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Sample records for corium

  1. Corium spreading issue; Le corium et son etalement

    Energy Technology Data Exchange (ETDEWEB)

    Cognet, G.; Brayer, C.; Cranga, M.; Journeau, C.; Laffont, G.; Splinder, B.; Veteau, J.M. [CEA Grenoble, Dept. de Thermohydraulique et de Physique (DPT), 38 (France)

    1999-07-01

    Safety is one of the major issues for nuclear power plants; its improvement is a constant R and D axis for the CEA. In the event of a highly unlikely core melt-down accident in Light Water Reactors, the Safety Authorities of several EU countries have requested the industries and utilities to consider severe accidents with reactor pressure vessel failure for the design of the next generation of nuclear power plants. The objective is to preserve the integrity of the containment as the main barrier of fission product release to the environment. This can only be achieved if the core melt mixture (called corium, essentially composed of UO{sub 2}, ZrO{sub 2}, Zr, Fe and fission products) is stabilized before it can penetrate the basement. Consequently, various core-catcher concepts are under investigation for future reactors in order to prevent basement erosion, and to stabilize and control the corium within the containment. In particular, in the EPR (European Pressurized Reactor) core-catcher concept, the corium is mixed with a special concrete, and the molten mixture spread over a large multi-layer surface cooled from the bottom; with subsequent cooling by flooding with water. Therefore, melt spreading requires intensive investigation in order to determine and quantify the key phenomena, which govern the spreading. For some years now, the Nuclear Reactor Division of the Atomic Energy Commission (CEA/DRN) has been conducting a large program to improve knowledge on corium behaviour and coolability. This program is based on experimental (with simulant and prototypic materials) and theoretical investigations, which are finally gathered into scenario and mechanistic computer codes. Within this framework, a large part is currently devoted to the study of corium spreading. After a reminder of the general objectives and a description of the DRn approach and facilities, this paper presents the most important results. (authors)

  2. Performance testing of engineered corium cooling systems

    Energy Technology Data Exchange (ETDEWEB)

    Lomperski, S., E-mail: lomperski@anl.gov [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439-4840 (United States); Farmer, M.T. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439-4840 (United States)

    2012-02-15

    Highlights: Black-Right-Pointing-Pointer Experiments tested two engineered corium cooling systems. Black-Right-Pointing-Pointer The systems passively inject water into corium from below. Black-Right-Pointing-Pointer These systems cool corium much faster than top flooding. - Abstract: The coolability of ex-vessel core debris continues to be an issue of concern in the realm of light water reactor safety. Extensive research into corium/concrete interaction phenomena has been unable to establish the certainty of melt quench and stabilization within the containment boundary for all credible cases of cooling restricted to top flooding. As a result, there has been continuing interest in engineered systems that can augment cooling. This paper describes the testing of two passive cooling concepts that inject water into corium from below via nozzles embedded within the basemat: one with porous concrete nozzles and the other with a type of composite nozzle. The latter supplements water injection with noncondensable gas to stabilize flow and suppress vapor explosions. Each test involved a 136 kg melt composed of 56/23/14 wt% UO{sub 2}/ZrO{sub 2}/siliceous concrete at an initial depth of 30 cm. The setup with the porous concrete nozzles successfully injected water into the melt at heads as low as 2.3 m. The composite nozzle test was partially successful, with three nozzles delivering coolant while a fourth was damaged by the melt and failed to inject water. The melts cooled twice as fast as similar ones tested in a top flooding configuration. These experiments confirmed earlier work at Forschungszentrum Karlsruhe and elsewhere indicating that cooling via bottom water injection is a particularly effective method for quenching ex-vessel corium melts.

  3. Analysis of the corium phases by X-ray diffraction; Analyses des phases du corium par diffraction des rayons X

    Energy Technology Data Exchange (ETDEWEB)

    Trillon, G

    2004-07-01

    In the framework of the severe accidents R and D studies led by CEA, the better knowledge of the corium behaviour, corium coming from the melting of a nuclear reactor, are fundamental stakes in order to master this kind of accident. Among the available physical properties of the corium, the nature of the final crystalline compounds which have been made during the, cooling gives information about its solidification and its stabilisation. X-Rays Diffraction is the reference method used in order to characterize the corium coming from the different facilities of the European platform PLINIUS of CEA-Cadarache. This work presents the scientific approach that has been followed in order to obtain information both qualitative and quantitative on corium, using X-Rays Diffraction. For instance, a specific method for identifying U{sub 1-x}Zr{sub x}O{sub 2} solid solutions has been developed, and the validity of quantitative analysis of corium crystalline phases using the Rietveld method (with an internal standard), has been tested. This last method has also permitted semi-quantitative measurements of amorphous phases within corium. For these studies, analysis of prototypical corium has been conducted on samples coming from the experiences led on the different facilities of the PLINIUS platform. These analysis allowed for the first time to obtain quantitative data of the corium crystalline phases in order to validate thermodynamic databases and has been used to estimate the thereto-physical properties of the corium. New information on crystalline phases of corium has also been found, especially for the UO{sub 2}-ZrO{sub 2} pseudo binary system. (author)

  4. Final synthesis of Sarnet (Phase 1) corium activities; Synthese finale des activites corium du reseau Sarnet (Phase 1)

    Energy Technology Data Exchange (ETDEWEB)

    Journeau, Ch. [CEA Cadarache, 13 - Saint Paul lez Durance (France); Steinbruck, M. [Forschungszentrum Juelich GmbH (Germany); Repetto, G.; Duriez, Ch.; Koundy, V. [Institut de Radioprotection et de Surete Nucleaire (IRSN), 92 - Clamart (France); Ma, W.M. [Kungliga Tekniska Hoegskolan (KTH), Stockholm (Sweden); Burger, M. [Institut fur Kernenergetik und Energiesysteme, Stuttgart (IKE) (Germany); Spindler, B. [CEA Grenoble, 38 (France)

    2009-07-01

    Within the SARNET Severe Accident Research Network of excellence, the Corium topic covers all the behaviour of corium (mixture formed by the molten materials arising from a postulated nuclear reactor severe accident) 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 to improve significantly the corium behaviour 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 and 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 analytical experiments, 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 report summarizes the progress that has been achieved in the frame of the networking activities for the four and half years of the FP6 project. (authors)

  5. A study on corium behaviour under external vessel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae Joon; Kim, Sang Baik; Kang, Kyung Ho; Koo, Kil Mo; Kim, Hee Dong

    2000-04-01

    This study presents the results of evaluation and analysis on the second phase of the RASPLAV project for three years between July 1, 1997 and June 30, 2000. In the RASPLAV Phase II study, two large-scale experiments of RASPLAV-AW-200-3, 4 were conducted to estimate the heat flux distribution in the corium and thermal interaction between the corium and the reactor vessel. Several small-scale experiments such as TULPAN, TF, and STF were conducted to analyze thermal stratification and additive effect of core materials on corium behavior. The Salt experiments were conducted to estimate the crust and the mushy region formation, as well as natural convection heat transfer in the corium. Material properties of the corium and the salt were measured in the RASPLAV project. During the RASPLAV-AW-200-3 test, approximately 22 kg of the corium leaked from the test furnace, because Fe from the FeO, which was additive to reduce the melting temperature of fuel pellet, interacted with Tungsten protector. It is concluded from the AW-200-3 test results that the oxidized U-Zr-O is not separated. From the RASPLAV-AW-200-4 test results, the C-32 fuel with the miscibility gap and low content of carbon was not separated thermally. The carbon is known as a dominant factor in the thermal stratification of the corium from the small and medium scale test results such as TULPAN, TF, and STF. The fuel composition, test method and condition in the RASPLAV-AW-2003,4 were selected using the small and medium scale test results. It is confirmed from the Salt test that the analytical model of the CONV code predicts heat transfer with crust formation in the molten pool very well.

  6. Effect of Multi-Layered Corium Formations on Integrity of Steel Components under Steam Explosion Condition

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Hyun; Kim, Tae Hyun; Chang, Yoon-Suk [Kyung Hee University, Yongin (Korea, Republic of); Cho, Yong-Jin [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    The object of the present study is to examine effect of multi-layered corium formations on the integrity of steel components under a representative steam explosion condition. In this context, multi-layered corium formation conditions are assumed based on a previous study. Subsequently, stress evaluation of steel components is performed by TNT (trinitrotoluene) model for the steam explosion analysis and their results are discussed. In this paper, comparative numerical analyses were carried out to examine effect of the multi-layered corium formations on integrity of steel components under a typical steam explosion condition and the following conclusions were derived. (1) The highest maximum von Mises stress was calculated at RPV. However, stress values of all components did not exceed their yield strengths. (2) Effect of the 3-layer corium formation was higher than 2-layer corium formation. Resulting von Mises stress increased 20% than that of no corium formation and 16% than that of 2-layer corium formation.

  7. Steam explosion triggering phenomena: stainless steel and corium-E simulants studied with a floodable arc melting apparatus. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, L.S.; Buxton, L.D.

    1978-05-01

    Laboratory-scale experiments on the thermal interaction of light water reactor core materials with water have been performed. Samples (10--35 g) of Type 304 stainless steel and Corium-E simulants were each flooded with approximately 1.5 litres of water to determine whether steam explosions would occur naturally. Many of the experiments also employed artificially induced pressure transients in an attempt to initiate steam explosions. Vigorous interactions were not observed when the triggering pulse was not applied, and for stainless steel the triggering pulse initiated only coarse fragmentation. Two-stage, pressure-producing interactions were triggered for an ''oxidic'' Corium-E simulant. An impulse-initiated gas release theory has been simulated to explain the initial sample fragmentation. Although the delayed second stage of the event is not fully understood, it does not appear to be readily explained with classical vapor explosion theory. Rather, some form of metastability of the melt seems to be involved.

  8. Steel oxidation phenomena during Molten Corium siliceous Concrete Interaction (MCCI)

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez-Brusset, Mathieu; Piluso, Pascal [CEA/DEN/Cadarache, SMTA/LPMA, 13108 St. Paul lez-Durance (France); Balat-Pichelin, Marianne [PROMES-CNRS Laboratory, 7 rue du four solaire, 66120 Font-Romeu Odeillo (France); Bottomley, Paul David; Wiss, Thierry [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe, German (Germany)

    2015-02-15

    Highlights: • Corium metallic phase oxidation during corium-concrete interaction is studied. • Steel is separated from the oxide melt or emulsified inside the oxide melt. • Oxidation layer depends on the nature of the interfaces and location in the corium. • Oxides formed are (Fe,Cr){sub 3}O{sub 4} and Cr{sub 2}O{sub 3}. • Concrete gases are not sufficient to explain the experimental steel oxidation. - Abstract: The VULCANO facility at CEA Cadarache is a Molten Corium Concrete Interaction (MCCI) installation for testing material reactions representative of the late stages of a nuclear reactor severe accident. The objectives of the VBS-U3 test were to study ablation phenomena and oxidation of the metallic phase when two liquid phases are present: oxide phase and metallic phase (steel). In this paper we describe the materials post-test analysis of the VULCANO VBS-U3 test performed at the Institute for Transuranium Elements in Karlsruhe (JRC-ITU) with the focus on the metallic phase oxidation of the corium. Post-test analyses show that the remaining metallic phase of the corium is under two forms: drops discontinuously dispersed in the oxide phase forming an emulsion and a continuous metallic ingot clearly separated from the oxide phase. In average, taking into account or not the metallic phase dispersed in the oxide phase, between 60% and 70% of the steel has been oxidized. The size of the drops and their proportion in the oxide phase is depending on their distance from horizontal and vertical walls of the concrete test section. Oxidation mechanisms are mainly depending on two parameters: nature of the metallic interface and localization in the test section. Calculations at thermodynamic equilibrium show that the only product from steel oxidation is (Fe,Cr){sub 3}O{sub 4}, Cr{sub 2}O{sub 3} is never formed. Moreover taking into account the two gaseous species coming from the concrete (CO{sub 2} and H{sub 2}O), considered up to now as being the only sources

  9. Results of recent KROTOS FCI tests. Alumina vs. corium melts

    Energy Technology Data Exchange (ETDEWEB)

    Huhtiniemi, I.; Magallon, D.; Hohmann, H. [Commission of the European Communities, Ispra (Italy). Joint Research Center

    1998-01-01

    Recent results from KROTOS fuel-coolant interaction experiments are discussed. Five tests with alumina were performed under highly subcooled conditions, all of these tests resulted in spontaneous steam explosions. Additionally, four tests were performed at low subcooling to confirm, on one hand, the suppression of spontaneous steam explosions under such conditions and, on the other hand, that such a system is still triggerable using an external initiator. The other test parameters in these alumina tests included the melt superheat and the initial pressure. All the tests in the investigated superheat range (150 K - 750 K) produced a steam explosion and no evidence of the explosion suppression by the elevated initial pressure (in the limited range of 0.1 - 0.375 MPa) was observed in the alumina tests. The corium test series include a test with 3 kg of melt under both subcooled and near saturated conditions at ambient pressure. Two additional tests were performed with subcooled water; one test was performed at an elevated pressure of 0.2 MPa with 2.4 kg of melt and another test with 5.1 kg of melt at ambient pressure. None of these tests with corium produced a propagating energetic steam explosion. However, propagating low energy (about twice the energy of the trigger pulse) events were observed. All corium tests produced significantly higher water level swells during the mixing phase than the corresponding alumina tests. Present experimental evidence suggests that the water depletion in the mixing zone suppresses energetic steam explosions with corium melts at ambient pressure and in the present pour geometry. Processes that could produce such a difference in void generation are discussed. (author)

  10. Experiments on the dispersion of corium; Experimente zur Dispersion von Corium

    Energy Technology Data Exchange (ETDEWEB)

    Gargallo, M.; Greulich, M.; Kirstahler, M.; Meyer, L.; Schwall, M.; Wachter, E.; Woerner, G. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Kern- und Energietechnik

    2000-08-01

    Experiments are being performed in a scaled annular cavity design, typical for the European Pressurized Reactor (EPR), to investigate melt dispersal from the reactor cavity when the reactor pressure vessel lower head fails at low system pressure of less than 20 bar. In the first part of the experimental program the fluid dynamics of the dispersion process has been studied using model fluids, water or bismuth alloy instead of corium, and nitrogen or helium instead of steam. The effects of different breach sizes, - central holes in the lower head in this study, and different failure pressures on the dispersion have been studied systematically. For central holes in the lower head, the dispersed liquid fractions can be correlated by a Kutateladze number based on the velocity in the annular space of the cavity. (orig.) [German] Das Versagen des RDB unter vollem Systemdruck wird durch systemtechnische Massnahmen verhindert. Ein Versagen bei einem zu unterstellenden Druck von 10 bis 20 bar kann u.U. noch zu starker Dispersion und Umverteilung der Schmelze in die Reaktorraeume und den Dom und damit zur direkten Aufheizung der Containmentatmosphaere fuehren. Auch muesste das fuer das Konzept der Kernschmelzekuehleinrichtung beruecksichtigt werden. Das Zusammenwirken der hierbei auftretenden Materialtransportprozesse sowie der termischen und chemischen (Wasserstofferzeugung und -verbrennung) Wechselwirkungen ist komplex und kann derzeit mit Computercodes noch nicht oder nur bedingt berechnet werden. Notwendig sind daher Einsichten und Daten aus experimentellen Untersuchungen, insbesondere auch integrale Experimente, deren Ergebnisse mit Modellen und Coderechnungen auf die grosse EPR-Anlage uebertragen werden koennen. Das Hauptiel der Untersuchungen ist die Bestimmung eines maximal zulaessigen Druckes zum Zeitpunkt des RDB-Versagens, bei dem keine relevante Dispersion der Schmelze auftritt, im Zusammenspiel mit einer fuer dieses Ziel evtl. zu optimierenden Grubengeometrie. Im

  11. Spreading of molten corium in MK I geometry following vessel melt-through

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, J.J.; Farmer, M.T.; Spencer, B.W.

    1988-01-01

    For Mk I boiling water reactor severe-accident sequences in which molten corium is postulated to melt through the reactor pressure vessel (RPV) lower head, an important question concerns the relocation of the corium material that drains from the vessel. After filling the sump pits located in the pedestal concrete floor beneath the RPV, the molten corium that collects on the pedestal floor is generally free to flow through the doorway, which provides personnel access to the pedestal, and to spread out over the concrete floor in the annular region between the pedestal wall and the steel liner of the containment shell. A significant issue is whether the corium, after exiting the doorway, can spread under gravity all the way to the liner where thermal attack on the liner steel might be postulated to occur. A computer code (MELTSPREAD) has been developed to investigate the spreading dynamics and thermal interactions of a molten corium layer flowing horizontally over an ablating concrete substrate that may be initially covered with water. The principal objective is to predict, for specific conditions of corium composition, mass, and temperature, the lateral penetration of the corium that drains from a localized hole in the lower head immediately following RPV failure.

  12. FCI experiments in the corium/water system

    Energy Technology Data Exchange (ETDEWEB)

    Huhtiniemi, I.; Hohmann, H.; Magallon, D.

    1995-09-01

    The KROTOS fuel coolant interaction (FCI) tests aim at providing benchmark data to examine the effect of fuel/coolant initial conditions and mixing on explosion energetics. Experiments, fundamental in nature, are performed in well-controlled geometries and are complementary to the FARO large scale tests. Recently, a new test series was started using 3 kg of prototypical core material (80 w/o UO{sub 2}, 20 w/o ZrO{sub 2}) which was poured into a water column of {le} 1.25 m in height (95 mm and 200 mm in diameter) under 0.1 MPa ambient pressure. Four tests have been performed in the test section of 95 mm in diameter (ID) with different subcooling levels (10-80K) and with and without an external trigger. Additionally, one test has been performed with a test section of 200 mm in diameter (ID) and with an external trigger. No spontaneous or triggered energetic FCIs (steam explosions) have been observed in these corium tests. This is in sharp contrast with the steam explosions observed in the previously reported Al{sub 2}O{sub 3} test series which had the same initial conditions of ambient pressure and subcooling. The post-test analysis of the corium experiments indicated that strong vaporisation at the melt/water contact led to a partial expulsion of the melt from the test section into the pressure vessel. In order to avoid this and to obtain a good penetration and premixing os the corium melt, an additional test has been performed with a larger diameter test section. In all the UO{sub 2}-ZrO{sub 2} tests an efficient quenching process (0.7-1.2 MW/kg-melt) with total fuel fragmentation (mass mean diameter 1.4-2.5 mm) was observed. Results from Al{sub 2}O{sub 3} tests under the same initial conditions are also presented for further confirmation of the observed differences in behaviour between Al{sub 2}O{sub 3} and UO{sub 2}-ZrO{sub 2} melts.

  13. Simulation of corium concrete interaction in 2D geometry

    Energy Technology Data Exchange (ETDEWEB)

    Cranga, M. [IRSN, DPAM, F-13115 St Paul Les Durance (France); Spindler, B.; Dufour, E. [CEA Grenoble, DEN, F-38000 Grenoble (France); Dimov, Dimitar [Bulgarian Acad Sci, Inst Nucl Res and Nucl Energy, NPPSAL, BU-1784 Sofia (Bulgaria); Atkhen, Kresna [EDF, SEPTEN, F-69628 Villeurbanne (France); Foit, Jerzy [Forschungszentrum Karlsruhe, D-76021 Karlsruhe (Germany); Garcia-Martin, M. [Univ Politecn Madrid, E-28006 Madrid (Spain); Sevon, Tuomo [Tech Res Ctr Finland VTT, FI-02044 Helsinki (Finland); Schmidt, W. [AREVA, D-91058 Erlangen (Germany); Spengler, C. [Gesell Anlagen and Reaktorsicherheit GRS mbH, D-50667 Cologne (Germany)

    2010-07-01

    Benchmarking work was recently performed for the issue of molten corium concrete interaction (MCCI). A synthesis is given here. It concerns first the 2D CCI-2 test with a homogeneous pool and a limestone concrete, which was used for a blind benchmark. Secondly, the COMET-L2 and COMET-L3 2D experiments in a stratified configuration were used as a post-test (L2) and a blind-test (L3) benchmark. More details are given here for the recent benchmark considering a matrix of four reactor cases, with both a homogeneous and a stratified configuration, and with both a limestone and a siliceous concrete. A short overview is given on the different models used in the codes, and the consistency between the benchmark actions on experiments and reactor situations is discussed. Finally, the major uncertainties concerning MCCI are also pointed out. (authors)

  14. Zirconium carbide coating for corium experiments related to water-cooled and sodium-cooled reactors

    Science.gov (United States)

    Plevacova, K.; Journeau, C.; Piluso, P.; Zhdanov, V.; Baklanov, V.; Poirier, J.

