Severe Accident Management Research Articles

Key elements for proper development and implementation of severe accident management guidance/guidelines – SAMG

Severe Accident Management Guidelines (SAMG) have been developed to support a nuclear power plant (NPP) in case of a severe accident, i.e. an (unlikely) accident where the core and/or the fuel in the Spent Fuel Pool (SFP) have been damaged and there is risk for radioactive releases. Various approaches exist, such as the SAMG developed by the PWR and BWR Owners Groups (PWROG, BWROG), by Electricité de France (EdF) for its fleet of plants, the Candu Owners Group (COG) and by several individual NPPs. Often, the developers of SAMG develop a generic type of SAMG, applicable for a certain type of NPP, and leave it to the individual plants to adapt these to their configuration and Emergency Response Organisation (ERO). That means, that quite some burden remains for the individual plants to finally implement the SAMG. This implementation includes not only technical items, such as defining proper strategies, but also many organisational issues, as handling a severe accident needs an effective ERO, in which the SAMG must be well embedded.Various plants have been reviewed, both the ways they have constructed their plant-specific SAMG, implemented them in their ERO and tested the performance in drills and exercises. Most recently, a review of SAMG at Swedish plants has been performed, after which it was decided to pull the various experiences together and summarize them in one compendium. Some of these also include aspects of the development of the generic SAMG.The result is a number of recommendations (∼30), both on the technical field as well as regarding the ERO. We could also define a number of good practices that deserve wider attention.A few words have been devoted to international support for a stricken plant (Epilogue).

Development of a surrogate model for quenching estimation of ex-vessel debris beds and its coupling with MELCOR

In the severe accident management (SAM) strategy for Nordic boiling water reactors (BWRs), a flooded reactor cavity is conceived to receive corium in case of vessel failure, with the hope that the discharged corium will fragment and form a coolable particulate debris bed in the deep water pool. The so-formed debris bed on the cavity basement is supposed to be very hot at the beginning and therefore its quenching is a prerequisite for long-term coolability. In previous study the coupled MELCOR/COCOMO simulation was employed to simulate quench process of ex-vessel debris beds in severe accident scenarios. Although it successfully extended the MELCOR capability, the calculation was dramatically slowed down by explosive computational cost of COCOMO. To overcome the limitation, the present study is to develop a surrogate model (SM) which can replace the mechanical code COCOMO and realize quick estimations of the quench process of ex-vessel debris beds. It was then coupled with MELCOR code for integral severe accident analyses of a Nordic BWR with cooling of ex-vessel debris beds. The SM was developed based on a database generated from COCOMO calculations of various one-dimension (1D) debris beds quenched in the reactor cavity, using artificial neural networks (ANNs). Finally, the coupled MELCOR/SM simulation was applied to safety analyses of postulated severe accident scenarios due to station blackout (SBO) in the BWR, where MELCOR performs integral analysis of accident progression while SM predicts the consequences (e.g. energy transfer) of debris bed quench. The simulation results show that the coupled MELCOR/SM simulation can predict the trends of containment pressure and pool temperature similar to those of the coupled MELCOR/COCOMO simulation. Compared with MELCOR standalone simulation, the coupled MELCOR/SM simulation predicted earlier pool saturation and containment venting.

A scoping study on remelting process of a debris bed in the lower head of reactor pressure vessel

