HGF-Programm Erneuerbare Energien

The Severe Accident Research Program at KIT

A. Miassoedov, Th. W. Tromm, Karlsruhe Institute of Technology

Introduction

The understanding of the plant behaviour under beyond design basis accidents as well as the interaction of the operators with the plant is the most important pre-requisite to develop proper strategies to both control the accident progression and to minimize the radiological risk that may derive from operating nuclear power plants. In view of the Fukushima accident, a review of many issues important to safety e.g. severe accident analysis methodologies and assumptions, emergency operational procedures, severe accident management procedures (SAM), decision lines of the emergency team, etc. is needed to draw conclusions in order to avoid a repetition of Fukushima-like accidents. In addition, situations like the “black control room” need to be reconsidered and a re-evaluation of the necessary instrumentation for hypothetical severe accident situations is urgently needed. If the real plant state during core meltdown accidents is unknown, no effective measures can be initiated by the emergency team in order to assure the integrity of the safety barriers and hence the release of radioactive material to the environment. The work performed in this area is integrated in the European Networks such as SARNET (Severe Accident Research Network) for the severe accidents, and for emergency management in the NERIS-TP. In future all the activities will be included in the NUGENIA platform. In the following a brief overview of the KIT activities together with the experimental test facilities is given.

Fig. 1: Degraded core reflood flow map for degradation (left) and hydrogen release (right) issues

1 QUENCH-facility In the QUENCH programme hydrogen source term and high-temperature materials behaviour during DBAs and BDBAs are investigated. Integral bundle tests are conducted in the unique QUENCH facility complemented by a wide range of setups in laboratory scale. Experimental data are used worldwide for model development and validation of computer codes for simulation of nuclear accidents. Degraded core reflood database for beyond design accidents To assess capabilities for successful accident mitigation measures, the core reflood map, see Fig.1, was updated with ongoing QUENCH experiments. The experiment QUENCH-11, which simulates the whole course of a core dryout accident, is used successfully for hands-on training purposes as part of the AREVA Nuclear Professional School course ANPS-007 (Beyond Design and Severe Core Damage Accidents).

The three US-NRC codes SCDAP/RELAP5, RELAP5, and TRACE have been compared on the basis of the experiment QUENCH-04, dedicated to investigate the behavior of an overheated LWR rod bundle during steam cool-down, especially hydrogen production. Shortcomings of R5 as well as of TRACE could be identified. Even though some modeling improvements will be done in future, the calculations show clearly the relevance of the related experiments for safety analyses.

QUENCH The bundle test QUENCH-16 including an air ingress phase was successfully conducted in the framework of the EC funded LACOMECO programme. An extended oxygen starvation phase resulted in pronounced formation of zirconium nitride during air oxidation (Fig. 2). Reflood with 50 g/s cold water caused a temporary thermal runaway with maximum temperatures of 2150°C which was not simulated by any of the pre-test calculations. A SARNET2 benchmark exercise on the two QUENCH air ingress experiments QUENCH-10 and QUENCH-16 coordinated by PSI is in progress.

Fig. 2: QUENCH-16: Formation of porous nitrides inside the cladding oxide layer on the end of an air ingress phase (left) with their following re-oxidation during reflood (outer porous oxide sub-layer on the right picture)

The first two bundle experiments of the new QUENCH-LOCA series aimed at the investigation of coolability and secondary hydrogen uptake by Zry cladding and supported by VGB were conducted. Commissioning test QUENCH-L0 confirmed the qualification of the QUENCH facility for DBA experiments including highly transient heatup of the bundle and ballooning and burst failure of the separately pressurized fuel rod simulators. QUENCH-L1 simulated a typical LOCA scenario of German NPPs. For the first time non-symmetrical hydrogen bands near the burst position were detected by neutron imaging methods and their effect on the mechanical properties was shown (Fig. 3).

