Finite element analyses of a BWR vessel and penetration under severe accident conditions

This study assesses the two-dimensional thermal response of a BWR vessel and drain line penetration to three types of debris bed; primarily metallic, primarily ceramic and metallic and ceramic layered, with sensitivity studies for the most severe case. Structural finite element analysis evaluates vessel elastic, plastic and creep response for two cases which bound the thermal challenge to the vessel.Thermal analysis results indicate that drain line failure does not occur for the case when metallic debris relocates to the lower head; structural analysis predicts that the vessel also remains intact for this case. In cases where ceramic debris relocates to the lower head, drain line temperatures peak near values where failure may occur within several minutes; whereas vessel failure is not predicted for 3.5 to 4.0 hours. Sensitivity studies indicate that large porosity debris or high heat removal rates from the vessel and drain line outer surfaces can preclude failure temperatures from occurring.     

小破口引发的严重事故工况及事故缓解的研究

利用MAAP4程序对方家山核电站进行建模,针对事故后果较为严重的小破口事件进行了计算分析,得到了假设事故下电厂系统的反应以及相应的严重事故现象.对事故中发生的DCH(安全壳直接加热)现象和安全壳失效以及裂变产物向环境的释放进行了分析.随后,本文根据相关的严重事故管理导则和该事故的特点,对缓解该事故的策略进行了研究和计算...     

Inerting and venting of mark II containments in BWR nuclear plants

An analysis of hydrogen control systems corroborates containment inerting as the only way of preventing hydrogen explosions which may jeopardize the integrity of BWR Mark II containments during severe accidents. A severe Large Break LOCA and a severe Stuck Open Relief Valve Accident are simulated by the MARCH 2.0 code to compare the advantages and disadvantages of pre-inerting and post-inerting, with or without venting, in BWR Mark II containments.     

EPR与CPR1000严重事故缓解措施比较

简述了EPR的严重事故缓解措施,包括严重事故专用卸压阀,安全壳内换料水箱(IRWST),可燃气体控制系统,堆芯熔融物捕集、稳定和冷却系统,严重事故下安全壳内热量导出系统,双层安全壳,严重事故专用仪表和控制系统,严重事故下不间断供电系统,严重事故运行策略等,并与CPR1000严重事故缓解措施比较,提出CPR1000严重事故缓解措施改进方向。     

Prevention and mitigation of severe accident developments and recriticalities in advanced fast reactor systems

GEN-IV nuclear systems, especially advanced sodium-cooled fast reactors (SFRs) are on the horizon and a key issue of their success is the promise of a higher and improved safety level. In Europe safety investigations are currently under way e.g. in the collaborative CP-ESFR project of the EU. Both on the prevention and mitigation side significant efforts are invested to fulfill the high safety goals. One route of assurance concentrates on the mitigation or even elimination of specific severe accident routes leading to core disruption and recriticalities. The accident phase with larger disrupted and molten fuel regions is coined the transition phase. A competition between fuel losses and in-pool material motion exists deciding over recriticalities and energetics potentials in this phase. To get a control of the transition phase recriticalities and energetics, ideas have been developed to install dedicated means in the core that enhance and guarantee a sufficient and timely fuel discharge - a controlled material relocation (CMR). Several proposals are under way to accomplish this CMR and especially in Japan significant theoretical and experimental work has been performed. In Europe the path to develop CMR measures was driven in the past by the development of the CAPRA reactors with a high Pu enrichment. In the current paper the status of analyses is described and some new concepts are discussed. The CMR measures are discussed along two accident scenarios, the unprotected loss of flow (ULOF) and the instantaneous blockage accident (TIB).     

