| 1. |
- 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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| 2. |
- 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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| 3. |
- 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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| 4. |
- Almyashev, V. I., et al.
(författare)
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Oxidation effects during corium melt in-vessel retention
- 2016
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Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 305, s. 389-399
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Tidskriftsartikel (refereegranskat)abstract
- In the in-vessel corium retention studies conducted on the Rasplav-3 test facility within the ISTC METCOR-P project and OECD MASCA program, experiments were made to investigate transient processes taking place during the oxidation of prototypic molten corium. Qualitative and quantitative data have been produced on the sensitivity of melt oxidation rate to the type of oxidant, melt composition, molten pool surface characteristics. The oxidation rate is a governing factor for additional heat generation and hydrogen release; also for the time of secondary inversion of oxidic and metallic layers of corium molten pool.
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| 5. |
- Asmolov, V.G., et al.
(författare)
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Crucible-type core catcher for VVER-1000 reactor
- 2005
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Ingår i: Proceedings of the American Nuclear Society - International Congress on Advances in Nuclear Power Plants 2005, ICAPP'05. - : Curran Associates, Inc.. - 9781604236934 ; , s. 1221-1227
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Konferensbidrag (refereegranskat)abstract
- For new designs of NPPs with VVER-1000 reactors a crucible-type core catcher has been developed to stabilize and cool down molten corium in the reactor pit. The paper addresses distinguishing features of the concept including the "sacrificial" material and the core catcher design. Main phenomena in the catcher have been analyzed.
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| 6. |
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| 7. |
- Asmolov, V.G., et al.
(författare)
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Partitioning of Zr, U and FP between Molten Oxidic and Metallic Corium
- 2004
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Ingår i: Proceeding of MASCA Seminar.
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Konferensbidrag (refereegranskat)abstract
- Interaction of molten corium and liquid iron/stainless steel has been studied in different tests of theMASCA-1 program. These tests utilized the technology of induction melting in a cold crucible. Themasses of tested corium were approximately 0.5, 2 and 100 kg. Representative quantities of Mo, Ru,SrO, BaO, CeO2 and La2O3 served as fission product simulants.After the suboxidized melt - steel interaction U and Zr have been found in the metallic phase.To quantify the partitioning of Zr, U and fission products an extensive experimental program has beenperformed. The following key parameters have been identified: oxygen potential in the melt (degree ofZr-oxidation), the corium/steel mass ratio and U/Zr ratio. The paper discusses the influence of theseparameters on the partitioning of the main species.
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| 8. |
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| 9. |
- Bakardjieva, S., et al.
(författare)
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Improvement of the European thermodynamic database NUCLEA
- 2010
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Ingår i: Progress in nuclear energy (New series). - : Elsevier BV. - 0149-1970 .- 1878-4224. ; 52:1, s. 84-96
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Tidskriftsartikel (refereegranskat)abstract
- Modelling of corium behaviour during a severe accident requires knowledge of the phases present at equilibrium for a given corium composition, temperature and pressure. The thermodynamic database NUCLEA in combination with a Gibbs Energy minimizer is the European reference tool to achieve this goal. This database has been improved thanks to the analysis of bibliographical data and to EU-funded experiments performed within the SARNET network, PLINIUS as well as the ISTC CORPHAD and EVAN projects. To assess the uncertainty range associated with Energy Dispersive X-ray analyses, a round-robin exercise has been launched in which a UO2-containing corium-concrete interaction sample from VULCANO has been analyzed by three European laboratories with satisfactorily small differences.
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| 10. |
- Bakardjieva, S., et al.
(författare)
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Improvement of the European thermodynamic database NUCLEA in the frame of EU-funded experiments
- 2008
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Ingår i: Proceedings of the 3<sup>rd</sup> European Review Meeting on Severe Accident Research (ERMSAR 2008).
-
Konferensbidrag (refereegranskat)abstract
- Modelling of corium behaviour during a severe accident requires knowledge of the phases present at equilibrium for a given corium composition, temperature and pressure. The thermodynamic database NUCLEA in combination with a Gibbs Energy minimizer is theEuropean reference tool to achieve this goal. Within SARNET, this database has beenimproved thanks to the analysis of bibliographical data and to EU0funded experimentsperformed within the SARNET network, PLINIUS as well as the ISTC CORPHAD and EVAN projects. To assess the uncertainty range associated with Energy Dispersive X-ray analyses, a round robin exercise has been launched in which a UO2-containing coriumconcrete interaction sample from VULCANO has been analyzed by three European laboratories with satisfactorily small differences.
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| 11. |
- Bakardjieva, S., et al.
(författare)
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Quality improvements of thermodynamic data applied to corium interactions for severe accident modelling in SARNET2
- 2014
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Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 74, s. 110-124
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Tidskriftsartikel (refereegranskat)abstract
- In a severe accident transient, corium composition and its properties determine its behaviour and its potential interactions both with the reactor vessel and in the later phases with the concrete basemat. This, in turn, requires a detailed knowledge of the phases present at temperature and how they are formed. Because it implies mainly the investigation of chemical systems at high temperature, these data are often difficult to obtain or are uncertain if it already exists. Therefore more data are required both to complete the thermodynamic databanks (such as NUCLEA) and to construct accurate equilibrium phase diagrams and to finally contribute to the improvement of the codes simulating these severe accident conditions. The MCCI work package (WP6) of the SARNET 2 Network of Excellence has been addressing these problems. In this framework in large facilities such as VULCANO tests have been performed on the interactions and ablation of UO2-containing melts with concrete. They have been completed by large scale MCCI testing such EPICOR on vessel steel corrosion. In parallel in major EU-funded ISTC projects co-ordinated with national institutes, such as the CORPHAD and PRECOS, smaller, single effect tests have been carried out on the more difficult phase diagrams. These have produced data that can be directly used by databanks and for modelling improvement/validation. From these data significant advances in the melt chemistry and pool behaviour have been made. A selection of experiments from participating institutes are presented in this paper and give hindsight into the major processes and so give clear indications for the future work, especially in light of the Fukushima accident.