2 Report for the Midterm Review of the KIT Programme NUKLEAR

Fig. 3: QUENCH-L0: Neutron tomography image of hydrogen bands in Zry cladding near the burst position and corresponding double brittle failure during tensile test. Neutron radiography was additionally applied in in-situ experiments up to 1400°C on hydrogen behaviour in zirconium alloys during oxidation in steam and delayed hydrogen cracking using the specially designed INRRO furnace with windows transparent for neutrons. Further separate- effects tests resulted in substantial progress of the phenomenological understanding of oxidation of zirconium alloys in various atmospheres as well as of other materials relevant for reactor applications. To assess capabilities for successful accident mitigation measures the core reflood map was updated with ongoing QUENCH experiments. The experiment QUENCH-11, which simulates the whole course of a core dryout accident is used successfully for hands-on training purposes as part of the AREVA Nuclear Professional School course ANPS-007 (Beyond Design and Severe Core Damage Accidents). The three codes SCDAP/RELAP5, RELAP5, and TRACE have been compared on the basis of the experiment QUENCH-04, dedicated to investigate the hydrogen production of steam cool- down of an overheated LWR rod bundles. Though some modelling improvements might be done, the calculations with S/R5 agree sufficiently well with experimental data that shortcomings of R5 as well as of TRACE could be identified. Further LOCA bundle experiments will be devoted to advanced cladding materials and behaviour of high burnup fuel rod claddings. In the framework of SARNET2 a very challenging bundle experiment on the formation and coolability of a debris bed in a degraded core is under preparation and will be conducted end 2012/begin 2013. Structure elements which should survive the test (shroud, heater rods) are made of hafnium; the debris bed will be formed by prototypical fully oxidized Zircaloy-4 cladding and pre-fragmented ZrO2 pellets. The high-temperature behaviour of core materials will be further investigated with special emphasis on corrosion, hydrogenation, mutual interactions and mechanical properties. The range of materials will be extended; thus a PhD thesis on SiC, which is one promising candidate for future accident tolerant cladding tubes, has started in 2012.

LIVE-facility The main objective of the LIVE programme is to study the late in-vessel core melt behaviour and core debris coolability both experimentally in large scale 2D and 3D geometry and in supporting separate-effects tests. The test series with insulated upper melt surface and KNO3- NaNO3 as melt simulant was completed. The obtained information resulted in the data base containing heat flux distribution along the reactor pressure vessel wall in transient and steady state conditions, crust growth velocity and evolution of the interface temperatures between the liquid pool and the solid crust. Supporting post-test analysis contributed to characterization of solidification processes of binary non-eutectic melts.

3 First test series to study the behaviour of a stratified molten pool in the lower head were carried out with molten salts. These tests are preparatory for the experiments with high-temperature V2O5-based melt in a stratified configuration. In all tests lower layer was volumetrically heated and the upper layer has no power dissipation (in one test 25% of the total heat was generated in the upper layer for comparison). Both layers were composed of 20% KNO3-20% NaNO3 non- eutectic mixture and were separated by a 2 mm copper plate located at 333 mm of a 436 mm- height pool. The information obtained from the experiments includes the distribution of heat flux through the vessel wall from the two layers, heat transfer from the lower layer to the upper part of the melt, melt pool temperature profiles and crust characteristics. LIVE-L6 test was selected as benchmark experiment for the evaluation of codes used by the SARNET2 partners engaged in the in-vessel melt retention studies. LIVE-L8B test, see Fig. 4, investigated temperature distribution in a debris bed in PWR lower head in dry-out conditions and the transition from the debris bed to the molten pool. The test was requested by the SARNET2 partners and was aimed at providing input data for analysis of different debris cooling concepts and severe accident management. The simulant material used for the debris bed and the liquid melt was a non-eutectic binary mixture of KNO3-NaNO3. The debris bed and the liquid melt account each 50% of the total core material in the lower head. The temperature distribution in the debris bed before melt relocation was interpreted by a 2D diagram showing the region of highest temperatures and the temperature gradient in the debris bed. After liquid melt relocation in the debris bed, the form and the volume of the liquid region during the melting process are evaluated. Additionally, the important stages of formation of a fully developed molten pool in the lower head were investigated. A compact crust layer with embedded debris was formed at the upper part of the vessel wall, whereas a loose debris layer at the vessel bottom can exist as long as the vessel wall is externally water-cooled.

