ALISA Project. Access to Large Infrastructures for Severe Accidents in Europe and in China

Transcription

1 ALISA Project. Access to Large Infrastructures for Severe Accidents in Europe and in China S. ZHANG 1, A. MIASSOEDOV 2, Y. LIAO 1, J.-M. RUGGIERI 3, F. SHEN 4, X. GAUS-LIU 2 1 CNPRI, Shenzhen (CN) 4 SNPTC, Beijing (CN) 2 KIT, Karlsruhe (DE) 3 CEA, Cadarache (FR) ABSTRACT Enhanced transnational access to large research infrastructures in Europe and in China is offered to allow the optimal use of the resources in the complex field of severe accident analysis for the existing power plants. This research involves very substantial human and financial resources and, in general, the research field is too wide to allow investigation of all phenomena by any national programme. To optimise the use of the resources, the collaboration between nuclear utilities, industry groups, research centres, safety authorities and technical safety organisations at both European and Chinese levels is very important. This is precisely the main objective of the ALISA project (Access to Large Infrastructures for Severe Accidents), which aims to provide these resources and to facilitate this collaboration by providing large scale experimental platforms in Europe and in China for transnational access. Large-scale facilities offered for access in ALISA project are designed to resolve some of the most important remaining severe accident safety issues, ranked with high or medium priority by the SARP group for SARNET NoE. Among others, these issues are coolability of a degraded core, corium coolability in the reactor pressure vessel, possible melt dispersion to the reactor cavity, molten corium concrete interaction and hydrogen mixing and combustion in the containment. The major aspect is to understand how these events affect the safety of existing and advanced reactors and how to deduce soundly-based accident management procedures. Activities within ALISA project will focus on the large scale experiments under prototypical conditions addressing the remaining R&D issues on severe accident management in light water reactors. ALISA offers a unique opportunity for European and Chinese research organisations to get involved in the networks and activities supporting safety of existing and advanced reactors and it allows European and Chinese researches to work together as equal partners so as to arrive at a wider common safety culture. 1 INTRODUCTION Severe accidents can cause significant damage to reactor fuel resulting in more or less complete core meltdown and threaten the containment integrity. They are the focus of considerable research, because the release of radioactive products into the environment would have serious consequences. This research also reflects a commitment to the defence-in-depth approach. As stated in the Strategic Research Agenda (SRA) [1] of the Sustainable Nuclear Energy Technology Platform (SNETP), needs for safety research are identified by both regulators and operators, from their respective perspective. As discussed in the SRA, safety research is still needed to support long-term operation of existing LWRs in Europe. Though the SRA can not integrate all national programmes on safety research carried out in Europe, the platform members agree on the issues that are of highest priority. Regarding the issues in severe accidents, the SRA refers to the work carried out in the 1/11 pages

2 framework of the Severe Accident Research Network of Excellence (SARNET) [2], [3] to conclude to a common view on the ranking of the research priorities in the field. The research priorities on severe accident management [4] were prepared by the SARNET SARP group. The objective of this group was to review and reassess the priorities of research issues and to propose the results as basis to harmonise and to re-orient research programmes, to define new ones, and to close if possible resolved issues on a common basis. The group obtained a consensus on six high priority issues on severe accident management on which research was considered as necessary. These issues are: - Core coolability during reflood and debris cooling; - Ex-vessel melt pool configuration during Molten Corium Concrete Interaction (MCCI), ex-vessel corium coolability by top flooding; - Melt relocation into water, ex-vessel Fuel Coolant Interaction (FCI); - Hydrogen mixing and combustion in containment; - Oxidising impact (Ruthenium oxidising conditions/air ingress for High Burn-up and Mixed Oxide fuel elements) on source term; - Iodine chemistry in Reactor Coolant System (RCS) and in containment. These phenomena are extremely complex; they generally demand the development of specific research, therefore optimised use of recourses and the collaboration at international level is very important. The main thrust of ALISA project is towards large scale tests under prototypical conditions. These will help the