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| 12. |
- Bandaru, S V Ravikumar, et al.
(författare)
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Multi-nozzle spray cooling of a reactor pressure vessel steel plate for the application of ex-vessel cooling
- 2021
-
Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 375
-
Tidskriftsartikel (refereegranskat)abstract
- Spray cooling is a versatile technology for various cooling applications involving high surface heat fluxes. Experimental facility was built to study heat transfer performance of an upward multi-nozzle array of water sprays impacting a surface of heated plate made of reactor vessel grade steel. The effect of inclination angles of the steel surface on the cooling performance was investigated to assess heat transfer in complex semispherical/ semielliptical geometry of large reactor lower head and to address possible application of spray cooling in severe accident management (SAM) of light water reactors (LWRs) based on In-vessel melt retention with external reactor vessel cooling (IVR-ERVC). Joule heating of SA302B steel foil of 0.15 mm thickness and surface area of 96 cm2 enabled prototypic heat fluxes to be evacuated during reactor accident. A 2×3 array of full jet narrow-coned pressure-swirl spray nozzles was used to reproduce multi-nozzle cooling. The tests were conducted as a series of consequent steady states realized at stepwise increasing power and surface heat fluxes up to the maximum values of 29 kW and 2.97 MW/m2 limited in the specific facility design. Seven surface inclinations, between 0o and 90o were tested and no significant variations in spray cooling performance with the inclination of the heated surface was found. The results indicated a promising prospect of using a multi-nozzle array system for cooling of large surface area of reactor lower head. Much higher heat fluxes can be safely extracted by spray cooling in comparison with the critical heat fluxes that appeared at RPV water pool cooling at natural convection. The maximum value of heat flux at direct spray impact zones and its drop-off slightly from the center to the periphery of the spray cone was detected in the tests. The water flow rate and liquid subcooling significantly influenced maximum steel surface temperature but had no noticeable effects on surface temperature uniformity. The spray-to-spray interaction had no observable effects on local surface temperatures, however, the colliding zones where four spray cones have visible effects on local surface temperatures due to poor liquid momentum. The results also showed that more uniform liquid film distribution could be obtained for some inclinations because of improved liquid drainage, which in turn leads to maintaining low surface temperatures.
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| 13. |
- Bandaru, S V Ravikumar, et al.
(författare)
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Upward-facing multi-nozzle spray cooling experiments for external cooling of reactor pressure vessels
- 2020
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Ingår i: International Journal of Heat and Mass Transfer. - : Elsevier BV. - 0017-9310 .- 1879-2189. ; 163
-
Tidskriftsartikel (refereegranskat)abstract
- Cooling by water spray is a well-known technology that can reach significantly higher Critical Heat Flux (CHF) compared to other cooling methods. For the light water reactor safety, the in-vessel retention (IVR) by external reactor vessel cooling (ERVC) is a comprehensive severe accident management strategy to arrest and confine the corium in the lower head of the reactor pressure vessel. Heat fluxes up to 1.5 MW/m2 have already been assumed attainable in low-power nuclear reactors while cooling required in high-power reactors is expected to reach 2.5 MW/m2. Instead of reactor lower head flooding and relying on cooling due to natural convection, a viable and more efficient alternative is to spray the external surface of the vessel. Given all the advantages of spray cooling reported in the literature, a lab-scale experimental facility was built to validate the efficiency of multi-nozzle spray cooling of a downward-facing heated surface inclined at different angles up to 90o. The facility employed a 2×3 matrix of spray nozzles to cool the FeCrAl alloy foil with an effectively heated surface area of 96 cm2 using water as the coolant. Heat loads and surface inclinations were varied parameters in the test matrix. The results show that no significant variations in spray cooling performance concerning the inclination of the heated surface. A surface heat flux of 2.5 MW/m2 was achieved at every inclination of the downward-facing surface. The results also indicate that more uniform liquid film distribution could be obtained for some inclinations, which in turn leads to maintaining low surface temperature. The obtained surface heat flux margin by spray cooling indicates that it is feasible to adopt IVR-ERVC strategy for a large power reactor.
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| 14. |
- Bechta, Sevostian, et al.
(författare)
-
Corium phase equilibria based on MASCA, METCOR and CORPHAD results
- 2008
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Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 238:10, s. 2761-2771
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Tidskriftsartikel (refereegranskat)abstract
- Experimental data on component partitioning between suboxidized corium melt and steel in the invessel melt retention (IVR) conditions are compared. The data are produced within the OECD MASCAprogram and the ISTC CORPHAD project under close-to-isothermal conditions and in the ISTC METCORproject under thermal gradient conditions. Chemical equilibrium in the U–Zr–Fe(Cr,Ni,. . .)–O system isreached in all experiments. In MASCA tests the molten pool formed under inert atmosphere has twoimmiscible liquids, oxygen-enriched (oxidic) and oxygen-depleted (metallic), resulting of the miscibilitygap of the mentioned system. Sub-system data of the U–Zr–Fe(Cr,Ni,. . .)–O phase diagram investigatedwithin the ISTC CORPHAD project are interpreted in relation with the MASCA results. In METCOR teststhe equilibrium is established between oxidic liquid and mushy metallic part of the system. Results ofcomparison are discussed and the implications for IVR noted.
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| 15. |
- Bechta, Sevostian, et al.
(författare)
-
CORPHAD and METCOR ISTC projects
- 2005
-
Ingår i: Proceedings of The first European Review Meeting on Severe Accident Research (ERMSAR-2005).