138 350 106 187 155 57,0 171 300 155 138 122 250 89,5 106 ) 122 m 171 89,5

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100 122 138 50 106 0 0 100 200 300 400 radius (mm)

Fig. 4: LIVE debris test and temperature development

Within the EU LACOMECO project one test addressing transient 2D corium-refractory material interaction for the core catcher design was performed together with CEA Grenoble. The objective was to simulate ablation process of a high-melting temperature refractory material by low-melting temperature corium. Due to the dissolution of the refractory material into the boundary area and in the bulk corium, the boundary layer and the bulk melt are gradually enriched in the refractory material, which leads to an increase of bulk melt melting temperature and boundary temperature. In the test a thick crust layer of pure KNO 3 (melting temperature of ~334°C) was generated along the semi-spherical vessel wall as the simulant of the refractory material, and the eutectic mixture of 50 mol% KNO3 - 50 mol% NaNO3 (melting temperature of ~220°C) simulating the corium was poured in the vessel afterwards. During the ablation process, the liquid melt is homogenous heated and its temperature was defined to below the melting temperature of the pure KNO3. Two ablation tests were performed. The heating powers in the first and the second test were 7 kW and 15 kW respectively. The refractory material and

4 Report for the Midterm Review of the KIT Programme NUKLEAR the corium simulant for the second test were the actual crust and melt at the end of the first test. The ablation tests provided the results of the evolution melt temperature and the interface temperature during the ablation process. Also the crust thickness during ablation as well as in the final state was measured. The heat flux distribution along the crust and the crust composition were analyzed.

To compare the experimental results of 3D experiments in LIVE-3D with 2D experiments, the LIVE-2D test facility, see Fig. 5, was constructed and commissioned and a series of tests have been completed. In the LIVE-2D facility the melt pool in a lower plenum of a RPV is modelled by a slice with a thickness of 12 cm and a radius of 0.5 m (the same radius as the hemisphere in LIVE-3D). To study the top/bottom distribution of the heat flux, the LIVE-2D test vessel can be covered at the top by an insulated or a cooled lid. The results of the LIVE-2D experiments regarding the upward heat transfer are compared with the earlier tests performed in the BALI and SIMECO facilities as well as with correlations developed from those tests. For low Rayleigh numbers (1012-1013) the results of LIVE-2D tests demonstrate higher heat transfer to the top of the melt. One of the objectives of the future tests will be to check if this deviation also persists for higher Rayleigh numbers (1014-1015) and to identify the main reasons for this behaviour. The comparison with the 3D experiments is still outstanding, because the relevant 3D experiments in the LIVE-3D facility will be performed in the near future.

Fig. 5: LIVE 2D test facility

DISCO facility The DISCO-H test facility was set up to perform scaled experiments that simulate melt ejection scenarios under low system pressure in severe accidents in PWRs such as for example German Konvoi plants. These experiments are designed to investigate the fluid-dynamic, thermal and chemical processes during melt ejection out of a breach in the lower head of a PWR pressure vessel at pressures around and below 2 MPa with an iron-alumina melt and steam. The main components of the facility are scaled about 1:18 linearly to a large European pressurized water reactor. Standard test results are: pressure and temperature history in the RPV, the cavity, the reactor compartment and the containment, post test melt fractions in all locations with size distribution of the debris, video film in reactor compartment and containment (timing of melt flow and hydrogen burning), and pre- and post test gas analysis in the cavity and the containment.