understanding of core degradation, melt formation and relocation as well as core coolability in real reactors by scaling-up and by providing data for the improvement and validation of computer codes applied for safety assessment and planning of accident mitigation concepts, such as ASTEC. The importance of the ALISA project for the European and Chinese severe accident research is reflected in three aspects: 1) The access to large scale experimental facilities is proposed to investigate most important processes from the early core degradation to late in-vessel phase pool formation in the lower head, continuation to ex-vessel melt situations and to the hydrogen behaviour in the containment. Therefore four high priority issues identified by the SARP group will be addressed in the project. 2) The results of the project will be applicable both to the European and Chinese reactor fleet taking into account the main LWR types. 3) The project offers a unique opportunity for Chinese experts to get an access to large scale facilities in Western research organisation and vice versa, to improve understanding of material properties and core behaviour under severe accident conditions, and to become familiar with the high level safety concepts in nuclear power plants. As for now, no facilities for investigation of corium behaviour with prototypic materials exist in China. From the other hand, no large-scale experimental facilities on the in-vessel retention, especially on the behaviour of water at the outer side of the RPV, exist in Europe. Therefore experiments in these facilities are of a great interest for both Chinese and European organisations. This will strengthen the scientific excellence and improve use of recourses and the collaboration at international level. 2 ALISA EXPERIMENTAL FACILITIES The experimental facilities offered in ALISAfor the transnational access are unique in providing the ability to perform experiments in specific fields of severe accident management. The facilities are operated by a team of experts who have long and recognised experience in the nuclear safety research. The experiments are designed to be 2/11 pages

3 complementary to other experimental platforms (e.g. CODEX) and projects (e.g. SARNET2), to form a coherent nuclear experimental network. They contribute to a better understanding of core melt sequences and thus improve safety of existing and, in the longterm, of future reactors. For Chinese research organisations access will be offered to two experimental platforms in Europe: LACOMECO at KIT and PLINIUS at CEA [5]. The LACOMECO experimental platform at KIT includes the following facilities: - QUENCH facility [6] is the only operating experimental facility in EU for investigations of the early and late phases of core degradation in prototypic geometry for different reactor designs and different cladding alloys, incl. analysis of the relocation of cladding and fuel and the formation and cooling of in-core debris beds to gain information on the characteristics of the created particles. - LIVE facility [7] concentrates on the investigation of the whole evolution of the invessel late phase of a severe accident, including e.g. formation and growth of the incore melt pool, characteristics of corium arrival in the lower head, and molten pool behaviour after the debris re-melting in large scale 3D geometry with emphasis on the transient behaviour. - HYKA experimental facilities [8] are among the largest available in the world. In combination with the high static and dynamic pressures the experimental facilities are designed for, a unique experimental centre especially for combustion experiments in confined spaces is available with HYKA. Due to the different orientations and sizes the set of large and strong experimental vessels offers a flexible basis for scientific experimental work on reactive hydrogen mixtures, so that a wide range of gas mixtures in prototypic conditions as expected in a severe accident can be studied. - DISCO is the only operating facility [9] available worldwide for integral DCH investigations. It is designed to perform scaled experiments that simulate melt ejection from the RPV to the reactor cavity after the RPV failure under low system pressure during severe accidents in LWRs. These experiments investigate the fluid-dynamic, thermal and chemical processes during melt ejection out of a breach in the lower head of an LWR pressure vessel (RPV) at pressures below 2 MPa. The PLINIUS experimental platform at CEA is composed of four facilities devoted to the corium behaviour and to physical properties studies: - VULCANO is a rotating plasma arc furnace able to melt about 80 kg of corium at temperatures of up to 3000 C (in- or ex-vessel corium) and to pour the melt according to different configurations: spreading, interaction, solidification, studies, etc [10]. - COLIMA is a 1.5 m 3 controlled atmosphere vessel [5], with