    2011-07-01

    Since the TMI and Chernobyl accidents the risk of nuclear severe accident is intensively studied for existing and future reactors. In case of a core melt-down accident in a nuclear reactor, a complex melt, called corium, forms. To be able to perform experiments with prototypic corium materials at high temperature, a coating which resists to different corium melts related to Generation I and II Water Reactors and Generation IV sodium fast reactor was researched in our experimental platforms both in IAE NNC in Kazakhstan and in CEA in France. Zirconium carbide was selected as protective coating for graphite crucibles used in our induction furnaces: VCG-135 and VITI. The method of coating application, called reactive wetting, was developed. Zirconium carbide revealed to resist well to the (U x, Zr y)O 2-z water reactor corium. It has also the advantage not to bring new elements to this chemical system. The coating was then tested with sodium fast reactor corium melts containing steel or absorbers. Undesirable interactions were observed between the coating and these materials, leading to the carburization of the corium ingots. Concerning the resistance of the coating to oxide melts without ZrO 2, the zirconium carbide coating keeps its role of protective barrier with UO 2-Al 2O 3 below 2000 °C but does not resist to a UO 2-Eu 2O 3 mixture.

  15. Lessons learnt from FARO/TERMOS corium melt quenching experiments

    Energy Technology Data Exchange (ETDEWEB)

    Magallon, D.; Huhtiniemi, I.; Hohmann, H. [Commission of the European Communities, Ispra (Italy). Joint Research Center

    1998-01-01

    The influence of melt quantity, melt composition, water depth and initial pressure on quenching is assessed on the basis of seven tests performed in various conditions in the TERMOS vessel of the FARO facility at JRC-Ispra. Tests involved UO{sub 2}-based melt quantities in the range 18-176 kg at a temperature of approximately 3000 K poured into saturated water. The results suggest that erosion of the melt jet column is an efficient contributor to the amount of break-up, and thus quenching, for large pours of corium melt. The presence of Zr metal in the melt induced a much more efficient quenching than in a similar test with no Zr metal, attributed to the oxidation of the Zr. Significant amounts of H{sub 2} were produced also in tests with pure oxidic melts (e.g. about 300 g for 157 kg melt). In the tests at 5.0 and 2.0 MPa good mixing with significant melt break-up and quenching was obtained during the penetration in the water. At 0.5 MPa, good penetration of the melt into the water could still be achieved, but a jump in the vessel pressurisation occurred when the melt contacted the bottom and part (5 kg) of the debris was re-ejected from the water. (author)

  16. Simulation of In-Vessel Corium Retention through External Reactor Vessel Cooling for SMART using SIMPLE

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Jin-Sung; Son, Donggun; Park, Rae-Joon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Thermal load analysis from the corium pool to the outer reactor vessel in the lower plenum of the reactor vessel is necessary to evaluate the effect of the IVR-ERVC during a severe accident for SMART. A computational code called SIMPLE (Sever Invessel Melt Progression in Lower plenum Environment) has been developed for analyze transient behavior of molten corium in the lower plenum, interaction between corium and coolant, and heat-up and ablation of reactor vessel wall. In this study, heat load analysis of the reactor vessel for SMART has been conducted using the SIMPLE. Transient behavior of the molten corium in the lower plenum and IVR-ERVC for SMART has been simulated using SIMPLE. Heat flux from the corium pool to the outer reactor vessel is concentrated in metallic layer by the focusing effect. As a result, metallic layer shows higher temperature than the oxidic layer. Also, vessel wall of metallic layer has been ablated by the high in-vessel temperature. Ex-vessel temperature of the metallic layer was maintained 390 K and vessel thickness was maintained 14 cm. It means that the reactor vessel integrity is maintained by the IVR-ERVC.

  17. Experimental simulation of the water cooling of corium spread over the floor of a BWR containment

    Energy Technology Data Exchange (ETDEWEB)

    Morage, F.; Lahey, R.T. Jr.; Podowski, M.Z. [Rensselaer Polytechnic Institute, Troy, NY (United States)

    1995-09-01

    This paper is concerned with an experimental investigation of the cooling effect of water collected on the surface of corium released onto the floor of a BWR drywell. In the present experiments, the actual reactor materials were replaced by simulant materials. Specifically, the results are shown for Freon-11 film boiling over liquid Wood`s metal spread above a solid porous surface through which argon gas was injected. An analysis of the obtained experimental data revealed that the actual film boiling heat transfer between a molten pool of corium and the water above the pool should be more efficient than predicted by using standard correlations for boiling over solid surfaces. This effect will be further augmented by the gas released due to the ablation of concrete floor beneath the corium and percolating towards its upper surface and into through the water layer above.

  18. Spreading of molten corium in Mk I geometry following vessel meltthrough

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, J.J.; Farmer, M.T.; Spencer, B.W.

    1988-01-01

    A one-dimensional, multicell, Eulerian computer code is under development to predict the gravity-driven spreading dynamics and thermal interactions of a molten corium layer flowing horizontally over a concrete substrate. The code is compared to recent experiments in which molten mixtures of iron and aluminum oxide flowed over concrete in the presence and absence of water. Results are presented from scoping calculations for the Mk I BWR system investigating the spreading-induced penetration immediately following the drainage of a predominantly oxide molten corium mixture from a localized breach in the reactor vessel. 12 refs., 7 figs.

  19. Corium spreading: hydrodynamics, rheology and solidification of a high-temperature oxide melt; L'etalement du corium: hydrodynamique, rheologie et solidification d'unbain d'oxydes a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Journeau, Ch

    2006-06-15

    In the hypothesis of a nuclear reactor severe accident, the core could melt and form a high- temperature (2000-3000 K) mixture called corium. In the hypothesis of vessel rupture, this corium would spread in the reactor pit and adjacent rooms as occurred in Chernobyl or in a dedicated core-catcher s in the new European Pressurized reactor, EPR. This thesis is dedicated to the experimental study of corium spreading, especially with the prototypic corium material experiments performed in the VULCANO facility at CEA Cadarache. The first step in analyzing these tests consists in interpreting the material analyses, with the help of thermodynamic modelling of corium solidification. Knowing for each temperature the phase repartition and composition, physical properties can be estimated. Spreading termination is controlled by corium rheological properties in the solidification range, which leads to studying them in detail. The hydrodynamical, rheological and solidification aspects of corium spreading are taken into account in models and computer codes which have been validated against these tests and enable the assessment of the EPR spreading core-catcher concept. (author)

  20. Study of the rheological behaviour of corium/concrete mixtures; Etude du comportement rheologique de melanges issus de l'interaction corium/beton

    Energy Technology Data Exchange (ETDEWEB)

    Ramacciotti, M

    1999-09-24

    In the hypothetical event of a severe accident in a Light Water Reactor, scenarios in which the reactor pressure vessel (RPV) fails and the core melt mixture (called corium) relocates into the reactor cavity, cannot be excluded. The viscosity (in fact, corium rheological behaviour) plays a major role in many phenomena such as core melt down, discharge from reactor pressure vessel, interaction with structural materials (concrete,...) and spreading in a core-catcher. For these reasons, it is important to be able to predict the rheological behaviour of corium melts of different compositions (essentially based on UO{sub 2}, ZrO{sub 2}, Fe{sub x}O{sub y} and Fe for in-vessel scenarios, plus SiO{sub 2} and CaO for ex-vessel scenarios) at temperatures above solidus temperature. In the case of corium-concrete mixtures, the increase of viscosity depends not only on the increase of particles in the melts but also on the increase of the residual liquid phase viscosity (due to the increase in silica contents). The Urban correlation is used to calculate the viscosity of the carrying liquid with silica. This model was tested and gave good agreements between measured and estimated viscosities of various basalts among which one contained 18 wt% of UO{sub 2}. Then, in the solidification range, the analysis of published data showed that the viscosity cannot be described by a suspension viscosity model of non-interactive spherical particles; consequently we proposed an Arrhenius type law with a multiplying factor such as {eta}{sub r} = exp(2.5 C{phi}) and the C factor value varies between 4 and 8. This factor is more important in the case of low shear rates and low cooling rates. The analysis of the samples structure after quenching shows a dependence of this factor on the particle morphology. Finally, for a value of 6.1 of the C factor, we obtained the best agreement with experimental data for a corium spreading test at 2100 K on a horizontal surface. (author)

  1. KATS experiments to simulate corium spreading in the EPR core catcher concept

    Energy Technology Data Exchange (ETDEWEB)

    Eppinger, B.; Fieg, G.; Schuetz, W.; Stegmaier, U. [Forschungszentrum Karlsruhe, Insitute fur Kern- und Energietechnik, Karlsruhe (Germany)

    2001-07-01

    In future Light Water Reactors special devices (core catchers) might be required to prevent containment failure by basement erosion after reactor pressure vessel melt-through during a core meltdown accident. Quick freezing of the molten core masses is desirable to reduce release of radioactivity. Several concepts of core catcher de-vices have been proposed based on the spreading of corium melt onto flat surfaces with subsequent cooling by flooding with water. Therefore a series of experiments to investigate high temperature melt spreading on flat surfaces has been carried out using alumina-iron thermite melts as a simulant. The oxidic thermite melt is conditioned by adding other oxides to simulate a realistic corium melt as close as possible. Spreading of oxidic and metallic melts have been performed in one- and two-dimensional geometry. Substrates were chemically inert ceramic layers, dry concrete and concrete with a shallow water layer on top. (authors)

  2. Simulation of transient behavior of corium pool in the lower plenum of RPV using COMPASS

    Energy Technology Data Exchange (ETDEWEB)

    Son, Dong Gun; Bae, Jun Ho; Park, Rae Jun; Kim, Dong Ha; Kim, Hwan Yeol [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Development of an integrated severe accident analysis code has been started by the collaboration of three institutes in Korea. KAERI (Korea Atomic Energy Research Institute) is responsible for developing modules related to the in-vessel phenomena, including the corium behavior in the lower plenum of RPV. We developed computational software called COMPASS (COre Meltdown Progression Accident Simulation Software). SIMPLE module was created with the mass and energy equations of particulate debris bed, metallic molten pool, oxidic molten pool. It receives thermo-hydraulic conditions of the lower plenum, then returns total heat to the coolant and surrounding structures. After relocation of the corium to the lower plenum, most of them were remain particulate debris bed. RPV wall ablation starts after the oxidic materials were relocated, and there is solidified crust where the oxidic pool contact with RPV wall.

  3. The modeling of core melting and in-vessel corium relocation in the APRIL code

    Energy Technology Data Exchange (ETDEWEB)

    Kim. S.W.; Podowski, M.Z.; Lahey, R.T. [Rensselaer Polytechnic Institute, Troy, NY (United States)] [and others

    1995-09-01

    This paper is concerned with the modeling of severe accident phenomena in boiling water reactors (BWR). New models of core melting and in-vessel corium debris relocation are presented, developed for implementation in the APRIL computer code. The results of model testing and validations are given, including comparisons against available experimental data and parametric/sensitivity studies. Also, the application of these models, as parts of the APRIL code, is presented to simulate accident progression in a typical BWR reactor.

  4. Corium Configuration and Penetration Tube Failure for Fukushima Daiichi Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    An, Sang Mo; Lee, Jae Bong; Kim, Hwan Yeol; Song, Jin Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    For the LWRs (light water reactors), the penetration tubes at the reactor vessel lower head are regarded as the most vulnerable structures along with a global vessel failure during a severe accident because they can be seriously damaged by a corium melt or debris relocated into the lower plenum of the vessel. The research on the penetration tube failure is of higher importance in the BWRs, as it could lead to melt discharge into the containment and subsequent release of radioactive materials to the environment due to the containment failure. There are more than one hundred of penetration tubes in the Fukushima Daiichi NPPs (nuclear power plants), such as ICM-GTs (in-core monitoring guide tubes), CRGTs (control rod guide tubes) and drain tubes. The ICM-GTs include SRMs (source range monitors), IRMs (intermediate range monitors), LPRMs (local power range monitors) and TIPs (traversing in-core probes), which are much thinner than other tubes. The experimental researches to investigate the corium configuration and the penetration tube failure for the Fukushima Daiichi NPPs were introduced and some meaningful results were summarized. It was shown that the corium ingot was separated into two layers, of which the upper layer was metal-rich while the lower one was oxide-rich. It seemed that B{sub 4}C would contribute to reducing the density of the metallic melt. The two-layered configuration will provide useful information to understand the core melt progression and post-recovery actions for the Fukushima Daiichi NPPs. In addition, we performed a large scale penetration tube failure experiment for the SRM/IRM guide tube, and showed high possibilities of large amount of corium discharge out of the reactor vessel lower head, which followed by the tube melting in a very short time. We are planning to perform the penetration tube failure experiments for another dry tube of ICM-GT (LPRM guide tube), and later for the wet tube (CRGT)

  5. Study on corium behavior in the reactor cavity during severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Song, Y. M.; Kim, D. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    The report contains following four results of studies on molten corium-concrete interaction, which has been recognized as important aspects of severe reactor accident; 1. MELCOR code modification has been performed for heat transfer model between ex-vessel molten corium and overlying water pool. The existing model do not consider debris particulation and water penetration in the ex-vessel debris cooling. The new model employs dryout heat flux in determining the heat removal from a debris bed by water penetration. 2. A parametric model which can evaluate ex-vessel concrete erosions has been developed. The model is expected to evaluate the concrete erosion in a limited error range with only a little effort. The model has been derived by the sensitivity studies using MELCOR and MAAP programs. 3. During the corium-concrete interaction, there is a temperature distribution inside basemat concrete. MELCOR calculates concrete response based on one-dimensional steady-state ablation, with no consideration given to conduction into the concrete or to decomposition in advance of the ablation front. Thus there is a necessity to improve the concrete decomposition model in MELCOR. In this report the transient conduction model and the methodology of implementation into MELCOR were suggested. 4. Major modeling assumptions and limits of MELTSPREAD-1, which is a transient one-dimensional computer code to predict the gravity-driven spreading of molten corium in the reactor cavity under severe accidents, are evaluated via review of general conservation equations and used models. The models being reviewed include heat transfer models at melt lower/upper surfaces, a concrete dryout model, and a shell heatup model. The evaluation results suggest the degree of MELTSPREAD-1 approximation compared with real spreading flow and the strong/weak points or restrictions of the code. 17 refs., 19 figs., 6 tabs. (Author)

  6. Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hoon; Chang, Soon Heung; Choi, Tae Hoon; Kim, Hyun Sop; Yang, Soo Hyung; Kim, Soo Hyung; Kim, Seung Hop; An Hyung Taek; Jeong, Yong Hoon; Huh, Gyun Young [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-03-15

    Cooling methodologies for the molten corium resulted from the severe accident of the nuclear power plant is suggested as one of most important items for the safety of the NPP. In this regard, considerable experimental and analytical works have been devoted. In the 1st phase of this project, present status related to the external reactor vessel cooling for the retention of the corium in the reactor vessel and corium at the reactor cavity have been investigated and preliminary studies have been accomplished for the detail evaluation of the each cooling methodology. The preliminary studies include the analysis and detail investigation of the possible phenomena, investigation of the heat transfer correlations and preliminary evaluation of the external reactor vessel cooling using the developed computer code.

  7. Oxidation effect on steel corrosion and thermal loads during corium melt in-vessel retention

    Energy Technology Data Exchange (ETDEWEB)

    Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Sulatsky, A.A.; Almjashev, V.I. [Alexandrov Scientific-Research Technology Institute (NITI), Sosnovy Bor (Russian Federation); Bechta, S.V. [KTH, Stockholm (Sweden); Gusarov, V.V. [SPb State Technology University (SPbGTU), St. Petersburg (Russian Federation); Barrachin, M. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), St Paul lez Durance (France); Bottomley, P.D., E-mail: paul.bottomley@ec.europa.eu [EC-Joint Research Centre, Institute for Transuranium Elements (ITU), Karlsruhe (Germany); Fischer, M. [AREVA GmbH, Erlangen (Germany); Piluso, P. [Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA), Cadarache, St Paul lez Durance (France)

    2014-10-15

    Highlights: • The METCOR facility simulates vessel steel corrosion in contact with corium. • Steel corrosion rates in UO{sub 2+x}–ZrO{sub 2}–FeO{sub y} coria accelerate above 1050 K. • However corrosion rates can also be limited by melt O{sub 2} supply. • The impact of this on in-vessel retention (IVR) strategy is discussed. - 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 UO{sub 2+x}–ZrO{sub 2}–FeO{sub y} 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 °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)

  8. Internal structure of an ex-vessel corium debris bed during severe accidents of LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eunho; Park, Jin Ho; Moriyama, Kiyofumi; Park, Hyun Sun [POSTECH, Daejeon (Korea, Republic of)

    2015-10-15

    In the aspect of the coolability assessment the configuration of the debris bed, including internal and external characteristics, has significant importance as boundary conditions for simulations, however, relatively little investigation of the sedimentation process. For the development of a debris bed, recently there have been several studies that focused on thermal characteristics of corium particles. Yakush et al. performed simulation studies and showed that two phase natural convection affects the particle settling trajectory and changes the final arrival location of particles to result more flattened bed. Those simulation results have been supported by the experimental studies of Kim et al. using simulant particles and air bubble injection. For the internal structure of a debris bed, there have been several simulation and experimental studies, which investigated the effect of internal structure on debris bed coolability. Magallon has reported the particle size distribution at three elevations of the debris bed of FARO L-31 case, where the mean particle size was bigger for the lower elevation. However, there is a lack of detailed information on the characteristics of the debris bed, including the local structure and porosity. In this study, we investigated the internal structure of the debris bed using a mixture of stainless steel particles and air bubble injection. Local particle sedimentation quantity, particle size distribution change in radial direction and axial direction, and bed porosity was measured to investigate a relationship between the internal structure and the accident condition. An experimental investigation was carried out for the internal structure of ex-vessel corium debris bed in the flooded cavity during sever accident. Moderate corium discharge in high flooding level was assumed for full fragmentation of melt jet. The test particle mixture was prepared by following an empirical correlation, which reflects the particle size distribution of

  9. Ex-Vessel corium coolability and steam explosion energetics in nordic light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, T.N.; Ma, W.M.; Karbojian, A.; Kudinov, P.; Tran, C.T.; Hansson, C.R. [Royal Institute of Technology (KTH), (Sweden)

    2008-03-15

    This report presents advances and insights from the KTH's study on corium pool heat transfer in the BWR lower head; debris bed formation; steam explosion energetics; thermal hydraulics and coolability in bottom-fed and heterogeneous debris beds. Specifically, for analysis of heat transfer in a BWR lower plenum an advanced threedimensional simulation tool was developed and validated, using a so-called effective convectivity approach and Fluent code platform. An assessment of corium retention and coolability in the reactor pressure vessel (RPV) lower plenum by means of water supplied through the Control Rod Guide Tube (CRGT) cooling system was performed. Simulant material melt experiments were performed in an intermediate temperature range (1300-1600K) on DEFOR test facility to study formation of debris beds in high and low subcooled water pools characteristic of in-vessel and ex-vessel conditions. Results of the DEFOR-E scoping experiments and related analyses strongly suggest that porous beds formed in ex-vessel from a fragmented high-temperature debris is far from homogeneous. Calculation results of bed thermal hydraulics and dryout heat flux with a two-dimensional thermal-hydraulic code give the first basis to evaluate the extent by which macro and micro inhomogeneity can enhance the bed coolability. The development and validation of a model for two-phase natural circulation through a heated porous medium and its application to the coolability analysis of bottom-fed beds enables quantification of the significant effect of dryout heat flux enhancement (by a factor of 80-160%) due to bottom coolant injection. For a qualitative and quantitative understanding of steam explosion, the SHARP system and its image processing methodology were used to characterize the dynamics of a hot liquid (melt) drop fragmentation and the volatile liquid (coolant) vaporization. The experimental results provide a basis to suggest that the melt drop preconditioning is instrumental to

  10. Evaluation of In-Vessel Corium Retention under a Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae-Joon; Kang, Kyung-Ho; Ha, Kwang-Soon; Kim, Jong-Tae; Koo, Kil-Mo; Cho, Young-Ro; Hong, Seong-Wan; Kim, Sang-Baik; Kim, Hee-Dong

    2008-02-15

    The current study on In-Vessel corium Retention and its application activities to the actual nuclear power plant have been reviewed and discussed in this study. Severe accident sequence which determines an initial condition of the IVR has been evaluated and late phase melt progression, heat transfer on the outer reactor vessel, and in-vessel corium cooling mechanism have been estimated in detail. During the high pressure sequence of the reactor coolant system, a natural circulation flow of the hot steam leads to a failure of the pressurizer surge line before the reactor vessel failure, which leads to a rapid decrease of the reactor coolant system pressure. The results of RASPLAV/MASCA study by OECD/NEA have shown that a melt stratification has occurred in the lower plenum of the reactor vessel. In particular, laver inversion has occurred, which is that a high density of the metal melt moves to the lower part of the oxidic melt layer. A method of heat transfer enhancement on the outer reactor vessel is an optimal design of the reactor vessel insulation for an increase of the natural circulation flow between the outer reactor vessel and the its insulation, and an increase of the critical Heat flux on the outer reactor vessel by using various method, such as Nono fluid, coated reactor vessel, and so on. An increase method of the in-vessel melt cooling is a development of the In-vessel core catcher and a decrease of focusing effect in the metal layer.

