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.     

The TONUS CFD code for hydrogen risk analysis: Physical models, numerical schemes and validation matrix

The French Atomic Energy Commission (CEA) and the Institute for Radiological Protection and Nuclear Safety (IRSN) are developing a hydrogen risk analysis code, called the TONUS code, which incorporates both lumped-parameter (LP) and computational fluid dynamics (CFD) formulations. In this paper we present the governing equations, numerical strategy and schemes used for the CFD approach.Several benchmark exercises based on experimental results obtained on large-scale facilities, such as MISTRA, TOSQAN and RUT, are presented. They have been used as verification and assessment procedures for modelling and numerical approaches of the code. Specific emphasis is given to the sensitivity analysis of the computed results with respect to numerical and physical parameters. The powerful Design-Of-Experiments technique for sensitivity analysis is successfully applied to the ISP47 MISTRA test case.The TONUS CFD code is presently used to support the hydrogen risk assessment for the European Pressurized Reactor (EPR) plant and to investigate the impact of the two-room concept on hydrogen distribution in the EPR containment.     

Ensuring the long-term functionality of passive auto-catalytic recombiners under operational containment atmosphere conditions—An interdisciplinary investigation

In order to sustain the structural integrity of the containment and other safety relevant components i.e. to avoid a detonation of the hydrogen-air mixture generated during a severe accident in light water reactors, passive auto-catalytic recombiners (PAR) are used for hydrogen removal in many European nuclear power plants (NPP). In 1999, the German NPP Emsland (KKE) was equipped with 58 PAR of AREVA design as an internal accident management measure for a beyond-design accident. Since that time the recombiners are in a stand-by state. As the catalyst elements are exposed to various airborne substances during normal plant operation their function is controlled periodically by testing selected catalyst sheets in a specially designed device. Under the conservative test conditions during this procedure some catalyst sheets showed a delayed responding behavior. First internal analysis gave indication of a beginning fouling on the catalytic surface.The aim of a precautionary investigation performed in cooperation between KKE, Forschungszentrum Juelich and RWTH Aachen University was to characterize the composition of the fouling and to correlate it with potential sources within the containment.In the framework of the investigation the reports of the periodic inspections were analyzed and appropriate sample sheets were selected from the installation. These samples were subjected to a comprehensive chemical surface analysis in order to identify effects like thermal sintering, poisoning or a blocking of the catalytic surface (Baerns, M., 2004. Basic Principles in Applied Catalysis, Springer Verlag). Along with the chemical analysis the catalytic activity of the samples was assessed in several test series in order to correlate the chemically quantified deposition on the catalyst samples with the characteristics of the start-up and the steady-state performance of the recombination reaction. In a final step, possible sources of the fouling were analyzed with regard to their possible contribution to the phenomena. According to the results achieved, measures have been implemented at KKE in order to optimize procedures and to enhance the performance of the PARs.     

严重事故下安全壳卸压箱隔间氢气浓度场模拟

根据MELCOR程序对全厂断电诱发的严重事故下安全壳内各隔间的氢气浓度分布的计算结果,参考美国联邦法规关于氢气控制和风险分析的标准,分析安全壳内氢气的燃烧风险。结果表明：安全壳内平均氢气浓度不会导致整体性氢气燃烧,但存在局部燃烧的风险。通过CFD程序对氢气浓度较高的卸压箱隔间进行氢气释放和空间气体流动过程的模拟,得到更细致的卸压箱隔间内氢气浓度场分布,给出氢气聚集区域的准确位置,为采取严重事故缓解措施,设计氢复合器布置方案提供了参考依据。     

Analysis of hydrogen risk mitigation with passive autocatalytic recombiner system in CPR1000 NPP during a hypothetical station blackout

Hydrogen safety has attracted extensive concern in severe accident analysis especially after the Fukushima accident. In this study, a similar station blackout as happened in Fukushima accident is simulated for CPR1000 nuclear power plant (NPP) model, with the computational fluid dynamic code GASFLOW. The hydrogen risk is analyzed with the assessment of efficiency of passive autocatalytic recombiner (PAR) system. The numerical results show that the CPR1000 containment may be damaged by global flame acceleration (FA) and local detonation caused by hydrogen combustion if no hydrogen mitigation system (HMS) is applied. A new condensation model is developed and validated in this study for the consideration of natural circulation flow pattern and presence of non-condensable gases. The new condensation model is more conservative in hydrogen risk evaluation than the current model in some compartments, giving earlier starting time of deflagration to detonation transition (DDT). The results also indicate that the PAR system installed in CPR1000 could prevent the occurrence of the FA and DDT. Therefore, HMS such as PAR system is suggested to be applied in NPPs to avoid the radioactive leak caused by containment failure.     