an internal pressure which can rise to 0.3 MPa. Induction heating can maintain some kilograms of corium at very high temperature (up to 3000 C) to study the thermal exchanges, aerosol release and the thermo-physical interaction studies. - VITI has been developed to perform viscosity and surface tension measurements on corium by aerodynamic levitation up to 2500 C [11]. Samples of a few milligrams of corium can be implemented. It is also used for small mass experiments for properties estimation or material compatibility tests. - KROTOS facility [12] is dedicated to steam explosion phenomenon studies. About 5 kg of corium at more than 2850 C are dropped in water. Thermal, optical and pressure instrumentation, along with fast imaging, constitute the instrumentation. The KROTOS facility for corium-water interaction tests has been transferred from JRC Ispra (Italy) and has been rebuilt at CEA Cadarache within the PLINIUS platform. 3/11 pages

4 For European research organisations access will be offered to the following experimental facilities in China: - IVR Test Facility (CNPRI). The in-vessel retention (IVR) test facility of CNPRI was designed to study coolability of corium inside the lower head using external reactor vessel cooling. In-vessel retention of core melt is a severe accident mitigation measure adopted by the CGNPC CPR1000+ nuclear power plants. To achieve this goal and assist in the licensing application, design modifications will be incorporated into CPR1000+ prior to safety regulators allowing external reactor vessel cooling as part of the licensing basis. The facility provides a full-height two-dimensional simulation of the down-ward facing boiling process on the outer surface of a hemispherical reactor vessel lower head using a two-dimensional metal slice with independently heated zones. The local critical heat flux along with its dependence on the surface orientation will be experimentally obtained using the IVR test facility. In addition, the test facility will be used to study the impact of several relevant CPR1000+ design modifications, for examples, study on flow path formed by vessel external surface and insulation to ensure that there is sufficient flow area to allow water ingress and steam egress, and on strength of insulation to ensure that insulation can withstand loads expected during reactor cavity flooding in severe accidents. The IVR test facility has been commissioned and the heat balance tests have been performed. The tests on critical heat flux have been performed in the middle of 2011 [13]. - WAFCORE (SJTU). This facility is dedicated to study passive containment cooling system (PCCS) of GEN-III plants. WAFCORE consists of a stainless plate (5m long, 2m wide) mounted on a metallic frame allowing for plate inclination. The surface of the plate undergoes a preparation by painting with organic zinc to maintain good wettability. The plate flat is heated by 16 electric heaters, the power of which is controlled independently by the voltage regulator. A visualization window is mounted parallel to the stainless steel plate, about 5~20 cm apart from it, in order to form a rectangular channel simulating the air gap existing between the containment steel envelope and the metallic baffles in the reference passive reactor design. A blower located at the bottom of the channel, simulates the natural circulation air flow occurring in the reference plant conditions. Variation of the reference air velocity approximately from 0 to 12 m/s is allowed. The water film distribution tank is mounted on the top of the facility to ensure the uniform water film distribution of the inlet. - WAFSEA (SJTU). Similarly to WAFCORE, this facility is dedicated to study PCCS of GEN- III plants. It consists of a stainless plate (5m long, 2m wide) fixed vertically on the ground. A visualization window made of Plexiglas is 20 cm apart from the stainless steel plate in parallel. The temperature and the mass flow rate of the water film are controlled by the preheater and the flowmeter. Essential measurements for this PCCS experimental facility are available, including the measurement of the water film thickness. Both facilities have been and are being applied to study mixed convection heat transfer and water film characteristics under various conditions, such as inclination, heat flux, inlet conditions and surface conditions. In addition, further investigations are ongoing in the frame of a national key project. - REPEC facilities (SJTU). REPEC is the first test facility and research project in China studying external reactor vessel cooling (ERVC) related phenomena under in-vessel retention (IVR) strategy for severe accident mitigation. The project as well as test facilities are divided into two phases, namely, REPEC-I and REPEC-II. REPEC-I facility is erected in the year of 2009, which has been successfully applied for preliminary investigation and understanding on ERVC flow boiling characteristics, natural circulation ability in the channel between RPV and the thermal insulation, as well as corresponding experimental techniques. REPEC-II facility is now under construction based on the data, features and experience acquired from REPEC-I. REPEC-II is of full- 4/11 pages