Coolability and retention of core melt (corium) in the lower head of reactor pressure vessel (RPV) has been accepted as a severe accident management strategy to maintain the reactor pressure vessel (RPV) integrity of a light water reactor. To qualify the in-vessel melt retention strategy, lots of studies have been performed to investigate the natural convection heat transfer of a melt pool in the lower head. However, little work has been attributed to the precursory phase of the melt pool, i.e., the remelting process of a debris bed which is formed in the lower head at the very beginning of corium relocation from the core the lower head. The present study is motivated to conduct an experimental study on the debris remelting process. For this purpose, a dedicated test facility named COREM (COrium REMelting) is conceived and constructed, which features internal heating of electromagnetic induction and visualization of debris remelting dynamics. The test section is a semicircular vessel representing a slice of scaled-down RPV lower head, whose front and back walls are made of transparent tempered glass which facilitate visualization and induction heating of the debris bed. Fiber probes with multiple optical temperature sensors are mounted in the semicircular wall of the test section to measure its temperature distribution. In the scoping test, n-octanol and Wood's metal are selected as the simulant materials of metallic and oxidic components of corium, respectively, and their particles ware loaded in the test section to form a debris bed. This paper presents the first two scoping tests which have been performed with the COREM facility so far, under different mixing ratios of two debris materials. The measured data include the photography of debris remelting processes and the temperatures in the debris bed and the semicircular wall. Based on the temperature distributions, heat flux profile along the semicircular vessel wall is also estimated. The scoping tests well reproduce the dynamic process of debris remelting, with two distinct stages of fusion of n-octanol and Wood’s metal, i.e., melting successively from low to high melting-point debris particles. During the remelting process, the vessel wall temperatures increase with the polar angle firstly and then decrease gradually, leading to the highest temperature appearing in the middle of the lower layer of the stratified molten pool which is finally formed.

MELCOR-To-MELCOR Coupling Method in Severe Accident Analysis Involving Core and Spent Fuel Pool

A lot of effort has been spent to prevent the occurrence of SA in nuclear plant and to develop Severe Accidents (SA) Management to mitigate the consequences of a SA. Those consequences are mainly related to limit the release of fission product to the environment. The core in the vessel is not the only source of fission products as the Spent Fuel Pool (SFP) hosting the fuel removed by the core is, in some NPP, inside the containment and SA conditions can also occur. This is especially important in reactors having proximity between the RPV and SFP such as the VVER-1200. This close proximity implies that any SA occurring in the SFP potentially affects the RPV and vice-versa. This potential combination might cause unexpected evolution in the SA progression to whom the safety systems are not able to contain. MELCOR code is a widely used, flexible powerful SA code but it is incapable (due to the uniqueness of the COR package use inside the same input) to reproduce a situation in which both the fuel in vessel core and the fuel in the SFP, inside the same containment, are going to experience a severe accident scenario. The current study presents a MELCOR-to-MELCOR coupling method to simulate simultaneously scenarios with both, core and SFP, as sources capable of H2 generation, fuel damage and FP release in a VVER-1200 NPP. The coupling is performed by running two simulations in parallel and with the data exchange supervised and managed by a dedicated Python coupling supervising script developed at NINE.

Backfitting of Dukovany NPP for design extension conditions

The contribution deals with the backfitting of Dukovany NPP for design extension conditions with emphasis on severe accidents. This NPP, located in Czech Republic, is equipped with 4 units of VVER-440/V213 reactors. During the last decades significant attention was paid to backfitting of these units. New safety systems aimed at mitigating the consequences of severe accidents have been designed and have been (are being) implemented. At first, special features of VVER-440/V213 reactor and containment equipped with bubbler condenser pressure suppression system are briefly described. The systematic process of plant backfitting for design extension condition is summarized further. Attention is paid to description of crucial severe accident management strategies that have been adopted and design and implementation of new safety systems that are necessary to fulfil these strategies. Than the requirements of Czech regulatory body (SUJB) set for practical elimination of early or large radioactive releases are discussed. Finally, the performances of severe accident management strategies are illustrated using the specific case of blackout scenario. For this purpose, an analysis was performed using ASTEC code developed by the French IRSN. The results of analysis are briefly described and discussed further. From the obtained results it follows that the high-pressure melting scenario is reliably avoided, corium is localised and stabilised inside the reactor vessel and the hydrogen in containment atmosphere is safely controlled by passive autocatalytic recombiners. Thus, stable plant state with preserved containment integrity is reached. Source term that has been estimated under these conditions indicates that the predicted releases are well below the criteria set by SUJB for early or large radioactive releases, and therefore long-term contamination of the environment is avoided.