-
Konferensbidrag (refereegranskat)abstract
- The ongoing CORPHAD Project (Phase Diagrams for Multicomponent SystemsContaining Corium and Products of its Interaction with NPP Materials) started in August2001. The main aim of the project is to experimentally determine the relevantphysicochemical data on phase diagrams of binary, ternary, quaternary and prototypic multicomponent systems, which are important for analysis and modelling of a severe accident (SA)and efficient planning of severe accident management (SAM) measures. The data should bedirectly used for the European NUCLEA database development and validation. The followingsystems are in the focus of the project: (1) UO2 – FeO, (2) ZrO2 – FeO, (3) SiO2– Fe2O3, (4)UO2 – SiO2, (5) UO2 – ZrO2-FeO, (6) UO2 – ZrO2-FeOy, (7) U-O-Fe, (8) Zr-O-Fe, (9) U-OZr, (10) U-Zr-Fe-O, (11) complex corium mixtures.The experimentally determined data of the listed diagrams include: coordinates ofcharacteristic points (eutectics, peritectics and others); liquidus and solidus concentrationcurves; component solubility limits in the solid phase; tie line coordinates and temperatureconcentration regions of the miscibility gap. Different methodologies are used for the phasediagram study. Classical methods of thermal analysis, like DTA and DSC are combined withmethods specifically developed for corium studies.The METCOR project (Investigation of Corium Melt Interaction with NPP ReactorVessel Steel) started in April 1999. The objectives of the project are to qualify and to quantifyphysico-chemical phenomena of corium melt interaction with reactor vessel steel cooled fromthe outside. The variable parameters of the interaction tests are: oxygen potential in thesystem, corium composition, interaction interface temperature and heat flux from corium tosteel. The medium scale tests with corium mass of about 2 kg are carried out by using highfrequency induction heating of the corium melt in a cold crucible.The METCOR & CORPHAD work-packages are performed by Russian partners inclose collaboration with leading European scientific institutes in the area of corium researchas well as with the European nuclear industry.This paper briefly describes the results obtained in both projects and their possibleapplication for SA analysis and SAM. The paper concludes with recommendations for futureresearch activities in the framework of METCOR and CORPHAD projects.
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| 16. |
- Bechta, Sevostian, et al.
(författare)
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Corrosion of vessel steel during its interaction with molten corium : Part 2. Model development
- 2006
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Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 236:13, s. 1362-1370
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Tidskriftsartikel (refereegranskat)abstract
- An experimental examination of the cooled vessel steel corrosion during the interaction with molten corium is presented. The experiments havebeen conducted on “Rasplav-2” test facility and followed up with physico-chemical and metallographic analyses of melt samples and coriumspecimeningots. The results discussed in the first part of the paper have revealed specific corrosion mechanisms for air and inert atmosphere abovethe melt. Models have been proposed based on this information and approximate curves constructed for the estimation of the corrosion rate orcorrosion depth of vessel steel in conditions simulated by the experiments.
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| 17. |
- Bechta, Sevostian, et al.
(författare)
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Corrosion of vessel steel during its interaction with molten corium : Part 1. Experimental
- 2006
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Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 236:17, s. 1810-1829
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Tidskriftsartikel (refereegranskat)abstract
- This paper is concerned with corrosion of a cooled vessel steel structure interacting with molten corium in air and neutral (nitrogen) atmospheresduring an in-vessel retention scenario. The data on corrosion kinetics at different temperatures on the heated steel surface, heat flux densities andoxygen potential in the system are presented. The post-test physico-chemical and metallographic analyses of melt samples and the corium–specimeningot have clarified certain mechanisms of steel corrosion taking place during the in-vessel melt interaction.
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| 18. |
- Bechta, Sevostian, et al.
(författare)
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Experimental studies of oxidic molten corium-vessel steel interaction
- 2001
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Ingår i: Nuclear Engineering and Design. - 0029-5493 .- 1872-759X. ; 210:1-3, s. 193-224
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Tidskriftsartikel (refereegranskat)abstract
- The experimental results of molten corium-steel specimen interaction with molten corium on the 'Rasplav-2' test facility are presented. In the experiments, cooled vessel steel specimens positioned on the molten pool bottom and uncooled ones lowered into the molten pool were tested. Interaction processes were studied for different corium compositions, melt superheating and in alternative (inert and air) overlying atmosphere. Hypotheses were put forward explaining the observed phenomena and interaction mechanisms. The studies presented in the paper were aimed at the detection of different corium-steel interaction mechanisms. Therefore certain identified phenomena are more typical of the ex-vessel localization conditions than of the in-vessel corium retention. Primarily, this can be referred to the phenomena of low-temperature molten corium-vessel steel interaction in oxidizing atmosphere.
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| 19. |
- 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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| 20. |
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| 21. |
- Bechta, Sevostian, et al.
(författare)
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Influence of corium oxidation on fission product release from molten pool
- 2010
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Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 240:5, s. 1229-1241
-
Tidskriftsartikel (refereegranskat)abstract
- Qualitative and quantitative determination of the release of low-volatile fission products and core materialsfrom molten oxidic corium was investigated in the EVAN project under the auspices of ISTC. Theexperiments carried out in a cold crucible with induction heating and RASPLAV test facility are described.The results are discussed in terms of reactor application; in particular, pool configuration, melt oxidationkinetics, critical influence of melt surface temperature and oxidation index on the fission productrelease rate, aerosol particle composition and size distribution. The relevance of measured high releaseof Sr from the molten pool for the reactor application is highlighted. Comparisons of the experimentaldata with those from the COLIMA CA-U3 test and the VERCORS tests, as well as with predictions fromIVTANTHERMO and GEMINI/NUCLEA codes are made. Recommendations for further investigations areproposed following the major observations and discussions.
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| 22. |
- 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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| 23. |
- Bechta, Sevostian, et al.
(författare)
-
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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| 24. |
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| 25. |
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| 26. |
- Bechta, Sevostian, et al.