Eight tests were performed in Konvoi geometry, see Fig. 6, with holes in the centre of the RPV lower head (1m diameter scaled), using an iron-alumina melt driven out of the RPV by steam, and an atmosphere of air, steam and 5% hydrogen with 0.2 MPa in the containment vessel. The geometries of the reactor pit of the Konvoi plant and the EPR are similar with two distinct differences regarding the flow paths out of the lower pit: In the EPR the flow path into the refuelling room above the reactor pit can be considered to be closed for all possible DCH cases,

5 while in the Konvoi reactor it is open if the overpressure in the pit is more than 0.2 MPa. The Konvoi cavity has a biological shield and pressure venting flaps behind the shield must be assumed to be open and serve as a large path into the neighbouring compartments, which could be blocked by water during an accident if its level is high enough.

Fig. 6: Disco facility and Konvoi geometry

For the base case of the tests series both of these flow paths were closed. The standard flow path out of the pit was that along the 8 main cooling lines leading into a subcompartment which is connected to the containment dome by large openings. To investigate the effect of the breach size one test was conducted with a smaller hole in the lower head (0.5 m diameter scaled). Two tests were performed with lower RPV pressures and one with a higher hydrogen concentration in the containment atmosphere. In one test the flow path into the refuelling room was open, and another test was conducted with open pressure venting flaps and water behind the bio-shield and in the subcompartment. With identical initial conditions in the EPR and the Konvoi plants similar results for the containment pressure can be expected, provided the flow paths into the refuelling room and behind the bio-shield stay closed and the space behind the bio-shield remains dry. Lower initial RPV pressures and smaller breach sizes lead to less melt dispersion out of the reactor pit and thus to lower containment pressure increase. If the flow path through the vessel support into the refuelling room is open, a mitigating effect occurs. A considerable amount of debris and hydrogen is trapped in this small compartment and does not contribute to pressure increase in the containment, provided that the concrete slaps, which cover this room, stay in place. However, if the overpressure in the pit is high enough to open the path at the vessel support it might also be high enough to lift the concrete plates in the refuelling room, and a direct path into the containment would be available, which means an enhancing effect for DCH. The open flow path behind the bio-shield together with the presence of water leads to low values of the containment pressure increase. About half of the melt dispersed out of the pit is entrained through the venting flaps behind the bio-shield and enters the water where it is quenched. Very little hydrogen is produced and little burned, and consequently the pressure increase is low. A higher initial hydrogen concentration in the containment leads to substantial higher containment pressures. The more hydrogen exists before the blowdown the higher is the fraction of burnt hydrogen. The total amount of hydrogen burnt correlates with the containment pressure increase. The maximum pressure increase in the containment measured was 0.4 MPa.

DISCO hydrogen combustion tests Hydrogen combustion tests are performed in two different size facilities to reproduce hydrogen effects during a severe accident with high pressure melt ejection and direct containment heating. The hydrogen is blown out of a pressure vessel into a constrained compartment, modelling the reactor pressure vessel and the reactor pit, respectively, and from there into a large vessel, modelling the containment. A number of distributed igniters simulate hot melt particles. Tests with and without steam and with concentrations of pre-existing hydrogen in the containment atmosphere between 0 and 8% are conducted. The results of the tests have shown so far that there is no scaling effect relative to the pressure increase in the containment. The pressure increase correlates with total hydrogen burned. The fraction of hydrogen that burns depends on the ratio of pre-existing to blow-down hydrogen and on the total amount of hydrogen and varies between 46% and 100%. Compartments may have an effect on the burnt fraction but this will be investigated in depth. The efficiency of combustion energy conversion into pressure varies between 42 and 71% and again may be affected by compartments and structures in the containment. These effects will be analysed by code calculations, for which the experimental results may serve as a data base for code modelling and validation.