Analysis of steam and hydrogen distributions with PAR mitigation in NPP containments

The 3-D-field code, GASFLOW is a joint development of Forschungszentrum Karlsruhe and Los Alamos National Laboratory for the simulation of steam/hydrogen distribution and combustion in complex nuclear reactor containment geometries. GASFLOW gives a solution of the compressible 3-D Navier–Stokes equations and has been validated by analysing experiments that simulate the relevant aspects and integral sequences of such accidents. The 3-D GASFLOW simulations cover significant problem times and define a new state-of-the art in containment simulations that goes beyond the current simulation technique with lumped-parameter models. The newly released and validated version, GASFLOW 2.1 has been applied in mechanistic 3-D analyzes of steam/hydrogen distributions under severe accident conditions with mitigation involving a large number of catalytic recombiners at various locations in two types of PWR containments of German design. This contribution describes the developed 3-D containment models, the applied concept of recombiner positioning, and it discusses the calculated results in relation to the applied source term, which was the same in both containments. The investigated scenario was a hypothetical core melt accident beyond the design limit from a large-break loss of coolant accident (LOCA) at a low release location for steam and hydrogen from a rupture of the surge line to the pressurizer (surge-line LOCA). It covers the in-vessel phase only with 7000 s problem time. The contribution identifies the principal mechanisms that determine the hydrogen mixing in these two containments, and it shows generic differences to similar simulations performed with lumped-parameter codes that represent the containment by control volumes interconnected through 1-D flow paths. The analyzed mitigation concept with catalytic recombiners of the Siemens and NIS type is an effective measure to prevent the formation of burnable mixtures during the ongoing slow deinertization process after the hydrogen release and has recently been applied in backfitting the operational German Konvoi-type PWR plants with passive autocatalytic recombiners (PAR).     

Efficacy assessment of independent severe accident mitigation measures in OPR1000 using MELCOR code

In-vessel and ex-vessel mitigation strategies have been revisited to improve the severe accident management (SAM) for operating nuclear power plants. Because independent mitigation measures tend to produce positive and adverse effects simultaneously, it is necessary to investigate the efficacy of individual measures by means of proper quantification. Thus, in the present study we investigated the overall efficacy of existing SA mitigation strategies prepared for the Optimized Power Reactor 1000 MWe (OPR1000) by means of MELCOR 1.8.6 code. The numerical evaluation showed that the Mitigation-01, feeding water into the steam generators, is the most effective among the other mitigations. In addition, Mitigation-02, reactor coolant system depressurization, could not mitigate the SA sufficiently when applied individually. Among the four ex-vessel mitigation strategies, execution of containment spray was effective in removing most of the aerosol fission product but also intensified hydrogen combustion by increasing the partial hydrogen pressure owing to steam condensation. Mitigation-07, operation of passive autocatalytic recombiners (PARs), could reduce the hydrogen concentration, though the catalytic reaction was predicted to increase the containment pressure. In conclusion, this study suggests that mitigation measures should be carefully selected, and that counteracting measures should be prepared to minimize potential adverse effects.     

Analysis of multidimensional and countercurrent effects in a BWR loss-of-coolant accident

The presence of parallel enclosed channels in a boiling water reactor (BWR) provides opportunities for multiple flow regimes in cocurrent and countercurrent flow under loss-of-coolant accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the steam sector test facility (SSTF), which simulated a full scale 30° sector of a BWR/6 reactor vessel. Both steady-state separate effects tests and integral transients with vessel blowdown and refill were performed. The presence of multidimensional and parallel-channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved.     

Spherical steel containments of pressurized water reactors under accident conditions: Investigation program and first results

The investigations will deal with the mechanical behavior of a free standing spherical containment shell built for the latest type of a German pressurized water reactor. The diameter of the containment shell is 56 m. The wall thickness is 38 mm. The material used is the ferritic steel 15MnNi63.The investigation program includes theoretical as well as experimental activities and concerns four different accidents which are beyond the scope of the common design and licensing practice: containment behavior under quasi-static pressure increase up to containment failure; containment behavior under high transient pressures; containment vibrations due to earthquake loadings (consideration of shell imperfections); containment buckling due to earthquake loadings. First results concerning the containment behavior under quasi-static pressure increase are presented. It turns out that the mechanical failure of the containment shell is controlled by plastic instability. A computer program to describe this problem has been developed and membrane tests to check the computational methods have been carried out.     