(författare)
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New experimental results on the interaction of molten corium with reactor vessel steel
- 2004
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Ingår i: Proceedings of the 2004 International Congress on Advances in Nuclear Power Plants, ICAPP'04. - : American Nuclear Society. - 0894486802 ; , s. 1072-1081
-
Konferensbidrag (refereegranskat)abstract
- In order to justify the concept of in-vessel core melt retention, it is necessary to understand the thermal and physico-chemical phenomena. Especially the interaction of the molten pool with the reactor vessel during outside cooling needs to be understood. These phenomena are very complex, in particular, where interactions with the oxidic melt are concerned. In the early stages of the retention process, the oxidic corium and the vessel steel interact under the conditions of low oxygen potential in the melt. These conditions can be simulated by a molten corium having the composition UO2/ZrO 2Zr, where the degree of Zr-oxidation is in the range between 30 % (C-30) and 100 % (C-100). Corresponding experiments with prototypic melts at low oxygen potentials are being performed in the ISTC METCOR project 2nd phase. These are: MC 5 of corium composition 71w%UO2-29w%ZrO 2 (C-100) in neutral atmosphere (argon), MC 6 of corium composition 76w%UO2-9w%ZrO2-15w%Zr (C∼30), also in argon. In test MC 5, the interaction of molten C-100 corium with a water-cooled steel specimen was studied for the following maximum temperatures at the specimen surface: 1075°C, 1180°C, 1315°C and 1435°C. The total duration of the experiment was ∼ 36 hours. The MC5 test serves as a reference test for determining the characteristics of the interaction between oxidic melt and steel specimen under the conditions of minimum chemical interaction potential. To investigate the effect of substoichiometry, test MC 6 was then performed with suboxidized molten corium C∼30. The maximum surface temperature of the cooled steel specimen was held at ∼ 1400°C. The test duration was ∼ 10 hours. The ablation phenomena were found to differ significantly from those observed both in the reference test, as well as in former tests with oxidized melts, as they involved the formation of a low-melting metallic phase at the interface which contains iron, zirconium and uranium. The paper summarizes the results of the experiments and of the performed posttest analysis for tests MC 5 and MC 6.
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| 27. |
- Bechta, Sevostian, et al.
(författare)
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On the EU-Japan roadmap for experimental research on corium behavior
- 2019
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Ingår i: Annals of Nuclear Energy. - : PERGAMON-ELSEVIER SCIENCE LTD. - 0306-4549 .- 1873-2100. ; 124, s. 541-547
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Tidskriftsartikel (refereegranskat)abstract
- A joint research roadmap between Europe and Japan has been developed in severe accident field of light water reactors, focusing particularly on reactor core melt (corium) behavior. The development of this roadmap is one of the main targets of the ongoing EU project SAFEST. This paper presents information about ongoing severe accident studies in the area of corium behavior, rationales and comparison of research priorities identified in different projects and documents, expert ranking of safety issues, and finally the research areas and topics and their priorities suggested for the EU Japan roadmap and future bilateral collaborations. These results provide useful guidelines for (i) assessment of long-term goals and proposals for experimental support needed for proper understanding, interpretation and learning lessons of the Fukushima accident; (ii) analysis of severe accident phenomena; (iii) development of accident prevention and mitigation strategies, and corresponding technical measures; (iv) study of corium samples in European and Japanese laboratories; and (v) preparation of Fukushima site decommissioning.
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| 28. |
- Bechta, Sevostian, et al.
(författare)
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Phase diagram of the UO2-FeO1+x system
- 2007
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 362:1, s. 46-52
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Tidskriftsartikel (refereegranskat)abstract
- Phase-relation studies of the UO2–FeO1+x system in an inert atmosphere are presented. The eutectic point has beendetermined, which corresponds to a temperature of (1335 ± 5) C and a UO2 concentration of (4.0 ± 0.1) mol.%. Themaximum solubility of FeO in UO2 at the eutectic temperature has been estimated as (17.0 ± 1.0) mol.%. Liquidus temperaturesfor a wide concentration range have been determined and a phase diagram of the system has been constructed.
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| 29. |
- Bechta, Sevostian, et al.
(författare)
-
Phase diagram of the ZrO2-FeO system
- 2006
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 348:1-2, s. 114-121
-
Tidskriftsartikel (refereegranskat)abstract
- The results on the ZrO2–FeO system studies in a neutral atmosphere are presented. The refined eutectic point has beenfound to correspond to a ZrO2 concentration of 10.3 ± 0.6 mol% at 1332 ± 5 C. The ultimate solubility of iron oxide inzirconia has been determined in a broad temperature range, taking into account the ZrO2 polymorphism. A phase diagramof the pseudobinary system in question has been constructed.
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| 30. |
- Bechta, Sevostian, et al.
(författare)
-
Phase Relations in the ZrO2–FeO System
- 2006
-
Ingår i: Russian Journal of Inorganic Chemistry. - 0036-0236 .- 1531-8613. ; 51:2, s. 325-331
-
Tidskriftsartikel (refereegranskat)abstract
- We present the results of the investigation of the ZrO2-FeO system under an inert atmosphere. We have refined the position of the eutectic point, which lies at 1332 +/- 5 degrees C and 10.3 +/- 0.6 mol % ZrO2. The iron oxide solubility boundaries in zirconium dioxide have been determined over a wide temperature range taking into account the polymorphism in ZrO2. A phase diagram for the system has been designed.
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| 31. |
- Bechta, Sevostian, et al.
(författare)
-
Phase transformation in the binary section of the UO2-FeO-Fe system
- 2007
-
Ingår i: Radiochemistry (New York, N.Y.). - 1066-3622 .- 1608-3288. ; 49:1, s. 20-24
-
Tidskriftsartikel (refereegranskat)abstract
- Phase transformations in the oxide binary section of the UO2-FeO-Fe ternary system were studied. The melting onset point of the UO2-FeO heterogeneous system (1335±5°C) was determined and the fusion curve of this system was constructed. The limiting solubility of FeO in the UO2 solid solution was measured. The changes in crystal parameters in formation of the solid solution were determined. Uranium dioxide was found to be insoluble in the wüstite phase (FeO).
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| 32. |
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| 33. |
- Bechta, Sevostian, et al.
(författare)
-
VVER steel corrosion during in-vessel retention of corium melt
- 2008
-
Ingår i: Proceedings of the 3<sup>rd</sup> European Review Meeting on Severe Accident Research (ERMSAR 2008).
-
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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| 34. |
- Bechta, Sevostian, et al.
(författare)
-
VVER vessel steel corrosion at interaction with molten corium in oxidizing atmosphere
- 2009
-
Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 239:6, s. 1103-1112
-
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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| 35. |
- Bechta, Sevostian, et al.