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.     

10 MW固态燃料熔盐实验堆安全分析关键技术初步研究

钍基熔盐反应堆(Thorium Molten Salt Reactor,TMSR)项目是中国科学院科技先导项目之一。基于10 MW热功率熔盐反应堆-固体燃料(Thorium Molten Salt Reactor-Solid Fuel,TMSR-SF)的设计,对TMSR的关键技术安全分析进行了初步研究。TMSR-SF与现有反应堆之间的差异对核安全审查提出挑战,TMSR-SF审查方法的研究将准备其安全审查的技术和要求。固态燃料熔盐实验堆安全分析关键技术初步研究包含4个方面:堆芯核设计关键安全限值、事故序列及验收准则、源项及其审评方法和验收准则、概率安全评价方法和始发事件。首先对其它类型反应堆的安全审查方法进行了研究,对其关键参数和重要规定做了概述,并借鉴了高温气体冷堆和钠冷却快堆的审评要求和方法;然后使用蒙特卡罗和其他方法、模型来计算TMSR-SF的关键参数。应用逻辑图方法讨论概率风险评价(Probabilistic Risk Assessment,PRA)方法和始发事件清单。在本研究中,计算了核心核设计安全限值,研究和讨论事故列表和分类,讨论了TMSR-SF的PRA框架和始发事件清单,该研究将支持TMSR-SF的安全审查和安全设计。     

Analyses for VVER-1000/320 reactor for spectrum of break sizes along with SBO

Severe accident analysis of a reactor is an important aspect for evaluation of source term. This in turn helps in emergency planning and severe accident management (SAM). Analyses have been carried out for VVER-1000 (V320) reactor following LOCA along with station blackout (SBO) to generate information on these aspects. Availability and unavailability of hydro-accumulators (HAs) are also considered for this study. Integral code ASTEC V1.3 (jointly developed by IRSN, France, and GRS, Germany) is used for analysing the transients. The predictions of different severe accident parameters like vessel rupture time, hydrogen and corium production and radioactivity release to containment have been compared for a spectrum of break sizes to provide information for probabilistic safety analysis (PSA) level-2 and severe accident management (SAM) guidelines.     

The design and evaluation of a passively cooled containment for a high-rating pressurized water reactor

An integrated pressurized water reactor (PWR) containment was conceptualized that allows heat to be rejected passively to the environment. The proposed containment is based on the demonstrated Ebasco Waterford 3 design. The secondary concrete shell was equipped with inlet and outlet vents that create an air-convection annulus. These vents also permit the submersion of the lower part of the primary containment into an external water pool. An internal water pool located at the bottom of the lower containment was added to increase in-containment heat storage. The performance of the proposed passively cooled containment was evaluated using a subdivided volume code, version 3.4e; the relative novelty of subdivided volume analyses for containment performance evaluation requires experimental verification of principal code predictions. Two experiments were carried out; one to test the performance of the external moat, and one to verify the code’s ability to predict thermal-stratification inside the containment. To improve the subdivided-volume simulation of convection-related parameters, a modeling technique (boundary layer flow approximation) was devised. Finally, the behavior of the proposed containment was evaluated for the worst-case large break loss of coolant accident and the worst-case main steam line break accident. Peak pressures remained below 0.45 MPa during both transients; internal wall pressure differences, equipment qualification temperatures, pressure restoration time also remained below design limits. The mitigation capability of hydrogen recombiners was also evaluated.     