Materials chemistry and transport modeling for severe accident analyses in light-water reactors II: Gap processes and heat release

Several consequences of steam starvation of the gas filling the internals of the core of a light-water reactor in the fuel-uncovery phase of a severe accident up to cladding melting are analysed. Emphasis is placed on processes that occur in the H₂-rich gas external to the fuel rod cladding; absorption of oxygen and hydrogen by the cladding; the composition and flow rates of gas in the fuel-cladding gap; and the response of the fuel to these conditions. The transport processes and chemical reactions in the cladding, and the fuel controlled by the behavior of the gas in the gap are modeled for a simple temperature transient characteristic of a severe fuel damage accident in a light-water reactor. Cladding burst is assumed to occur at 1273 K at the midplane elevation of the fuel rod, permitting the gas in the gap to come into contact with that external to the fuel rod. The results of the analysis include the following. Steam ingress is restricted to a few centimeters from the failure site by the gettering action of the metal-water reaction on the cladding inner wall. Hydrogen moves axially into the gap only a few times further than steam by diffusion in the Xe-He mixture. The chief process restricting H₂ ingress is the backflow resulting from thermal expansion of the gas in the fuel rod as the temperature rises. When the protective ZrO₂ scale on the outer surface of the cladding disappears by dissolution in the metal, hydrogen permeation through the cladding wall rapidly replaces the inert gas in the gap with H₂. Hydrogen uptake by the cladding draws gas into the core region from the upper plenum and augments the heat release by the metal-water reaction. Exposure of the fuel to this H₂-rich gas results in minor fuel reduction and accompanying cladding oxidation.     

The severe accident mitigation concept and the design measures for core melt retention of the European Pressurized Reactor (EPR)

For the mitigation of severe accidents, the European Pressurized Water Reactor (EPR) has adopted and improved the defense-in-depth approaches of its predecessors, the French “N4” and the German “Konvoi” plants. Beyond the corresponding evolutionary changes, the EPR includes a new, 4th level of defense-in-depth that is aimed at limiting the consequences of a postulated severe accident with core melting. It involves a strengthening of the confinement function and the avoidance of large early releases. The latter requires the prevention of scenarios and events that can result in high loads on the containment, e.g., a failure of the Reactor Pressure Vessel (RPV) at high internal pressure. This is achieved by dedicated design measures.

The paper gives an short overview of the general concept and the strategies for: primary circuit depressurization, H₂ mitigation and the avoidance of energetic Fuel Coolant Interactions (FCIs). It then describes, in detail, the conceptual solution for the stabilization and long-term cooling of the molten core.

The EPR melt retention strategy supports itself on the use of an ex-vessel core catcher located in a compartment lateral to the pit. The related spatial and functional separation isolates the core catcher from the various loads during RPV failure and, at the same time, avoids risks resulting from an unintended initiation of the system during power operation.

Within the core catcher, the melt will be passively flooded with water from the Internal Refueling Water Storage Tank (IRWST). Due to the effective cooling of the melt from all sides a stable state will be reached within hours and complete solidification of the melt is achieved after a few days. The core catcher can optionally be supplied by the Containment Heat Removal System (CHRS). In this active mode of operation, the water levels inside spreading compartment and reactor pit rise and the pools become subcooled, so further steaming is avoided. This results in a depressurization of the containment in the long-term.     

Evaluation of uncertainties in risk emanating from a number of severe accident issues

A study has been performed to estimate, for a particular pressurized water reactor, the uncertainty in risk associated with a number of key phenomenological issues. A second objective was to distinguish the individual importance of the various issues as contributors to the overall uncertainty in risk. The issues considered touched upon the areas of system behavior, containment loading, containment performance, and fission product source term behavior. It was found that the most important source of uncertainty for the plant in question (Surry) was direct containment heating (i.e., the transfer of heat from the core debris to the containment atmosphere when the debris is ejected at high pressure from the reactor vessel and dispersed throughout the atmosphere). Other significant issues included hydrogen burning, containment failure pressure, aerosol agglomeration uncertainties, the frequency of check valve failures leading to a loss-of-coolant accident (LOCA) outside containment, and the potential for having a LOCA induced by high temperatures in the reactor coolant system.     