5 height. It has also been designed on the basis of corresponding integrated, hierarchical scaling analysis results, according to the facility function. The main object is to study in detail feasibility of IVR-ERVC strategy and its behavior in specific large-scale reactors. REPEC-II facility is functionally designed and expected to investigate limits of external cooling, i.e. CHF distribution on RPV wall, under various non-uniform melt heating conditions, flow characteristics within specific natural circulation channel (including flow instabilities), influences of real plant and severe accident conditions on ERVC abilities (e.g. considering such factors as RPV surface conditions, chemistry of cooling water, dynamic load on thermal insulation, etc.), along with possible, practical enhanced heat removal approaches. REPEC-II facility will be commissioned in the middle of 2011 and the first test report will be available by the end of the year. - MELAHT(SNPTC). Metal layer heat transfer characteristic experiment facility is built to verify the correlations used in metal layer and to obtain the appropriate correlations in the real plant condition with high Ra number. The experiment will use water instead of liquid metal as working substance. The total experimental facility includes test section, recirculating cooling water system, electricity supply, measuring instrument, control console and so on. The heating power of the test section is 2-25kW/m 2, and the range of the Ra number in the experiment is COPAD (XJTU). This facility can be used to investigate the coolability and capillary effect of particulate core-metal debris bed [14]. An electrically heated SS304 block is employed to simulate the decay heat. The block is put on the particulate debris bed deposited in a water layer keeping a constant depth. The heights of the debris bed are higher than the depth of the water layer. The temperature distribution in the heated block is measured to evaluate the heat flux. The influence of particle size, height of particle debris bed and water level on coolability and capillary can be profoundly studied. - GADOC (XJTU). The facility can be used to investigate the natural flow and heat transfer behaviour of water in narrow spherical or curving gap with downward-facing circular heated surface. The heat source was modelled using a metal plate which is directly heated by DC power supply system. Quartz glass is fabricated in GADOC facility so that the bubble dynamic, co-current flow limitation, local dry-out phenomenon can be visually investigated. The ALISA project starts on July 1, 2012 and has duration of 48 months. The project includes two work packages. - WP1 SAM: Severe accident management, and - WP2 PALI: Portal to access to large and unique infrastructures The timing of the work packages and the related tasks is illustrated in Figure 1. Tentative dates for the experiments are also given. The project will start with a nine months preparation phase. One of the important tasks during this period will be preparation of rules of access to the ALISA facilities. This document will contain technical description of all facilities included in the ALISA project, sample application form for submission of proposal and description of services offered by facility providers. It will be available for download on the project website and will be used as a guide for user groups interested in participation in the ALISA project. After publishing the rules of access to the ALISA facilities, three calls for proposals will be widely announced attracting the interested users to specify the experimental requirements and conditions. Moreover, further active participants who have enough interest in particular tests and agree to provide funding to perform the experiments within the ALISA project will be also invited. 5/11 pages

6 Following each call for proposals, the User Selection Panel will evaluate the proposals and select a short-list of user groups that should benefit under this call for proposals. Together with facility operators the user groups will prepare, perform, analyse and document the experiments. WP No.: Activity Year 1 Year 2 Year 3 Year 4 Q1 Q2 Q3 Q4 Q1 Q2 Q3 Q4 Q1 Q2 Q3 Q4 Q1 Q2 Q3 Q4 WP2: Project preparation WP2: 1 st call for proposals WP1: Experiments WP2: 2 nd call for proposals WP1: Experiments WP2: 3 rd call for proposals WP1: Experiments WP2: Evaluation of proposals 1 st User Selection Panel Meeting 2 nd User Selection Panel Meeting 3 rd User Selection Panel Meeting All WPs: Final report All WPs: Review meetings 1 st review meeting 2 nd review meeting Final project meeting Preparation of experiments Performance and analysis of experiments Preparation of experimental reports Figure 1: Timescale and milestones of the ALISA project. 