Review of current Severe Accident Management (SAM) approaches for Nuclear Power Plants in Europe

The Fukushima accidents highlighted that both the in-depth understanding of such sequences and the development or improvement of adequate Severe Accident Management (SAM) measures are essential in order to further increase the safety of the nuclear power plants operated in Europe. To support this effort, the CESAM (Code for European Severe Accident Management) R&D project, coordinated by GRS, started in April 2013 for 4 years in the 7th EC Framework Programme of research and development of the European Commission. It gathers 18 partners from 12 countries: IRSN, AREVA NP SAS and EDF (France), GRS, KIT, USTUTT and RUB (Germany), CIEMAT (Spain), ENEA (Italy), VUJE and IVS (Slovakia), LEI (Lithuania), NUBIKI (Hungary), INRNE (Bulgaria), JSI (Slovenia), VTT (Finland), PSI (Switzerland), BARC (India) plus the European Commission Joint Research Center (JRC). The CESAM project focuses on the improvement of the ASTEC (Accident Source Term Evaluation Code) computer code. ASTEC,, jointly developed by IRSN and GRS, is considered as the European reference code since it capitalizes knowledge from the European R&D on the domain. The project aims at its enhancement and extension for use in severe accident management (SAM) analysis of the nuclear power plants (NPP) of Generation II-III presently under operation or foreseen in near future in Europe, spent fuel pools included. In the frame of the CESAM project one of the tasks consisted in the preparation of a report providing an overview of the Severe Accident Management (SAM) approaches in European Nuclear Power Plants to serve as a basis for further ASTEC improvements. This report draws on the experience in several countries from introducing SAMGs and on substantial information that has become available within the EU “stress test”. To disseminate this information to a broader audience, the initial CESAM report has been revised to include only public available information. This work has been done with the agreement and in collaboration with all the CESAM project partners. The result of this work is presented here.JRC.F.5-Nuclear Reactor Safety Assessmen

Validation of ASTEC v2.0 corium jet fragmentation model using FARO experiments

ASTEC is an integral code for the prediction of Severe Accidents in Nuclear Power Plants. As such, it needs to cover all physical processes that could occur during accident progression, yet keeping its models simple enough for the ensemble to stay manageable and produce results within an acceptable time. The present paper is concerned with the validation of the Corium jet fragmentation model of ASTEC v2.0 rev3 by means of a selection of six experiments carried out within the FARO facility. The different conditions applied within these six experiments help to analyse the model behaviour in different situations and to expose model limits. In addition to comparing model outputs with experimental measurements, sensitivity analyses are applied to investigate the model. Results of the paper are (i) validation runs, accompanied by an identification of situations where the implemented fragmentation model does not match the experiments well, and discussion of results; (ii) its special attention to the models calculating the diameter of fragmented particles, the identification of a fault in one model implemented, and the discussion of simplification and ad hoc modification to improve the model fit; and, (iii) an investigation of the sensitivity of predictions towards inputs and parameters. In this way, the paper offers a thorough investigation of the merit and limitation of the fragmentation model used in ASTEC.JRC.F.5-Nuclear Reactor Safety Assessmen

Analysis of LFW & LBLOCA scenarios for a PWR 900 MWe NPP using the integral computer codes ASTEC2.0 and MAAP4.0.8

The present work has been carried out in the context of the development and validation of a set of “reference” ASTEC input decks for the main generic types of nuclear power plants (NPPs) in Europe; in the CESAM (Code for European Severe Accident Management) project, where JRC is contributing in several of the six working packages (WPs) in which it is divided. The JRC team have used the PWR900 model (delivered by IRSN to CESAM partners in February 2014) to simulate two accidents, a LBLOCA (Large-Break, Loss-Of-Coolant Accident) and a LFW (Loss of Feed Water), in order to confirm the ASTEC v2.0r3 capabilities. The assessment was done comparing the evolutions of these accidents vs. MAAP 4.0.8 code. To perform such kind of comparison, JRC developed a specific MAAP4 input deck for a pressurized water reactor (PWR) of 900 MWe. This input deck is based on a generic MAAP4 input deck, which has been adapted to correspond to the best extent possible to the ASTEC V2.0r3 PWR900 model. The comparisons focus only on in-vessel thermal hydraulics phenomena, core degradation and corium behaviour in the lower head. The calculations stop as soon as the reactor pressure vessel fails. The comparisons of the main thermo-hydraulic variables between the two code results are detailed in this paper and are satisfactory overall.JRC.F.5-Nuclear Reactor Safety Assessmen