Uncertainty quantification for a severe accident sequence in a SFP in the frame of the H-2020 project MUSA: First outcomes

The Management and Uncertainties of Severe Accidents (MUSA) project, funded in HORIZON 2020 and coordinated by CIEMAT (Spain), aims at consolidating a harmonized approach for the analysis of uncertainties and sensitivities associated with Severe Accidents (SAs) focusing on Source Term (ST). In this framework, the objectives of the Innovative Management of Spent Fuel Pool Accidents (IMSFP – WP6), led by IRSN (France), are to quantify and rank the uncertainties affecting accident analyses in a Spent Fuel Pool (SFP), to review existing and contemplated SA management measures and systems and to assess their possible benefits in terms of reduction of radiological consequences.To quantify the propagation of the uncertainties of the input parameters to the output uncertainties of severe accident codes (ASTEC, MELCOR, RELAP/SCDAP), a diverse set of uncertainty quantification (UQ) tools (DAKOTA, RAVEN, SUNSET, SUSA) are used. The statistical framework used by the different UQ-tools is similar e.g. pure random (Monte Carlo) and Latin hypercube sampling (LHS).Fourteen partners from three different world regions are involved in the WP6 activities. The target of this paper is to describe the achievements during the first three years of the project. In a first part, a description is given of the SFP accidental scenario, of the key target variables and radionuclides chosen as ST Figures of Merit (FoM) and of the identified uncertainty sources in models and input parameters. A key element when defining the SFP scenario has been the consideration (or not) of the reactor building, as it is expected to significantly affect analyses. In a second part, the first insights coming out from the calculation phase of the project are presented. The review of existing SA management measures is also exposed, as well as systems whose benefits will be assessed in the second phase of the project. Finally, challenges that arise from such an exercise are discussed, as well as major difficulties found when applying UQ methodologies to SFP scenarios and solutions adopted.

Influence of guide tubes on configuration and heat-up behaviour for PHWR specific suspended debris bed – An analytical assessment

The estimation of accident source term and hydrogen generation is dependent on the core heat-up under postulated accident scenarios. The uncertainty of such estimations increases as the accident severity increases due to complexity in the determination of i) evolving degraded core configurations, and ii) associated suitable physical models for heat-up and degradation simulation. Analysis of postulated severe accidents for PHWR is essential for emergency planning and severe accident management guidelines from a plant licensing perspective. The study presents the order of uncertainty that is present as of today in deciding the size and shape of the suspended debris bed, which in turn decides the exposed core heat up, hydrogen generation as well as fission product release. The study reveals that the location and integrity of reactivity mechanism guide tubes significantly impact the exposed core heat-up, channel disassembly pattern and shape and size of the suspended debris bed. Nevertheless, the experimental study on the guide tube behaviour on asymmetric heat-up condition and their failure will further substantiate the postulation of suspended debris bed configuration.

MELCOR 2.2-ASTEC V2.2 crosswalk study reproducing SBLOCA and CSBO scenarios in a PWR1000-like reactor part II: Analysis of containment thermal-hydraulics and ex-vessel phenomena