Computational fluid dynamics analysis of spallation target for experimental accelerator-driven transmutation system

One of the key milestones in the roadmap of the European accelerator-driven transmutation system (ADS) is the design and construction of the European experimental ADS (XADS). The window spallation target unit in the lead–bismuth eutectic (LBE) cooled reactor system is one of the basic options considered in the preliminary design study of XADS (PDS-XADS). This paper presents the computational fluid dynamics (CFD) analysis and the main results achieved for this option focusing on the coolability of the window. Steady-state as well as transient behavior, including beam interrupts and three major accident scenarios, has been analyzed using the CFD code CFX 5.6 with an advanced turbulence model. The required boundary conditions were provided by a one-dimensional system code. Based on the CFD analysis, the window geometry was modified in order to achieve sufficient cooling capability of the window under normal operating conditions. The transient behavior of the window temperature under beam trip conditions shows the importance of the beam interrupt duration to the thermal stress load of the window structural material. Further transient analysis of three major accidental scenarios, i.e., beam focusing, loss of heat sink, and beam intensity jump, indicates that the beam focusing accident gives the most serious safety concern. In this case, window failure occurs in less than 1 s after the start of the beam focusing.     

Evaluation of computational fluid dynamic methods for reactor safety analysis (ECORA)

The objective of the ECORA project is the evaluation of computational fluid dynamics (CFD) software for reactor safety applications, resulting in best practice guidelines (BPG) for an efficient use of CFD for reactor safety problems. The project schedule is as follows: (i) establishment of BPGs for use of CFD codes, for judgement of CFD calculations and for assessment of experimental data; (ii) assessment of CFD simulations for three-dimensional flows in LWR primary systems and containments; (iii) quality-controlled CFD simulations for selected UPTF and SETH PANDA test cases; and (iv) demonstration of CFD code customisation for PTS analysis by implementation and validation of improved turbulence and two-phase flow models.The project started in October 2001 and is for a period of 36 months. The project consortium consists of 12 partners combining thermal-hydraulic experts, code developers, safety experts and engineers from nuclear industry and research organizations. At mid-term, the following results were achieved: (i) BPGs are available for simulations of reactor safety relevant flows. These BPGs have found interest in the European projects FLOMIX-R, ASTAR and ITEM; (ii) important flow phenomena for PTS and containment flows have been identified; (iii) experimental data featuring these phenomena have been selected and described in a standardised manner suitable for simulation with CFD methods; (iii) surveys of existing CFD calculations and experimental data for containment and primary loop flows have been performed and documented; (iv) first results for simulations of PTS-relevant single-phase and two-phase flow cases are available.Documentation is available via the internet at http://domino.grs.de/ecora/ecora.nsf. The models developed within the project are implemented in industrial and commercial CFD software packages and are therefore accessible by industry and research institutions.     

百万千瓦级压水堆严重事故下局部隔间氢气风险分析

核安全法规要求控制严重事故下核电厂安全壳内的氢气浓度。除安全壳整体外,局部隔间的氢气浓度同样是关注的重点。本文采用一体化严重事故分析程序对百万千瓦级压水堆核电厂安全壳局部隔间进行建模,分析了不同事故下的氢气风险。结果表明,严重事故下部分隔间短时间内可能存在燃烧风险。本文对降低燃烧风险的方法进行分析计算和筛选,得出的结论可以为安全壳隔间的设计优化提供参考依据。     

大功率非能动压水堆严重事故工况堆芯熔毁进程研究

应用MELCOR 2.1程序,建立了大功率非能动反应堆主要回路、非能动安全系统及安全壳的热工水力模型,并以热段小破口叠加ADS 1阀门失效和内置换料水箱失效触发严重事故为研究对象,对事故进程进行模拟,对堆芯熔毁进程进行了分析。分析结果表明:1)锆合金和不锈钢氧化释热功率在蒸汽充足的情况下高于燃料的衰变功率,将加速堆芯的恶化;2)约13.1%的不锈钢和27.1%的锆合金被氧化,共产生550.99kg氢气;3)堆芯构件的熔化主要依赖于材料自身的熔点和有无构件支撑,堆芯支撑板能够延缓熔融物跌落进入下封头的进程;4)熔池形成后若外部冷却的不足将很快导致下封头应力失效。     

Semi-analytical modeling of a PWR lower head failure under severe accident conditions using an axisymetrical shell theory