Iodine behaviour under LWR accident conditions: Lessons learnt from analyses of the first two Phebus FP tests

The International Phebus Fission Product programme, initiated in 1988 and performed by the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN), investigates through a series of in-pile integral experiments, key phenomena involved in light water reactor (LWR) severe accidents. The tests cover fuel rod degradation and the behaviour of fission products released via the primary coolant circuit into the containment building.The results of the first two tests, called FPT0 and FPT1, carried out under low pressure, in a steam rich atmosphere and using fresh fuel for FPT0 and fuel burned in a reactor at 23 GWdt⁻¹ for FPT1, were immensely challenging, especially with regard to the iodine radiochemistry. Some of the most important observed phenomena with regard to the chemistry of iodine were indeed neither predicted nor pre-calculated, which clearly shows the interest and the need for carrying out integral experiments to study the complex phenomena governing fission product behaviour in a PWR in accident conditions. The three most unexpected results in the iodine behaviour related to early detection during fuel degradation of a weak but significant fraction of volatile iodine in the containment, the key role played by silver rapidly binding iodine to form insoluble AgI in the containment sump and the importance of painted surfaces in the containment atmosphere for the formation of a large quantity of volatile organic iodides.To support the Phebus test interpretation small-scale analytical experiments and computer code analyses were carried out. The former, helping towards a better understanding of overall iodine behaviour, were used to develop or improve models while the latter mainly aimed at identifying relevant key phenomena and at modelling weaknesses. Specific efforts were devoted to exploring the potential origins of the early-detected volatile iodine in the containment building. If a clear explanation has not yet been found, the non-equilibrium chemical processes favoured in the primary coolant circuit and the early radiolytic oxidation of iodides in the condensed water films are at present the most likely explanations. Models that were modified or developed and embodied in the computer codes for organic iodide formation/destruction in the gas phase and Ag–I reactions in the sump lead, in agreement with the Phebus findings respectively to greatly enhanced organic iodide formation kinetics and long term concentration in the containment atmosphere on one hand and, in the conditions of Phebus experiments, to significantly limited molecular iodine volatilisation from the sump in so far as silver was in excess compared to iodine, on the other hand. Organic iodides then quickly gain in importance and become the predominant volatile iodine species at long term.     

Sustainable integration of EU research in severe accident phenomenology and management

In order to optimise the use of the available means and to constitute sustainable research groups in the European Union, the Severe Accident Research NETwork of Excellence (SARNET) has gathered, between 2004 and 2008, 51 organizations representing most of the actors involved in severe accident (SA) research in Europe plus Canada. This project was co-funded by the European Commission (EC) under the 6th Euratom Framework Programme. Its objective was to resolve the most important pending issues for enhancing, in regard of SA, the safety of existing and future nuclear power plants (NPPs).SARNET tackled the fragmentation that existed between the national R&D programmes, in defining common research programmes and developing common computer codes and methodologies for safety assessment. The Joint Programme of Activities consisted in:

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Implementing an advanced communication tool for accessing all project information, fostering exchange of information, and managing documents;

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Harmonizing and re-orienting the research programmes, and defining new ones;

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Analyzing the experimental results provided by research programmes in order to elaborate a common understanding of relevant phenomena;

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Developing the ASTEC code (integral computer code used to predict the NPP behaviour during a postulated SA) by capitalizing in terms of physical models the knowledge produced within SARNET;

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Developing scientific databases, in which the results of research experimental programmes are stored in a common format;

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Developing a common methodology for probabilistic safety assessment of NPPs;

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Developing short courses and writing a text book on severe accidents for students and researchers;

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Promoting personnel mobility amongst various European organizations.