(författare)
-
Water boiling on the corium melt surface under VVER severe accident conditions
- 2000
-
Ingår i: Nuclear Engineering and Design. - 0029-5493 .- 1872-759X. ; 195:1, s. 45-56
-
Tidskriftsartikel (refereegranskat)abstract
- Experimental results are presented on the interaction of corium melt with water supplied on its surface. The tests were conducted in the `Rasplav-2' experimental facility. Corium melt was generated by induction melting in the cold crucible. The following data were obtained: heat transfer at boiling water-melt surface interaction, gas and aerosol release, post-interaction solidified corium structure. The corium melt charge had the following composition, mass%: 60% UO2+x-16% ZRO2-15% Fe2O3-6% Cr2O3-3% Ni2O3. The melt surface temperature ranged within 1920-1970 K.
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| 36. |
- Bian, Boshen
(författare)
-
CFD Study of Molten Pool Convection in a Reactor Vessel during a Severe Accident
- 2023
-
Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- During severe accidents in nuclear reactors, the core and internal structures can melt down and relocate into the reactor pressure vessel (RPV) lower head (LH) forming there a stratified molten corium pool. The pool generally consists of superheated oxidic and metallic liquid layers imposing thermo-mechanical loads on the RPV wall. The in-vessel retention (IVR) strategy employs external cooling with water to maintain RPV integrity. Investigating the thermo-fluid behaviour of corium and predicting heat flux distribution on the vessel wall are crucial. The molten pool exhibits natural convection, which can typically consist of two stratified layers. There exists internally heated (IH) natural convection in the oxidic layer and Rayleigh-Bénard (RB) convection in the surface metallic layer.This study starts by illustrating the mathematical models that involve the numerical study of natural convection flow in molten corium. A verification work of the model has been done using a previous direct numerical simulation (DNS) study, and the results show good agreement. In addition, a scaling theory of the natural convection flow is demonstrated to facilitate the pre-estimation based on the Rayleigh number (Ra) and Prandtl number (Pr). After that, the numerical approaches involved in the numerical simulation of the corium are illustrated, especially focusing on the DNS method. A DNS mesh strategy is proposed in the form of a pipeline from the pre-estimation to the post-check. A scalability study of Nek5000 is performed on four different HPC clusters based on a DNS case of the IH molten convection in a hemispherical geometry with Ra=1.6×1011. The results show a super-liner speedup property of Nek5000 on each cluster within a certain range.Then, three numerical studies focusing on turbulent natural convection flow within both the oxidic and metallic layers of corium are demonstrated and discussed. Through these simulations, the thermos-fluid behaviour of the system is examined in detail, including flow configuration, temperature distribution, heat flux profiles on cooling boundaries, and turbulent quantities.1. A DNS investigation is performed on the IH molten pool convection within a hemispherical domain, employing a Rayleigh number of 1.6×1011 and a Prandtl number of 0.5. The results show a turbulent flow characterized by three distinct regions, consistent with the observation from the BALI experiments. Detailed information regarding turbulence, including turbulent kinetic energy (TKE), turbulent heat flux (THF), and temperature variance, is presented. Furthermore, the study offers comprehensive 3D heat flux distributions along the boundaries, showing heat flux fluctuations along the top boundary due to nearby turbulent eddies and a nonlinear increase in heat flux along the curved boundary from bottom to top.2. A numerical study investigates the effect of Prandtl number on the natural convection of an IH molten pool in a 3D semi-circular test section. Prandtl numbers of 3.11, 1.0, and 0.5 are considered, with a Ra= 6.54×1011. Smaller Prandtl numbers result in more vigorous turbulent motion and a thicker layer of intense turbulent mixing in the upper region. The descending flow extends further down the bottom, creating a stronger circulation at the bottom with smaller Pr. Additionally, smaller Pr leads to more thermal stripping structures and less stable stratification layers. Comparing heat fluxes on the top and curved walls reveals higher fluctuation frequency with smaller Pr for heat fluxes to the top boundary. However, the maximum heat fluxes to the side walls are lower with smaller Pr.3. A numerical study investigates the turbulent natural convection in a 3D fluid layer based on the BALI-Metal 8U experiment. Different methods, including DNS and three Reynolds-averaged Navier-Stokes (RANS) models, are employed. The results are compared with experimental data, and the performance of the RANS models is evaluated using DNS as a reference. DNS reproduces a two-distinct region flow structure observed in experiments, while the k-ω SST model exhibits similar flow patterns and TKE profiles. However, all simulations overpredict temperature compared to experimental data, with DNS providing the closest results. The DNS results also achieve better agreement with experimental data in terms of heat flux distribution and energy balance, specifically capturing the transient maximum heat flux on the lateral cooling wall. This transient behaviour plays a crucial role in accurately estimating the ‘focusing effect’.
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| 37. |
- Bottomley, D., et al.
(författare)
-
Severe accident research in the core degradation area : An example of effective international cooperation between the European Union (EU) and the Commonwealth of Independent States (CIS) by the International Science and Technology Center
- 2012
-
Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 252, s. 226-241
-
Tidskriftsartikel (refereegranskat)abstract
- The International Science and Technology Center (ISTC) was set up in Moscow to support non-proliferation of sensitive knowledge and technologies in biological, chemical and nuclear domains by engaging scientists in peaceful research programmes with a broad international cooperation. The paper has two following objectives: to describe the organization of complex, international, experimental and analytical research of material processes under extreme conditions similar to those of severe accidents in nuclear reactors and, to inform briefly about some results of these studies. The main forms of ISTC activity are Research Projects and Supporting Programs. In the Research Projects informal contact expert groups (CEGs) were set up by ISTC to improve coordination between adjacent projects and to encourage international collaboration. The European Commission was the first to use this. The CEG members - experts from the national institutes and industry - evaluated and managed the projects' scientific results from initial stage of proposal formulation until the final reporting. They were often involved directly in the project's details by joining the Steering Committees of the project. The Contact Expert Group for Severe Accidents and Management (CEG-SAM) is one of these groups, five project groups from this area from the total of 30 funded projects during 10 years of activity are detailed to demonstrate this: (1) QUENCH-VVER from RIAR, Dimitrovgrad and IBRAE, Moscow, and PARAMETER projects (SF1-SF4) from LUCH, Podolsk and IBRAE, Moscow; these concerned a detailed study of bundle quenching from high temperature; (2) Reactor Core Degradation; a modelling project simulating the fuel rod degradation and loss of geometry from IBRAE, Moscow; (3) METCOR projects from NITI, St. Petersburg on the interaction of core melt with reactor vessel steel; (4) INVECOR project, NNE Kurchatov City, Kazakhstan; this is a large-scale facility to examine the vessel steel retention of 60 kg corium during the decay heat; and finally, (5) CORPHAD and PRECOS projects, NITI, St. Petersburg undertook a systematic examination of refractory ceramics relevant to in-vessel and ex-vessel coria, particularly examining poorly characterised, limited data or experimentally difficult systems.