  11. High temperature reaction between sea salt deposit and (U,Zr)O2 simulated corium debris

    Science.gov (United States)

    Takano, Masahide; Nishi, Tsuyoshi

    2013-11-01

    In order to clarify the possible impacts of seawater injection on the chemical and physical state of the corium debris formed in the severe accident at Fukushima Daiichi Nuclear Power Plants, the high temperature reaction between sea salt deposit and (U,Zr)O2 simulated corium debris (sim-debris) was examined in the temperature range from 1088 to 1668 K. A dense layer of calcium and sodium uranate formed on the surface of a sim-debris pellet at 1275 K under airflow, with the thickness of over 50 μm. When the oxygen partial pressure is low, calcium is likely to dissolve into the cubic sim-debris phase to form solid solution (Ca,U,Zr)O2+x. The diffusion depth was 5-6 μm from the surface, subjected to 1275 K for 12 h. The crystalline MgO remains affixed on the surface as the main residue of salt components. A part of it can also dissolve into the sim-debris.

  12. Low pressure corium dispersion experiments in the DISCO test facility with cold simulant fluids

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, L.; Gargallo, M.; Kirstahler, M.; Schwall, M.; Wachter, E.; Woerner, G.

    2006-08-15

    In a severe accident special pressure relief valves in the primary circuit of German Pressurized Water Reactors (PWR) will transfer a high pressure accident into a low pressure scenario. However, there may be a time window during late in-vessel reflooding scenarios where the pressure is in the order of 1 or 2 MPa at the moment of the reactor vessel rupture. A failure in the bottom head of the reactor pressure vessel, followed by melt expulsion and blowdown of the reactor cooling system, might disperse molten core debris out of the reactor pit, even at such low pressures. The mechanisms of efficient debris-to-gas heat transfer, exothermic metal/oxygen reactions, and hydrogen combustion may cause a rapid increase in pressure and temperature in the reactor containment. Integral experiments are necessary to furnish data for modeling these processes in computer codes, that will be used to apply these result to the reactor case. The acquired knowledge can lead to realize additional safety margins for existing or future plants. The test facility DISCO-C (DIspersion of Simulant COrium - Cold) models the annular reactor cavity and the subcompartments of a large European reactor in a scale 1:18. The fluid dynamics of the dispersion process was studied using model fluids, water or bismuth alloy instead of corium, and nitrogen or helium instead of steam. The effects of different breach sizes and locations, and different failure pressures on the dispersion were studied, specifically by testing central holes, lateral holes, horizontal rips, and complete ripping of the bottom head. 22 experiments were performed in a basic cavity geometry with holes at the bottom of the lower head to study the similarity relations. Variables were the hole diameter, the initial pressure in the RPV and the fluids used. The only flow path out of the reactor pit was the annular gap between the inner wall of the reactor pit and the RPV, and then along the main coolant lines into the subcompartments

  13. Experimental investigation of 150-KG-scale corium melt jet quenching in water

    Energy Technology Data Exchange (ETDEWEB)

    Magallon, D.; Hohmann, H.

    1995-09-01

    This paper compares and discusses the results of two large scale FARO quenching tests known as L-11 and L-14, which involved, respectively, 151 kg of W% 76.7 UO{sub 2} + 19.2 ZrO{sub 2} + 4.1 Zr and 125 kg of W% 80 UO{sub 2} + 20 ZrO{sub 2} melts poured into 600-kg, 2-m-depth water at saturation at 5.0 MPa. The results are further compared with those of two previous tests performed using a pure oxidic melt, respectively 18 and 44 kg of W% 80 UO{sub 2} + 20 ZrO{sub 2} melt quenched in 1-m-depth water at saturation at 5.0 MPa. In all the tests, significant breakup and quenching took place during the melt fall through the water. No steam explosion occurred. In the tests performed with a pure oxide UO{sub 2}-ZrO{sub 2} melt, part of the corium (from 1/6 to 1/3) did not breakup and reached the bottom plate still molten whatever the water depth was. Test L-11 data suggest that full oxidation and complete breakup of the melt occurred during the melt fall through the water. A proportion of 64% of the total energy content of the melt was released to the water during this phase ({approximately}1.5 s), against 44% for L-14. The maximum temperature increase of the bottom plate was 330 K (L-14). The mean particle size of the debris ranged between 2.5 and 4.8mm.

  14. Development of an ex-vessel corium debris bed with two-phase natural convection in a flooded cavity

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eunho; Lee, Mooneon; Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr; Moriyama, Kiyofumi; Park, Jin Ho

    2016-03-15

    Highlights: • For ex-vessel severe accidents in LWRs with wet-cavity strategy, development of debris bed with two-phase natural convection flow due to thermal characteristics of prototypic corium particles was investigated experimentally by using simulant particles and local air bubble control system. • Based on the experimental results of this study, an analytical model was established to describe the spreading of the debris bed in terms of two-phase flow and the debris injection parameters. • This model was then used to analyze the formation of debris beds at the reactor scale, and a sensitivity analysis was carried out based on key accident parameters. - Abstract: During severe accidents of light water reactors (LWRs), the coolability of relocated corium from the reactor vessel is a significant safety issue and a threat to the integrity of containment. With a flooded cavity, a porous debris bed is expected to develop on the bottom of the pool due to breakup and fragmentation of the melt jet. As part of the coolability assessment under accident conditions, the geometrical configuration of the debris bed is important. The Debris Bed Research Apparatus for Validation of the Bubble-Induced Natural Convection Effect Issue (DAVINCI) experimental apparatus facility was constructed to investigate the formation of debris beds under the influence of a two-phase flow induced by steam generation due to the decay heat of the debris bed. Using this system, five kilograms of stainless steel simulant debris were injected from the top of the water level, while air bubbles simulating the vapor flow were injected from the bottom of the particle catcher plate. The airflow rate was determined based on the quantity of settled debris, which will form a heat source due to the decay of corium. The radial distribution of the settled debris was examined using a ‘gap–tooth’ approach. Based on the experimental results of this study, an analytical model was established to

  15. Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hoon; Chang, Soon Heung; Kim, Soo Hyung; Kim, Kee Poong; Lee, Hyoung Wook; Jang, Kwang Keol; Jeong, Yong Hoon; Kim, Sang Jin; Lee, Seong Jin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Park, Jae Hong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    2001-03-15

    In this work, assessment system for methodology for reactor pressure vessel integrity is developed. Assessment system is make up of severe accident assessment code which can calculate the conditions of plant and structural analysis code which can assess the integrity of reactor vessel using given plant conditions. An assessment of cavity flooding using containment spray system has been done. As a result, by the containment spray, cavity can be flooded successfully and CCI can be reduced. The technical backgrounds for external vessel cooling and corium cooling on the cavity are summarized and provided in this report.

  16. Formation of a nuclear reactor's molten core bath in a crucible-type corium catcher for a nuclear power station equipped with VVER reactors

    Science.gov (United States)

    Beshta, S. V.; Vitol', S. A.; Granovskii, V. S.; Kalyago, E. K.; Kovtunova, S. V.; Krushinov, E. V.; Sulatskaya, M. B.; Sulatskii, A. A.; Khabenskii, V. B.; Al'Myashev, V. I.; Gusarov, V. V.

    2011-05-01

    Results from a calculation study on analyzing the formation of a melt bath in a crucible-type catcher for the conditions of a severe accident at a nuclear power station equipped with VVER-1000 reactors are presented. It is shown that the heat loads exerted on the water-cooled walls of the corium catcher shell are limited to a permissible level at which the necessary margins to nucleate boiling crisis and to destruction are ensured under the conditions of thermal and mechanical loading of the shell. An important role of sacrificial material in the efficient operation of the corium catcher is pointed out.

  17. Synthetic analyses of the LAVA experimental results on in-vessel corium retention through gap cooling

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Cho, Young Ro; Koo, Kil Mo; Park, Rae Joon; Kim, Jong Hwan; Kim, Jong Tae; Ha, Kwang Sun; Kim, Sang Baik; Kim, Hee Dong

    2001-03-01

    LAVA(Lower-plenum Arrested Vessel Attack) has been performed to gather proof of gap formation between the debris and lower head vessel and to evaluate the effect of the gap formation on in-vessel cooling. Through the total of 12 tests, the analyses on the melt relocation process, gap formation and the thermal and mechanical behaviors of the vessel were performed. The thermal behaviors of the lower head vessel were affected by the formation of the fragmented particles and melt pool during the melt relocation process depending on mass and composition of melt and subcooling and depth of water. During the melt relocation process 10.0 to 20.0 % of the melt mass was fragmented and also 15.5 to 47.5 % of the thermal energy of the melt was transferred to water. The experimental results address the non-adherence of the debris to the lower head vessel and the consequent gap formation between the debris and the lower head vessel in case there was an internal pressure load across the vessel abreast with the thermal load induced by the thermite melt. The thermal behaviors of the lower head vessel during the cooldown period were mainly affected by the heat removal characteristics through this gap, which were determined by the possibilities of the water ingression into the gap depending on the melt composition of the corium simulant. The enhanced cooling capacity through the gap was distinguished in the Al{sub 2}O{sub 3} melt tests. It could be inferred from the analyses on the heat removal capacity through the gap that the lower head vessel could effectively cooldown via heat removal in the gap governed by counter current flow limits(CCFL) even if 2mm thick gap should form in the 30 kg Al{sub 2}O{sub 3} melt tests, which was also confirmed through the variations of the conduction heat flux in the vessel and rapid cool down of the vessel outer surface in the Al{sub 2}O{sub 3} melt tests. In the case of large melt mass of 70 kg Al{sub 2}O{sub 3} melt, however, the infinite

  18. Interface temperature between solid and liquid corium in severe accident situations: A comprehensive study of characteristic time delay needed for reaching liquidus temperature

    Energy Technology Data Exchange (ETDEWEB)

    Combeau, H.; Appolaire, B. [Institut Jean Lamour, Departement SI2 M, CNRS - Nancy-Universite - UPV-Metz, Ecole des Mines de Nancy, Parc de Saurupt CS 14234, F-54042 Nancy Cedex (France); Seiler, J.M., E-mail: jean-marie.seiler@cea.f [CEA/DEN Grenoble, DTN/SE2T/LPTM, 17 rue des Martyrs, 38054 Grenoble Cedex 9 (France)

    2010-08-15

    T{sub liquidus} was proposed as the interface temperature, for various severe accident situations for thermalhydraulic steady state. This proposal was made on the basis of the analysis of solidification front stability in thermalhydraulic steady state for volumetrically heated corium pools and was extended to reactor transients with slow solidification rates that are controlled by the long-term decrease in residual power. The conclusions were corroborated by prototypic corium and variable solidification rates obtained by experimental approaches for corium containing small amounts of silica or none at all. When the concentration in silica increases (approximately above 10 wt%), it was concluded from the experiments that a plane-front situation could not be obtained. The present work offers a theoretical approach to the maximum time delay that is necessary for mass transfer and full phase-segregation in volumetrically heated liquid pools bounded by a crust. It is concluded that full segregation is obtained for in-vessel situations within time delays that are shorter or of the same order of magnitude as the characteristic time for the corium pool to form and evolve to a quasi-steady-state situation. The characteristic time delay for mass transfer associated with simulant material experiments is also determined. Phase segregation can also be obtained for corium-concrete interaction, provided that the silica content is less than approximately 10 wt%. However in the latter case, more complex phenomena occur at the interface due to the interaction with sparging gas (such as porous medium formation) which requires a different model approach.

  19. Possible in-vessel corium progression way in the Unit 1 of Fukushima Dai-ichi nuclear power plant using a phenomenological analysis

    Directory of Open Access Journals (Sweden)

    Payot Frédéric

    2015-01-01

    Full Text Available In the field of severe accident, the description of corium progression events is mainly carried out by using integral calculation codes. However, these tools are usually based on bounding assumptions because of high complexity of phenomena. The limitations associated with bounding situations ([J.M. Seiler, B. Tourniaire, A phenomenological analysis of melt progression in the lower head of a pressurized water reactor, Nucl. Eng. Des. 268, 87 (2014] e.g. steady state situations and instantaneous whole core relocation in the lower head led CEA to develop an alternative approach in order to improve the phenomenological description of melt progression. The methodology used to describe the corium progression was designed to cover the accidental situations from the core meltdown to the molten core concrete interaction. This phenomenological approach is based on available data (including learnings from TMI2, on physical models and knowledge about the corium behavior. It provides emerging trends and best estimated intermediate situations. As different phenomena are unknown, but strongly coupled, uncertainties at large scale for the reactor application must be taken into account. Furthermore, the analysis is complicated by the fact that these configurations are most probably three dimensional, all the more so because 3D effects are expected to have significant consequences for the corium progression and the resulting vessel failure. Such an analysis of the in-vessel melt progression was carried out for the Unit 1 of the Fukushima Dai-ichi nuclear power plant. The core uncovering kinetics governs the core degradation and impacts the appearance of the first molten corium inside the core. The initial conditions used to carry out this analysis are based on available results derived from codes like MELCOR calculation code [R. Ganntt, D. Kalinich, J. Cardoni, J. Phillips, A. Goldmann, S. Pickering, M. Francis, K. Robb, L. Ott, D. Wang, C. Smith, S. St. Germain

  20. The Plinius/Colima CA-U3 test on fission-product aerosol release over a VVER-type corium pool; L'essai Plinius/Colima CA-U3 sur le relachement des aerosols de produits de fission au-dessus d'un bain de corium de type VVER

    Energy Technology Data Exchange (ETDEWEB)

    Journeau, Ch.; Piluso, P.; Correggio, P.; Godin-Jacqmin, L

    2007-07-01

    In a hypothetical case of severe accident in a PWR type VVER-440, a complex corium pool could be formed and fission products could be released. In order to study aerosols release in terms of mechanisms, kinetics, nature or quantity, and to better precise the source term of VVER-440, a series of experiments have been performed in the Colima facility and the test Colima CA-U3 has been successfully performed thanks to technological modifications to melt a prototypical corium at 2760 C degrees. Specific instrumentation has allowed us to follow the evolution of the corium melt and the release, transport and deposition of the fission products. The main conclusions are: -) there is a large release of Cr, Te, Sr, Pr and Rh (>95%w), -) there is a significant release of Fe (50%w), -) there is a small release of Ba, Ce, La, Nb, Nd and Y (<90%w), -) there is a very small release of U in proportion (<5%w) but it is one of the major released species in mass, and -) there is no release of Zr. The Colima experimental results are consistent with previous experiments on irradiated fuels except for Ba, Fe and U releases. (A.C.)

  1. Premixing of corium into water during a Fuel-Coolant Interaction. The models used in the 3 field version of the MC3D code and two examples of validation on Billeau and FARO experiments

    Energy Technology Data Exchange (ETDEWEB)

    Berthoud, G.; Crecy, F. de; Duplat, F.; Meignen, R.; Valette, M. [CEA/Grenoble, DRN/DTP, 17 Avenue des Martyrs, 38054 Grenoble Cedex 9 (France)

    1998-01-01

    This paper presents the <> application of the multiphasic 3D computer code MC3D. This application is devoted to the premixing phase of a Fuel Coolant Interaction (FCI) when large amounts of molten corium flow into water and interact with it. A description of the new features of the model is given (a more complete description of the full model is given in annex). Calculations of Billeau experiments (cold or hot spheres dropped into water) and of a FARO test (<> corium dropped into 5 MPa saturated water) are presented. (author)

  2. 熔融物与下封头间球形窄缝通道内 CHF理论研究%Theoretical Research of CHF in Hemispherical Narrow Gap Channel Between Molten Corium and Lower Head

    Institute of Scientific and Technical Information of China (English)

    封坤; 田文喜; 巫英伟; 苏光辉; 秋穗正; 余红星

    2013-01-01

    严重事故下熔融物与下封头间球形窄缝通道的存在对于下封头结构的完整性有一定的积极意义。本工作通过理论分析,在汽液两相间逆向对流限制机理的基础上提出了球形窄缝通道内的CHF机理模型和预测关系式,预测结果与实验数据符合较好,验证了所建模型的正确性,并进一步分析了系统压力、熔融物半径、间隙尺寸等关键参数对临界热流密度的影响规律。利用本工作的预测模型对三哩岛(TMI-2)事故后堆芯熔融物特性进行了计算分析,结果表明,熔融物与下封头内壁面间的球形窄缝可有效带走堆芯余热,保证了下封头的完整性。%The existence of the hemispherical narrow gap between the molten corium and the lower head during severe accidents is very significative to the integrity of the lower head .Based on the counter current flow limitation between the vapour phase and liquid phase ,a mechanism model and a predicting formula of CHF in hemispherical narrow gap were proposed .The established model was validated by comparing the predicted results with the test data ,which showed good agreement between them .Furthermore ,the effects of key parameters including the system pressure ,the radius of molten corium and the gap size on the critical heat flux were also analyzed .Besides ,the accident of TMI-2 was also calculated by using the predicting formula .The results indicate that the narrow gap between the molten corium and the lower head can take away the afterheat of the reactor core effectively and ensure the integrity of the lower head .