This paper presents the major achievements after four and a half years of operation of the network, in terms of knowledge gained, of improvements of the ASTEC reference code, of dissemination of results and of integration of the research programmes conducted by the various partners.Most initial objectives were reached but the continuation of the SARNET network, co-funded by EC in the 7th Framework Programme (SARNET2 project that started in April 2009 for 4 years), will consolidate the first assets and focus mainly on the highest priority pending issues as determined during the first period. The objective will be also to make the network evolve towards a complete self-sustainability.     

The effects of debris bed's prototypical characteristics on corium coolability in a LWR severe accident

This paper is concerned with coolability assessment of a debris bed formed in fuel coolant interactions (FCIs) during a hypothetical severe accident in a light water reactor (LWR). The focus is placed the potential effect of the bed's prototypical characteristics on its coolability, in terms of (i) porosity range, (ii) multi-dimensionality, (iii) inhomogeneity, (iv) particle morphology, and (v) heat generation method (e.g. volumetric heating vs. local heaters). The analysis results indicate availability of substantial coolability margins compared to previous assessments based on models and experiments using an idealized bed configuration (e.g. 1D homogenous debris layer). Notably, high porosity (up to 70%) of debris beds, obtained in experiments and expected to be the case of prototypical debris beds, could increase the dryout heat flux by 100% and more, depending on particle size, compared with the dryout heat flux predicted for debris beds with traditionally assumed porosity of approximately 40%. Bed inhomogeneity represented by micro-channels in a mini bed is predicted to enhance the dryout heat flux by up to ∼50%, even if the micro-channels occupy only a small volume fraction (e.g., less than 4%) of the bed. The effect of coolant side ingress into a multidimensional bed is predicted to enhance the dryout heat flux by up to 40% for the beds analyzed.     

An integrated structure and scaling methodology for severe accident technical issue resolution: Development of methodology

Scaling has been identified as a particularly important element of the Severe Accident Research Program because of its relevance not only to experimentation, but also to analyses based on code calculations or special models. Recognizing the central importance of severe accident scaling issues, the United States Regulatory Commission implemented a Severe Accident Scaling Methodology (SASM) development program involving a lead laboratory contractor and a Technical Program Group to guide the development and to demonstrate its practicality via a challenging application. The Technical Program Group recognized that the Severe Accident Scaling Methodology was an integral part of a larger structure for technical issue resolution and, therefore, found the need to define and document this larger structure, the Integrated Structure for Technical Issue Resolution (ISTIR). The larger part of the efforts have been devoted to the development and demonstration of the Severe Accident Scaling Methodology, which is Component II of the ISTIR. The ISTIR and the SASM have been tested and demonstrated, by their application to a postulated direct containment heating scenario. The ISTIR objectives and process are summarized in this paper, as is its demonstration associated directly with the SASM. The objectives, processes and demonstration for the SASM are also summarized in the paper. The full body of work is referenced.     

The development and demonstration of integrated models for the evaluation of severe accident management strategies—SAMEM

This study is concerned with the further development of integrated models for the assessment of existing and potential severe accident management (SAM) measures. This paper provides a brief summary of these models, based on Probabilistic Safety Assessment (PSA) methods and the Risk Oriented Accident Analysis Methodology (ROAAM) approach, and their application to a number of case studies spanning both preventive and mitigative accident management regimes. In the course of this study it became evident that the starting point to guide the selection of methodology and any further improvement is the intended application. Accordingly, such features as the type and area of application and the confidence requirement are addressed in this project. The application of an integrated ROAAM approach led to the implementation, at the Loviisa NPP, of a hydrogen mitigation strategy, which requires substantial plant modifications. A revised level 2 PSA model was applied to the Sizewell B NPP to assess the feasibility of the in-vessel retention strategy. Similarly the application of PSA based models was extended to the Barseback and Ringhals 2 NPPs to improve the emergency operating procedures, notably actions related to manual operations. A human reliability analysis based on the Human Cognitive Reliability (HCR) and Technique For Human Error Rate (THERP) models was applied to a case study addressing secondary and primary bleed and feed procedures. Some aspects pertinent to the quantification of severe accident phenomena were further examined in this project. A comparison of the applications of PSA based approach and ROAAM to two severe accident issues, viz hydrogen combustion and in-vessel retention, was made. A general conclusion is that there is no requirement for further major development of the PSA and ROAAM methodologies in the modelling of SAM strategies for a variety of applications as far as the technical aspects are concerned. As is demonstrated in this project, the generic modelling framework was refined to enable a number of applications. Some recommendations have also been made regarding the applicability of these approaches to existing operating reactors and future reactors. The need for further research and development in the area of human reliability quantification was identified.     