Summary and conclusions from the International Seminar on In-Vessel Retention Strategy

On the 6-7 June 2016, IRSN hosted an international workshop about the « Strategy of In-Vessel Melt Retention: Knowledge and Perspectives ». The workshop was co-organized by JRC and IRSN, with the sponsorship of ETSON. With panel discussions and technical sessions, the workshop covered all the important issues related to in-vessel corium retention, from the physical understanding to regulatory frames. The major points of the safety demonstration were discussed. Some industrial aspects were also addressed. One of the objectives was to provide an orientation of R&D projects to strengthen IVR strategies, such as the H2020 IVMR project, coordinated by IRSN. The current approach followed by most experts for IVR is a compromise between a deterministic approach using the significant knowledge gained during the last two decades and a probabilistic approach to take into account large uncertainties due to lack of data for some phenomena and due to excessive simplifications of models. It was concluded that a harmonization of the positions of safety authorities on the IVR strategy is necessary to allow decision making based on scientific knowledge. For this, a consensus on several issues should be reached between R&D experts. This includes in particular the issues of the transient evolution of oxide and metal layers in the lower plenum and of the long term mechanical behavior of the thin “cold shell” resulting from vessel ablation.JRC.G.I.4-Nuclear Reactor Safety and Emergency Preparednes

CESAM – Code for European Severe Accident Management, EURATOM project on ASTEC improvement

The CESAM FP7 project of EURATOM has been conducted from April 2013 until March 2017 in the aftermath of the Fukushima Dai-ichi accidents. Nineteen international partners from Europe and India, including the European Joint Research Centre, have been participating under the coordination of GRS and with a strong involvement of IRSN that were both ASTEC code developers. The Project objectives were in priority an improved understanding of all relevant phenomena during the Fukushima Dai-ichi accidents and their importance for Severe Accident Management (SAM) measures as well as the improvement of the ASTEC computer code to simulate plant behaviour throughout accident sequences including SAM measures. One starting step was the analysis of current SAM measures implemented in European nuclear power plants. In order to achieve these goals, simulations of relevant experiments that allow a solid validation of the ASTEC code against single and separate effect tests have been conducted. Covered validation topics in the CESAM project have been grouped in 9 different areas among which are re-flooding of degraded cores, pool scrubbing, hydrogen combustion, or spent fuel pool behaviour. Furthermore, modelling improvements have been implemented in the current ASTEC V2.1 series for the estimation of source term consequences in the environment and the prediction of plant status in emergency centres. Finally, ASTEC reference input decks have been created for all reactor types operated in Europe today as well as for spent fuel pools. These reference input decks generically describe plant types like PWR, VVER, PHWR, and BWR without defining proprietary data of a special plant and they account for the best recommendations from code developers and users. In addition, a generic input deck for a spent fuel pool was elaborated. These input decks can be used as basis by all (and especially new) ASTEC users in order to understand code basic requirements and model features and to implement the specificities of their own NPP type. Based on these generic inputs, benchmark calculations have been performed with other codes (such as MELCOR, MAAP, ATHLET-CD, COCOSYS…) with a focus on applicability of ASTEC models to currently implemented SAM measures. This article provides a final summary of the CESAM project. Therefore, an overview of the improved modelling capabilities of the recent ASTEC V2.1 version is given followed by the validation status of ASTEC V2.1 as concluded after CESAM. Further, plant applications performed by CESAM partners will be summarized with a special focus on simulation of SAM measures in various NPP types and finally, insights gained on SAM measures will be derived.JRC.G.I.4-Nuclear Reactor Safety and Emergency Preparednes

Status of the IVMR project: First steps towards a new methodology to assess In-Vessel Retention Strategy for high-power reactors

