| 1. |
- Almjashev, V.I., et al.
(författare)
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Ternary eutectics in the systems FeO-UO2-ZrO2 and Fe2O3-U3O8-ZrO21
- 2011
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Ingår i: Radiochemistry. - 1066-3622. ; 53:1, s. 13-18
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Tidskriftsartikel (refereegranskat)abstract
- The systems FeO–UO2–ZrO2 (in inert atmosphere) and Fe2O3–U3O8–ZrO2 (in air) were studied. Forthe FeO–UO2–ZrO2 system, the eutectic temperature was found to be 1310°С, with the following componentconcentrations (mol %): 91.8 FeO, 3.8 UO2, and 4.4 ZrO2. For the Fe2O3–U3O8–ZrO2 system, the eutectictemperature was found to be 1323°С, with the following component concentrations (mol %): 67.4 FeO1.5,30.5 UO2.67, and 2.1 ZrO2. The solubility limits of iron oxides in the phases based on UO2(ZrO2,FeO) andUO2.67(ZrO2,FeO1.5) were determined
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| 2. |
- Bechta, Sevostian, et al.
(författare)
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Interaction between molten corium UO2+X-ZrO2-FeO y and VVER vessel steel
- 2009
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Ingår i: Proceeding of International Conference on Advances in Nuclear Power Plants, ICAPP 2008. - : Curran Associates, Inc.. - 9781605607870 ; , s. 210-218
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Konferensbidrag (refereegranskat)abstract
- In case of an in-vessel corium retention (1VR) the deterioration of vessel steel properties can be caused both by the steel melting and by its physicochemical interaction with corium. The interaction behavior has been studied in the medium-scale experiments with a prototypic corium within the METCOR project. The resulting experimental data give an insight into the steel corrosion during its interaction with U02+x- Zr02- FeOy melt in air and steam. It has been observed that the corrosion rate is almost the same in air and steam atmosphere; if the temperature on the interaction interface increases beyond a certain level, corrosion intensifies, which is explained by the formation of liquid phases in the interaction zone. The available experimental data have been used for developing a correlation of corrosion rate versus temperature and heat flux.
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| 3. |
- Bechta, Sevostian, et al.
(författare)
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INTERACTION BETWEEN MOLTEN CORIUM UO2+x-ZrO2-FeOy AND VVER VESSEL STEEL
- 2010
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Ingår i: Nuclear Technology. - : American Nuclear Society. - 0029-5450 .- 1943-7471. ; 170:1, s. 210-218
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Tidskriftsartikel (refereegranskat)abstract
- In case of in-vessel corium retention during a severe accident in a light water reactor, weakening of the vessel wall and deterioration of the vessel steel properties can be caused both by the melting of the steel and by its physicochemical interaction with corium. The interaction behavior has been studied in medium-scale experiments with prototypic corium. The experiments yielded data for the steel corrosion rate during interaction with UO2+x-ZrO2-FeOy melt in air and steam at different steel surface temperatures and heat fluxes from the corium to the steel. It has been observed that the corrosion rates in air and steam atmosphere are almost the same. Further, if the temperature at the interface increases beyond a certain level, corrosion intensifies. This is explained by the formation of liquid phases in the interaction zone. The available experimental data have been used to develop a correlation for the corrosion rate as afunction of temperature and heat flux.
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| 4. |
- Bechta, Sevostian, et al.
(författare)
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VVER vessel steel corrosion at interaction with molten corium in oxidizing atmosphere
- 2009
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Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 239:6, s. 1103-1112
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Tidskriftsartikel (refereegranskat)abstract
- The long-term in-vessel corium retention (IVR) in the lower head bears a risk of the vessel wall deterioration caused by steel corrosion. The ISTC METCOR Project has studied physicochemical impact of prototypic coria having different compositions in air and steam and has generated valuable experimental data on vessel steel corrosion. It is found that the corrosion rate is sensitive to corium composition, but the composition of oxidizing above-melt atmosphere (air, steam) has practically no influence on it. A model of the corrosion process that integrates the experimental data, is proposed and used for development of correlations.
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| 5. |
- Khabensky, V. B., et al.
(författare)
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Effect of temperature gradient on chemical element partitioning in corium pool during in-vessel retention
- 2018
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Ingår i: Nuclear Engineering and Design. - : Elsevier Ltd. - 0029-5493 .- 1872-759X. ; 327, s. 82-91
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Tidskriftsartikel (refereegranskat)abstract
- The paper presents some results of the ISTC (International Science and Technology Center)-financed project ‘Investigation of Corium Melt Interaction with NPP Reactor Vessel Steel’ (METCOR). In the METCOR experiments the metallic phase of a two-liquid system was produced by the interaction between hot suboxidized corium and cooled VVER vessel steel, with the steel being corroded. Models of corrosion mechanisms in the considered conditions are used to systematize data on the limiting temperature of corrosion/(dissolution) of the vessel steel. A considerable influence of thermal gradient conditions is shown, which has to be taken into account in the analysis of molten pool behaviour.