MOCKA facility Even though extensive research has been undertaken over several years in the area of core- concrete interaction, several subjects need further investigations. An important issue concerns the distribution of the heat flux to the concrete in the lateral and axial directions during the long- term 2-dimensional concrete erosion by a core melt. The knowledge of this partition is important in the evaluation of the consequences of a severe reactor accident.

Within the MOCKA programme a series of experiments have been performed to study the process of concrete erosion by an oxide and metal melt in a stratified configuration. The internal heat generation in the oxide phase is simulated by a succession of additions of pure thermite and Zr metal to the melt from the top being the first of a kind heating method realized for high temperature melts worldwide. The additional enthalpy generated by the thermite reaction and exothermal oxidation reactions of Zr is mainly deposited in the oxide phase which is representative for reactor situations. In small-scale MOCKA experiments siliceous concrete crucibles with an inner diameter of 25 cm were used. The initial melt consisted of 39 kg Fe together with 3 or 4 kg Zr, overlaid by 70 kg oxide melt (Al2O3, CaO). The initial melt temperature at start of interaction was approximately 2173 K. The long-term axial erosion by the metallic phase was a factor of 2-3 higher than the lateral ablation. Similar results were obtained in former BETA and COMET-L experiments. In contrast to the findings in BETA and COMET-L experiments significant lateral concrete erosion by the oxide melt was observed. The more pronounced downward erosion seems to be inherent to the erosion by metal melts. To address the scaling issues a large-scale MCCI test, see Fig. 7, was performed in a siliceous cylindrical crucible with 1 m inner diameter, wall thickness of 25 cm and 2.5 m height. The melt is generated in the crucible by thermite reaction, resulting in steel melt (1426 kg Fe, collapsed melt height of 28 cm) and oxide melt (1167 kg Al2O3, 907 kg CaO). The initial melt temperature is ~1900 °C. To extend the duration of the interaction with the concrete and allow for significant concrete erosion by the oxide as well as by the metal melt, 1342 kg thermite and 450 kg Zr are added to the melt from the top within 15 min. The overall downward erosion by the metal melt is of the same order than the sideward one and amounted to ~6 cm. The lateral concrete erosion by the overlaid oxide melt in the upper part of the crucible is significantly higher than that of the metal melt. The maximum erosion depth is ~18 cm. The interpretation and benchmarking activities will be carried out in close cooperation with international partners within the OECD-MCCI project and SARNET code benchmarks. One of the still unresolved issues is the long-term interaction of a melt with a reinforced concrete and it will be addressed in the future MOCKA programme.

7 Fig. 7: MOCKA large scale experiments and Zr-feeding system for decay heat simulation

HYKA-facility The Fukushima accident demonstrated drastically the importance of hydrogen mitigation measures in nuclear power plants. Unfortunately, none of the mitigation measures which have been implemented decades ago in German nuclear power plants, like filtered venting systems and passive autocatalytic recombiners (PAR) were installed there. Moreover, the thin walled reactor building could hardly withstand any fast hydrogen combustion. On the other hand, however, this accident reminds us again to take a closer look on the accuracy of our own predictions of hydrogen risks since our codes are used worldwide today to predict the consequences of hydrogen release, combustion and detonation in safety assessments for nuclear facilities. For the improved simulation of hydrogen behaviour under severe accident conditions, new models or improvements of existing models have been developed by KIT and implemented in the specialised CFD codes GASFLOW, COM3D and DET3D. The developments and improvements concern physical phenomena, in particular water sprays, heat transfer with phase transition, real gas behaviour, turbulent combustion, particle transport and hybrid hydrogen/dust combustion. Also the more technical or empirical models for the passive autocatalytic recombiners have been adapted to new experimental data. All this work has been based on reference experiments, either conducted in the KIT own facilities HYKA and based on own financial resources, or has jointly been performed in national projects, or contributed by international partners in the OECD benchmark exercises. Important contributions have also been provided to or received from European networks, in particular SARNET and HySafe.