  3. High temperature chemical reactivity in the system (U, Zr,Fe, O). A contribution to the study of zirconia as a ``core catcher``; Reactivite chimique a haute temperature dans le systeme (U, Zr, Fe, O) contribution a l`etude de la zircone comme recuperateur de ``corium``

    Energy Technology Data Exchange (ETDEWEB)

    Maurizi, A. [CEA Centre d`Etudes Nucleaires de Saclay, 91 -Gif-sur-Yvette (France)]|[CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France). Direction des Technologies Avancees

    1996-12-11

    Within the framework of the improvement of nuclear reactor safety, a device to recover corium is proposed to be installed under the reactor vessel to limit the consequences of a core melting. According to our bibliographic study, stabilised zirconia seems to be the best refractory material to play this role and to support the physicochemical, mechanical and thermal requirements imposed to the corium catcher. The nature of the chemical interactions between zirconia and iron of high temperature were established and experimental data on the (U, Fe, Zr, O) quaternary system which stands for the corium were determined. First of all, the Knudsen effusion mass-spectrometric method was used to establish the liquidus position for a (U, Zr, O) alloy representative of the corium (U/Zr = 1,5) at 2000 deg C. The oxygen solubility limit in a (U, Zr, O) liquid alloy is about 7 atomic %. In oxidising conditions, the reaction between zirconia and iron leads to the formation of a stabilised zirconia-iron oxide solid solution. Up to 10 atomic % of iron can be incorporated in the structure, leading to the stabilisation of cubic zirconia and a modification of lattice constants. The valence and localisation of those iron measured as a function of time and temperature from 1500 to 2400 deg C, after high frequency inductive heating, both on laboratory materials are commercial bricks. The reaction rate is governed by an activation energy of about 80 kJ/mol. Our results demonstrate that stabilised zirconia is able to efficiently absorb oxidised iron. (author). 169 refs.

  4. Cooling and spreading of corium during its fall into water in a pressurised water nuclear plant severe accident: description of mechanical and thermal interactions in a three phase flow during spreading of cold or heated spheres in a liquid pool; Refroidissement et dispersion du corium lors de sa chute dans l'eau pendant un accident severe de reacteur nucleaire a eau pressurisee: description des interactions mecaniques et thermiques en ecoulement triphasique lors de la dispersion de spheres solides froides ou chaudes dans un bain liquide

    Energy Technology Data Exchange (ETDEWEB)

    Duplat, F

    1998-10-26

    In the frame of nuclear safety studies about corium and water interactions, we address spreading and cooling stage of corium fragments in a liquid pool. Considering the complexity of encountered flow regimes and mechanical and thermal interactions coupling, modelling validation is based on a thermal-hydraulic computer code (MC3D). A bibliographical study shows that classical modelling of three phase flow is based on constitutive laws already established in the case of two phase flow. The present study states a complete analysis of BILLEAU experiments and defines a characterisation method for a sphere cloud. Some complementary QUEOS experiments are also described. Mechanical interaction terms such as added mass, lift and turbulent dispersion have been presented in the frame of a three phase flow and their influence has been tested in numerical simulations of BILLEAU tests. The effect of film vapour overheat, as well as particle diameter evolution have been studied. Moreover a radiative heat transfer modelling developed in Karlsruhe research centre (FZK) has been analysed and completed. Numerical simulations achieved for this study show that mechanical and thermal behaviour of the system are actually coupled. Taking into account lift and turbulent dispersion terms as well as heat transfer modifications all wed better results. This study also presents some considerations about flow regimes identification as a preliminary for studies about numerical diffusion that was already estimated in the present state of the computer code MC3D. (author)

  5. Analyses of corium spreading in Mark I containment geometry

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, J.J.; Chu, C.C.; Farmer, M.T.

    1991-12-31

    An assessment of melt spreading in the Mark I system has been carried out using the MELTSPREAD-1 computer code together with supporting analyses. Application of MELTSPREAD-1 confirms the calculation of shell survival in a wet containment for the most probable melt release conditions from NUREG/CR-5423. According to MELTSPREAD-1, a dry containment also may not be threatened by melt spreading. This reflects the heat losses undergone by the melt in the process of spreading to the shell conservatively neglected in NUREG/CR-5423. However, there exist parameter ranges outside the most probable set where shell failure may be calculated. Accounting for the breakup and quenching of melt relocating through a deep layer of subcooled water also conservatively neglected in NUREG/CR-5423 can reduce the set of parameter variations for which containment failure is calculated in the wet case.

  6. Analyses of corium spreading in Mark I containment geometry

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, J.J.; Chu, C.C.; Farmer, M.T.

    1991-01-01

    An assessment of melt spreading in the Mark I system has been carried out using the MELTSPREAD-1 computer code together with supporting analyses. Application of MELTSPREAD-1 confirms the calculation of shell survival in a wet containment for the most probable melt release conditions from NUREG/CR-5423. According to MELTSPREAD-1, a dry containment also may not be threatened by melt spreading. This reflects the heat losses undergone by the melt in the process of spreading to the shell conservatively neglected in NUREG/CR-5423. However, there exist parameter ranges outside the most probable set where shell failure may be calculated. Accounting for the breakup and quenching of melt relocating through a deep layer of subcooled water also conservatively neglected in NUREG/CR-5423 can reduce the set of parameter variations for which containment failure is calculated in the wet case.

  7. Coolability of corium debris under severe accident conditions in light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rahman, Saidur

    2013-11-15

    The debris bed which may be formed in different stages of a severe accident will be hot and heated by decay heat from the radioactive fission products. In order to establish a steady state of long-term cooling, this hot debris needs to be quenched at first. If quenching by water ingression into the dry bed is not rapid enough then heat-up by decay heat in still dry regions may again yield melting. Thus, chances of coolability must be investigated considering quenching against heat-up due to decay heat, in the context of reactor safety research. As a basis of the present investigations, models for simulation of two phase flow through porous medium were already available in the MEWA code, being under development at IKE. The objective of this thesis is to apply the code in essential phases of severe accidents and to investigate the chances, options and measures for coolability. Further, within the tasks, improvements to remove weaknesses in modeling and implementation of extensions concerning missing parts are included. It was identified previously that classical models without explicit considering the interfacial friction, can predict dryout heat flux (DHF) well under top fed condition but under-predict DHF values under bottom flooding conditions. Tung and Dhir introduced an interfacial friction term in their model, but this model has deficits for smaller particles considered as relevant for reactor conditions. Therefore, some modification of Tung and Dhir model is proposed in the present work to extent it for smaller particles. A significant improvement with the new friction description (Modified Tung and Dhir, MTD) is obtained considering the aim of a unified description for both top and bottom flooding conditions and for broad bandwidth of bed conditions. Calculations for reactor conditions are carried out in order to explore whether or to which degree coolability can be concluded, how strong the trend to coolability is and where major limits occur. The general result from the various calculations in this work is that there exist significant cooling margins and strong trends to coolability which is achieved due to multidimensional cooling options, especially lateral and bottom ingression of water, established in the core region through an intact rod or bypass region, in the lower head through the wall and in the cavity due to the shape (heap) of the bed. These cooling options together with cooling effects of steam flow through a hot dry zone provide mechanisms to facilitate and support quenching processes. Limits also have been obtained, mainly with significant piling up of particles, cake parts with very low porosities and bed with very small particles. The initial temperature distribution inside the bed has a major influence on the coolability behavior of the bed, no matter if the bed is located in the lower head or in the flooded cavity. Previously, quenching calculations were only possible for given debris configurations starting from assumed initial temperatures. However, assuming the whole bed at a uniform initial temperature strongly misses the real process in which settling of partly solidified melt drops occurs simultaneously with water inflow and quenching. Therefore, in the frame of this work, the MEWA models have been extended i.e. coupled to jet breakup and mixing model (JEMI) to treat the combined process. This improved the capabilities of realistic analysis significantly and showed significant effects on cooling in the calculations. Another important step for the improvement of overall modeling of coolability is undertaken by introducing the porosity formation in liquid melt layers through the supply of water from the bottom (COMET concept) in the MEWA model. The related modeling is implemented for situations where liquid melt arrives un-fragmented at the cavity floor due to incomplete breakup of melt.

  8. Experimental investigation on molten pool representing corium composition at Fukushima Daiichi nuclear power plant

    Science.gov (United States)

    An, Sang Mo; Song, Jin Ho; Kim, Jong-Yun; Kim, HwanYeol; Naitoh, Masanori

    2016-09-01

    A configuration of molten core in the Fukushima Daiichi NPP (nuclear power plant) was investigated by a melting and solidification experiment. About 5 kg of a mixture, whose composition in terms of weight is UO2 (60%), Zr + ZrO2 (25%), stainless steel (14%), B4C (1%), was melted in a cold crucible using an induction heating technique. It was shown that the solidified melt consists of upper crust and lower solidified ingot. The solidified ingot was separated into two layers. A physical and chemical analysis was performed for the samples taken from the solidified melt to investigate the morphology and chemical characteristics. It was found that the solidified ingot consists of a metal-rich layer on the top and an oxide-rich layer at the bottom. In addition, the oxide layer at the bottom has composition close to the initial charge composition and surrounded by a thin crust layer. It turned out that B4C was more concentrated in the upper metal-rich layer. These findings provide important insights for understanding the core melt progression and taking proper post-accident recovery actions for the Fukushima Daiichi NPP.

  9. Agglomeration and size distribution of debris in DEFOR-A experiments with Bi{sub 2}O{sub 3}–WO{sub 3} corium simulant melt

    Energy Technology Data Exchange (ETDEWEB)

    Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se; Karbojian, Aram, E-mail: aram@safety.sci.kth.se; Tran, Chi-Thanh, E-mail: thanh@safety.sci.kth.se; Villanueva, Walter, E-mail: walter@safety.sci.kth.se

    2013-10-15

    Highlights: • Debris agglomeration in case of melt pouring into a coolant is experimentally investigated. • The effects of jet diameter, melt superheat and water subcooling are addressed. • Most influential factor which can significantly increase fraction of agglomerates is melt superheat. • Rapid decrease of the fraction of agglomerates as a function of water depth is obtained in all cases. • Provided data is valuable for model development and code validation. -- Abstract: Flooding of lower drywell has been adopted as a cornerstone of severe accident management strategy in Nordic type Boiling Water Reactors (BWR). It is assumed that the melt ejected into a deep pool of water will fragment, quench and form a porous debris bed coolable by natural circulation. If debris bed is not coolable, then dryout and possibly re-melting of the debris can occur. Melt attack on the containment basemat can threaten containment integrity. Agglomeration of melt debris and formation of solid “cake” regions provide a negative impact on coolability of the porous debris bed. In this work we present results of experimental investigation on the fraction of agglomerated debris obtained in the process of hot binary oxidic melt pouring into a pool of water. The Debris Bed Formation and Agglomeration (DEFOR-A) experiments provide data about the effects of the pool depth and water subcooling, melt jet diameter, and initial melt superheat on the fraction of agglomerated debris. The data presents first systematic study of the debris agglomeration phenomena and facilitates understanding of underlying physics which is necessary for development and validation of computational codes to enable prediction of the debris bed coolability in different scenarios of melt release.

  10. Contribution of prototypic material tests on the Plinius platform to the study of nuclear reactor severe accident; Contribution des essais en materiaux prototypiques sur la plate-forme Plinius a l'etude des accidents graves de reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Journeau, Ch

    2008-01-15

    The PLINIUS experimental platform at CEA Cadarache is dedicated to the experimental study of nuclear reactor severe accidents thanks to experiments between 2000 and 3500 K with prototypic corium. Corium is the mixture that would be formed by an hypothetical core melting and its mixing with structural materials. Prototypical corium has the same chemical composition as the corium corresponding to a given accident scenario but has a different isotopic composition (use of depleted uranium,...). Research programs and test series have been performed to study corium thermophysical properties, fission product behaviour, corium spreading, solidification and interaction with concrete as well as its coolability. It was the frame of research training of many students and was realized within national, European and international collaborations. (author)

  11. A status of the art report for OECD RASPLAV program

    Energy Technology Data Exchange (ETDEWEB)

    Nho, Ki Man; Kim, Sang Baik; Bang, Kwang Hyun; Park, Jong Hwa; Kim, Hee Dong; Suh, Kun Yeol [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-06-01

    The objective of current study is to summarize the work of OECD RASPLAV technical reports, which include investigation of natural convection in the corium, chemical interaction between corium and reactor vessel, solidification of corium crust during severe accident such as TMI-2 accident in the United States and Chernobyl accident in the USSR. The experimental data and technique will be used when designing a large scale experimental facility for the second phase of the project. 7 tabs., 11 figs., 14 refs. (Author).

  12. Analytical Chemistry Laboratory progress report for FY 1992

    Energy Technology Data Exchange (ETDEWEB)

    Green, D.W.; Heinrich, R.R.; Graczyk, D.G.; Lindahl, P.C.; Boparai, A.S.; Bass, D.A.

    1992-12-01

    The ACL activities covered IFR fuel reprocessing, corium-concrete interactions, environmental samples, wastes, WIPP support, Advanced Photon Source, H-Tc superconductors, EBWR vessel, soils, illegal drug detection, quality control, etc.

  13. Environ: E00525 [KEGG MEDICUS

    Lifescience Database Archive (English)

    Full Text Available E00525 Chicken stomach lining Galli stomachichum corium Crude drug Gallus gallus [T...AX:9031] Phasianidae Gallus gallus stomach lining (dried) Crude drugs [BR:br08305] Animals Birds E00525 Chicken stomach lining ...

  14. Molten pool-lower head integrity. Heat transfer models including advanced numerical simulations (DNS)

    Energy Technology Data Exchange (ETDEWEB)

    Seiler, J.M.; Bonnet, J.M.; Bernaz, L. [CEA Grenoble (France)

    2001-07-01

    Extensive studies have been performed to investigate the heat transfer within a molten corium pool (homogeneous, stratified and with miscibility gap): Synthesis of heat transfer correlations in molten pool (homogeneous and stratified), Focusing effect in stratified metal layer, DNS analysis of Rayleigh Benard instabilities at the top boundary; interpretation of the different convection regimes and exponents affecting the Rayleigh number in the heat transfer correlations, Molten pool model for corium presenting a miscibility gap. Condition for de-stratification. (authors)

  15. COTELS project (1): overview of project to study FCI and MCCI during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Nagasaka, Hideo; Kato, Masami; Sakaki, Isao [Nuclear Power Engineering Corp., Tokyo (Japan). System Safety Dept.; Cherepnin, Y.; Vasilyev, Y.; Kolodeshnikov, A.; Zhdanov, V.; Zuev, V. [National Nuclear Center, Kurchatov (Kazakhstan)

    2000-05-01

    Fuel coolant interaction (FCI) and molten core concrete interaction (MCCI) have been studied experimentally within the framework of COTELS project from 1995 as a joint study between NUPEC (Japan) and NNC (Republic of Kazakhstan) using one of the testing complex at NNC. The testing complex includes three experimental facilities ''SLAVA'', ''LAVA'' and ''LAVA-M'' for debris coolability tests. Three types of experiments were carried out. To get the molten corium, the electric induction melting furnace (EMF) was used. The EMF produced {proportional_to}60 kg of corium containing UO{sub 2}, stainless steel, Zr and ZrO{sub 2}. The temperature of the produced melt was about 3200 K. The melt was discharged into the water pool in test A or onto the concrete trap in test B/C. The corium in the concrete trap was heated in test B/C by another induction melt heater. Prior to main test A and test B/C, several supporting experiments were conducted. Integrity of graphite crucible with TaC sheet during producing UO{sub 2} corium was confirmed experimentally. The induction melt heater was calibrated and the efficiency for the induction heater of ''LAVA-M'' facility was determined as 47%. The thermal conductivity and thermal diffusivity of concrete up to about 1073 K, and melting-solidification points of eutectics generated from corium components were determined experimentally. Discharge corium behavior, using UO{sub 2} corium, was also observed by speed cameras in test 01. (orig.)

  16. 20-MHz B-mode ultrasound in monitoring the course of localized scleroderma (morphea).

    Science.gov (United States)

    Hoffmann, K; Gerbaulet, U; el-Gammal, S; Altmeyer, P

    1991-01-01

    Ultrasonographic methods have recently provided us with the means for objective and non-invasive monitoring of the dynamics of chronic skin diseases. We examined 34 patients with localized scleroderma (morphea) using a 20-MHz B-mode ultrasound scanner (DUB 20, Taberna pro Medicum, Lüneburg). In patients with plaque-type and linear band-type localized scleroderma intraindividual comparison of sclerotic skin with corresponding areas of healthy skin showed thickening of the corium. The increase in corium thickness was between 2% and 251%. The extent of the difference in corium thickness between sclerotic and healthy skin depended on the location-originally thin skin showed a greater degree of sclerosis. We also frequently found enhanced reflexes in the lower corium and hyperechoic, widened bands of connective tissue traversing the subcutaneous fatty tissue from the corium-subcutis border in the direction of the muscle fascia. 20 patients were examined several times in the course of one year. In nine patients we found ultrasonographic evidence of regression (decrease in thickness 26%) and in nine the ultrasound examination showed progression (increase in thickness 28%). 20-MHz B-mode ultrasound imaging is a suitable non-invasive method for monitoring the course and treatment of localized scleroderma. Its routine use is strongly recommended.

  17. In-vessel melt retention as a severe accident management strategy for the Loviisa Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kymaelaeinen, O.; Tuomisto, H. [IVO International Ltd., Vantaa (Finland); Theofanous, T.G. [Univ. of California, Santa Barbara, CA (United States)

    1997-02-01

    The concept of lower head coolability and in-vessel retention of corium has been approved as a basic element of the severe accident management strategy for IVO`s Loviisa Plant (VVER-440) in Finland. The selected approach takes advantage of the unique features of the plant such as low power density, reactor pressure vessel without penetrations at the bottom and ice-condenser containment which ensures flooded cavity in all risk significant sequences. The thermal analyses, which are supported by experimental program, demonstrate that in Loviisa the molten corium on the lower head of the reactor vessel is coolable externally with wide margins. This paper summarizes the approach and the plant modifications being implemented. During the approval process some technical concerns were raised, particularly with regard to thermal loadings caused by contact of cool cavity water and hot corium with the reactor vessel. Resolution of these concerns is also discussed.

  18. Development of computer-controlled ultrasonic image processing system for severe accidents research

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Kil Mo; Kang, Kyung Ho; Kim, Jong Tai; Kim, Jong Whan; Cho, Young Ro; Ha, Kwang Soon; Park, Rae Jun; Kim, Sang Baik; Kim, Hee Dong; Sim, Chul Moo

    2000-07-01

    In order to verify in-vessel corium cooling mechanism, LAVA(Lower-plenum Arrested Vessel Attack) experiment is being performed as a first stage proof of principle test. The aims of this study are to find a gap formation between corium melt and reactor lower head vessel, to verify the principle of the gap formation and to analyze the effect of the gap formation on the thermal behavior of corium melt and the lower plenum. This report aims at developing a computer controlled image signal processing system which is able to improve visualization and to measure the gap distribution with 3-dimensional planar image using a time domain signal analysis method as a part of the ultrasonic pulse echo methods and a computerized position control system. An image signal processing system is developed by independently developing an ultrasonic image signal processing technique and a PC controlled position control system and then combining both systems.

  19. The SAM software system for modeling severe accidents at nuclear power plants equipped with VVER reactors on full-scale and analytic training simulators

    Science.gov (United States)

    Osadchaya, D. Yu.; Fuks, R. L.

    2014-04-01

    The architecture of the SAM software package intended for modeling beyond-design-basis accidents at nuclear power plants equipped with VVER reactors evolving into a severe stage with core melting and failure of the reactor pressure vessel is presented. By using the SAM software package it is possible to perform comprehensive modeling of the entire emergency process from the failure initiating event to the stage of severe accident involving meltdown of nuclear fuel, failure of the reactor pressure vessel, and escape of corium onto the concrete basement or into the corium catcher with retention of molten products in it.

  20. Neutron transport in random media

    Energy Technology Data Exchange (ETDEWEB)

    Makai, M. [KFKI Atomic Energy Research Institute, Budapest (Hungary)

    1996-08-01

    The survey reviews the methods available in the literature which allow a discussion of corium recriticality after a severe accident and a characterization of the corium. It appears that to date no one has considered the eigenvalue problem, though for the source problem several approaches have been proposed. The mathematical formulation of a random medium may be approached in different ways. Based on the review of the literature, we can draw three basic conclusions. The problem of static, random perturbations has been solved. The static case is tractable by the Monte Carlo method. There is a specific time dependent case for which the average flux is given as a series expansion.