The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and isolate corium and fission products in the reactor pressure vessel and in the primary circuit. This type of Severe Accident Management (SAM) strategy has already been incorporated in the SAM guidance (SAMG) of several operating small size LWR (reactor below 500 MWe (like VVER440)) and is part of the SAMG strategies for some Gen III+ PWRs of higher power like the AP1000 or the APR1400. However, the demonstration of IVR feasibility for high power reactors requires using less conservative models as the safety margins are reduced. In Europe, the IVMR project aims at providing new experimental data and a harmonized methodology for IVR. During the first year of the project, the work was mostly dedicated to methodology and computer code analysis. A synthesis of the methodology applied to demonstrate the efficiency of IVR strategy for VVER-440 in Europe (Finland, Slovakia, Hungary and Czech Republic) was made. It showed very consistent results, following quite comparable methodologies. The main weakness of the demonstration was identified in the evaluation of the heat flux that could be reached in transient situations, e.g. under the “3-layers” configuration. Analyses have also started for various designs of reactors with a power between 900 and 1300 MWe. Large discrepancies of results were observed, which were due to the use of very different models for the description of the molten pool: homogeneous, stratified with fixed configuration, stratified with evolving configuration. The last type of model provides the highest heat fluxes (above 3 MW/m2) whereas the first type provides the lowest heat fluxes (around 500 MW/m2). Obviously, there is an urgent need to reach a consensus about best estimate practice to be used in the major codes for safety analysis, such as ASTEC, MELCOR, SOCRAT, MAAP, SCDAP/RELAP, etc. Despite the discovered model discrepancies, and leaving aside the unrealistic case of homogeneous pool, the average calculated heat fluxes in many cases are well above 1 MW/m2 which would threaten the integrity of the vessel considerably and require a detailed mechanical analysis. Therefore, it is clear that the safety demonstration of IVR for high power reactors requires a more careful evaluation of the situations which can lead to formation of either a very thin top metal layer provoking focusing effect or significantly overheated metal, e.g. after oxide and metal layer inversion. An international workshop was organized at the end of the first year of the IVMR project in order to gather the positions of several safety authorities and research organizations from all over the world. As a conclusion, it appears that the current approach followed by most experts for IVR is a compromise between a deterministic analysis using the significant knowledge gained during the last two decades and a probabilistic analysis to take into account large uncertainties due to the lack of data for some physical phenomena (as listed above) and due to excessive simplifications of models. A harmonization of the positions of safety authorities on the IVR strategy is necessary to allow decision making based on shared scientific knowledge. Currently, the acceptance criteria of a safety demonstration for IVR may be differently defined from one country to the other and the differences should be further discussed to reach harmonization on this important topic. This includes the accident scenarios to be considered in the demonstration and the modelling of the phenomena in the vessel. Such harmonization is one of the goals of IVMR project. The project will now focus on providing new experimental data for situations of interest like the inversion of stratification and the kinetics of growth of the top metal layer. The project will also provide new data about external vessel cooling from full-scale facilities: CERES for VVER-440 and a new facility built by UJV for VVER1000. It will also include an activity on innovations dedicated to increase the efficiency of the IVR strategy such as delaying corium arrival in the lower plenum, increasing the mass of molten steel or implementing measures for simultaneous in-vessel water injection.JRC.G.I.4-Nuclear Reactor Safety and Emergency Preparednes

Some considerations to improve the methodology to assess In-Vessel Retention strategy for high-power reactors