KIT coordinated the first international standard problem ISP49 dedicated to hydrogen combustion in containment like geometries and performed a unique programme with large scale experiments for the combustion of semi-open flat layers. For the latter case, KIT cooperated with the SME partner Pro-Science, with GRS and with Technical University of Munich. In the HYKA facility A1, premixed layers of hydrogen and air were introduced beneath a flat ceiling with a carefully designed system, consisting of premixing vessels, set of release nozzles and protective foils. These premixed layers have been 9 m long, 3 m wide and up to 0,6 m thick. Another set of large scale combustion experiments with and without nitrogen and steam inertisation and with or without vertical mixing gradients have been conducted in the HYKA vessels A2 and A3 in the frame of the EC project LACOMECO, also coordinated by KIT. The results have been partly disclosed to the SARNET-2 consortium for benchmarking purposes. The PAR models implemented in GASFLOW have been validated by participation in national as well as in international benchmarks based on recent experiments performed in the Thai facility of Becker Technology. The issues addressed in these investigations have been ignition

8 Report for the Midterm Review of the KIT Programme NUKLEAR probabilities induced by the PARs themselves and behaviour of the PAR under oxygen depletion conditions. Further large scale flame stability tests have been prepared for the A2 vessel. Tests are scheduled for end of 2012, beginning of 2013. These tests will help to improve the combustion models and provide insights into the different mechanisms for flame acceleration and flame extinction. The experimental programme with the flat layer geometries, see Fig. 8, has been recently extended. Broad variations of the configurations including vertical mixing gradients and containment-like internal structures, like grids, will be studied in a four years project. The experiments have been and will be simulated in detail with different reactive CFD codes - the code COM3D has been applied at KIT for this purpose. Primary objective of these investigations will be to derive more robust, simplified engineering correlations and criteria for flame acceleration and deflagration-detonation transition. These criteria will be used in particular in lumped parameter codes like MELCOR or COCOSYS. Also motivated by the accident phenomena in Fukushima, vented explosions will be assessed in the European project “HyInDoor”. KIT contributes with the broadest experimental sub- programme and with many numerical simulations to this project. Although not explicitly accommodated in a nuclear research framework, the investigated phenomena and achieved results are applicable to nuclear and non-nuclear hydrogen related accident scenarios, where venting is a mitigation strategy.

P, I P, I P, I P, I P, I P, I P, I P, I P, I I  CH2

Fig. 8: Experimental layout for measurement of flame acceleration and deflagration-detonation transition in flat layers with and without vertical concentration gradients in the HYKA A1 vessel

Plant applications and contributions to PSA-II methodologies To model the complete progression of a severe accident in a light-water reactor nuclear power plant fully integrated, engineering-level computer codes such as MELCOR and ASTEC are used. Both codes were applied to accident sequence studies, uncertainty and sensitivity studies and support to experiments. Probabilistic safety assessments (PSA) for NPPs are conducted to yield insight into the design and performance of plants and their potential environmental effects. For this reason, the behavior of the NPP in case of certain initiating events (LOCAs, Transients etc.) is investigated taking into account several kinds of uncertainties. A PSA-2 covers possible events occurring in the range from the beginning of core degradation and ending with the release of radioactive material into the environment or with a stable state, in which a long term retention of the radioactive material inside the containment can be ensured. KIT takes part in

9 the further qualification of the integral tools MELCOR an ASTEC for use in PSA-2. Broad spectra of severe accident phenomena in both boiling and pressurized water reactors are treated. These include thermal-hydraulic phenomena related to the containment such as hydrogen production, transport and combustion. On the other hand, core heatup, degradation, and relocation in vessel and core-concrete interaction in the reactor cavity are investigated. The way is to qualify the models in these codes for best estimate calculations. In ASTEC, especially models for core degradation and H2-production are tested. In MELCOR, the work is focused on models for H2 distribution in the containment and for description of the behavior of a core melt in the lower plenum. For this, the competence, existing due to the theoretical and experimental investigations at the Quench- and LIVE facilities, are used. Another goal of KIT is to develop a tool package, which allows statistical sensitivity analyses with arbitrary codes. It is foreseen to use this tool in a first step for sensitivity analyses with MELCOR and ASTEC. The need for such tools is given, because the existing tool SNAP, delivered by SANDIA, does not allow the implementation of new mathematical methods as well as the handling of large result files.