  1. Description of premixing with the MC3D code including molten jet behavior modeling. Comparison with FARO experimental results

    Energy Technology Data Exchange (ETDEWEB)

    Berthoud, G.; Crecy, F. de; Meignen, R.; Valette, M. [CEA-G, DRN/DTP/SMTH, 17 rue des Martyrs, 38054 Grenoble Cedex 9 (France)

    1998-01-01

    The premixing phase of a molten fuel-coolant interaction is studied by the way of mechanistic multidimensional calculation. Beside water and steam, corium droplet flow and continuous corium jet flow are calculated independent. The 4-field MC3D code and a detailed hot jet fragmentation model are presented. MC3D calculations are compared to the FARO L14 experiment results and are found to give satisfactory results; heat transfer and jet fragmentation models are still to be improved to predict better final debris size values. (author)

  2. Three-dimensional numerical study on the mechanism of anisotropic MCCI by improved MPS method

    Energy Technology Data Exchange (ETDEWEB)

    Li, Xin, E-mail: lixin@fuji.waseda.jp; Yamaji, Akifumi

    2017-04-01

    Highlights: • 3-D simulation of a MCCI test was presented with improved moving particle method. • The influence of thermally stable silica aggregates on MCCI has been investigated. • The mechanisms for isotropic/anisotropic ablation have been clarified mechanistically. - Abstract: In two-dimensional (2-D) molten corium-concrete interaction (MCCI) experiments with prototypic corium and siliceous concrete, the more pronounced lateral concrete erosion behavior than that in the axial direction, namely anisotropic ablation, has been a research interest. However, the knowledge of the mechanism on this anisotropic ablation behavior, which is important for severe accident analysis and management, is still limited. In this paper, 3-D simulation of 2-D MCCI experiment VULCANO VB-U7 has been carried out with improved Moving Particle Semi-implicit (MPS) method. Heat conduction, phase change, and corium viscosity models have been developed and incorporated into MPS code MPS-SW-MAIN-Ver.2.0 for current study. The influence of thermally stable silica aggregates has been investigated by setting up different simulation cases for analysis. The simulation results suggested reasonable models and assumptions to be considered in order to achieve best estimation of MCCI with prototypic oxidic corium and siliceous concrete. The simulation results also indicated that silica aggregates can contribute to anisotropic ablation. The mechanisms for anisotropic ablation pattern in siliceous concrete as well as isotropic ablation pattern in limestone-rich concrete have been clarified from a mechanistic perspective.

  3. Four Helvella (Ascomycota: Pezizales: Helvellaceae species from the Cold Desert of Leh, Ladakh, Jammu and Kashmir, India

    Directory of Open Access Journals (Sweden)

    K. Dorjey

    2013-03-01

    Full Text Available The present paper deals with four Helvella species namely Helvella acetabulum, H.corium, H.queletii and H.macropus from Leh district in Ladakh region of the state of Jammu and Kashmir. Of these, the first three species are new fungus records for India while H.macropus constitutes first authentic record from Ladakh.

  4. Transplantation of artificial gelatin-co-bletillastriata gelatin/Salvia ...

    African Journals Online (AJOL)

    structure of the repaired skin was similar to that of natural corium. ... Skin, the largest organ of human being, plays a vital role as a ... the vascularized effects of dermal grafts were ... chemicals used in our study were analytical ... wound became blurred and hair growth was ... the control group, the color of the graft in the.

  5. Fast increase of proteinase inhibitors in necrotic collagenous tissue.

    Science.gov (United States)

    Oehmichen, M

    1989-01-01

    Using the PAP immunohistochemical technique, accumulation of two proteinase inhibitors, alpha-1-antichymotrypsin and alpha-2-macroglobulin, can be detected at the edges of collagenous fibers in the corium after slash wounds of the skin. This accumulation was observed within a survival time of 10-30 min. It, however, is not detectable in postmortally inflicted trauma.

  6. Consequences of material effects on in-vessel retention

    Energy Technology Data Exchange (ETDEWEB)

    Seiler, Jean Marie [CEA Grenoble, DTN/SE2T/LPTM, 17 rue des Martyrs, 38054 Grenoble Cedex 9 (France); Tourniaire, Bruno [CEA Grenoble, DTN/SE2T/LPTM, 17 rue des Martyrs, 38054 Grenoble Cedex 9 (France)]. E-mail: bruno.tourniaire@cea.fr; Defoort, Francoise [CEA Grenoble, DTN/SE2T/LPTM, 17 rue des Martyrs, 38054 Grenoble Cedex 9 (France); Froment, Karine [CEA Grenoble, DTN/SE2T/LPTM, 17 rue des Martyrs, 38054 Grenoble Cedex 9 (France)

    2007-09-15

    In-vessel retention (IVR) consists in cooling the corium contained in the reactor vessel by natural convection and reactor cavity flooding. This strategy of severe accident management enables the corium to be kept inside the second confinement barrier: the reactor vessel. The general approach which is used to study IVR problems is a 'bounding' approach which consists in assuming a specified corium stratification in the vessel and then demonstrating that the vessel can cope with the resulting thermal and mechanical loads. Thermal loading on the vessel is controlled by the convective heat transfer inside the molten corium in the lower head. If there is no water in the vessel and if the corium pool is overlaid by a liquid steel layer, then the heat flux might focus on the vessel in front of the steel layer ('focusing effect') and exceed the dry-out heat flux (CHF or DHF). One of the critical points of these studies is linked to the determination of the height of the molten steel layer that can stratify above the oxidic pool. The MASCA experiments have highlighted that part of molten steel may stratify under the oxidic corium which reduces the thickness of the steel layer on top of the pool. This behavior can be explained by chemical interaction between the oxide and metallic phases of the pool which confirms that these materials cannot be treated as inert species. Following these conclusions, a methodology which couples physicochemical effects and thermalhydraulics has been developed to address the IVR issue. The main purpose of this paper is to present this methodology and its application for given corium mass inventories. Attention focuses on the influence of parameters such as the ratio U/Zr and oxidation ratio of zirconia. For a 1000 MW PWR, approximately 10 t of steel stratify at the bottom of the vessel for 40% Zr oxidation, and 25 t for 30% Zr oxidation. This leads to a 25-50% increase of the mass of molten steel that is required for

  7. Fuel-Coolant Interaction visualization in TROI test facility

    Energy Technology Data Exchange (ETDEWEB)

    Na, Young Su; Hong, Seong-Ho; Song, Jin Ho; Hong, Seong-Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    It is necessary to observe the FCI (Fuel-Coolant Interaction) phenomena at the condition of vessel failure to IVR. We carried out a visualization test on the interaction of a corium melt and water to observe the premixing phase without a free fall of a melt jet in a gas phase before contacting the cooling water. This paper is based on the previous study presented at Ninth Korea-Japan Symposium on Nuclear Hydraulics and Safety, we added the results on sieved debris distribution. The visualization test on the FCI without a free fall of a corium melt jet in a gas phase was conducted carefully in the TROI test facility. A prototypic corium consisting of uranium oxide and zirconium oxide with a weight ratio of UO{sub 2} to ZrO{sub 2} of 80 to 20, respectively, was heated up using the induction heating method. It was observed that a corium melt jet penetrated into water with 1000 mm in depth, and it took about 0.6 seconds from opening the releasing valve, which was confirmed by the sequential variation of the temperature measured by the sacrificial thermocouples installed in the direction of a falling melt jet. The cumulative mass fraction of the debris smaller than 1.0 mm was 15%, and the mass mean diameter of the debris was 2.9 mm. This visualization test can generate the valuable information such as the behavior of the corium melt jet and the size of mixing zone for validating the computer code.

  8. Phytochemical screening of twelve species of phytoplankton isolated from Arabian Sea coast

    Directory of Open Access Journals (Sweden)

    Sushanth Vishwanath Rai

    2015-11-01

    Full Text Available Objective: To analyze the phytochemicals in twelve species of marine phytoplankton. Methods: Total phenolic content of methanol extract was estimated by the Folin-Ciocalteu method. Total flavonoid content of the methanol extarct was determined by aluminium chloride method. Chlorophylls, β-carotene and astaxanthin were estimated by acetone extraction method. Vitamin C was determined by dinitrophenyl-hydrazine method. Phycobiliproteins such as allophycocyanin, phycocyanin and phycoerythrin in the aqueous extracts were determined. Results: Total phenolics varied from 5.41 mg gallic acid equivalents/g dry weight (DW in Phormidium corium (P. corium to 17.37 mg gallic acid equivalents/g DW in Oscillatoria fremyii (O. fremyii. Total flavonoids ranged between 0.74 mg quercetin equivalent/g DW in P. corium and 9.87 mg quercetin equivalent/g DW in Nannochloropsis oceanica. Chlorophyll-a pigment was high in Chaetoceros calcitrans (C. calcitrans (15.51 mg/g DW and low in P. corium (1.08 mg/g DW. Chlorophyll-c ranged between 0.07 mg/g DW in Nannochloropsis oceanica and 4.62 mg/g DW in C. calcitrans. High contents of β-carotene and astaxanthin were found in C. calcitrans and low in P. corium which ranged from 0.33 to 10.03 mg/g DW and 0.18 to 3.85 mg/g DW, respectively. Vitamin C content varied from 0.50 mg/g DW in C. calcitrans to 1.51 mg/g DW in Phormidium tenue. O. fremyii showed highest total phycobiliproteins of 317.05 mg/g DW. High contents of allophycocyanin and phycocyanin were found in O. fremyii, whereas high contents of phycoerythrin were found in Oscillatoria sancta. All the three phycobiliproteins were low in Chroococcus turgidus. Conclusions: Marine phytoplankton are one of the natural sources providing novel biologically active compounds with potential for pharmaceutical applications.

  9. Development of MPS Method for Analyzing Melt Spreading Behavior and MCCI in Severe Accidents

    Science.gov (United States)

    Yamaji, Akifumi; Li, Xin

    2016-08-01

    Spreading of molten core (corium) on reactor containment vessel floor and molten corium-concrete interaction (MCCI) are important phenomena in the late phase of a severe accident for assessment of the containment integrity and managing the severe accident. The severe accident research at Waseda University has been advancing to show that simulations with moving particle semi-implicit (MPS) method (one of the particle methods) can greatly improve the analytical capability and mechanical understanding of the melt behavior in severe accidents. MPS models have been developed and verified regarding calculations of radiation and thermal field, solid-liquid phase transition, buoyancy, and temperature dependency of viscosity to simulate phenomena, such as spreading of corium, ablation of concrete by the corium, crust formation and cooling of the corium by top flooding. Validations have been conducted against experiments such as FARO L26S, ECOKATS-V1, Theofanous, and SPREAD for spreading, SURC-2, SURC-4, SWISS-1, and SWISS-2 for MCCI. These validations cover melt spreading behaviors and MCCI by mixture of molten oxides (including prototypic UO2-ZrO2), metals, and water. Generally, the analytical results show good agreement with the experiment with respect to the leading edge of spreading melt and ablation front history of concrete. The MPS results indicate that crust formation may play important roles in melt spreading and MCCI. There is a need to develop a code for two dimensional MCCI experiment simulation with MPS method as future study, which will be able to simulate anisotropic ablation of concrete.

  10. A phenomenological analysis of melt progression in the lower head of a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Seiler, J.M., E-mail: jean-marie.seiler@cea.fr [CEA, DEN, DTN, F-38054 Grenoble (France); Tourniaire, B. [EDF/Septen, Lyon (France)

    2014-03-15

    Highlights: • We propose a phenomenological description of melt progression into the lower head. • We examine changes in heat loads on the vessel. • Heat loads are more severe than emphasized by the bounding situation assumption. • Both primary circuit and ex-vessel reflooding are necessary for in-vessel retention. • Vessel failure conditions are examined. - Abstract: The analysis of in-vessel corium cooling (IVC) and retention (IVR) involves the description of very complex and transient physical phenomena. To get round this difficulty, “bounding” situations are often emphasized for the demonstration of corium coolability, by vessel flooding and/or by reactor pit flooding. This approach however comes up against its own limitations. More realistic melt progression scenarios are required to provide plausible corium configurations and vessel failure conditions. Work to develop more realistic melt progression scenarios has been done at CEA, in collaboration with EDF. Development has concentrated on the French 1300 MWe PWR, considering both dry scenarios and the possibility of flooding of the RPC (reactor primary circuit) and/or the reactor pit. The models used for this approach have been derived from the analysis of the TMI2 accident and take benefit from the lessons derived from several programs related to pool thermal hydraulics (BALI, COPO, ACOPO, etc.), material interactions (RASPLAV, MASCA), critical heat flux (CHF) on the external surface of the vessel (KAIST, SULTAN, ULPU), etc. Important conclusions of this work are as follows: (a)After the start of corium melting and onset of melt formation in the core at low pressure (∼1 to 5 bars), it seems questionable that RPV (reactor pressure vessel) reflooding alone would be sufficient to achieve corium retention in the vessel; (b)If the vessel is not cooled externally, it may fail due to local heat-up before the whole core fuel inventory is relocated in the lower head; (c)Even if the vessel is

  11. OECD MMCI Small-Scale Water Ingression and Crust Strength tests (SSWICS) SSWICS-1 final data report, Rev. 1 February 10, 2003.; Report, Rev. 1

    Energy Technology Data Exchange (ETDEWEB)

    Lomperski, S.; Farmer, M. T.; Kilsdonk, D.; Aeschlimann, B. (Nuclear Engineering Division)

    2011-05-23

    The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core/concrete interaction (MCCI) when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. However, due to the integral nature of these tests, several questions regarding the crust freezing behavior could not be adequately resolved. These questions include: (1) To what extent does water ingression into the crust increase the melt quench rate above the conduction-limited rate and how is this affected by melt composition and system pressure; and (2) What is the fracture strength of the corium crust when subjected to a thermal-mechanical load and how does it depend upon the melt composition? A series of separate-effects experiments are being conducted to address these issues. The first employs an apparatus designed to measure the quench rate of a pool of corium ({approx}{phi}30 cm; up to 20 cm deep). The main parameter to be varied in these quench tests is the melt composition since it is thought to have a critical influence on the crust cracking behavior which, in turn, alters quench rate. The issue of crust strength will be addressed with a second apparatus designed to mechanically load the crust produced by the quench tests. This apparatus will measure the fracture strength of the crust while under a thermal load created by a heating element beneath the crust. The two apparatuses used to measure the melt quench rate and crust strength are jointly referred to as SSWICS (Small-Scale Water Ingression and Crust Strength). This report describes results of the first water ingression test, designated SSWICS-1. The test investigated the quench behavior of a 15 cm deep

  12. RARE CASE OF SYSTEMIC SCLEROSIS IN A CHILD AGED 4 MONTHS

    Directory of Open Access Journals (Sweden)

    S.S. Postnikov

    2007-01-01

    Full Text Available The article provides a clinical and morphologic description of a rare case of systemic sclerosis along with the beginning of the diseases during the infancy. In the clinical picture, the researchers identified occurrences of the systemic vasculitis: abundant cyanotic and red spotty rash with atrophy in the middle, thick edemas of legs and ankles, necrosis of the nail bone of the left little finger, banti's syndrome. In the histological picture, most characteristic peculiarities were: 3 stages of systemic sclerosis process development — inflammation, hardening and atrophy; disorganization of collagenous corium fibers; nidi of calcification along the borderline of corium and hypoderm; multiple ulcers of small and large intestines, perforation of one of which caused peritonitis and fatal outcome of the patient.Key words: infants, vasculitis, systemic sclerosis.

  13. Development of ultrasonic high temperature system for severe accidents research

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Kil Mo; Kang, Kyung Ho; Kim, Young Ro and others

    2000-07-01

    The aims of this study are to find a gap formation between corium melt and the reactor lower head vessel, to verify the principle of the gap formation and to analyze the effect of the gap formation on the thermal behavior of corium melt and the lower plenum. This report aims at suggesting development of a new high temperature measuring system using an ultrasonic method which overcomes the limitations of the present thermocouple method used for severe accident experiments. Also, this report describes the design and manufacturing method of the ultrasonic system. At that time, the sensor element is fabricated to a reflective element using 1mm diameter and 50 mm and 80 mm long tungsten alloy wires. This temperature measuring system is intended to measure up to 2800 deg C.

  14. Leather structure determination by small-angle X-ray scattering (SAXS): cross sections of ovine and bovine leather.

    Science.gov (United States)

    Basil-Jones, Melissa M; Edmonds, Richard L; Allsop, Timothy F; Cooper, Sue M; Holmes, Geoff; Norris, Gillian E; Cookson, David J; Kirby, Nigel; Haverkamp, Richard G

    2010-05-12

    SAXS has been applied to structural determination in leather. The SAXS beamline at the Australian Synchrotron provides 6 orders of magnitude dynamic range, enabling a rich source of structural information from scattering patterns of leather sections. SAXS patterns were recorded for q from 0.004 to 0.223 A(-1). Collagen d spacing varied across ovine leather sections from 63.8 nm in parts of the corium up to 64.6 nm in parts of the grain. The intensity of the collagen peak at q = 0.06 A(-1) varied by 1 order of magnitude across ovine leather sections with the high-intensity region in the corium and the low intensity in the grain. The degree of fiber orientation and the dispersion of the orientation has been quantified in leather. It is shown how the technique provides a wealth of useful information that may be used to characterize and compare leathers, skin, and connective tissue.

  15. Estimates of early containment loads from core melt accidents. Draft report for comment

    Energy Technology Data Exchange (ETDEWEB)

    None

    1985-12-01

    The thermal-hydraulic processes and corium debris-material interactions that can result from core melting in a severe accident have been studied to evaluate the potential effect of such phenomena on containment integrity. Pressure and temperature loads associated with representative accident sequences have been estimated for the six various LWR containment types used within the United States. Summaries distilling the analyses are presented and an interpretation of the results provided. 13 refs., 68 figs., 39 tabs.

  16. [Chronic cough and worsening dyspnea: a case of idiopathic tracheal stenosis].

    Science.gov (United States)

    Conti, Valentina; Calia, Nunzio; Pasquini, Claudio; Zardi, Silvia; Finetti, Cinzia; Stomeo, Francesco; Ravenna, Franco

    2013-04-01

    We report a case of idiopathic tracheal stenosis in a 75-year-old woman, who presented to our observation with a diagnosis of asthmatic bronchitis characterized by cough and exertional dyspnea, later complicated by the appearance of tirage. Biopsy of the lesion showed focal squamous metaplasia of the epithelium lining, multiple sclerosis and chronic inflammatory infiltration of the corium. The patient was treated with endoscopic destruction via rigid bronchoscopy, through the combined action of YAG laser and mechanical debulking.

  17. Collagenous skeleton of the rat mystacial pad.

    Science.gov (United States)

    Haidarliu, Sebastian; Simony, Erez; Golomb, David; Ahissar, Ehud

    2011-05-01

    Anatomical and functional integrity of the rat mystacial pad (MP) is dependent on the intrinsic organization of its extracellular matrix. By using collagen autofluorescence, in the rat MP, we revealed a collagenous skeleton that interconnects whisker follicles, corium, and deep collagen layers. We suggest that this skeleton supports MP tissues, mediates force transmission from muscles to whiskers, facilitates whisker retraction after protraction, and limits MP extensibility.