The Accident Source Term Evaluation Code (ASTEC) and Methods of Estimation of Leakages and Consequences of Releases (MELCOR) system codes are both used by Tractebel to perform Severe Accident (SA) analyses for the Belgian Nuclear Power Plants (NPPs).This paper contains the second part of the comparative study between these two integral software focused on the SA progression in the containment after the Vessel Failure (VF).The main phenomena characterize this phase of the accident usually called ex-vessel phase are the Molten Corium Concrete Interaction (MCCI) and ex-vessel corium/debris coolability, which determine the containment Thermal-Hydraulics (TH) response and can cause its late failure. Despite their importance from a safety point of view, chemical and physical processes behind these phenomena are not yet well understood, partly because of their extreme complexity and partly because of the limitations of current experimental capabilities in reproducing them. As result, models developed and implemented in the codes to describe the MCCI and debris coolability are extremely simplified because based on limited data availability.The benchmark study consists of a comparison of the results calculated by the ASTEC and MELCOR codes during the ex-vessel phase following two different scenarios a Complete Station Blackout (CSBO) and a Small Break Loss Of Coolant Accident (SBLOCA).For each scenario, two different Severe Accident Management Strategies (SAMS) are adopted, the first one is to ensure that the corium, once fully relocated in the cavity, has top-flooding conditions throughout the transient, while the second one is to keep the reactor cavity completely dry. Furthermore, for every case reproduced the reference and best-practice models are investigated.The results show that in case of dry cavity conditions, despite the different VF timing calculated by the two codes, the final results provide very similar indications even though some dissimilarity is observed regarding the amount of fission product (FP) released through the Containment Filtering Venting System (CFVS).In contrast, in the case of wet cavity conditions, the accident progression predicted is significantly different regarding number of venting operations, TH containment conditions, and amount of concrete ablated.The discrepancies observed, should be sought in the corium and debris coolability models, which given their parametric nature and oversimplification can lead to these results.

Severe Accident Management at NPP-2006 Reactor Plants (Leningrad NPP-2)

Severe Accident Management at NPP-2006 Reactor Plants (Leningrad NPP-2)

Numerical Analysis of Heating Technique in Corium Melt Pool Convection Flow Field & Thermal Interaction in a Volumetrically Heated Molten Pool

In-Vessel Retention (IVR) is one of the existing strategies of severe accident management of LWR, which intends to stabilize and isolate corium & fission products inside the reactor pressure vessel (RPV) and primary containment structure. Since it has become an important safety objective for nuclear reactors, it is therefore needed to model and evaluate relevant phenomena of IVR strategy in assessing safety of nuclear power reactors. One of the relevant phenomena during accident progression in the oxidic pool is non-uniform high heat generation occurring at large scale. Consequently, direct experimental studies at these scales are not possible. The role computer codes and models are therefore important in order to transpose experimental results to reactor safety applications. In this paper, the state-of-the-art ANSYS FLUENT CFD code is used to simulate Non-uniform heat generation in the lower plenum by the application of Cartridge heating under severe accident conditions to derive the basic accident scenario. However, very few studies have been performed to simulate non-uniform decay heat generation by Cartridge heaters in a pool corresponding lower plenum of power reactor. The current investigation focuses on non-uniform heating in the fluid domain by Cartridge heaters, which has been done using ANSYS FLUENT CFD code by K-epsilon model. The computed results are based on qualitative assessment in the form of temperature and velocity contour and quantitative assessment in terms of temperature and heat flux distribution to assess the impact of heating method on natural convective fluid flow and heat transfer.

Assessment of safety injection in severe accident management following BDBA scenarios in a Swedish PWR

Analytical simulation using best-estimate codes were suggested to be extended for the elaboration and improvement of SAMG in current PWRs. This work is about to perform an assessment work using MELCOR for the effectiveness of safety injection to achieve the PBF strategy in SAM following a BDBA scenario, that is LOCA with concurrent SBO. In the simulations, the safety injection is assumed to be retrieved with the postulated power recovery at different timing during core relocation. The simulation results illustrates that the grace period of preventing vessel failure varies with LOCA break size and locations. The safety injection implemented in grace period is capable of retarding or ceasing the core relocation, sequentially avoiding the massive core relocation into lower plenum, mitigating the hydrogen generation and fission product release from core. Meanwhile, the injection later than grace period would be failed to prevent RPV failure, and it negatively affects hydrogen generation in some scenarios. The results also indicate that the smallest injection capacity of HPSI system in Swedish PWR is sufficient to achieve the effective mitigation.