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| 38. |
- Chen, Yangli, et al.
(författare)
-
A sensitivity study of MELCOR nodalization for simulation of in-vessel severe accident progression in a boiling water reactor
- 2019
-
Ingår i: Nuclear Engineering and Design. - : ELSEVIER SCIENCE SA. - 0029-5493 .- 1872-759X. ; 343, s. 22-37
-
Tidskriftsartikel (refereegranskat)abstract
- This paper presents a sensitivity study of MELCOR nodalization for simulation of postulated severe accidents in a Nordic boiling water reactor, with the objective to address the nodal effect on in-vessel accident progression, including thermal-hydraulic response, core degradation and relocation, hydrogen generation, source term release, melt behavior and heat transfer in the lower head, etc. For this purpose, three meshing schemes (coarse, medium and fine) of the COR package of MELCOR are chosen to analyze two severe accident scenarios: station blackout (SBO) accident and large break loss-of-coolant accident (LOCA) combined with station blackout. The comparative results of the MELCOR simulations show that the meshing schemes mainly affect the core degradation and relocation to the lower head of the reactor pressure vessel: the fine mesh leads to a delayed leveling process of a heap-like debris bed in the lower head, and a later breach of the vessel. The simulations with fine mesh also provide more detailed distributions of corium mass and temperature, as well as heat flux which is an important parameter in qualification assessment of the In-Vessel Melt Retention (IVR) strategy.
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| 39. |
- Deng, Yucheng, et al.
(författare)
-
An experimental study on the effect of coolant salinity on steam explosion
- 2024
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 201
-
Tidskriftsartikel (refereegranskat)abstract
- The steam explosion plays an essential role in the safety analysis of light water reactors (LWRs). Some studies have demonstrated that the occurrence of steam explosions is dependent on many factors such as melt and coolant temperatures, melt and coolant properties, non -condensable gases, etc. After the Fukushima accident, seawater as an emergency coolant and its impact on fuel coolant interactions are receiving attention. However, there is still little knowledge on the impact of seawater on steam explosion. The present study is intended to examine the effect of coolant salinity on steam explosion through a series of tests with single molten droplet falling in different coolant pools (DI water, and seawater at different salinities from 7.7 g/kg to 35 g/kg). The experimental results reveal that the salinity of coolant significantly influences the probability of spontaneous steam explosion of molten tin droplets. The probability of steam explosion generally increases with increasing salinity from 0 to 17.5 g/kg. The molten droplet in seawater experiences more pronounced deformation at same depth before the vapor film of the droplet collapses. What's more, the peak pressure generated by steam explosion in seawater is notably higher than that in DI water. The fragmentation of molten tin droplet after the explosion is enhanced accordingly.
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| 40. |
- Deng, Yucheng, 1993-
(författare)
-
Experimental Study on Steam Explosions in Chemical Solutions and Seawater
- 2024
-
Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- Steam explosions may be encountered in severe accidents of light water reactors (LWRs), which are thermal detonations caused by rapid and intense vaporization of the coolant upon its direct contact with the core melt (corium). Motivated by the interest in understanding and mitigation of severe accident progression, many studies have been conducted to investigate the steam explosion phenomena during severe accidents. However, most of the previous studies did not consider the effect of chemical additives in the coolant of nuclear power plants, such as additions of H3BO3, NaOH and Na3PO4 for water chemistry control, and direct utilization of seawater (NaCl additive) under an extreme condition like the Fukushima accident. The present thesis work is motivated to fill the knowledge gap concerning the impacts of chemical additives (H3BO3, NaOH, Na3PO4, and NaCl) on steam explosions. The primary objective of the present research is to obtain characteristics of steam explosions in seawater and chemical solutions of H3BO3/NaOH/Na3PO4 with prototypical concentrations. To achieve this goal, a series of experiments have been carried out in the MISTEE experimental platform at KTH, involving single droplet and multiple droplets falling into a variety of coolant pool filled with seawater or chemical solutions of H3BO3/NaOH/Na3PO4 additives. The thesis work consists of four parts as follows.The first part is a description of the experimental methodology developed in the present study. Two experimental facilities, dubbed MISTEE-CE and MISTEE-SEA of respective mechanical plug and aerodynamic levitation for melt delivery, were designed on the MISTEE platform. Both setups were equipped with high-speed cameras for visualization, a pressure sensor for dynamic pressure measurement, and a fragment catcher for debris collection. A double-crucible design was employed to enable induction heating while avoiding melt contamination. The aerodynamic levitation system was implemented in MISTEE-SEA to reduce the disturbance of the mechanical plug. All chemical solutions were prepared in the laboratory with degassed deionized water. Tin (Sn) was chosen as the melt material due to benign properties suitable for safe handling in the laboratory.The second part is the presentation of visual observations and parameters selected to characterize steam explosions. The visualization includes the phenomena of droplet-coolant interactions and steam explosion occurrences. A molten single droplet falling into the coolant pool with deionized water or chemical solution might experience one of the three typical phenomena: deformation without fragmentation, minor fragmentation, or spontaneous steam explosion. In contrast, a multi-droplet test might involve merging and multiple explosions of droplets, resulting in a more complex set of phenomena. The quench depth and the lateral deformation ratio were defined and used to analyze the dynamic process of a single droplet in the coolant, while the peak pressure was employed to compare steam explosion energetics. In addition, the size distribution of debris particles was scrutinized.The third part is a summary and highlights of the experimental study on single-droplet steam explosion in different chemical solutions, using 1g of melt sample. The results revealed that the H3BO3 additive had little impact on steam explosion when the H3BO3 concentration was lower than 1.2 wt.%, but the risk of steam explosion in 3.2 wt.% H3BO3 solution was higher. The addition of NaOH and Na3PO4 to an H3BO3 solution significantly offset the influence on steam explosion. This suggests that the presence of PO43- and H+ ions play a significant role in spontaneous steam explosions. Additionally, seawater enhanced the occurrence of spontaneous steam explosions, with a clear correlation between increasing salinity and a higher likelihood of steam explosion. Compared to deionized water, chemical solutions (including seawater) caused more pronounced deformation in molten droplets at equivalent depths prior to direct contact of melt with coolant. Furthermore, the peak pressures of steam explosions in chemical solutions had the potential to reach notably higher values than those in deionized water. The chemical solutions except for the one of 1.2 wt.