  18. CO2 laser biopsies of oral mucosa: an immunocytological and histological comparative study

    Science.gov (United States)

    Vitale, Marina C.; Botticelli, Annibale R.; Zaffe, Davide; Martignone, Alessandra; Cisternino, Aurelia; Vezzoni, Franco; Scarpelli, Francesco

    2001-04-01

    The relationship between bioptic technique and tissue preservation has been studied in 18 oral biopsies of young patients obtained by electro surgery or CO2 laser surgery. Biopsies were formalin fixed, paraffin embedded and histologically, histochemically and immunocytochemically treated. All the biopsies show inflammatory cell infiltration, epithelial spongiosis, trichocariosis, supra basal small blisters, and epithelial clefts with lamina detaching from the corium. Histochemistry shows both the presence of edema and acid mucopolysaccharides inside the corium, and variable glycogen content in epithelial cells. Trichocariotic cells show a positive MiB1/Ki67 expression, when they are present. Nevertheless, laser biopsies show a lower amount of basophilic fibrous tissue and of bc12 bodies detection, connected with a higher amount of glycogen, Cytokeratin and MiB1/Ki67 expression in epithelial cells, compared to bovie biopsies. The result show a higher degree of damages in particular at the epithelial level, in electro surgery biopsies rather than laser biopsies. The best epithelial and corium preservation showed by laser biopsies suggest a chance of reversible condition, which can lead to a complete recovery due to its higher capability of restoring tissues.

  19. Analysis of the KROTOS KFC test by coupling X-Ray image analysis and MC3D calculations

    Energy Technology Data Exchange (ETDEWEB)

    Brayer, C.; Charton, A.; Grishchenko, D.; Fouquart, P.; Bullado, Y.; Compagnon, F.; Correggio, P.; Cassiaut-Louis, N.; Piluso, P. [Commissariat a l' Energie Atomique et Aux Energies Alternatives, CEA Cadarache, DEN, F-13108 Saint-Paul-Les-Durance (France)

    2012-07-01

    During a hypothetical severe accident sequence in a Pressurized Water Reactor (PWR), the hot molten materials (corium) issuing from the degraded reactor core may generate a steam explosion if they come in contact with water and may damage the structures and threaten the reactor integrity. The SERENA program is an international OECD project that aims at helping the understanding of this phenomenon also called Fuel Coolant Interaction (FCI) by providing data. CEA takes part in this program by performing tests in its KROTOS facility where steam explosions using prototypic corium can be triggered. Data about the different phases in the premixing are extracted from the KROTOS X-Ray radioscopy images by using KIWI software (KROTOS Image analysis of Water-corium Interaction) currently developed by CEA. The MC3D code, developed by IRSN, is a thermal-hydraulic multiphase code mainly dedicated to FCI studies. It is composed of two applications: premixing and explosion. An overall FCI calculation with MC3D requires a premixing calculation followed by an explosion calculation. The present paper proposes an alternative approach in which all the features of the premixing are extracted from the X-Ray pictures using the KIWI software and transferred to an MC3D dataset for a direct simulation of the explosion. The main hypothesis are discussed as well as the first explosion results obtained with MC3D for the KROTOS KFC test. These results are rather encouraging and are analyzed on the basis of comparisons with the experimental data. (authors)

  20. Phytochemical screening of twelve species of phytoplankton isolated from Arabian Sea coast

    Institute of Scientific and Technical Information of China (English)

    Sushanth Vishwanath Rai; Madaiah Rajashekhar

    2015-01-01

    Objective:To analyze the phytochemicals in twelve species of marine phytoplankton. Methods: Total phenolic content of methanol extract was estimated by the Folin-Ciocalteu method. Total flavonoid content of the methanol extarct was determined by aluminium chloride method. Chlorophylls,β-carotene and astaxanthin were estimated by acetone extraction method. Vitamin C was determined by dinitrophenyl-hydrazine method. Phycobiliproteins such as allophycocyanin, phycocyanin and phycoerythrin in the aqueous extracts were determined. Results: Total phenolics varied from 5.41 mg gallic acid equivalents/g dry weight (DW) in Phormidium corium (P. corium) to 17.37 mg gallic acid equivalents/g DW inOscillatoria fremyii(O. fremyii). Total flavonoids ranged between 0.74 mg quercetin equivalent/g DW inP. corium and 9.87 mg quercetin equivalent/g DW inNannochloropsis oceanica. Chlorophyll-a pigment was high inChaetoceros calcitrans(C. calcitrans)(15.51 mg/g DW) and low inP. corium (1.08 mg/g DW). Chlorophyll-c ranged between 0.07 mg/g DW inNannochloropsis oceanica and 4.62 mg/g DW inC. calcitrans. High contents ofβ-carotene and astaxanthin were found inC. calcitrans and low inP. corium which ranged from 0.33 to 10.03 mg/g DW and 0.18 to 3.85 mg/g DW, respectively. Vitamin C content varied from 0.50 mg/g DW inC. calcitrans to 1.51 mg/g DW inPhormidium tenue.O. fremyii showed highest total phycobiliproteins of 317.05 mg/g DW. High contents of allophycocyanin and phycocyanin were found inO. fremyii, whereas high contents of phycoerythrin were found inOscillatoria sancta. All the three phycobiliproteins were low inChroococcus turgidus. Conclusions: Marine phytoplankton are one of the natural sources providing novel biologically active compounds with potential for pharmaceutical applications.

  1. OECD MCCI Small-Scale Water Ingression and Crust Strength Tests (SSWICS) SSWICS-3 test data report : thermal Hydraulic results, Rev. 0 February 19, 2003.

    Energy Technology Data Exchange (ETDEWEB)

    Lomperski, S.; Farmer, M. T.; Kilsdonk, D.; Aeschlimann, B. (Nuclear Engineering Division)

    2011-05-23

    The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core/concrete interaction (MCCI) when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. However, due to the integral nature of these tests, several questions regarding the crust freezing behavior could not be adequately resolved. These questions include: (1) To what extent does water ingression into the crust increase the melt quench rate above the conduction-limited rate and how is this affected by melt composition and system pressure and (2) What is the fracture strength of the corium crust when subjected to a thermal-mechanical load and how does it depend upon the melt composition? A series of separate-effects experiments are being conducted to address these issues. The first employs an apparatus designed to measure the quench rate of a pool of corium ({approx}{phi}30 cm; up to 20 cm deep). The main parameter to be varied in these quench tests is the melt composition since it is thought to have a critical influence on the crust cracking behavior which, in turn, alters quench rate. The issue of crust strength will be addressed with a second apparatus designed to mechanically load the crust produced by the quench tests. This apparatus will measure the fracture strength of the crust while under a thermal load created by a heating element beneath the crust. The two apparatuses used to measure the melt quench rate and crust strength are jointly referred to as SSWICS (Small-Scale Water Ingression and Crust Strength). This report describes results of the third water ingression test, designated SSWICS-3. This test investigated the quenching behavior of a fully

  2. Small-Scale Water Ingression and Crust Strength Tests (SSWICS) SSWICS-6 test data report : thermal hydraulic results, Rev. 0.

    Energy Technology Data Exchange (ETDEWEB)

    Lomperski, S.; Farmer, M. T.; Kilsdonk, D.; Aeschlimann, B. (Nuclear Engineering Division)

    2011-06-28

    The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core/concrete interaction (MCCI) when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. However, due to the integral nature of these tests, several questions regarding the crust freezing behavior could not be adequately resolved. These questions include: (1) To what extent does water ingression into the crust increase the melt quench rate above the conduction-limited rate and how is this affected by melt composition and system pressure? (2) What is the fracture strength of the corium crust when subjected to a thermal-mechanical load and how does it depend upon the melt composition? A series of separate-effects experiments are being conducted to address these issues. The first employs an apparatus designed to measure the quench rate of a pool of corium ({approx} {phi} 30 cm; up to 20 cm deep). The main parameter to be varied in these quench tests is the melt composition since it is thought to have a critical influence on the crust cracking behavior which, in turn, alters quench rate. The issue of crust strength is being addressed with a second apparatus designed to mechanically load the crust produced by the quench tests. This apparatus measures the fracture strength of the crust while it is either at room temperature or above, the latter state being achieved with a heating element placed below the crust. The two apparatuses used to measure the melt quench rate and crust strength are jointly referred to as SSWICS (Small-Scale Water Ingression and Crust Strength). This report describes results of the sixth water ingression test, designated SSWICS-6. This test

  3. R and D relative to the serious accidents in the PWR type reactors: assessment and perspectives; R and D relative aux accidents graves dans les reacteurs a eau pressurisee: bilan et perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Bentaib, A.; Bonneville, H.; Caroli, H.; Chaumont, B.; Clement, B.; Cranga, M.; Koundy, V.; Laurent, B.; Micaelli, J.C.; Meignen, R.; Pichereau, F.; Plassart, D.; Van-Dorsselaere, P. [Institut de Radioprotection et de Surete Nucleaire (IRSN), 92 - Clamart (France); Ducros, G.; Journeau, Ch.; Magallon, D. [CEA Cadarache, 13 - Saint Paul lez Durance (France); Durin, M.; Studer, E. [CEA Saclay 91 - Gif sur Yvette (France); Seiler, J.M. [CEA Grenoble, 38 (France); Ranval, W. [Electricite de France (EDF), 75 - Paris (France)

    2006-07-01

    This document presents the current state of the research relative to the grave accidents realized in France and abroad. It aims at giving the most exhaustive possible and objective vision of this original field of research. He allows to contribute to the identification and to the hierarchical organization of the needs of R and D, this hierarchical organization in front of, naturally, to be completed by a strong lighting on needs in terms of safety analyses associated with the different risks and the physical phenomena, in particular with the support of probability evaluations of safety level 2, whose the level of sharpness must be sufficient not to hide, by construction, physical phenomena of which the limited knowledge leads to important uncertainties. Let us note that neither the safety analyses, nor the E.P.S. 2 are presented in this document. This report presents the physical phenomena which can arise during a grave accident, in the reactor vessel and in the reactor containment, their chain and the means allowing to ease the effects. The corresponding scenarios are presented to the chapter 2. The chapter 3 is dedicated to the progress of the accident in the reactor vessel; the degradation of the core in reactor vessel (3.1), the behavior of the corium in bottom of reactor vessel (3.2) the break of the reactor vessel (3.3) and the fusion in pressure (3.4) are thus handled there. The chapter 4 concerns the phenomena which can lead to a premature failure of the containment, namely the direct heating of gases of the containment (4.1), the hydrogen risk (4.2) and the vapor explosion (4.3). The phenomenon which can lead to a delayed failure from the containment, namely the interaction corium-concrete, is approached on the chapter 5. The chapter 6 is dedicated to the problems connected to the keeping back and to the corium cooling in reactor vessel and out of reactor vessel, namely the keeping back in reactor vessel by re-flooding of the primary circuit or by re

  4. Fuel and fission product behaviour in early phases of a severe accident. Part II: Interpretation of the experimental results of the PHEBUS FPT2 test

    Energy Technology Data Exchange (ETDEWEB)

    Dubourg, R. [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); Barrachin, M., E-mail: marc.barrachin@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); Ducher, R. [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); Gavillet, D. [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); De Bremaecker, A. [Institute for Nuclear Materials Sciences, SCK-CEN, Boeretang 200, B-2400 Mol (Belgium)

    2014-10-15

    One objective of the FPT2 test of the PHEBUS FP Program was to study the degradation of an irradiated UO{sub 2} fuel bundle and the fission product behaviour under conditions of low steam flow. The results of the post-irradiation examinations (PIE) at the upper levels (823 mm and 900 mm) of the test section previously reported are interpreted in the present paper. Solid state interactions between fuel and cladding have been compared with the characteristics of interaction identified in the previous separate-effect tests. Corium resulting from the interaction between fuel and cladding was formed. The uranium concentration in the corium is compared to analytical tests and a scenario for the corium formation is proposed. The analysis showed that, despite the rather low fuel burn up, the conditions of temperature and oxygen potential reached during the starvation phase are able to give an early very significant release fraction of caesium. A significant part (but not all) of the molybdenum was segregated at grain boundaries and trapped in metallic inclusions from which they were totally removed in the final part of the experiment. During the steam starvation phase, the conditions of oxygen potential were favourable for the formation of simple Ba and BaO chemical forms but the temperature was too low to provoke their volatility. This is one important difference with out-of-pile experiments such as VERCORS for which only a combination of high temperature and low oxygen potential induced a significant barium release. Finally another significant difference with analytical out-of-pile experiments comes from the formation of foamy zones due to the fission gas presence in FPT2-type experiments which give an additional possibility for the formation of stable fission product compounds.

  5. Fuel and fission product behaviour in early phases of a severe accident. Part II: Interpretation of the experimental results of the PHEBUS FPT2 test

    Science.gov (United States)

    Dubourg, R.; Barrachin, M.; Ducher, R.; Gavillet, D.; De Bremaecker, A.

    2014-10-01

    One objective of the FPT2 test of the PHEBUS FP Program was to study the degradation of an irradiated UO2 fuel bundle and the fission product behaviour under conditions of low steam flow. The results of the post-irradiation examinations (PIE) at the upper levels (823 mm and 900 mm) of the test section previously reported are interpreted in the present paper. Solid state interactions between fuel and cladding have been compared with the characteristics of interaction identified in the previous separate-effect tests. Corium resulting from the interaction between fuel and cladding was formed. The uranium concentration in the corium is compared to analytical tests and a scenario for the corium formation is proposed. The analysis showed that, despite the rather low fuel burn up, the conditions of temperature and oxygen potential reached during the starvation phase are able to give an early very significant release fraction of caesium. A significant part (but not all) of the molybdenum was segregated at grain boundaries and trapped in metallic inclusions from which they were totally removed in the final part of the experiment. During the steam starvation phase, the conditions of oxygen potential were favourable for the formation of simple Ba and BaO chemical forms but the temperature was too low to provoke their volatility. This is one important difference with out-of-pile experiments such as VERCORS for which only a combination of high temperature and low oxygen potential induced a significant barium release. Finally another significant difference with analytical out-of-pile experiments comes from the formation of foamy zones due to the fission gas presence in FPT2-type experiments which give an additional possibility for the formation of stable fission product compounds.

  6. FARO base case post-test analysis by COMETA code

    Energy Technology Data Exchange (ETDEWEB)

    Annunziato, A.; Addabbo, C. [Joint Research Centre, Ispra (Italy)

    1995-09-01

    The paper analyzes the COMETA (Core Melt Thermal-Hydraulic Analysis) post test calculations of FARO Test L-11, the so-called Base Case Test. The FARO Facility, located at JRC Ispra, is used to simulate the consequences of Severe Accidents in Nuclear Power Plants under a variety of conditions. The COMETA Code has a 6 equations two phase flow field and a 3 phases corium field: the jet, the droplets and the fused-debris bed. The analysis shown that the code is able to pick-up all the major phenomena occurring during the fuel-coolant interaction pre-mixing phase.

  7. Material effects on multiphase phenomena in late phases of severe accidents of nuclear reactors; Effets des materiaux sur les phenomenes multiphasiques se produisant lors des phases avancees d'accident grave de reacteur nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Seiler, J.M.; Froment, K

    2003-07-01

    This paper reviews and presents work carried out in the French Atomic Energy Commission (CEA) on the subject of nuclear severe accidents, i.e. those which are accompanied by melting of the nuclear core material. The emphasis is on the (crucial) thermodynamic and material behaviour of corium melts in the solidus-liquidus temperature interval, which is linked to the thermal hydraulic description. A global model approach is proposed. The work is presented in the context of the overall international effort in the area. (authors)

  8. The ASN considers the Fessenheim-1 reactor able to operate 10 years more; L'ASN juge le reacteur 1 de Fessenheim apte a fonctionner 10 annees de plus

    Energy Technology Data Exchange (ETDEWEB)

    Anon

    2011-07-15

    On July 4., 2011 the Authority of Nuclear Safety (ASN) allowed the unit 1 of the Fessenheim plant to operate 10 years more under 2 major conditions. First the reinforcement of the foundation mat in order to improve the reactor resistance to corium and secondly the installation of an emergency system for the evacuation of the residual heat in case of the loss of the heat sink. The ASN stresses that this decision does not take into account the conclusions (expected at the end of 2011) of the complementary safety assessment (ECS) that was launched following the Fukushima accident. (A.C.)

  9. The MELTSPREAD-1 computer code for the analysis of transient spreading in containments

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M.T.; Sienicki, J.J.; Spencer, B.W.

    1990-01-01

    A one-dimensional, multicell, Eulerian finite difference computer code (MELTSPREAD-1) has been developed to provide an improved prediction of the gravity driven spreading and thermal interactions of molten corium flowing over a concrete or steel surface. In this paper, the modeling incorporated into the code is described and the spreading models are benchmarked against a simple dam break'' problem as well as water simulant spreading data obtained in a scaled apparatus of the Mk I containment. Results are also presented for a scoping calculation of the spreading behavior and shell thermal response in the full scale Mk I system following vessel meltthrough. 24 refs., 15 figs.

  10. Evaluation of Ablation rate by the change of Sacrificial Material for PECS in EU-APR

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Do Hyun; Kim, Yong Soo; Lee, Keun Sung [KHNP-CRI, Daejeon (Korea, Republic of)

    2015-05-15

    EU-APR, modified and improved from its original design of APR1400, has been developed to comply with European Utility Requirements (EUR) and nuclear design requirements of the European countries. In EU-APR, Severe Accident Mitigation Systems are dedicated to providing an independent defense line from that of Engineered Safety Feature (ESF) and Diverse Safety Feature (DSF). They consist of Emergency Reactor Depressurization System (ERDS), Passive Ex-vessel corium retaining and Cooling System (PECS), Severe Accident Containment Spray System (SACSS), Hydrogen Mitigation System (HMS) and Containment Filtered Vent System (CFVS). The PECS, so called core catcher, was introduced to prevent the Molten Core Concrete Interaction (MCCI) after Reactor Vessel (RV) failure. The PECS has experienced a lot of changes from its original design. Recently, the most significant change was that as a SM, limestone concrete is installed on PECS's body wall instead of previous sacrificial material rich in Fe{sub 2}O{sub 3}. The main reason of this design change is to overcome the issue that the sacrificial material is ablated rather too fast when reacting with corium that contains a large fraction of Zr metal. Other changes in the geometry of PECS's wall and downcomer design are considered as minor ones. In this paper, the comparison of ablation rates between previous SM and limestone concrete is carried out using MAAP5 code with respective MCCI model according to the material. In this paper, major improvements of MAAP5 model for PECS in EU-APR are presented and the evaluation of ablation rate for the previous SM model and the new LC model is carried out by means of ablation depths with LBLOCA sequence. Two models have respective unique ablation process. The ablation of LC model proceeds at a constant rate regardless of water while the ablation of SM model proceeds at a faster rate before the arrival of cooling water for corium and SM mixture. The change of sacrificial material

  11. Research at the CEA in the field of safety in 2nd and 3rd generation light water reactors

    Science.gov (United States)

    Billot, Philippe

    2012-05-01

    The research programs at the CEA in the field of safety in nuclear reactors are carried out in a framework of international partnerships. Their purpose is to develop studies on: The methods allowing for the determination of earthquake hazards and their consequences; The behaviour of fuel in an accident situation; The comprehension of deflagration and detonation phenomena of hydrogen and the search for effective prevention methods involving an explosion risk; The cooling of corium in order to stop its progression in and outside the vessel thereby reducing the risk of perforating the basemat; The behaviour of the different fission product families according to their volatility for the UO2 and MOX fuels.

  12. OECD MCCI project Small-Scale Water Ingression and Crust Strength Tests (SSWICS) SSWICS-1 test data report : thermal hydraulic results. Rev. 0 September 20, 2002.