3 OBJECTIVES OF THE EXPERIMENTS 3.1 WP1 SAM: Severe accident management At present, knowledge of various core melt sequences and the consequences of possible operator actions are not yet sufficient as they are too dependent on specific characteristics of the power plant under consideration. ALISA aims to provide the resources for a better understanding of possible scenarios of core quenching, different core melt sequences, molten corium-concrete interaction and hydrogen behaviour for different reactor designs in Europe and in China. The aim is not only to understand the physical background of severe accidents but to provide the underpinning knowledge that can help to reduce the severity of the consequences thus leading to improved severe accident management measures, which are essential for reactor safety. This work-package is divided into three sub-wps, each one addressing a specific issue: WP1.1 In-vessel corium coolability. Core coolability during reflood, corium debris cooling and behaviour of the stratified corium melt pool in the lower head are still critical issues in understanding of PWR core meltdown accidents. They were ranked as high and medium priority issues by the SARP group of the SARNET NoE. The major motivation is to reduce or possibly solve the remaining uncertainties on the possibility of cooling structures and materials during severe accidents, either in the core or the vessel bottom head or in the reactor cavity, so as to limit the progression of the accident. This could be achieved either by ensuring corium retention within the vessel or at least a slow corium progression and small flow rates of corium release into the cavity. The experiments will provide important information on 6/11 pages

7 - main factors governing the quantity of hydrogen production and melt formation during quenching of an overheated rod bundle, - physico-chemical behaviour of overheated fuel elements under different reflood conditions, - formation and stability of melt pools in the core region, - mechanisms of corium arrival to the lower head and corium interaction with lower head structures, - melt pool behaviour in the lower head, - heat flux distribution along the reactor pressure vessel wall in transient and steady state conditions, - thickness and stability of the melt crust depending on the power density and the external cooling mode and its influence on the transient and steady-state heat fluxes through the vessel wall, - retention of the corium melt in the lower head of the reactor pressure vessel. WP1.2 Ex-vessel corium coolability. In the case of a severe accident with vessel meltthrough, the containment is the ultimate barrier between the corium and the environment. This sub-wp will address the situation where the reactor pit is initially dry but water injection may occur later during MCCI. This issue was ranked as high priority by the SARP group of the SARNET NoE. The major objective of this sub-wp is to provide new data and understanding of corium-concrete interaction, by using innovative approaches to address more prototypical phenomena and to achieve significant progress toward the closure of this issue closure will be very significant. The experiments will provide important information on - the effect of concrete nature on 2D ablation, in view of achieving predictive modelling of the partition between lateral and vertical concrete ablation, - the role of metallic layer on the molten corium-concrete interaction, - scaling and geometry effects. WP1.3 Containment behaviour. The issue considered in this sub-wp is the threat to the containment integrity due to direct containment heating, ex-vessel fuel-coolant interaction, hydrogen combustion and study of passive containment cooling systems for power plants. These issues are ranked as high and medium priority issues by the SARP group of the SARNET NoE. Steam explosions may be caused by ex-vessel fuel-coolant interaction due to a reactor pressure vessel failure and pouring of the reactor core melt in the flooded reactor cavity. Hydrogen combustion (deflagration and detonation) may be caused by ignition of a gas mixture with high local hydrogen concentrations, which may be due to the imperfect mixing of the containment atmosphere. In case of accidents, water in the storages tank above the containment flows down by the gravity to the outer surface of the inside steel containment, forms water film and removes heat via water film evaporation. In addition, natural convention of air and thermal radiation heat transfer happen in the annular gap. Therefore, a full understand of different effects such as air mixed convection, water film flow and heat transfer is essential. Essential insights and results from the experiments performed in this sub-wp should be applicable to different reactor designs in Europe and in China. The experiments will provide important information on - clarification of basic FCI phenomenal like melt jet fragmentation, void generation, fine fragmentation, - final distribution of corium debris in the containment, - loads on the reactor pit and the containment with respect to pressure and temperature, - fast deflagration and detonation at high initial pressures, - behaviour of hydrogen jets in mixed air-steam-hydrogen atmospheres, 7/11 pages