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| 6. |
- Journeau, C., et al.
(författare)
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Safest roadmap for corium experimental research in Europe
- 2018
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Ingår i: ASCE-ASME J of Risk & Uncertainty in Engineering Systems Part B. - : ASME Press. - 2332-9017 .- 2332-9025. ; 4:3
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Tidskriftsartikel (refereegranskat)abstract
- Severe accident facilities for European safety targets (SAFEST) is a European project networking the European experimental laboratories focused on the investigation of a nuclear power plant (NPP) severe accident (SA) with reactor core melting and formation of hazardous material system known as corium. The main objective of the project is to establish coordinated activities, enabling the development of a common vision and severe accident research roadmaps for the next years, and of the management structure to achieve these goals. In this frame, a European roadmap on severe accident experimental research has been developed to define research challenges to contribute to further reinforcement of Gen II and III NPP safety. The roadmap takes into account different SA phenomena and issues identified and prioritized in the analyses of severe accidents at commercial NPPs and in the results of the recent European stress tests carried out after the Fukushima accident. Nineteen relevant issues related to reactor core meltdown accidents have been selected during these efforts. These issues have been compared to a survey of the European SA research experimental facilities and corium analysis laboratories. Finally, the coherence between European infrastructures and R&D needs has been assessed and a table linking issues and infrastructures has been derived. The comparison shows certain important lacks in SA research infrastructures in Europe, especially in the domains of core late reflooding impact on source term, reactor pressure vessel failure and molten core release modes, spent fuel pool (SFP) accidents, as well as the need for a large-scale experimental facility operating with up to 500 kg of chemically prototypic corium melt.
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| 7. |
- Klein-Hessling, W., et al.
(författare)
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Conclusions on severe accident research priorities
- 2014
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 74, s. 4-11
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Tidskriftsartikel (refereegranskat)abstract
- The objectives of the SARNET network of excellence are to define and work on common research programs in the field of severe accidents in Gen. II-III nuclear power plants and to further develop common tools and methodologies for safety assessment in this area. In order to ensure that the research conducted on severe accidents is efficient and well-focused, it is necessary to periodically evaluate and rank the priorities of research. This was done at the end of 2008 by the Severe Accident Research Priority (SARP) group at the end of the SARNET project of the 6th Framework Programme of European Commission (FP6). This group has updated this work in the FP7 SARNET2 project by accounting for the recent experimental results, the remaining safety issues as e.g. highlighted by Level 2 PSA national studies and the results of the recent ASAMPSA2 FP7 project. These evaluation activities were conducted in close relation with the work performed under the auspices of international organizations like OECD or IAEA. The Fukushima-Daiichi severe accidents, which occurred while SARNET2 was running, had some effects on the prioritization and definition of new research topics. Although significant progress has been gained and simulation models (e.g. the ASTEC integral code, jointly developed by IRSN and GRS) were improved, leading to an increased confidence in the predictive capabilities for assessing the success potential of countermeasures and/or mitigation measures, most of the selected research topics in 2008 are still of high priority. But the Fukushima-Daiichi accidents underlined that research efforts had to focus still more to improve severe accident management efficiency.
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| 8. |
- Almjashev, V.I., et al.
(författare)
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Eutectic crystallization in the FeO(1.5)-UO(2+x)-ZrO(2) system
- 2009
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 389:1, s. 52-56
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Tidskriftsartikel (refereegranskat)abstract
- Results of the investigation of the FeO(1.5)-UO(2+x)-ZrO(2) system in air are presented. The eutectic position and the content of the phases crystallized at this point have been determined. The temperature and the composition of the ternary eutectic are 1323 +/- 7 degrees C and 67.4 +/- 1.0 FeO(1.5), 30.5 +/- 1.0 UO(2+x), 2.1 +/- 0.2 ZrO(2) mol.%, respectively. The solubilities of FeO(1.5) and ZrO(2) in the UO(2+x)(FeO(1.5), ZrO(2)) solid solution correspond to respectively 3.2 and 1.1 mol.%. The solubilities of UO(2) and ZrO(2) in FeO(1.5) are not significant. The existence of a solid solution on the basis of U(Zr)FeO(4) compound is found. The ZrO(2) Solubility in this solid solution is 7.0 mol.%.