    Energy Technology Data Exchange (ETDEWEB)

    Lomperski, S.; Farmer, M. T.; Kilsdonk, D. J.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division); (NRC)

    2011-05-23

    The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core/concrete interaction (MCCI) when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. However, due to the integral nature of these tests, several questions regarding the crust freezing behavior could not be adequately resolved. These questions include: (1) To what extent does water ingression into the crust increase the melt quench rate above the conduction-limited rate and how is this affected by melt composition and system pressure and (2) What is the fracture strength of the corium crust when subjected to a thermal-mechanical load and how does it depend upon the melt composition? A series of separate-effects experiments are being conducted to address these issues. The first employs an apparatus designed to measure the quench rate of a pool of corium ({approx}{phi}30 cm; up to 20 cm deep). The main parameter to be varied in these quench tests is the melt composition since it is thought to have a critical influence on the crust cracking behavior which, in turn, alters quench rate. The issue of crust strength will be addressed with a second apparatus designed to mechanically load the crust produced by the quench tests. This apparatus will measure the fracture strength of the crust while under a thermal load created by a heating element beneath the crust. The two apparatuses used to measure the melt quench rate and crust strength are jointly referred to as SSWICS (Small-Scale Water Ingression and Crust Strength). This report describes results of the first water ingression test, designated SSWICS-1. The report includes a description of the test apparatus, the

  13. OECD MMCI Small-Scale Water Ingression and Crust Strength Tests (SSWICS) SSWICS-2 test data report : thermal hydraulic results, Rev. 0 September 20, 2002.

    Energy Technology Data Exchange (ETDEWEB)

    Lomperski, S.; Farmer, M. T.; Kilsdonk, D. J.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division); (NRC)

    2011-05-23

    The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core/concrete interaction (MCCI) when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. However, due to the integral nature of these tests, several questions regarding the crust freezing behavior could not be adequately resolved. These questions include: (1) To what extent does water ingression into the crust increase the melt quench rate above the conduction-limited rate and how is this affected by melt composition and system pressure and (2) What is the fracture strength of the corium crust when subjected to a thermal-mechanical load and how does it depend upon the melt composition? A series of separate-effects experiments are being conducted to address these issues. The first employs an apparatus designed to measure the quench rate of a pool of corium ({approx}{phi}30 cm; up to 20 cm deep). The main parameter to be varied in these quench tests is the melt composition since it is thought to have a critical influence on the crust cracking behavior which, in turn, alters quench rate. The issue of crust strength will be addressed with a second apparatus designed to mechanically load the crust produced by the quench tests. This apparatus will measure the fracture strength of the crust while under a thermal load created by a heating element beneath the crust. The two apparatuses used to measure the melt quench rate and crust strength are jointly referred to as SSWICS (Small-Scale Water Ingression and Crust Strength). This report describes results of the second water ingression test, designated SSWICS-2. The report includes a description of the test apparatus, the

  14. Development of Scaling Approach for Prediction of Terminal Spread Thickness of Melt Poured into a Pool of Water

    OpenAIRE

    Konovalenko, Alexander; Kudinov, Pavel

    2012-01-01

    Corium melt stabilization and long term cooling in a pool of water located beneath reactor vessel is adopted in several existing designs of light water reactors (LWRs) as an element in severe accident (SA) mitigation strategy. At certain conditions of melt release into the pool (e.g. large ratio of the vessel breach size to the pool depth), liquid melt can spread under water and reach a coolable configuration. Coolability of the melt is contingent on terminal spread thickness of the melt laye...

  15. Pressure Load Analysis during Severe Accidents for the Evaluation of Late Containment Failure in OPR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. Y.; Ahn, K. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The MAAP code is a system level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a level 2 probabilistic safety assessment or severe accident management strategy developments. The code employs lots of user-options for supporting a sensitivity and uncertainty analysis. The present application is mainly focused on determining an estimate of the containment building pressure load caused by severe accident sequences. Key modeling parameters and phenomenological models employed for the present uncertainty analysis are closely related to in-vessel hydrogen generation, gas combustion in the containment, corium distribution in the containment after a reactor vessel failure, corium coolability in the reactor cavity, and molten-corium interaction with concrete. The phenomenology of severe accidents is extremely complex. In this paper, a sampling-based phenomenological uncertainty analysis was performed to statistically quantify uncertainties associated with the pressure load of a containment building for a late containment failure evaluation, based on the key modeling parameters employed in the MAAP code and random samples for those parameters. Phenomenological issues surrounding the late containment failure mode are highly complex. Included are the pressurization owing to steam generation in the cavity, molten corium-concrete interaction, late hydrogen burn in the containment, and the secondary heat removal availability. The methodology and calculation results can be applied for the optimum assessment of a late containment failure model. The accident sequences considered were a loss of coolant accidents and loss of offsite accidents expected in the OPR-1000 plant. As a result, uncertainties addressed in the pressure load of the containment building were quantified as a function of time. A realistic evaluation of the mean and variance estimates provides a more complete

  16. On the Analysis and Evaluation of Direct Containement Heating with the Multidimensional Multiphase Flow Code MC3D

    Directory of Open Access Journals (Sweden)

    Tanguy Janin

    2010-01-01

    Full Text Available In the course of a postulated severe accident in an NPP, Direct Containment Heating (DCH may occur after an eventual failure of the vessel. DCH is related to dynamical, thermal, and chemical phenomena involved by the eventual fine fragmentation and dispersal of the corium melt out of the vessel pit. It may threaten the integrity of the containment by pressurization of its atmosphere. Several simplified modellings have been proposed in the past but they require a very strong fitting which renders any extrapolation regarding geometry, material, and scales rather doubtful. With the development of multidimensional multiphase flow computer codes, it is now possible to investigate the phenomenon numerically with more details. We present an analysis of the potential of the MC3D code to support the analysis of this phenomenon, restricting our discussion to the dynamical processes. The analysis is applied to the case of French 1300 MWe PWR reactors for which we derive a correlation for the corium dispersal rate for application in a Probabilistic Safety Analysis (PSA level 2 study.

  17. Ex-Vessel Core Melt Modeling Comparison between MELTSPREAD-CORQUENCH and MELCOR 2.1

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Farmer, Mitchell [Argonne National Lab. (ANL), Argonne, IL (United States); Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-03-01

    System-level code analyses by both United States and international researchers predict major core melting, bottom head failure, and corium-concrete interaction for Fukushima Daiichi Unit 1 (1F1). Although system codes such as MELCOR and MAAP are capable of capturing a wide range of accident phenomena, they currently do not contain detailed models for evaluating some ex-vessel core melt behavior. However, specialized codes containing more detailed modeling are available for melt spreading such as MELTSPREAD as well as long-term molten corium-concrete interaction (MCCI) and debris coolability such as CORQUENCH. In a preceding study, Enhanced Ex-Vessel Analysis for Fukushima Daiichi Unit 1: Melt Spreading and Core-Concrete Interaction Analyses with MELTSPREAD and CORQUENCH, the MELTSPREAD-CORQUENCH codes predicted the 1F1 core melt readily cooled in contrast to predictions by MELCOR. The user community has taken notice and is in the process of updating their systems codes; specifically MAAP and MELCOR, to improve and reduce conservatism in their ex-vessel core melt models. This report investigates why the MELCOR v2.1 code, compared to the MELTSPREAD and CORQUENCH 3.03 codes, yield differing predictions of ex-vessel melt progression. To accomplish this, the differences in the treatment of the ex-vessel melt with respect to melt spreading and long-term coolability are examined. The differences in modeling approaches are summarized, and a comparison of example code predictions is provided.

  18. Numerical analysis of crust formation in molten core-concrete interaction using MPS method

    Energy Technology Data Exchange (ETDEWEB)

    Seiichi, Koshizuka; Shoji, Matsuura; Mizue, Sekine; Yoshiaki, Oka [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.; Hiroyuki, Obata [Japan Atomic Power Co., Tokyo (Japan)

    2001-07-01

    A two-dimensional code is developed for molten core-concrete interaction (MCCI) based on Moving Particle Semi-implicit (MPS) method. Heat transfer is calculated without any specific correlations. A particle can be changed to a moving (fluid) or fixed (solid) particle corresponding to its enthalpy, which provide the phase change model for particles. The phase change model is verified by one-dimensional test calculations. Nucleate boiling and radiation heat transfers are considered between the core debris and the water pool. The developed code is applied to SWISS-2 experiment in which stainless steel is used as the melt material. Calculated heat flux to the water pool agrees well with the experiment, though the ablation speed in the concrete is a little slower. A stable crust is formed in a short time after water is poured in and the heat flux to the water pool rapidly decreases. MACE-M0 using corium is also analyzed. The ablation speed of concrete is slower than that of SWISS-2 because of low heat conduction in corium. An unlimited geometry is analyzed by setting the cyclic boundary condition on the sides. When the crust is broken by the decomposition gas, heat transfer to the water pool is kept high for a longer time because the crust re-formation is delayed. (author)

  19. COTELS project (4) : structural investigation of solidified debris in MCCI

    Energy Technology Data Exchange (ETDEWEB)

    Zhdanov, V.; Vasilyev, Y.; Kolodeshnikov, A.; Cherepnin, Y. [National Nuclear Center, Kurchatov (Kazakhstan). Inst. of Atomic Energy; Sakaki, Isao; Nagasaka, Hideo [Nuclear Power Engineering Corp., Tokyo (Japan). Systems Safety Dept.

    2000-05-01

    Cross section of concrete trap along with solidified debris tested in COTELS test B/C, in which the interaction among core melt, water and concrete was simulated, were structurally investigated. In 6 tests out of 10 tests, particulate debris bed was formed above continuous ingot debris. The size distribution of the particulate debris was well correlated by Rosin-Rammler equation. Large amount of smallest diameter particles was obtained due to the entrainment of molten corium, decomposed concrete and oxidation of metallic components in corium associated with molten core concrete interaction (MCCI) generated gas. The upper region of the solidified debris included more concrete compositions. The concrete erosion depth, concrete degradation condition and the structure of solidified debris were evaluated to clarify the basic difference between COTELS and former tests results. Concrete erosion depth was less than that observed in MACE, WETCOR, SWISS tests. The major differences of COTELS results compared with the former test results were: 1) absence of strong adhesion of crust to melt trap side wall: 2) water penetration into debris through both eroded side wall and channels inside ingot debris: 3) absence of large void inside ingot debris: and 4) formation of pebble bed below ingot debris. All of these promoted the suppression of MCCI. (orig.)

  20. Recent severe accident research synthesis of the major outcomes from the SARNET network

    Energy Technology Data Exchange (ETDEWEB)

    Van Dorsselaere, J.-P., E-mail: jean-pierre.van-dorsselaere@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), Saint-Paul-lez-Durance (France); Auvinen, A. [VTT Technical Research Centre, Espoo (Finland); Beraha, D. [Gesellschaft für Anlagen- und Reaktorsicherheit mbH (GRS), Köln (Germany); Chatelard, P. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), Saint-Paul-lez-Durance (France); Herranz, L.E. [Centro de Investigaciones Energéticas MedioAmbientales y Tecnológicas (CIEMAT), Madrid (Spain); Journeau, C. [Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA), Paris (France); Klein-Hessling, W. [Gesellschaft für Anlagen- und Reaktorsicherheit mbH (GRS), Köln (Germany); Kljenak, I. [Jozef Stefan Institute (JSI), Ljubljana (Slovenia); Miassoedov, A. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Paci, S. [University of Pisa, Pisa (Italy); Zeyen, R. [European Commission Joint Research Centre, Institute for Energy (JRC/IET), Petten (Netherlands)

    2015-09-15

    Highlights: • SARNET network of excellence integration mid-2013 in the NUGENIA Association. • Progress of knowledge on corium behaviour, hydrogen explosion and source term. • Further development of ASTEC integral code to capitalize knowledge. • Ranking of next R&D high priority issues accounting for international research. • Dissemination of knowledge through education courses and ERMSAR conferences. - Abstract: The SARNET network (Severe Accident Research NETwork of excellence), co-funded by the European Commission from 2004 to 2013, has allowed to significantly improve the knowledge on severe accidents and to disseminate it through courses and ERMSAR conferences. The major investigated topics, involving more than 250 researchers from 22 countries, were in- and ex-vessel corium/debris coolability, molten-core–concrete-interaction, steam explosion, hydrogen combustion and mitigation in containment, impact of oxidising conditions on source term, and iodine chemistry. The ranking of the high priority issues was updated to account for the results of recent international research and for the impact of Fukushima nuclear accidents in Japan. In addition, the ASTEC integral code was further developed to capitalize the new knowledge. The network has reached self-sustainability by integration in mid-2013 into the NUGENIA Association. The main activities and outcomes of the network are presented.

  1. Preparation and in vitro characterization of dexamethasone-loaded poly(D,L-lactic acid) microspheres embedded in poly(ethylene glycol)-poly({varepsilon}-caprolactone)-poly(ethylene glycol) hydrogel for orthopedic tissue engineering.

    Science.gov (United States)

    Fan, Min; Guo, QingFa; Luo, JingCong; Luo, Feng; Xie, Ping; Tang, XiaoHai; Qian, ZhiYong

    2013-08-01

    The corium is decreased to about half of its thickness in skin defects and wrinkles due to gravity and environment. In this study, dexamethasone/poly(d,l-lactic acid) (Mn = 160,000) microspheres were incorporated into poly(ethylene glycol)-poly(ε-caprolactone)-poly(ethylene glycol) (Mn = 3300) hydrogel to prepare an injectable hydrogel composite. The composite was designed to increase the thickness of the corium. Dexamethasone/poly(d,l-lactic acid) microspheres were prepared by oil-in-water emulsion/solvent evaporation technique. The properties of microspheres were investigated by size distribution measurement, scanning electron microscope and x-ray diffraction. Drug loading, encapsulation efficiency, and drug delivery behavior of microspheres were also studied in detail. Cell adhesion of microspheres was investigated by NIH3T3 cell in vitro. The properties of hydrogel composite were investigated by scanning electron microscope, rheological measurements and methyl thiazolyl tetrazolium assay. Drug release from composite was determined by HPLC-UV analysis. These results suggested that poly(d,l-lactic acid) microspheres encapsulating dexamethasone embedded in poly(ethylene glycol)-poly(ε-caprolactone)-poly(ethylene glycol) hydrogel might have prospective application in orthopedic tissue engineering field.

  2. CAREM-25 RPV thermal regime evaluation during the application of in-vessel retention strategies

    Energy Technology Data Exchange (ETDEWEB)

    Pomier Baez, Lazaro E.; Baron, Jorge H.; Nunez Mac Leod, Juan E. [Universidad Nacional de Cuyo, Mendoza (Argentina). Facultad de Ingenieria. Instituto CEDIAC

    2002-07-01

    The structural integrity of the reactor vessel is a key question in the analysis of the possibility for retaining the melted materials inside the pressure vessel as a severe accident management (SAM) strategy. The pressure of the system and the thermohydraulic behavior of relocated materials that determine the loads, stresses and the displacements of reactor vessel determine the vessel failure mode and the time until rupture. In-Vessel Retention (IVR) strategy analyses are carried on as part of the advanced design CAREM-25 SAM evaluations. One of the most promising initiatives in this area, is the development of an in-vessel metallic core catcher (IVCC) to arrest reactor vessel meltdown sequences during a severe accident. The concept working principle consists in the mixing of the catcher metallic material (so-called sacrificial material) with the corium relocating fragments in the reactor lower head after the initiation of relocation process. The catcher will limit the catcher-corium mixture temperature by boiling the sacrificial material. For this purpose, a low-boiling point material is chosen. The analysis methodology presented in this paper is designed to evaluate the integrity of CAREM-25 pressure vessel during a severe accident sequence with complete core damage when the IVR strategy is employed. CAREM-25 is a multipurpose small advanced reactor design being developed by CNEA (Comision Nacional de Energia Atomica) and INVAP S. E. in Argentina. (author)

  3. A study on the correlations development for film boiling heat transfer on spheres

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yong Hoon; Baek, Won Pil; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    Film boiling is the heat transfer mechanism that can occurs when large temperature differences exist between a cold liquid and hot material. In the nuclear reactor safety analysis, film boiling has become an important issue in recent years. During severe accident, hot molten corium fall into relatively cool water, and fragment into spheres or sphere-like particles. If the steam explosion is triggered, the thermal energy of corlium is converted into the mechanical energy that can threaten the integrity of reactor vessel or reactor cavity. One of the important concerns in the heat transfer analysis during pre-mixing stage is the film boiling heat transfer between the corium and water/steam two-phase flow. Until now, considerable works on film boiling have been performed. However, there is no available correlation adequate for severe accident analysis. In this study, film boiling heat transfer correlations have been developed, and their applicable ranges have been enlarged and their prediction accuracy has been enhanced. 7 refs., 5 figs., 5 tabs. (Author)

  4. Research and development with regard to severe accidents in pressurised water reactors: Summary and outlook

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-07-01

    This document reviews the current state of research on severe accidents in France and other countries. It aims to provide an objective vision, and one that's as exhaustive as possible, for this innovative field of research. It will help in identifying R and D requirements and categorising them hierarchically. Obviously, the resulting prioritisation must be completed by a rigorous examination of needs in terms of safety analyses for various risks and physical phenomena, especially in relation to Level 2 Probabilistic Safety Assessments. PSA-2 should be sufficiently advanced so as not to obscure physical phenomena that, if not properly understood, might result in substantial uncertainty. It should be noted that neither the safety analyses nor PSA-2 are presented in this document. This report describes the physical phenomena liable to occur during a severe accident, in the reactor vessel and the containment. It presents accident sequences and methods for limiting impact. The corresponding scenarios are detailed in Chapter 2. Chapter 3 deals with in-vessel accident progression, examining core degradation (3.1), corium behaviour in the lower head (3.2), vessel rupture (3.3) and high-pressure core meltdown (3.4). Chapter 4 focuses on phenomena liable to induce early containment failure, namely direct containment heating (4.1), hydrogen risk (4.2) and steam explosions (4.3). The phenomenon that could lead to a late containment failure, namely molten core-concrete interaction, is discussed in Chapter 5. Chapter 6 focuses on problems related to in-vessel and ex-vessel corium retention and cooling, namely in-vessel retention by flooding the primary circuit or the reactor pit (6.1), cooling of the corium under water during the corium-concrete interaction (6.2), corium spreading (6.3) and ex-vessel core catchers (6.4). Chapter 7 relates to the release and transport of fission products (FP), addressing the themes of in-vessel FP release (7.1) and ex-vessel FP release (7

  5. Study of diluting and absorber materials to control reactivity during a postulated core melt down accident in Generation IV reactors; Etude des materiaux sacrificiels absorbants et diluants pour le controle de la reactivite dans le cas d'un accidnet hypothetique de fusion du coeur de reacteurs de quatrieme generation

    Energy Technology Data Exchange (ETDEWEB)

    Plevacova, K.