8 - influence of the combustion of hydrogen produced by oxidation during melt ejection from RPV on the containment pressurisation, - evaporation heat transfer characteristics, - distribution of the water film thickness. 3.2 WP2 PALI: Portal to access to large infrastructures The main objective of this work package is to provide the necessary support for the overall project management. It will cover all administrative, legal and financial issues of the project: - facilitate collaborative working, - establish clearly defined links with existing Euratom projects like SARNET2, - contact with EC and CAEA, - follow-up of contractual obligations, - keeping administrative and financial accounting records, - control of budget, reimbursement of costs, - monitor the progress of the project and propose corrective actions in order to reach the ALISA objectives. - prepare and issue Coordination Agreement between the European and Chinese organisations, - prepare and issue the rules of access to the experimental facilities, - prepare and publish the calls for proposals for the project, - organise the User Selection Panel meetings providing a technical knowledge of the facility, - coordinate the work programme and monitor the progress within the work packages, - review and approve the release of the contractually required reports and deliverables and survey milestones, - organise the project review meetings, - organise technical meetings and prepare the minutes of the meetings, - anticipate and examine possible difficulties in the execution of the project, - elaborate and update budget of the project, - set-up and update the project website, the address lists, etc. - disseminate the results to the interested EU and CN organisations and to the public. 4 ALISA PROJECT PARTNERS The ALISA consortium includes six organizations: Karlsruhe Institute of Technology (KIT, Germany): KIT is one of the largest science and engineering research institutions in Europe, funded jointly by the Federal Republic of Germany and the State of Baden-Württemberg. It has a profound knowledge and experience in R&D of nuclear reactor safety and in national and international co-operation. It has long experience in the performance of large- and small-scale experiments in the specific fields of core damage initiation and progression such as the in-vessel and ex-vessel melt progression phenomena and hydrogen behaviour in the containment. A unique combination of facilities and expertise is available at KIT. The expertise is maintained by participating in collaborative research with laboratories in the EU Member and FP7 Associated States and other countries. It is considered as a powerful internationallyrecognised centre of reference for nuclear research. Commissariat à l'energie Atomique et aux Energies Alternatives (CEA, France): CEA is a science and engineering research institution, funded by the French Government and industrial contractors. The CEA is organized in four research and development sectors: nuclear energy (DEN), technological research (DRT), fundamental research (DSM and DSV) and defence (DAM). The nuclear energy programs are devoted to the support for nuclear power stations in operation, design of systems for the future (in particular Gen-IV 8/11 pages

9 reactors), studies for waste management and dismantling of obsolete installations. CEA has recognized and long scientific and technological experience in the field of severe accident studies. The PLINIUS platform lies within the Nuclear Energy Division, Nuclear Technology Department, Service of industrial reactor technologies, Severe Accident mastering Laboratory (DEN/DTN/STRI/LMA) at the CEA Cadarache site. It is devoted to the study of the behavior of prototypical corium. China Nuclear Power Technology Research Institute (CNPRI, China):CNPRI is an engineering application and research oriented institute with 1,000 researchers. CNPRI is one of the subsidiary companies and the technology center of China Guangdong Nuclear Power Group. The headquarters of CNPRI is in Shenzhen. The research of CNPRI covers nuclear and conventional power technologies. CNPRI s research focuses on reactor core design, advanced fuel management, safety analysis, source term calculation and evaluation, environment impact evaluation, severe accident analysis, nuclear fuel cycle research, probability safety analysis, nuclear