% H3BO3 tended to produce higher fractions of finer debris particles. The fourth part is about the experimental results of an investigation on steam explosion involving multiple droplets falling into deionized water and chemical solutions, using 5 g and 20 g of melt samples, respectively. It was found that under identical test conditions, the peak pressure of steam explosion increased with melt sample mass, resulting in a noticeably higher fraction of fine debris particles in the case of 20 g melt sample. The steam explosion location was concentrated within a shallower range when using chemical solutions instead of deionized water. In contrast to single-droplet experiments, the influence of the chemical solutions on the steam explosion was diminishing in the tests with multiple droplets.
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| 41. |
- Dietrich, P., et al.
(författare)
-
Coupling of melcor with the pecm for improved modelling of a core melt in the lower plenum
- 2015
-
Ingår i: International Conference on Nuclear Engineering, Proceedings, ICONE. - : JSME.
-
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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| 42. |
- Dietrich, P., et al.
(författare)
-
Extension of the MELCOR code for analysis of late in-vessel phase of a severe accident
- 2015
-
Ingår i: IYCE 2015 - Proceedings. - : IEEE conference proceedings. - 9781467371728
-
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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| 43. |
- Fichot, F., et al.
(författare)
-
A revised methodology to assess in-vessel retention strategy for high-power reactors
- 2018
-
Ingår i: PROCEEDINGS OF THE 26TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, 18, VOL 7. - : The American Society of Mechanical Engineers. - 9780791851517
-
Konferensbidrag (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 design and SAM guidances (SAMGs) of several operating small and medium capacity LWRs (reactors below 500 MWe, e.g. VVER440) and is part of the SAMG strategies for some Gen III+ PWRs of higher power such as the AP1000 or the APR1400. However, the demonstration of IVR feasibility for high power reactors requires using less conservative models as the safety margins are reduced. In Europe, the IVMR project aims at providing new experimental data and a harmonized methodology for IVR. A synthesis of the methodology applied to demonstrate the efficiency of IVR strategy for VVER-440 in Europe (Finland, Slovakia, Hungary and Czech Republic) was made. It showed very consistent results, following quite comparable methodologies. The main weakness 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 W/m(2)) whereas the first type provides the lowest heat fluxes (around 500 kW/m(2)) but this model is not realistic due to the immiscibility of molten steel with oxide melt. Obviously, there is a need to reach a consensus about best estimate practices for IVR assessment to be used in the major codes used 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 can reach, in many cases, values which are well above 1 MW/m(2). This could reduce the residual thickness of the vessel considerably and threaten its strength and integrity. Therefore, it is clear that the safety demonstration of IVR in 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 the focusing effect or significantly overheated metal, e.g. after oxide and metal layer inversion. Both situations are illustrated in this paper. The demonstration also requires an accurate thermo-mechanical analysis of the ablated vessel. The standard approach based on "yield stress" (plastic behaviour) is compared with more detailed calculations made on realistic profiles of ablated vessels. The validity of the standard approach is discussed. The current approach followed by many 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, e.g. associated with molten pool transient behaviour, 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. Some elements that might help to reach such harmonization are proposed in this paper, with a preliminary revision of the methodology that could be used to address the IVR issue. In the proposed revised methodology, the safety criterion is not based on a comparison of the heat flux and the Critical Heat Flux (CHF) profiles as in the current approaches but on the minimum vessel thickness reached after ablation and the maximum pressure load that is applied to the vessel during the transient. The main advantage of this revised criterion is in consideration of both steady-state and transient loads on the RPV. Another advantage is that this criterion is more straightforward to be used in a deterministic approach.
|
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| 44. |
- Fichot, F., et al.
(författare)
-
Some considerations to improve the methodology to assess In-Vessel Retention strategy for high-power reactors
- 2018
-
Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 119, s. 36-45
-
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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| 45. |
- Fischer, M., et al.
(författare)
-
Core melt stabilization concepts for existing and future LWRs and associated R&D needs
- 2015
-
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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| 46. |
- Fischer, Manfred, et al.
(författare)
-
Core Melt Stabilization Concepts for Existing and Future LWRs and Associated Research and Development Needs
- 2016
-
Ingår i: NUCLEAR TECHNOLOGY. - : American Nuclear Society. - 0029-5450. ; 196:3, s. 524-537
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Tidskriftsartikel (refereegranskat)abstract
- In the event of a severe accident in a nuclear power plant with the core melting, 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 to internal structures. The related failure modes may result in significant long-term radiological consequences and related high costs. Because of this, the licensing frameworks of several countries now include a requirement to implement mitigative core melt stabilization measures. This applies not only to new builds but also to existing light water reactors. 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 the molten core-concrete interaction with top flooding, water injection from the bottom (COMET), and retention in an outside-cooled crucible structure. This overview covers the physical background and functional principles of these concepts, as well as their validation status and, if applicable, the remaining open issues and research and development needs. For the concepts based on melt retention inside a cooled crucible that have reached sufficient maturity to be implemented in current Generation III+ designs, like the VVER-1000/1200 and the European Pressurized Water Reactor, 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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| 47. |
- Granovsky, V. S., et al.