LWR severe accident simulation: synthesis of the results and interpretation of the first Phebus FP experiment FPT0

The international Phebus Fission Product (FP) programme, initiated in 1988 and performed by the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN), investigates key phenomena of severe water reactor accidents. Six in-pile experiments were planned. Four have been successfully performed in 1993, 1996, 1999 and 2000.The first experiment, called FPT0, used uranium dioxide fuel of 4.5% enrichment in-situ irradiated for 9 days to a burn-up of 230 MWd t⁻¹. It was designed to reach significant fuel melting and to study low pressure fission products release and transport through the primary cooling system including a non-condensing steam generator and into the containment vessel. As the first test of the programme, FPT0 was intended to demonstrate the adequacy of the new, complex Phebus facility to simulate the anticipated phenomena and was the first attempt in using the new experimental results for verifying codes.The scientific results from FPT0 were sufficiently challenging that they deserve to be documented and interpreted. Since some of them did not correspond to the predicted and pre-calculated behaviour, the post-test analysis and interpretation period was rather long. Three years later, the second experiment FPT1, rather similar in its boundary conditions but using a fuel burned in a reactor (23 GWd t⁻¹), confirmed certain FPT0 results, helping in their final interpretation and removing doubts about possible fundamental shortcomings of the Phebus facility. More detailed experimental results of the test are available in the final test report deliverable on CD-ROM support. It can be obtained upon request from IRSN.¹This report retraces the history of FPT0 and its general programme context, and briefly describes the layout of the facility, supporting separate effect tests and computational tools. It then presents the synthesis of the results and of the international understanding reached concerning their interpretation, with emphasis on fuel and fission product behaviour.Finally, conclusions are presented about the impact of FPT0 on severe accident modelling with implications on source term evaluation and on accident prevention and mitigation studies.     

Advanced treatment of zircaloy cladding high-temperature oxidation in severe accident code calculations: Part I. Experimental database and basic modeling

Progress in the treatment of zirconium oxidation kinetics in different severe accident (SA) codes and convergence towards an agreed data base are required for the reliable verification of sophisticated LWR core degradation models.Focused on the comprehensive experimental studies the available information on the high-temperature oxidation kinetics of zircaloy (Zry) is evaluated. Important discrepancies between results for the high-temperature range are interpreted in terms of different experimental and evaluation procedures. The critical assessment identifies the following items, which require separate consideration in the simplified convention of using fixed reaction rate correlations for the high and low temperature ranges: the co-existence of two oxide phase sub-layers gives rise to a transitional kinetic response in an intermediate temperature range. Towards higher temperatures the validity of the correlations approach (assuming reaction rate control as in a semi-infinite solid-state diffusion system) is restricted further, on the one hand by kinetic control within the gas phase (initial oxidation range), on the other hand by metal matrix consumption (final oxidation range). The analytical treatment of the two last-mentioned concomitant phenomena is given, and the consideration allows to reconcile apparent discrepancies between experimental results. A reasonable base of combined data is thus identified, for a moment with certain uncertainty, which is not suitable for direct application.In Part II, the kinetic base will be further verified and refined by statistical evaluation of complementary experimental data and transferred to the form, allowing implementation in codes. In Part III, the satisfactory application in ICARE2 code calculations of separate-effect and bundle experiments will be presented.