accident emergency management, and many others. CNPRI is a professional platform of nuclear power technology innovation and services, integrates thirty-year nuclear power experiences and ability of CGNPC and now is promoting development of China nuclear power. CNPRI s severe accident branch has poured huge and long-term manpower and material resources into the research and management of severe accident, has trained a R&D team which has numerous experience on research and engineering, has successfully developed and implemented the severe accident management guidance on Daya Bay and LingAo phase 1 NPP, completed the analysis of severe accident management strategy for HongYanHe, NingDe and YangJiang NPPs, completed the preliminary design of the passive hydrogen removal system for CPR1000, and is now responsible for several design modifications to be incorporated into CGNPC CPR1000+ nuclear power plants and for IVR critical heat flux experiments to provide the licensing basis of external reactor vessel cooling for CPR Shanghai Jiao Tong University (SJTU, China): SJTU is one of the elite universities in China and the only university in South-eastern China with a department of nuclear engineering, which was established in The School of Nuclear Science and Engineering (SNSE) was established in 2006, including three departments, i.e. Department of Reactor Engineering, Department of Nuclear Material & Fuel, and Department of Radiation Protection & Nuclear Technology Application. State Nuclear Power Technology Corporation (SNPTC, China): SNPTC is a state-owned corporation, dedicated to leading the development of nuclear power industry with innovations, and by starting from AP1000, a state-of-the-art nuclear power technology, promoted the development of China s nuclear power industry. In the field of nuclear energy research and development, some experimental facilities, codes development, codes verification and validation are proposed by SNPTC and supported by China s government. Xi an Jiao Tong University (XJTU, China): XJTU is one of the most important universities in China with the department of nuclear engineering which was established in The School of Nuclear Science and Technology are devoted to the education and research of Nuclear Engineering. There are two departments in the school, one is the department of Nuclear Engineering and the other is Nuclear Technology and application. Four research organisations and two universities are participating actively in ALISA, providing the necessary technological competence and scientific excellence to successfully fulfil the objectives of the project. Some of the ALISA partners are covering a wide range of competence though not complete, whereas others are specialized in very specific areas and thus complementarities are developing. Overall, it is estimated that the critical mass of competence for performing and analysing experiments needed in the severe accident research domains for most NPP types in Europe and in China is met. 9/11 pages

10 5 CONCLUSIONS Enhanced transnational access to large research infrastructures will allow the optimal use of the R&D resources in Europe and in China in the complex field of severe accident analysis for existing and future power plants. This research involves substantial human and financial resources and, in general, the research field is too wide to allow investigation of all phenomena by any national programme. To optimise the use of the resources, the collaboration between nuclear utilities, industry groups, research centres and safety authorities, at both European and Chinese levels is very important. This is precisely the main objective of the ALISA project, which aims to provide these resources and to facilitate this collaboration by providing state-of-the-art large-scale experimental platforms in Europe and in China for transnational access. Large-scale facilities of the ALISA project are designed to resolve the most important remaining severe accident safety issues, ranked with high or medium priority by the SARP group for SARNET NoE. These issues are coolability of a degraded core, corium coolability in the RPV, possible melt dispersion to the reactor cavity, molten corium concrete interaction and hydrogen mixing and combustion in the containment. The major aspect is to understand how these events affect the safety of existing reactors and how to deduce soundly-based accident management procedures. The aim is not only to understand the physical background of severe accidents but to provide the underpinning knowledge that can help to reduce the severity of the consequences. It is crucially important to understand the whole core melt sequences and identify opportunities to lower the risk. This can be done by: - altering the timing or magnitude of reflooding the degraded core, - in-vessel melt retention in the lower plenum of the reactor pressure vessel, - ensuring the