The USNRC/SNL OLHF program was carried out within the framework of an OECD project. This program consisted of four one-fifth scale experiments of a reactor pressure vessel (RPV) lower head failure (LHF) under well controlled internal pressure and large throughwall temperature differentials; the objectives were to characterize the mode, timing and size of a possible PWR lower head failure in the event of a core meltdown accident. These experiments should also lead to a better understanding of the mechanical behavior of the reactor vessel lower head, which is of importance both in severe accident assessment and the definition of accident mitigation strategies. A well-characterized failure of the lower head is of prime importance for the evaluation of the quantity of core material that can escape into the containment, since this defines the initial conditions for all ex-vessel events. A large quantity of escaping corium may lead to direct heating of the containment or ex-vessel steam explosion. These are important issues due to their potential to cause early containment failure. The experiments also provide data for model development and validation. For our part, as one of the program partners, a 2D semi-analytical model has been developed and used to simulate these experiments. The aim of this effort is to develop a simplified but well predicting code that can be then implemented in European integral severe accident computer codes (ASTEC, ICARE/CATHARE). This paper presents the detailed mathematical formulation of this simplified method which is used to interpret the experimental results. The axi-symmetric shell theory under internal pressure proposed by Timoshenko has been utilised. The solution to the equilibrium equations is presented, with particular attention to the Rabotnov analytical formula. The radius and the polar angle of the deformed structure have been written as analytical expressions in order to take the large displacements and large strains into account using our mathematical formulation. The Norton type creep law and the Kachanov damage law have been used. Several failure criteria were used in the calculations and their effect on the numerical results is discussed. This 2D semi-analytical model gives very satisfactory results when compared, with the experimental and numerical results that were presented recently in the Benchmark calculations based on the first test of the OLHF program. The performance of this model is also illustrated by its capacity to accurately simulate the deformation of the lower head, including the variation of wall thickness.     

Comparison of hydrogen generation for TVSM and TVSA fuel assemblies for water water energy reactor (VVER)-1000

This paper presents the results received during investigation of hydrogen generation for both types fuel assemblies—the old modernistic type of fuel assemblies (TVSM) and recently installed new one alternative type of fuel assemblies (TVSA) in case of severe accident. There are some differences between both types FAs. They have different geometry as well as different burnable poisons. To investigate behavior of new fuel assemblies during the severe conditions it have been performed comparison of fuel behavior of old type TVSM fuel assembly to new one TVSA.To perform this investigation it has been used MELCOR “input model” for Kozloduy Nuclear Power Plant (KNPP) VVER 1000. The model was developed by Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS) for investigation of severe accident scenarios and Probabilistic Safety Analyses (PSA) level 2. The model provides a significant analytical capability for the Bulgarian technical specialists, working in the field of the NPP safety, for analysis of core and containment damaged states and the estimation of radionuclides release outside fuel cladding.It was accepted criteria for vessel integrity about hydrogen concentration to be 8%. This criterion was based on the decision of RSK (Germany commission for reactor safety).Generally based on the received results it was made conclusion that using both types of fuel assemblies it was not disturbance safety conditions of NPP.     

Effect of spray on performance of the hydrogen mitigation system during LB-LOCA for CPR1000 NPP

During the course of the hypothetical large break loss-of-coolant accident (LB-LOCA) in a nuclear power plant (NPP), hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel (RPV). It is then ejected from the break into the containment along with a large amount of steam. Management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen mitigation system (HMS) and spray system in CPR1000 NPP. The computational fluid dynamics (CFD) code GASFLOW is utilized in this study to analyze the spray effect on the performance of HMS during LB-LOCA. Results show that as a kind of HMS, deliberate igniter system (DIS) could initiate hydrogen combustion immediately after the flammability limit of the gas mixture has been reached. However, it will increase the temperature and pressure drastically. Operating the DIS under spray condition could result in hydrogen combustion being suppressed by suspended droplets inside the containment. Furthermore, the droplets could also mitigate local the temperature rise. Operation of a PAR system, another kind of HMS, consumes hydrogen steadily with a lower recombination rate which is not affected noticeably by the spray system. Numerical results indicate that the dual concept, namely the integrated application of DIS and PAR systems, is a constructive improvement for hydrogen safety under spray condition during LB-LOCA.     