In case of a severe accident, the containment is the ultimate barrier to the environment. Therefore, reliable simulations tools for containment thermal hydraulics, including hydrogen distribution are indispensable. The simulation of the behavior of the containment atmosphere under severe accident conditions with a postulated source term of water, steam and hydrogen was performed under the frame of the VGB contract, using the detailed 3D CFD code GASFLOW and a lumped parameter code MELCOR in order to compare and assess their modeling capabilities. A simplified generic containment including all important components was used as a test case. The calculated pressure histories, mass and energy balances, convective flow as well as steam and hydrogen distributions were analysed. Integral values were modeled in good agreement by both codes. The overall flow was reasonably predicted. However discrepancies in the calculated steam and hydrogen concentrations were observed. Several QUENCH and LIVE tests were calculated and are currently under preparation with ASTEC, mostly in the frame of SARNET. Using MELCOR and ASTEC, KIT takes part in the OECD project BETMI-2, in the frame of which three different severe accident sequences are being examined starting from a previous benchmark. The impact on hydrogen production, core coolability, corium relocation into the lower plenum and vessel failure will be addressed.

By an adequate modeling of the events in the lower plenum of a boiling-water reactor (SWR) during a severe accident and the further-reaching modeling of the phenomena of the containment cooling, it will be possible to calculate the complete accident event from the triggering event, core destruction, the aerosol spreading in the containment up to the source term to the surrounding environment. The severe accident codes ASTEC and MELCOR being strengthened and improved for BWR conditions will be primarily used for a comparison with the GRS codes ATHLET CD and COCOSYS. Moreover, relevant experiments will be analysed in detail for validation and improvements if necessary. With these refinements, accident management measures for selected scenarios will be analysed and optimized for German plant types.

The coupling of MELCOR and GASFLOW is expected to be finalized in October 2013 and proofed by calculations of relevant accident sequences. This will be the basis for further investigations of the H2-distribution in the containment and an improved evaluation of the Reco- concept in german NPPs.

The currently developed sensitivity tool will be tested for accident sequences, which are currently being calculated in the frame of the OECD project BETMI-2.

Emergency Management

Thanks to the Japanese emergency management system SPEEDY, the population around Fukushima could be informed daily about the expected radioactive dose rates during the accident and people could be evacuated in time to be protected safely. KIT was supporting the daily predictions with our Real-time On-line DecisiOn Support system RODOS and informed the public via the KIT internet site. As a consequence, different from the consequences of the Chernobyl accident, the emergency management system could practically avoid any secondary radioactive damage of the population. Today, emergency management systems are an integral part of the safety culture in the area of nuclear activities. This is reflected in national and international research activities. In particular harmonization on the European level was and is an important aspect of these activities. KIT is and was always involved in international projects having coordinated the European project EURANOS (European approach to nuclear and radiological emergency management and rehabilitation strategies) and is coordinating the European project NERIS-TP (Towards a self sustaining European Technology Platform on Preparedness for Nuclear and Radiological Emergency Response and Recovery). The objective of the latter on is to close those gaps being identified at the end of the EURANOS project. One of the activities not covered by the two projects mentioned above is the optimisation of monitoring networks, an important task nowadays, because many monitoring systems installed following the Chernobyl accident had to be revised. A further important aspect of our activity is to bring both the nuclear and non-nuclear world together, thus learning on the one hand side from conventional emergency management and on the other side use our experience in operational decision support systems in the conventional area. KIT performed research in the following areas:  Development of a planning tool to optimise the design of monitoring networks in European member States