upper bound of system pressure at vessel failure by dedicated depressurisation valves, - installation of devices or implementing accident management procedures to mitigate melt dispersion into the containment, - delaying basement penetration or even by regaining melt cooling before containment failure or basement penetration occur, - implementing hydrogen mitigation measures in the containment (ignitors, recombiners etc.). At present, knowledge of various core melt sequences and the consequences of possible operator actions are not yet sufficient as they are too dependent on specific characteristics of the power plant under consideration. ALISA aims to provide the resources for a better understanding of possible scenarios of core quenching, different core melt sequences and hydrogen behaviour for different reactor designs by offering state-of-theart experimental facilities for the transnational access. This knowledge shall lead to improved severe accident management measures, which are essential for reactor safety and in addition offer competitive advantages for the nuclear industry. The experimental results will be used for the development of models and their implementation in the severe accident codes such as ASTEC. This helps to capitalise the knowledge obtained in the field of severe accident research in the severe accident codes and scientific databases, thus preserving and diffusing this knowledge to a large number of current and future end-users both in Europe and in China. Due to strong links to other European projects, ALISA offers a unique opportunity for all partners to get involved in the networks and activities supporting safety of existing and advanced reactors and to get access to large-scale experimental facilities in Europe and in China to enhance understanding reactor core behaviour under severe accident conditions. The ALISA project allows European and Chinese researches to work together as equal partners so as to arrive at a wider common safety culture. By doing this, the project will 10/11 pages

11 also help to diffuse the European culture of research and innovation as a support to regulatory and industrial decision making both in Europe and in China. ACKNOWLEDGEMENTS The authors gratefully acknowledge funding by Euratom and CAEA to support the work within ALISA project. REFERENCES [1] SNETP: Sustainable Nuclear Energy Technology Platform, Strategic Research Agenda, May [2] T. Albiol et al., SARNET: Severe accident research network of excellence, Progress in Nuclear Energy Volume 52, Issue 1, January 2010, pp [3] Jean-Pierre Van Dorsselaere et al., Status of the SARNET network on severe accidents, Proceedings of the 2010 International Congress on Advances in Nuclear Power Plants, ICAPP'10: embedded topical meeting, June 13-17, 2010, San Diego, California, LaGrange Park, American Nuclear Society, 2010, pp [4] B. Schwinges et al., Ranking of severe accident research priorities, Progress in Nuclear Energy Volume 52, Issue 1, January 2010, pp [5] A. Miassoedov et al., LACOMECO and PLINIUS Experimental Platforms at KIT and CEA, 4 th European Review Meeting on Severe Accident Research (ERMSAR-2010), Bologna (Italy), May 11-12, [6] M. Steinbrück, M. Große, L. Sepold, J. Stuckert, Synopsis and outcome of the QUENCH experimental program, Nucl. Eng. Design 240, 2010, pp [7] A. Miassoedov, T, Cron, X. Gaus-Liu, A. Palagin, S. Schmidt-Stiefel, T. Wenz, LIVE experiments on melt behavior in the reactor pressure vessel lower head. 8th Internat. Conf. on Heat Transfer, Fluid Mechanics and Thermodynamics (HEFAT 2011), Pointe aux Piments, MS, July 11-13, [8] T. Jordan, W. Breitung, FZK methodology for analysis of hydrogen behaviour in containments, Proc. of Conf. on Numerical Flow Models for Controlled Fusion, Porquerolles, France, April 16-20, [9] L. Meyer, G. Albrecht, C. Caroli, I. Ivanov, Direct containment heating integral effects tests in geometries of European nuclear power plants, Nuclear Engineering and Design, Vol. 239, 2009, pp [10] C. Journeau et al., European Experiments on 2D Corium-Concrete Interaction: HECLA and VULCANO, Nuclear Technology, 170, 2010, pp [11] Piluso, J. Monerris, C. Journeau, G. Cognet, Viscosity measurements of ceramic oxides by aerodynamic levitation, Int. J. Thermophys, 23,2002, pp [12] I. Huhtiniemi, D. Magallon, Insight into Steam Explosions with Corium Melts in KROTOS, Nuclear Engineering and Design, 204, 2001, pp [13] X. Chen, H. Zhang, J. Zhang, S. Zhang, J. Lin, Study on in-vessel retention(ivr) strategy for CPR1000, Proc. of 14th International Topical Meeting on Nuclear Reactor Thermalhydraulics, NURETH-14, Toronto, Canada, September 25-30, [14] G. Su, K. Sugiyama, H. Aoki, I. Kimura, Experimental study on coolability of particulate core-metal debris bed, with Oxidization, (II) Fragmentation and enhanced heat transfer in zircaloy debris bed, Journal of Nuclear Science and Technology, 43, 2006, pp /11 pages