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| 9. |
- Almjashev, V.I., et al.
(författare)
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Phase equilibria in the FeO(1+x)-UO(2)-ZrO(2) system in the FeO(1+x)-enriched domain
- 2010
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 400:2, s. 119-126
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Tidskriftsartikel (refereegranskat)abstract
- Experimental results of the investigation of the FeO(1+x)UO(2)-ZrO(2) system in neutral atmosphere are presented. The ternary eutectic position and the composition of the phases crystallized at this point have been determined. The phase diagram is constructed for the FeO(1+x)-enriched region and the onset melting temperature of 1310 degrees C probably represents a local minimum and so will be a determining factor in this system and its application to safety studies in nuclear reactors.
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| 10. |
- Bechta, Sevostian, et al.
(författare)
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Experimental study of interactions between suboxidized corium and reactor vessel steel
- 2006
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Ingår i: Proceedings of the 2006 International Congress on Advances in Nuclear Power Plants, ICAPP'06. - 0894486985 - 9780894486982 ; , s. 1355-1362
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Konferensbidrag (refereegranskat)abstract
- One of the critical factors in the analysis of in-vessel melt retention is the vessel strength. It is, in particular, sensitive to the thickness of intact vessel wall, which, in its turn, depends on the thermal conditions and physicochemical interactions with corium. Physicochemical interaction of prototypic UO2-ZrO2-Zr corium melt and VVER vessel steel was examined during the 2nd Phase of the ISTC METCOR Project. Rasplav-3 test facility was used for conducting four tests, in which the Zr oxidation degree and interaction front temperature were varied; in one of the tests, stainless steel was added to the melt. Direct experimental measurements and posttest analyses were used for determining corrosion kinetics and maximum corrosion depth (i.e. the physicochemical impact of corium on the cooled vessel steel specimens), as well as the steel temperature conditions during the interaction, and finally the structure and composition of crystallized ingots, including the interaction zone. The minimum temperature on the interaction front boundary, which determined its final position and maximum corrosion depth was ∼ 1090°C. An empirical correlation for calculation of corrosion kinetics has been derived.
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| 11. |
- Bechta, Sevostian, et al.
(författare)
-
VVER steel corrosion during in-vessel retention of corium melt
- 2008
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Ingår i: Proceedings of the 3<sup>rd</sup> European Review Meeting on Severe Accident Research (ERMSAR 2008).
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Konferensbidrag (refereegranskat)abstract
- Physicochemical phenomena taking place at the corium-steel interaction during theexternal cooling of reactor vessel can result in high-temperature steel corrosion and thinningof the vessel wall. The ISTC METCOR project's experimental studies have shown that themain factors influencing corrosion depth and kinetics are oxygen potential, melt compositionand steel interfacial temperature but also melt – vessel heat flux.Experimental data are used for building a model for VVER vessel steel corrosion undercorium thermochemical loads and for correlations to quantitatively analyze the influence ofcorrosion on the rector vessel thinning. The finite-element calculations, in which thedeveloped models of corrosion and heat transfer in corium pool were used, were able toreproduce the temperature and stress-and-strain vessel condition.
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| 12. |
- Buck, M., et al.
(författare)
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The LIVE program : Results of test L1 and joint analyses on transient molten pool thermal hydraulics
- 2010
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Ingår i: Progress in nuclear energy (New series). - : Elsevier BV. - 0149-1970 .- 1878-4224. ; 52:1, s. 46-60
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Tidskriftsartikel (refereegranskat)abstract
- The development of a corium pool in the lower head and its behaviour is still a critical issue. This concerns, in general, the understanding of a severe accident with core melting, its course, major critical phases and timing, and the influence of these processes on the accident progression as well as, in particular, the evaluation of in-vessel melt retention by external vessel flooding as an accident mitigation strategy. Previous studies were especially related to the in-vessel retention question and often just concentrated on the quasi-steady state behaviour of a large molten pool in the lower head, considered as a bounding configuration. However, non-feasibility of the in-vessel retention concept for high power density reactors and uncertainties e.g. due to layering effects even for low or medium power reactors, turns this to be insufficient. Rather, it is essential to consider the whole evolution of the accident, including e.g. formation and growth of the in-core melt pool, characteristics of corium arrival in the lower head, and molten pool behaviour after the debris re-melting. These phenomena have a strong impact on a potential termination of a severe accident. The general objective of the LIVE program at FZK is to study these phenomena resulting from core melting experimentally in large-scale 3D geometry and in supporting separate-effects tests, with emphasis on the transient behaviour. Up to now, several tests on molten pool behaviour have been performed within the LIVE experimental program with water and with non-eutectic melts (KNO3-NaNO3) as simulant fluids. The results of these experiments, performed in nearly adiabatic and in isothermal conditions, allow a direct comparison with findings obtained earlier in other experimental programs (SIMECO, ACOPO, BALI, etc.) and will be used for the assessment of the correlations derived for the molten pool behaviour. Complementary to other international programs with real corium melts, the results of the LIVE activities also provide data for a better understanding of in-core corium pool behaviour. The experimental results are being used for the development and validation of mechanistic models for the description of molten pool behaviour, In the present paper, a range of different models is used for post-test calculations and comparative analyses. This includes simplified, but fast running models implemented in the severe accident codes ASTEC and ATHLET-CD. Further, a computational tool developed at KTH (PECM model implemented in Fluent) is applied. These calculations are complemented by analyses with the CFD code CONV (thermal hydraulics of heterogeneous, viscous and heat-generating melts) which was developed at IBRAE (Nuclear Safety Institute of Russian Academy) within the RASPLAV project and was further improved within the ISTC 2936 Project.