 Adaptation of the RODOS (Real-time On-line DecisiOn Support) system to the new ICRP recommendations on emergency management (ICRP-103)

 Coupling of the RODOS system to early notification messages from the IAEA

 Application of RODOS world wide

 Design of an integrated decision support system for large scale crisis events with affected critical infrastructures

Within the European project DETECT (Design of optimised systems for monitoring of radiation and radioactivity in case of a nuclear or radiological emergency in Europe) a planning and optimisation tool has been developed that allows the end users to test and develop environmental radiological monitoring strategies for their specific needs. The DETECT Optimisation Tool (DOT) is based on a comprehensive library of simulations of radioactive plumes from 64 sources (threats) in Europe that were identified to be most important by the users. The simulations cover whole Europe, so the tool allows evaluation and optimisation for all EU countries as well as evaluation of fencing sensors around the sources. The ICRP 103 recommendations, issued 2007 change the way how to define countermeasure strategies. The new concept requires an integrated treatment of all exposure pathways for accidental and existing exposure situations thus differing considerably from the existing concept of single exposure pathways resulting in actions such as sheltering, evacuation and distribution of stable iodine tablets. Within the NERIS-TP project, we have extended the simulation models of RODOS with a so called screening model that allows defining integrated countermeasure strategies with the ultimate target of a residual dose value in the range of 20 to 100 mSv in the first year.

11 As part of the NERIS-TP project, the RODOS system will be coupled to the early notification system of the IAEA allowing to estimate the consequences of a potential or indicated release automatically. For this purpose, RODOS has to be extended to allow operation everywhere in the world. This will be achieved by connecting freely available numerical weather prediction data from the Global Forecast Model System GFS to the RODOS system. In between, the WRF module will allow the nesting from global model to the local scale of interest. Work has been started in 2011 forced by the Fukushima incident. KIT has collected the necessary data sets for Japan and installed the meteorological forecast chain GFS and WRF together with the support of the “Wettergefahren Frühwarnzentrale”. This allowed KIT to perform daily assessments of the local radiological situation, based on source terms provided by GRS in Cologne, see e.g. Fig. 9.

Fig. 9: RODOS result: Contamination of Cesium 137 in Bq/m2

Summary

The understanding of major processes and the resulting numerical safety analysis codes for the assessment of the plant response and behaviour under design basis or beyond design basis situations still have to be further developed. Within SARNET (Severe Accident Research Network of Excellence) and the follow–up SARNET 2 a European consensus on a limited number of specific open issues in this field was obtained. These remaining topics include corium coolability, hydrogen mixing and combustion in the containment, and aerosol behaviour in the containment. KIT with partly unique expertise will continue to carry out experimental investigations to answer these remaining open issues of thermal-hydraulics and physico- chemical phenomena during the accident situations to further improve the simulation tools such as ASTEC and MELCOR and to derive appropriate science-based countermeasures.

The decision support system RODOS for off-site emergency management is installed in more than 10 European countries. The development and the operational use resulted in a large competence in decision support at KIT – on the nuclear sector. Conventional large crisis events are also challenging for decision making, in particular in case of critical infrastructures are affected. Based on our experience with RODOS, we intend to develop the design of an integrated decision support system for large crisis events. This activity is supported by projects running under the BMBF supported “Security Research” Topic and will be supported further by the Cross Cutting Activity “Security” of the Helmholtz Association. It is expected to have a first version of the software design and methodologies used by end of 2014. However, a final product cannot be expected with the resourced that can be allocated to that topic. Nevertheless, decision support for large crisis events with affected critical infrastructures is an extremely challenging but also extremely important topic as our society is becoming more and more vulnerable when Critical Infrastructures fail.