process for the CANDU 6 NPPs in Canada. These analyses provided insights into the accident progression and basis to develop the SAM guidelines.     

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.     

Uncertainty analysis of CsI distribution for typical CANDU6 station blackout accident

This paper illustrates an application of a severe accident analysis code, ISAAC (Integrated Severe Accident Analysis Code for the CANDU plants), to the uncertainty analysis of fission product behaviors during a severe reactor accident. The ISAAC code is a system-level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, and whose main purpose is to support a level 2 probabilistic safety assessment or severe accident management strategy developments. The code employs lots of user options for supporting sensitivity and uncertainty analyses. The present application is mainly focused on determining an estimate of the fission products in the release and transport processes and the relative importance of the dominant contributors to the predicted fission products. The key modeling parameters and phenomenological models employed for the present uncertainty analysis are closely related to the fission product release correlations, vapor–aerosol equilibrium, vapor–surface equilibrium for a revaporization calculation, and aerosol decontamination factors. A typical CANDU6 type plant, the Wolsong nuclear power plant, was used as a reference plant for the analysis.     

Overview of high-temperature fuel behaviour with relevance to CANDU fuel

This paper provides an overview of high-temperature phenomena in nuclear fuel elements and bundles, with particular relevance to the CANDU fuel design. The paper describes heat generation, fuel thermal response, and thermophysical properties of the fuel and sheath that can affect the thermal and mechanical response of the fuel element. Sources of chemical heat that can arise during accident conditions in the fuel element are also detailed. Specific phenomena associated with fuel restructuring, fuel sheath deformation, fuel-to-sheath heat transfer, fuel sheath failure criteria, oxidation, hydriding and embrittlement of the Zircaloy sheath, gap transport processes in failed elements, fuel/sheath interaction and fuel dissolution by molten cladding are detailed as important phenomena that can impact reactor safety analysis. Fuel behaviour during a power pulse and fuel bundle behaviour that occurs during a severe reactor accident are further considered. The review also points out areas of further research that are needed for a more complete understanding.     

TACR排管容器研究中多孔介质法的应用

以钍基先进重水堆(TACR:Thorium Based Advanced CANDU Reactor) 慢化剂系统排管容器作为研究对象,提出了一种适用于计算排管容器内流场和温度场的非结构网格多孔介质方法,并将其计算结果与精细数值计算法所获得的结果进行了比较,二者符合较好,由此验证了其方法与计算结果的正确性.     

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.     

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.