Effects of elevated temperature steam oxidation and subsequent quenching on light water reactor clad

Following a postulated severe accident, the temperature of fuel pins of light water reactor escalates because of decay heat and devoid of coolant. The presence of steam inside the core causes exothermic oxidation of the cladding surface at elevated temperatures. The in-vessel injection system of severe accident management guidelines injects the cold water inside the hot uncovered core to arrest the escalation of temperature rise and terminate the accident progression. However, the heated clad may be subjected to thermal shock during the quenching process, which inevitably assists in the failure and fragmentation of the severely oxidised fuel pins column.In this paper, the effect of thermal quenching shock on the severely oxidised fuel pins column during the reflooding of the hot uncovered core is studied experimentally. A 2ˣ2 square lattice arrangement of zircaloy-4 clads is used for experiments. First, the clads are oxidised in the presence of steam in the temperature range of 1100–1250 °C, and then the test-section is flooded with water at ambient temperature from its bottom. The violent heat transfer between the clad and cold water has been observed during the reflooding. Furthermore, the clad has failed with a circumferential crack at multiple locations along the clad length during this period; however, longitudinal cracks are not observed on the clad surface. The scanning electron micrographs of the clad samples show a significant oxide layer over the clad surface. At the same time, energy dispersive x-ray spectroscopy reveals significant oxygen diffusion inside the zirconium metal matrix, particularly near the outer tube surface.Additional experiments are performed to study the clad failure mechanism during the reflooding of the severely oxidised clad. The set of experiments reveals that the clad fails under the thermal quench shock (>7.41 °C/s) for severely oxidised cladding (ECR>25%) at an average temperature above 1100 °C. The failed clad is likely to create debris. The average quench front velocity is 0.6 cm/s while the reflood velocity is 1.13 cm/s. Quench velocity below the reflood velocity indicates excessive water evaporation and dispersed type flow ahead of the quench front. The total analytical H2 generated before reflooding is 3.75 g.

Influence of active and passive equipment for advanced pressurized water reactor on thermal hydraulic and source term behavior in severe accidents

Extensive studies have been carried out on the behavior of core degradation and fission products of common pressurized water reactors (PWRs). However, few of them have investigated the relationship between thermal hydraulic and fission product behavior in advanced passive PWRs. Due to the impact of thermal hydraulic behaviors in different accident sequences on the release and transportation of fission products, an integrated severe accident analysis (ISAA) code with highly coupled thermal hydraulic and source term calculations is required to simultaneously analyze thermal hydraulic and source term behavior. For advanced passive PWRs, important safety systems that may affect the behavior of the core and fission products should be considered. It is therefore necessary to simulate the thermal hydraulic and fission product behavior of advanced passive PWRs. In this study, the ISAA code is adopted to simulate the occurrence of a hypothetical double ended cold leg LBLOCA of HPR1000 in three scenarios of equipment failure. The results show that the high-temperature fuel rods and cladding materials exhibit delayed failure at the lower position of the active core, whereas earlier failure at higher position during the reflooding. Active and passive equipment affects fuel temperature, the oxidation conditions of the fuel, the interaction of fission products and structural materials, and the state of the fuel, thereby affecting the release of fission products in the fuel. HPR1000 only relies on passive equipment to relieve the core degradation in severe accidents, realize the in-vessel retention of melt, and eliminate the ex-vessel release possibility of fission product. It is hoped that the results can provide references for HPR1000 to formulate the severe accident management guidelines (SAMG).

CINEMA-SMART development and its application to SBLOCA analysis for Korean small integral reactor

The Korean small integral reactor of SMART (System-integrated Modular Advanced ReacTor) was developed, which contained core, reactor coolant pumps, steam generators and pressurizer within a single reactor pressure vessel. In the code development of severe accident sequence analysis for the SMART, a CINEMA (Code for INtegrated severe accidEnt Management Analysis)-SMART has been developing from an initiation event to a containment failure including fission product behavior in the vessel and the containment. The CINEMA-SMART is composed of CSPACE, SACAP (Severe Accident Containment Analysis Package), and SIRIUS (SImulation of Radioactive nuclide Interaction Under Severe accident), which are capable of core melt progression with thermal hydraulic analysis of the RCS (Reactor Coolant System), severe accident analysis of the containment, and fission product analysis, respectively. The main sequence of a SBLOCA (Small Break Loss Of Coolant Accident) for the SMART has been analyzed using CINEMA-SMART computer code. This analysis is focused on an evaluation of core melt progression and containment thermal hydraulic behavior during a severe accident. The CINEMA results have shown that the integrity of the containment can be maintained during the SBLOCA accident in the SMART.