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| 13. |
- Dietrich, P., et al.
(författare)
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Coupling of melcor with the pecm for improved modelling of a core melt in the lower plenum
- 2015
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Ingår i: International Conference on Nuclear Engineering, Proceedings, ICONE. - : JSME.
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Konferensbidrag (refereegranskat)abstract
- MELCOR contains a coupling interface based on the MPI-Standard, which enables the communication to other codes such as RELAP5 or GASFLOW. However a detailed knowledge of this coupling interface in MELCOR is necessary to use this possibility. Therefore, at the KIT the software tool DINAMO (Direct Interface for Adding Models) has been developed. This program contains the coupling routines as well as an interface to communicate with other programs. Using DINAMO it is also possible to utilize new or enhanced models for phenomena, which occur during a severe accident in a nuclear power plant, in MELCOR without modification of the MELCOR source code. In the present work MELCOR calculations of experiments in the LIVE-Facility are presented. The LIVE-Facility is used to simulate the behavior of a melt in the lower plenum of a reactor pressure vessel (RPV). For these calculations we coupled MELCOR via DINAMO with the Phase-Change Effective Convectivity Model (PECM), which has been developed at the KTH in Stockholm. Using the PECM it is possible to improve the prediction of a core melt in the lower plenum of a RPV in case of a core melt accident.
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| 14. |
- Dietrich, P., et al.
(författare)
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Extension of the MELCOR code for analysis of late in-vessel phase of a severe accident
- 2015
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Ingår i: IYCE 2015 - Proceedings. - : IEEE conference proceedings. - 9781467371728
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Konferensbidrag (refereegranskat)abstract
- The simulation of severe accidents in nuclear power plants with system codes is a powerful tool to improve the safety measures to prevent severe accidents. The further development of severe accident codes is part of current research. MELCOR, as the leading nuclear safety code, provides the possibility to be coupled to other codes. A detailed knowledge of this coupling interface is necessary to use this possibility. Therefore, the software tool DINAMO, which contains the coupling routines and an interface to communicate with other programs, was developed. Using DINAMO it is possible to utilize new models for specific phenomena in MELCOR. In the present work the Phase-Change Effective Convectivity Model was coupled using the CFD-software OpenFOAM and DINAMO to MELCOR to improve the prediction of molten core material in the lower plenum of a reactor pressure vessel. The simulation results were compared to the experimental findings of the LIVE-facility.
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| 15. |
- Journeau, C., et al.
(författare)
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European Research on the Corium issues within the SARNET network of excellence
- 2008
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Ingår i: International Conference on Advances in Nuclear Power Plants, ICAPP 2008. - 9781605607870 ; , s. 1172-1181
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Konferensbidrag (refereegranskat)abstract
- Within SARNET, the corium topic covers all the behaviors of corium from early phase of core degradation to in or ex-vessel corium recovery with the exception of corium interaction with water, direct containment heating and fission product release. The corium topic regroups in three work packages the critical mass of competence required to improve significantly the corium behavior knowledge. The spirit of the SARNET networking is to share the knowledge, the facilities and the simulation tools for severe accidents, so to reach a better efficiency and to rationalize the R&D effort at European level. Extensive benchmarking has been launched in most of the areas of research. These benchmarks were mainly dedicated to the recalculation of experiments, while, in the next periods, a larger focus will be given to integral experiments or reactor applications. Eventually, all the knowledge will be accumulated in the ASTEC severe accident simulation code through physical model improvements and extension of validation database. This paper summarizes the progress that has been achieved in the frame of the networking activities. A special focus is placed on the melt pool and debris coolability and corium-concrete interaction, in which, the effects due to multidimensional geometries and heterogeneities has been shown, during SARNET, to play a crucial role and for which further research is still needed.