The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and isolate corium and fission products in the reactor pressure vessel and in the primary circuit. This type of Severe Accident Management (SAM) strategy has already been incorporated in the SAM guidance (SAMG) of several operating small size LWR (reactor below 500 MWe (like VVER440)) and is part of the SAMG strategies for some Gen III+ PWRs of higher power like the AP1000 or the APR1400. However, the demonstration of IVR feasibility for high power reactors requires using less conservative models as the safety margins are reduced. In Europe, the IVMR project aims at providing new experimental data and a harmonized methodology for IVR. A synthesis of the methodology applied to demonstrate the efficiency of IVR strategy for VVER-440 in Europe (Finland, Slovakia, Hungary and Czech Republic) was made. It showed very consistent results, following quite comparable methodologies. The main weakness of the demonstration was identified in the evaluation of the heat flux that could be reached in transient situations, e.g. under the “3-layers” configuration, where the “focusing effect” may cause higher heat fluxes than in steady-state (due to transient “thin” metal layer on top). Analyses of various designs of reactors with a power between 900 and 1300 MWe were also made. Different models for the description of the molten pool were used: homogeneous, stratified with fixed configuration, stratified with evolving configuration. The last type of model provides the highest heat fluxes (above 3 MW/m2) whereas the first type provides the lowest heat fluxes (around 500 MW/m2) but is not realistic due to the non-miscibility of steel with UO2. Obviously, there is a need to reach a consensus about best estimate practices for IVR assessment to be used in the major codes for safety analysis, such as ASTEC, MELCOR, SOCRAT, MAAP, ATHLET-CD, SCDAP/RELAP, etc. Despite the model discrepancies, and leaving aside the unrealistic case of homogeneous pool, the average calculated heat fluxes in many cases are well above 1 MW/m² which could reduce the residual thickness of the vessel considerably and threaten its integrity. Therefore, it is clear that the safety demonstration of IVR for high power reactors requires a more careful evaluation of the situations which can lead to formation of either a very thin top metal layer provoking focusing effect or significantly overheated metal, e.g. after oxide and metal layer inversion. It also requires an accurate mechanical analysis of the ablated vessel. The current approach followed by most experts for IVR is a compromise between a deterministic analysis using the significant knowledge gained during the last two decades and a probabilistic analysis to take into account large uncertainties due to the lack of data for some physical phenomena (such as transient effects) and due to excessive simplifications of models. A harmonization of the positions of safety authorities on the IVR strategy is necessary to allow decision making based on shared scientific knowledge. Currently, the acceptance criteria of a safety demonstration for IVR may be differently defined from one country to the other and the differences should be further discussed to reach harmonization on this important topic. This includes the accident scenarios to be considered in the demonstration and the modelling of the phenomena in the vessel. Such harmonization is one of the goals of IVMR project. A revised methodology is proposed, where the safety criterion is not based on a comparison of the heat flux and the Critical Heat Flux (CHF) profiles as in current approaches but on the minimum vessel thickness reached after ablation and the maximum pressure load that is applied to the vessel during the transient. The main advantage of this revised criterion is to include both steady-state configurations and transient states of the corium pool. Another advantage is that this criterion may be used in both probabilistic and deterministic approaches.JRC.G.I.4-Nuclear Reactor Safety and Emergency Preparednes

Identification and Categorisation of Safety Issues for ESNII Reactor Concepts. Part I: Common phenomena related to materials

With the aim to develop a joint proposal for a harmonised European methodology for safety assessment of advanced reactors with fast neutron spectrum, SARGEN_IV (Safety Assessment for Reactors of Gen IV) Euratom coordination action project gathered together 22 partners’ safety experts from 12 EU Member States. The group consisted of eight European Technical Safety Organisations involved in the European Technical Safety Organisation Network (ETSON), European Commission’s Joint Research Centre (JRC), system designers, industrial vendors as well as research & development (R&D) organisations. To support the methodology development, key safety features of four fast neutron spectrum reactor concepts considered in Deployment Strategy of the Sustainable Nuclear Energy Technology Platform (SNETP) were reviewed. In particular, outcomes from running European Sustainable Nuclear Industrial Initiative (ESNII) system projects and related Euratom collaborative projects for Sodium-cooled Fast Reactors, Lead-cooled Fact Reactors, Gas-cooled Fast Reactors, and the lead-bismuth eutectic cooled Fast Spectrum Transmutation Experimental Facility were gathered and critically assessed. To allow a consistent build-up of safety architecture for ESNII reactor concepts, the safety issues were further categorised to identify common phenomena related to materials. Outcomes of the present work also provided guidance for identification and prioritisation of further R&D needs respective to the identified safety issues.JRC.F.5-Nuclear Reactor Safety Assessmen