Assessment of Filtered Containment Venting Strategies for Mitigating Severe Accident Consequences at a Generic Single-Unit CANDU

Abstract In many jurisdictions, filtered containment venting systems are an integral part of severe accident management guidelines but are normally only used as a last resort, to be manually activated when the containment pressure approaches the containment failure limit. This approach fails to utilize the capability of the filtered venting system to better control containment pressure throughout an accident and to create capacity for possible future pressurization based on the accident progression. In this paper, smart venting strategies are explored, using a generic single-unit Canada Deuterium Uranium (CANDU) (660 MWel) plant model in MAAP-CANDU, as a means to mitigate the consequences of a severe accident. These strategies are able to account for both current conditions as well as possible future pressurization. The use of strategic venting offers an improvement in maintaining containment integrity, as well as reducing I-131 releases to the environment by about 10% to 31% compared to last-resort venting while being feasible for manual operator actions. Overall, this comprehensive analysis demonstrates the capability of a smart venting strategy paired with a filtered containment venting system as an active mitigation tool.

Full scope 3D analysis of a VVER-1000 containment pressurization during a LB-LOCA by employing AutoCAD and GOTHIC code

One of the main goals of severe accident management strategies is to maintain containment integrity to prevent radioactive release into the environment. As a result, containment internal loads need to be investigated during a postulated accident. In the present study, short-term containment thermal-hydraulic (TH) parameters of a VVER-1000/V446 reactor are analysed during a LB-LOCA. To achieve this goal, in the first step as-built 3D structure of VVER-1000/V446 containment was modelled in detail by using AutoCAD. The AutoCAD model has been processed to be prepared for GOTHIC 3D input. Meanwhile, an equivalent GOTHIC lumped parameter (LP) model is also prepared to validate the modelling procedure and results against the FSAR. Finally, LP profiles and 3D TH contour results of GOTHIC code were presented and discussed. LP results show a close agreement with FSAR reference and can approve the accuracy of the simulation procedure. 3D contours present all-coordinate detailed TH parameters versus time inside the containment. Spatial distribution of TH parameters and minor short-term effects of spray as Engineering Safety Features (ESFs) to tackle the containment pressurization can be evaluated by employing these contours. 3D simulation results can provide advantages for the precise locating and installation of ESFs in design and operation to achieve the highest efficiency in case of containment accident.

An investigation of characteristics of flow and heat transfer in the two-layer molten pool under swing motion condition

In-Vessel Retention (IVR) of molten corium is an important strategy for management of severe accidents adopted by some active service generation II plus reactors and advanced generation III reactors like AP1000, Hualong one and APR1400. One of the most important factors contributing to IVR analysis can be attributed to the characteristics of heat transfer in the molten pool in the lower head of reactor pressure vessel (RPV), which directly determines the heat load imposed on the vessel wall of the RPV lower head. Besides, IVR is also an important strategy for management of severe accidents of Ocean Floating Reactor (OFR), whose flow and heat transfer characteristics are influenced by the ocean motion conditions. However, since all previous molten pool experiments were conducted under static condition, characteristics of the heat transfer in the molten pool thus obtained may be different from those under ocean motion conditions, thus not applicable to the latter. To address this issue, in this study, an experimental system was established, which includes a two-layer molten pool test facility and a swing motion platform. Based upon the system, the effects of the swing angle and swing period on the characteristics of heat transfer in the two-layer melt pool were investigated. In addition, a scaling method was proposed to select simulant materials for the two-layer corium pool. Results of the experiment show that the swing period and angle have a significant effect on the temperature field of the molten pool and the rate of heat transfer. We thus suggest that, in future studies, new heat transfer correlations, models and IVR analysis method for ocean floating reactors should be researched and developed to evaluate the IVR design of ocean floating reactors, which is critical for management of severe accidents of the ocean floating reactors.