(författare)
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Modeling of melt retention in EU-APR1400 ex-vessel core catcher
- 2012
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Ingår i: International Congress on Advances in Nuclear Power Plants 2012, ICAPP 2012. - 9781622762101 ; , s. 1412-1421
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Konferensbidrag (refereegranskat)abstract
- A core catcher is adopted in the EU-APR1400 reactor design for management and mitigation of severe accidents with reactor core melting. The core catcher concept incorporates a number of engineering solutions used in the catcher designs of European EPR and Russian WER-1000 reactors, such as thin-layer corium spreading for better cooling, retention of the melt in a water-cooled steel vessel, and use of sacrificial material (SM) to control the melt properties. SM is one of the key elements of the catcher design and its performance is critical for melt retention efficiency. This SM consists of oxide components, but the core catcher also includes sacrificial steel which reacts with the metal melt of the molten corium to reduce its temperature. The paper describes the required properties of SM. The melt retention capability of the core catcher can be confirmed by modeling the heat fluxes to the catcher vessel to show that it will not fail. The fulfillment of this requirement is demonstrated on the example of LBLOCA severe accident. Thermal and physicochemical interactions between the oxide and metal melts, interactions of the melts with SM, sacrificial steel and vessel, core catcher external cooling by water and release of non-condensable gases are modeled.
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| 48. |
- Granovsky, V. S., et al.
(författare)
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Oxidation effect on steel corrosion and thermal loads during corium melt in-vessel retention
- 2014
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Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 278, s. 310-316
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Tidskriftsartikel (refereegranskat)abstract
- During a severe accident with core meltdown, the in-vessel molten core retention is challenged by the vessel steel ablation due to thermal and physicochemical interaction of melt with steel. In accidents with oxidizing atmosphere above the melt surface, a low melting point UO2+x-ZrO2-FeOy corium pool can form. In this case ablation of the RPV steel interacting with the molten corium is a corrosion process. Experiments carried out within the International Scientific and Technology Center's (ISTC) METCOR Project have shown that the corrosion rate can vary and depends on both surface temperature of the RPV steel and oxygen potential of the melt. If the oxygen potential is low, the corrosion rate is controlled by the solid phase diffusion of Fe ions in the corrosion layer. At high oxygen potential and steel surface layer temperature of 1050 degrees C and higher, the corrosion rate intensifies because of corrosion layer liquefaction and liquid phase diffusion of Fe ions. The paper analyzes conditions under which corrosion intensification occurs and can impact on in-vessel melt retention (IVR).
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| 49. |
- Grishchenko, Dmitry, et al.
(författare)
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Insight into steam explosion in stratified melt-coolant configuration
- 2013
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Ingår i: 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics.
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Konferensbidrag (refereegranskat)abstract
- Release of core melt from failed reactor vessel into a pool of water is adopted in several existing designs of light water reactors (LWRs) as an element of severe accident mitigation strategy. When vessel breach is large and water pool is shallow, released corium melt can reach containment floor in liquid form and spread under water creating a stratified configuration of melt covered by coolant. Steam explosion in such stratified configuration was long believed as of secondary importance for reactor safety because it was assumed that considerable mass of melt cannot be premixed with the coolant. In this work we revisit these assumptions using recent experimental observations from the stratified steam explosion tests in PULiMS facility. We demonstrate that (i) considerable melt-coolant premixing layer can be formed in the stratified configuration with high temperature melts, (ii) mechanism responsible for the premixing is apparently more efficient than previously assumed Rayleigh-Taylor or Kelvin-Helmholtz instabilities. We also provide data on measured and estimated impulses, energetics of steam explosion, and resulting thermal to mechanical energy conversion ratios.
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| 50. |
- Grishchenko, Dmitry, et al.
(författare)
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Sensitivity Study of Steam Explosion Characteristics to Uncertain Input Parameters Using TEXAS-V Code
- 2014
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Ingår i: NUTHOS10, Paper-1293, Okinawa, Japan, 2014. - : Atomic Energy Society of Japan.
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Konferensbidrag (refereegranskat)abstract
- Release of core melt from failed reactor vessel into a pool of water is adopted in several existing designs of light water reactors (LWRs) as an element of severe accident mitigation strategy. Corium melt is expected to fragment, solidify and form a debris bed coolable by natural circulation. However, steam explosion can occur upon melt release threatening containment integrity and potentially leading to large early release of radioactive products to the environment. There are many factors and parameters that could be considered for prediction of the fuel-coolant interaction (FCI) energetics, but it is not clear which of them are the most influential and should be addressed in risk analysis. The goal of this work is to assess importance of different uncertain input parameters used in FCI code TEXAS-V for prediction of the steam explosion energetics. Both aleatory uncertainty in characteristics of melt release scenarios and water pool conditions, and epistemic uncertainty in modeling are considered. Ranges of the uncertain parameters are selected based on the available information about prototypic severe accident conditions in a reference design of a Nordic BWR. Sensitivity analysis with Morris method is implemented using coupled TEXAS-V and DAKOTA codes. In total 12 input parameters were studied and 2 melt release scenarios were considered. Each scenario is based on 60,000 of TEXAS-V runs. Sensitivity study identified the most influential input parameters, and those which have no statistically significant effect on the explosion energetics. Details of approach to robust usage of TEXAS-V input, statistical enveloping of TEXAS-V output and interpretation of the results are discussed in the paper. We also provide probability density function (PDF) of steam explosion impulse estimated using TEXAS-V for reference Nordic BWR. It can be used for assessment of the uncertainty ranges of steam explosion loads for given ranges of input parameters.
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