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| 16. |
- Fichot, F., et al.
(författare)
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Some considerations to improve the methodology to assess In-Vessel Retention strategy for high-power reactors
- 2018
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Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 119, s. 36-45
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Tidskriftsartikel (refereegranskat)abstract
- 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, for high power reactors, estimations using current level of conservatism show that RPV failure caused by thermo-mechanical rupture takes place in some cases. A better estimation of the residual risk (probability of cases with vessel rupture) requires the use of models with a lower level of conservatism. 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 based not only on a comparison of the heat flux and the Critical Heat Flux (CHF) profiles as in current approaches but also on the minimum vessel thickness reached after ablation and the maximum integral loads that is applied to the vessel during the transient. The main advantage of this revised criterion is in consideration of both steady-state and transient loads on the RPV. Another advantage is that this criterion may be used in both probabilistic and deterministic approaches, whereas the current approaches are mostly deterministic (with deterministic calculations used only for estimates of uncertainty ranges of input parameters).
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| 17. |
- Fischer, M., et al.
(författare)
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Core melt stabilization concepts for existing and future LWRs and associated R&D needs
- 2015
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Ingår i: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015. - 9781510811843 ; , s. 7578-7592
-
Konferensbidrag (refereegranskat)abstract
- In the event of a severe accident with core melting in a NPP the stabilization of the molten corium is an important mitigation issue, as it can avoid late containment failure caused by basemat penetration, overpressure, or severe damage of internal structures. The related failure modes may result in significant long-term radiological consequences and high related costs. Because of this, the licensing framework of several countries now includes the request to implement mitigative core melt stabilization measures. This does not only apply to new builds but also to existing LWR plants. The paper gives an overview of the ex-vessel core melt stabilization strategies developed during the last decades. These strategies are based on a variety of physical principles like: melt fragmentation in a deep water pool or during molten core concrete interaction with top-flooding, water injection from the bottom (COMET concept), and retention in an outside-cooled crucible structure. The provided overview covers the physical background and functional principles of these concepts, as well as their status of validation and, if applicable, the remaining open issues and R&D needs. For concepts based on melt retention inside a cooled crucible that reached sufficient maturity to be implemented in current Gen-III+ designs, like the VVER-1000/1200 and the EPR™, more detailed descriptions are provided, which include key aspects of the related technical realization. The paper is compiled using contributions from the main developers of the individual concepts.
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| 18. |
- Miassoedov, A., et al.
(författare)
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Corium and debris coolability studies performed in the severe accident research network of excellence (SARNET2)
- 2012
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Ingår i: Proceedings Of The 20th International Conference On Nuclear Engineering And The ASME 2012 Power Conference - 2012, Vol 2. - : ASME Press. ; , s. 383-392
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Konferensbidrag (refereegranskat)abstract
- The motivation of the work performed within the work package "Corium and Debris Coolability" of the Severe Accident Research Network of Excellence (SARNET) is to reduce or possibly solve the remaining uncertainties on the efficiency of cooling reactor core structures and materials during severe accidents, either in the core, in the vessel lower head or in the reactor cavity, so as to limit the progression of the accident. This can be achieved either by ensuring corium retention within the reactor pressure vessel or at least by limiting the corium progression and the rate of corium release into the cavity. These issues are to be covered within the scope of accident management for existing reactors and within the scope of design and safety evaluation of future reactors. The specific objectives are to create and enhance the database on debris formation, debris coolability and corium behavior in the lower head, to develop and validate the models and computer codes for simulation of in-vessel debris bed and melt pool behavior, to perform reactor scale analysis for in-vessel corium coolability and to assess the influence of severe accident management measures on in-vessel coolability. The work being performed within this work package comprises experimental and modeling activities with strong cross coupling between the tasks. Substantial knowledge and understanding of governing phenomena concerning coolability of intact rod-like reactor core geometry was obtained in previous projects. Hence the main thrust of experimental and modeling efforts concentrates mainly on the study of formation and cooling of debris beds in order to demonstrate effective cooling modes, cooling rates and coolability limits. Modeling efforts have been aimed at assessing and validating the models in system-level and detailed codes for core degradation, oxidation and debris behavior. The paper describes the work performed up to now and summarizes the main results achieved so far.
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| 19. |
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