Potential Impact of Reactor Core Damage on Severe Accident Management Actions in Vicinity of Spent Fuel Pool

Fukushima Daiichi NPP accident showed that plant technical support center (TSC) in an extreme and rare external event (design extended condition (DEC)) can have a problem in the case of coincident loss of decay heat removal from the core (possibly resulting in significant core damage) and loss of decay heat removal from spent fuel pool. From the point of view of prioritizing severe accident management strategies it looks like the priority mitigation action should be to reestablish the emergency core cooling in the reactor pressure vessel. The reason is the longer time window available before the water inventory in the spent fuel pool would be evaporated and spent fuel exposed to overheating. However, if such actions would not be successful and reactor core would, consequently, be damaged, potential design basis leakage (or even greater leakage) from the containment to the fuel handling building (FHB) can affect already established TSC measures or operator accessibility to FHB, or it can jeopardize functioning of the systems, structures and components due to radioactive releases and presence of hydrogen (independently of the fact that containment atmosphere can be inerted by steam or that containment may be equipped with passive autolytic recombiners (PARs)). Paper describes an engineering evaluation of possible hydrogen presence in the containment annulus, its flammability and leakages through the penetrations toward FHB in the case of long term station blackout (SBO) without successful restoration of the core cooling in the reactor pressure vessel. SBO accident sequence progression and amount of produced hydrogen is evaluated by MAAP code.

Enhancements to PWR SAMG since Fukushima

Plant specific Severe Accident Management Guidelines are provided to give guidance to plant staff to respond to a severe accident, to protect containment, minimise fission product releases and prioritise equipment recovery efforts while bringing the plant to a controlled stable condition. This paper presents and discusses the enhancements that have been made particularly to widely implemented SAMG based on the Westinghouse Owners Group generic model (now part of the PWR Owners Group) since the Fukushima accident. Although U.S. and European responses to the accident may seem at first sight to be different, in fact, the enhancements needed to SAMG in order to address lessons learned were quite consistent. How to deal with loss of d.c. power and / or instrumentation, long term loss of a.c. power (much longer than had previously been considered), multiple unit severe accidents, accidents in the fuel pool, accidents from shutdown initial plant conditions and other issues and conditions that occurred during the event and which we must learn from? This paper presents results of a number of projects aimed at providing SAMG enhancements to address these issues, some originating in the U.S. and some in Europe. Main enhancements addressing lessons learned from Fukushima are described. The paper also summarises and describes an ongoing project to develop a fully revised generic SAMG specifically applicable to PWR plants in Europe (and also applicable to other designs, such as VVER) which benefits from all these efforts and will provide a single integrated and updated basis for plants with mitigation systems typical in Europe (often different from the US) to update their SAMG.

Accident Management under Extreme Events

Most nuclear power plants have extensive sets of Emergency Operating Procedures and Severe Accident Management Guidelines. These offer protection for a large series of events, both inside and outside the licensed design basis of the plant. For Extreme Events, which are characterised by a large destruction on-site and may include loss of command and control, damage to multiple units on-site, loss of communication both on-site and to off-site centres, staff members wounded or killed, such protection may not be enough. Examples of Extreme Events are air plane crash, site flooding, large earthquake plus possible tsunamis, etc. This paper describes what additional procedures, guidelines, hardware and organisational issues are needed to protect a site against such events. It is based on lessons learned from large destructive events in the past, such as the 9/11 attacks in the USA in 2001 and the tsunami at the Fukushima-Daiichi plants in 2011.