Accident management following loss of residual heat removal during mid-loop operation in a Westinghouse two-loop PWR

The possibility of an accident or component failure during mid-loop operation has been identified in probabilistic studies as a major contributor to core melt frequency and source term risk during shutdown conditions. The wide range of plant states encountered and the unavailability of certain safety features make it difficult to guarantee that safety systems operation will always be sufficient to terminate the accident evolution. In this context analyses are performed using MELCOR 1.8.5 for loss of residual heat removal (RHR) at various times during mid-loop operation of a Westinghouse two-loop PWR. In the absence of recovery of RHR or other accident management (AM) measures, the sequences necessarily lead to a long term core uncovery, heat-up and degradation, loss of geometry and eventual failure of the reactor pressure vessel (RPV). The results show an extensive time window before uncovery and additionally before core damage, which increase progressively with increasing time after shutdown at which loss of RHR occurs. Significant oxidation of the cladding may result in concentrations of hydrogen sufficient for deflagration. The slow evolution implies an opportunity for the plant operators to initiate AM measures even after core uncovery has started. The analyses indicate a substantial time window during the uncovery within which the injection can recover the core without damage. The upper end of the window is determined by the temperature at which heat from cladding oxidation becomes a dominant factor, marking a critical point for the effectiveness of this recovery mode. The results provide confidence in the inherent robustness of the plant with respect to accident sequences of this type.     

Application of a Zircaloy/Steam Oxidation Model to a CFD Code and its Validation against a CANDU Fuel Channel Experiment: CS28-2

Oxidation of a Zircaloy cladding exposed to high-temperature steam is an important phenomenon in the safety analysis of CANDU reactors during a postulated loss-of-coolant accident (LOCA), since a Zircaloy/steam reaction is highly exothermic and results in hydrogen production. As part of a computational fluid dynamics (CFD) simulation of the CS28-2 high-temperature experiment for this accident analysis, two Zircaloy/steam reaction models based on a parabolic rate law are implemented in a commercial CFD code (CFX-10) through a user FORTRAN. It is confirmed that the present oxidation models for the CFX-10 reproduce the results of each empirical correlation in the verification tests well. Then the CFX-10 predictions of a temperature rise and hydrogen production due to Zircaloy/steam oxidation are compared with the results of the CS28-2 experiment. From these validation processes, it is shown that the Urbanic-Heidrick model, which is widely used in CANDU fuel channel codes, is also applicable to a CFX-10 simulation of Zircaloy/steam oxidation in a CANDU fuel channel.     

Fluid mixing and flow distribution in a primary circuit of a nuclear pressurized water reactor—Validation of CFD codes

The EU project FLOMIX-R was aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity.This report will focus on the computational fluid dynamics (CFD) code validation. Best practice guidelines (BPG) were applied in all CFD work when choosing computational grid, time step, turbulence models, modelling of internal geometry, boundary conditions, numerical schemes and convergence criteria. The strategy of code validation based on the BPG and a matrix of CFD code validation calculations have been elaborated. CFD calculations have been accomplished for selected experiments with two different CFD codes (CFX, FLUENT). The matrix of benchmark cases contains slug mixing tests simulating the start-up of the first main circulation pump which have been performed with three 1:5 scaled facilities: the Rossendorf coolant mixing model ROCOM, the Vattenfall test facility and a metal mock-up of a VVER-1000 type reactor. Before studying mixing in transients, ROCOM test cases with steady-state flow conditions were considered. Considering buoyancy driven mixing, experimental results on mixing of fluids with density differences obtained at ROCOM and the FORTUM PTS test facility were compared with calculations. Methods for a quantitative comparison between the calculated and measured mixing scalar distributions have been elaborated and applied. Based on the “best practice CFD solutions”, conclusions on the applicability of CFD for turbulent mixing problems in PWR were drawn and recommendations on CFD modelling were given. The results of the CFD calculations are mostly in-between the uncertainty bands of the experiments. Although no fully grid-independent numerical solutions could be obtained, it can be concluded about the suitability of applying CFD methods in engineering applications for turbulent mixing in nuclear reactors.