    2010-12-16

    In order to limit the consequences of a hypothetical core meltdown accident in Generation IV Sodium Fast Reactors, absorber materials in or near the core, such as boron carbide B{sub 4}C, and diluting materials in the core catcher will be used to prevent recriticality within the mixture of molten oxide fuel and molten structures called corium. The aim of the PhD thesis was to select materials of both types and to understand their behaviour during their interaction with corium, from chemical and thermodynamic point of view. Concerning B{sub 4}C, thermodynamic calculations and experiments agree with the formation of two immiscible phases at high temperature in the B{sub 4}C - UO{sub 2} system: one oxide and one boride. This separation of phases can reduce the efficiency of the neutrons absorption inside the molten fuel contained in the oxide phase. Moreover, a volatilization of a part of the boron element can occur. According to these results, the necessary quantity of B{sub 4}C to be introduced should be reconsidered for postulated severe accident sequence. Other solution could be the use of Eu{sub 2}O{sub 3} or HfO{sub 2} as absorber material. These oxides form a solid solution with the oxide fuel. Concerning the diluting materials, mixed oxides Al{sub 2}O{sub 3} - HfO{sub 2} and Al{sub 2}O{sub 3} - Eu{sub 2}O{sub 3} were preselected. These systems being completely unknown to date at high temperature in association with UO{sub 2}, first points on the corresponding ternary phase diagrams were researched. Contrary to Al{sub 2}O{sub 3} - Eu{sub 2}O{sub 3} - UO{sub 2} system, the Al{sub 2}O{sub 3} - HfO{sub 2} - UO{sub 2} mixture presents only one eutectic and thus only one solidification path which makes easier forecasting the behaviour of corium in the core catcher. (author) [French] Resume: Afin de limiter les consequences d'un accident grave avec la fusion du coeur dans un reacteur a neutrons rapides de generation IV refroidi au sodium, la recriticite doit

  6. 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

    Energy Technology Data Exchange (ETDEWEB)

    Bottomley, D., E-mail: paul.bottomley@ec.europa.eu [ITU Institut fuer Transurane, PO box 2340, 76125 Karlsruhe (Germany); Stuckert, J.; Hofmann, P. [KIT Campus Nord, Hermann-von-Helmholtz Pl. 1, 76344 Eggenstein-Leopoldshafen (Germany); Tocheny, L. [ISTC Krasnoproletarskaya 32-34, PO Box 20, 127473 Moscow (Russian Federation); Hugon, M. [European Commission DG - Research and Tech. Development, Sq. de Meeus, B-1049 Brussels (Belgium); Journeau, C. [CEA, DEN, Cadarache, F13108 St Paul lez Durance (France); Clement, B. [IRSN PSN-RES/SAG Cadarache, BP3 F13115, St Paul lez Durance (France); Weber, S. [GRS Muenchen, Thermal Hydraulics Div., Garching 85748,Germany (Germany); Guentay, S. [PSI NES/LTH OHSA C11, 5232 Villigen (Switzerland); Hozer, Z. [AEKI Fuel Department, P.O. Box 49, Budapest H-1525 (Hungary); Herranz, L. [CIEMAT, Energy -Nuclear Fission Division, Complutense 40, 28040 Madrid (Spain); Schumm, A. [EDF - R and D, SINETICS, Avenue du General de Gaulle 1, Clamart 92140 (France); Oriolo, F. [Pisa University, Ing. Mecc. Nucl. Prod., Largo Lazarino 2, Pisa 56126 (Italy); Altstadt, E. [HZDR Structural Matls, Rossendorf, Postfach 51 01 19, 01314 Dresden (Germany); Krause, M. [AECL - Reactor Safety, Chalk River, Ontario, Canada K0J 1J0 (Canada); Fischer, M. [AREVA NP GMBH, Dept. PEPA-G, 91058 Erlangen (Germany); Khabensky, V.B. [Alexandrov Institute of Technologies (NITI), Sosnovy Bor (Russian Federation); Bechta, S.V. [Kungliga Tekniska Hoegskolan (KTH), AlbaNova University Centre, Roslagstullsbacken 21, SE-106 91 Stockholm (Sweden); Veshchunov, M.S. [Nuclear Safety Institute (IBRAE), Russian Academy of Sciences, 52 B. Tulskaya, Moscow 115191 (Russian Federation); Palagin, A.V. [KIT Campus Nord, Hermann-von-Helmholtz Pl. 1, 76344 Eggenstein-Leopoldshafen (Germany); and others

    2012-11-15

    Highlights: Black-Right-Pointing-Pointer ISTC supported successful nuclear safety projects between EU and Russian institutes. Black-Right-Pointing-Pointer Two-tier project monitoring has proved to be very successful and flexible. Black-Right-Pointing-Pointer Examples are reactor degradation, corium steel corrosion, and corium thermodynamics. - 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: Bullet 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, Bullet 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

  7. OECD MCCI Small-Scale Water Ingression and Crust Strength tests (SSWICS) design report, Rev. 2 October 31, 2002.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M.; Lomperski, S.; Kilsdonk, D.; Aeschlimann, B.; Pfeiffer, P. (Nuclear Engineering Division); (NRC)

    2011-05-23

    The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core/concrete interaction (MCCI) when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. However, due to the integral nature of these tests, several questions regarding the crust freezing behavior could not be adequately resolved. These questions include: (1) To what extent does water ingression into the crust increase the melt quench rate above the conduction-limited rate and how is this affected by melt composition and system pressure and (2) What is the fracture strength of the corium crust when subjected to a thermal-mechanical load and how does it depend upon the melt composition? A series of separate-effects experiments are planned to address these issues. The first employs an apparatus designed to measure the quench rate of a pool of corium ({approx}{phi}30 cm; up to 20 cm deep). The main parameter to be varied in these quench tests is the melt composition since it is thought to have a critical influence on the crust cracking behavior which, in turn, alters quench rate. A description of the test apparatus, instrumentation, data reduction, and test matrix are the subject of the first portion of this report. The issue of crust strength will be addressed with a second apparatus designed to mechanically load the crust produced by the quench tests. This apparatus will measure the fracture strength of the crust while under a thermal load created by a heating element beneath the crust. The introduction of a thermal gradient across the crust is thought to be important for these tests because of uncertainty in the magnitude of the thermal stresses and thus their

  8. Separate effects tests on hydrogen combustion during direct containment heating events

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, L.; Albrecht, G.; Kirstahler, M.; Schwall, M.; Wachter, E.

    2008-01-15

    In the frame of severe accident research for light water reactors Forschungszentrum Karlsruhe (FZK/IKET) operates the facilities DISCO-C and DISCO-H since 1998, conceived to investigate the direct containment heating (DCH) issue. Previous DCH experiments have investigated the corium dispersion and containment pressurization during DCH in different European reactor geometries using an iron-alumina melt and steam as model fluids. The analysis of these experiments showed that the containment was pressurized by the debris-to-gas heat transfer but also to a large part by hydrogen combustion. The need was identified to better characterize the hydrogen combustion during DCH. To address this issue separate effect tests in the DISCO-H facility were conducted. These tests reproduced phenomena occurring during DCH (injection of a hot steam-hydrogen mixture jet into the containment and ignition of the air-steam-hydrogen mixture) with the exception of corium dispersion. The effect of corium particles as igniters was simulated using sparkler systems. The data will be used to validate models in combustion codes and to extrapolate to prototypic scale. Tests have been conducted in the DISCO-H facility in two steps. First a small series of six tests was done in a simplified geometry to study fundamental parameters. Then, two tests were done with a containment geometry subdivided into a subcompartment and the containment dome. The test conditions were as follows: As initial condition in the containment an atmosphere was used either with air or with a homogeneous air-steam mixture containing hydrogen concentrations between 0 and 7 mol%, temperatures around 100 C and pressure at 2 bar (representative of the containment atmosphere conditions at vessel failure). Injection of a hot steam-hydrogen jet mixture into the reactor cavity pit at 20 bar, representative of the primary circuit blow down through the vessel and hydrogen produced during this phase. The most important variables

  9. A study on the regulatory approach of safety issues for Korean next generation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Keun Seon; Choi, Jeong Tae; Kim, In Joon [Sunmoon Univ., Asan (Korea, Republic of); Lee, Sang Hoon [Korea Association of Nuclear Technology, Taejon (Korea, Republic of); Kang, Gee Sik [Korea Power Engineering Co., Inc., Seoul (Korea, Republic of); Kim, Han Gon [Korea Electric Power Corporation, Seoul (Korea, Republic of); Park, Jae Uk; Kim, Yun Il; Yang, Soo Hyeong [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-10-15

    The project aims to provide the regulatory direction of major technical issues related to the quantitative criteria for probabilistic risk analysis, establishment of containment performance criteria, design of additional water source for long-term corium cooling of core debris, protection against common mode failure of digital I and C system, criteria for safety-related operator action, quantitative reliability targets, classification of plant conditions and acceptance criteria and development of graded QA. These issues are parts of major technical issues resulted from the safety regulation R and D on the next generation reactor. Regulatory directions to be applied to the Korea Next Generation Reactors in this study are developed by analyzing the state-of-the-art of the development of local and foreign regulatory requirements, research trends, and the design features and safety goals of advanced reactors.

  10. Upgrade of High-Energy X-Ray real-time radioscopy for KROTOS experiment

    Energy Technology Data Exchange (ETDEWEB)

    Estre, N.; Payan, E.; Cassiaut-Louis, N.; Compagnon, F.; Valerian, M.; Mallet, R. [CEA-Cadarache (France)

    2015-07-01

    As part of its R and D programs on severe accidents, in particular on understanding of corium-water interaction, CEA is commissioning an update of the KROTOS experiment at Cadarache. The Xray imaging setup (high energy real-time radioscopy) is upgraded to provide the best performances for the new experimental program. In order to fit the performance needs (faster acquisition, smaller detection limit and higher field of view), two radioscopy setups, with two linear accelerators (linacs 9 MV and 6 MV), are placed in the irradiation cell. Having discussed the expected performances in terms of frequency, detection limit and field-of-view, this article details each stage of both radioscopy chains: principles and technical characteristics. Then, linacs and cameras synchronization (at few hundred Hertz), data flows and storage setups are detailed. Finally, experimental characterizations and performance validations on phantom are presented. (authors)

  11. OECD MCCI project enhancing instrumentation for reactor materials experiments, Rev. 0 September 3, 2002.

    Energy Technology Data Exchange (ETDEWEB)

    Lomperski, S.; Basu, S. (Nuclear Engineering Division); (NRC)

    2011-05-23

    Reactor safety experiments for studying the reactions of a molten core (corium) with water and/or concrete involve materials at extremely high temperature. Such high temperature severely restricts the types of sensors that can be employed to measure characteristics of the corium itself. Yet there is great interest in improving instrumentation so that the state of the melt can be established with more precision. In particular, it would be beneficial to increase both the upper range limit and accuracy of temperature measurements. The poor durability of thermocouples at high temperature is also an important issue. For experiments involving a water-quenched melt, direct measurements of the growth rate of the crust separating the melt and water would be of great interest. This is a key element in determining the nature of heat transfer between the melt and coolant. Despite its importance, no one has been able to directly measure the crust thickness during such tests. This paper considers three specialized sensors that could be introduced to enhance melt characterization: (1) A commercially fabricated, single point infrared temperature measurement with the footprint of a thermowell. A lens assembly and fiber optic cable linked to a receiver and amplifier measures the temperature at the base of a tungsten thermowell. The upper range limit is 3000 C and accuracy is {+-}0.25% of the reading. (2) In-house development of an ultrasonic temperature sensor that would provide multipoint measurements at temperatures up to {approx}3000 C. The sensors are constructed from tungsten rods and have a high temperature durability that is superior to that of thermocouples. (3) In-house development of an ultrasonic probe to measure the growth rate of the corium crust. This ultrasonic sensor would include a tungsten waveguide that transmits ultrasonic pulses up through the corium melt towards the crust and detects reflections from the melt/crust interface. A measurement of the echo time

  12. Fuel and fission product behaviour in early phases of a severe accident. Part I: Experimental results of the PHEBUS FPT2 test

    Energy Technology Data Exchange (ETDEWEB)

    Barrachin, M., E-mail: marc.barrachin@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); Gavillet, D. [Paul Scherrer Institute, Würenlingen and Villigen, CH-5232 Villigen PSI (Switzerland); Dubourg, R. [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); De Bremaecker, A. [Institute for Nuclear Materials Sciences, SCK-CEN, Boeretang 200, B-2400 Mol (Belgium)

    2014-10-15

    One objective of the FPT2 test of the PHEBUS FP Program was to study the degradation of an irradiated UO{sub 2} fuel test section and the fission product behaviour under conditions of low steam flow. The results of the post-irradiation examinations (PIE) at the upper levels (823 and 900 mm) of the 1-m long test section are presented in this paper. Material interactions leading to local corium formation were identified: firstly between fuel and Zircaloy-4 cladding, notably at 823 mm, where the cladding melting temperature was reached, and secondly between fuel and stainless steel oxides. Regarding fission products, molybdenum left so-called metallic precipitates mainly composed of ruthenium. Xenon and caesium behave similarly whereas barium and molybdenum often seems to be associated in precipitates.

  13. A view of treatment process of melted nuclear fuel on a severe accident plant using a molten salt system

    Energy Technology Data Exchange (ETDEWEB)

    Fujita, R.; Takahashi, Y.; Nakamura, H.; Mizuguchi, K. [Power and Industrial Research and Development Center, Toshiba Corporation Power Systems Company, 4-1 Ukishima-cho, Kawasaki-ku, Kawasaki 210-0862 (Japan); Oomori, T. [Chemical System Design and Engineering Department, Toshiba Corporation Power Systems Company, 8 Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan)

    2013-07-01

    At severe accident such as Fukushima Daiichi Nuclear Power Plant Accident, the nuclear fuels in the reactor would melt and form debris which contains stable UO2-ZrO2 mixture corium and parts of vessel such as zircaloy and iron component. The requirements for solution of issues are below; -) the reasonable treatment process of the debris should be simple and in-situ in Fukushima Daiichi power plant, -) the desirable treatment process is to take out UO{sub 2} and PuO{sub 2} or metallic U and TRU metal, and dispose other fission products as high level radioactive waste; and -) the candidate of treatment process should generate the smallest secondary waste. Pyro-process has advantages to treat the debris because of the high solubility of the debris and its total process feasibility. Toshiba proposes a new pyro-process in molten salts using electrolysing Zr before debris fuel being treated.

  14. Heating of reactor pressure vessel bottom head and penetrations in a severe reactor accident; Reaktoripaineastian pohjan ja laepivientien kuumeneminen sydaemen sulamisonnettomuudessa

    Energy Technology Data Exchange (ETDEWEB)

    Ikonen, K. [VTT Energy, Espoo (Finland). Nuclear Energy

    1997-10-01

    The report describes the fundamentals of heat conductivity and convection and numerical methods like finite difference and control volume method for calculation of the thermal history of a reactor pressure vessel bottom head and penetrations. Phase changes from solids to liquids are considered. Time integration is performed by explicit or implicit method. Developed computer codes for thermal conductivity and convection analyses and codes for graphical visualization are described. The codes are applied to two practical cases. They deal with analyses of Swiss CORVIS-experiments and analyses of control rod and instrument penetrations in a BWR bottom head. A model for calculation of effective thermal conductivity of granular corium is developed. The work is also related to EU MVI-project (Core Melt-Pressure Vessel Interactions During a Light Water Reactor Severe Accident), whose coordinator is Prof. B. R. Sehgal at Royal Institute of Technology in Stockholm. (orig.) (11 refs.).

  15. Boiling in porous media; Ebullition en milieux poreux

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-11

    This conference day of the French society of thermal engineers was devoted to the analysis of heat transfers and fluid flows during boiling phenomena in porous media. This book of proceedings comprises 8 communications entitled: `boiling in porous medium: effect of natural convection in the liquid zone`; `numerical modeling of boiling in porous media using a `dual-fluid` approach: asymmetrical characteristic of the phenomenon`; `boiling during fluid flow in an induction heated porous column`; `cooling of corium fragment beds during a severe accident. State of the art and the SILFIDE experimental project`; `state of knowledge about the cooling of a particulates bed during a reactor accident`; `mass transfer analysis inside a concrete slab during fire resistance tests`; `heat transfers and boiling in porous media. Experimental analysis and modeling`; `concrete in accidental situation - influence of boundary conditions (thermal, hydric) - case studies`. (J.S.)

  16. The Fukushima accident; Accident nucleaire a Fukushima

    Energy Technology Data Exchange (ETDEWEB)

    Delbecq, D.

    2012-02-15

    The Fukushima accident is characterized by a sequence of natural disasters: earthquake and tsunamis that deprived simultaneously 3 reactors from cooling and electrical power for quite a long time. A series of hydrogen explosion has added to the mess. Experts agree to say that certainly nuclear fuel has melt to form corium in all 3 reactors. The accident has contaminated tens of thousand acres of land around the plant and has jeopardized local coastal fishery. The human toll is unexpectedly low: no direct casualty in the population but several suicides among the people that was forced to leave their home. 5 people from the plant staff died certainly from the consequences of the tsunami. (A.C.)

  17. Experimental Results on Pouring and Underwater Liquid Melt Spreading and Energetic Melt-coolant Interaction

    OpenAIRE

    Konovalenko, Alexander; Karbojian, Aram; Kudinov, Pavel

    2012-01-01

    In a hypothetical light water reactor (LWR) core-melt accident with corium release from the reactor  vessel,  the  ultimate  containment  integrity  is  contingent  on  coolability  of  the decay-heated core debris. Pouring of melt into a pool of water located in the reactor cavity is considered in several designs of existing and new LWRs  as a part of severe accident (SA) management strategies. At certain conditions of melt release into the pool (e.g. large ratio of the  vessel  breach  size...

  18. Integrated hydrogen control solutions for severe accidents using passive autocatalytic recombiners

    Energy Technology Data Exchange (ETDEWEB)

    Bauer, M.; Tietsch, W.; Sabate Farnos, R.

    2012-07-01

    In a severe accident or a beyond-design-basis-accident, the reaction of water with zirconium alloy cladding, radiolysis of water, corium-concrete reactions and other corrosion phenomena generate hydrogen (H2). The detonation of this H2 in containment or in auxiliary buildings can result in damage to structures or loss of containment integrity. Identifying the generation and special distribution of hydrogen and controlling its concentration with Passive Autocatalytic Recombiners (PARs) solves this concern. Westinghouse's approach for hydrogen management starts by defining the quantities and transport/distribution of H{sub 2} in-containment and out of containment with analysis tools such as MAAP, MELCOR, GASFLOW or FATE. Based on the results of these analyses, an optimized H2 Control Strategy is proposed in terms of number and location of PARs, and efficient integration with other H{sub 2} management devices like e.g. existing igniters, H{sub 2} monitors, etc.

  19. Modeling in fast dynamics of accidents in the primary circuit of PWR type reactors; Modelisation en dynamique rapide d'accidents dans le circuit primaire des reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F

    2003-12-01

    Two kinds of accidents, liable to occur in the primary circuit of a Pressurized Water Reactor and involving fast dynamic phenomena, are analyzed. The Loss Of Coolant Accident (LOCA) is the accident used to define the current PWR. It consists in a large-size break located in a pipe of the primary circuit. A blowdown wave propagates through the circuit. The pressure differences between the different zones of the reactor induce high stresses in the structures of the lower head and may degrade the reactor core. The primary circuit starts emptying from the break opening. Pressure decreases very quickly, involving a large steaming. Two thermal-hydraulic simulations of the blowdown phase of a LOCA are computed with the Europlexus code. The primary circuit is represented by a pipe-model including the hydraulic peculiarities of the circuit. The main differences between both computations concern the kind of reactor, the break location and model, and the initialization of the accidental operation. Steam explosion is a hypothetical severe accident liable to happen after a core melting. The molten part of the core (called corium) falls in the lower part of the reactor. The interaction between the hot corium and the cold water remaining at the bottom of the vessel induces a massive and violent vaporization of water, similar to an explosive phenomenon. A shock wave propagates in the vessel. what can damage seriously the neighbouring structures or drill the vessel. This work presents a synthesis of in-vessel parametrical studies carried out with the Europlexus code, the linkage of the thermal-hydraulic code Mc3d dedicated to the pre-mixing phase with the Europlexus code dealing with the explosion, and finally a benchmark between the Cigalon and Europlexus codes relative to the Vulcano mock-up. (author)

  20. A state of the art on penetration failure estimation under external vessel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Min, B. T.; Park, R. J.; Kang, K. H.; Cho, Y. R.; Kim, J. W.; Kim, S. B.; Park, S. Y.; Lee, K. Y

    2000-04-01

    A state of the art on penetration failure was reviewed and analyzed to establish the direction of the experimental program in the KNGR and to decide the test section design. The interaction between the corium and the reactor vessel and the corium behavior in the lower plenum of the reactor vessel were analyzed to investigate the penetration effect on severe accident progression, and the TMI-2 accident was investigated in the point of penetration failure. Theoretical model and experiment results on penetration failure under the severe accident were investigated and reviewed to establish the direction of the experimental program on the estimation of the penetration failure in the KNGR. These results were compared with the TMI-2 results. The existing test facilities on penetration failure were investigated and reviewed to decide the test section design. It can be said from the state of the art review that penetration in the lower plenum of the reactor vessel is a week point in the reactor vessel failure under the severe accident, but the reactor vessel may not be failed by penetration failure in condition with the coolant supply to the penetration. Since the penetration is different with reactor types and there is no study on estimation of the penetration welding, it is necessary to investigate failure or not of the penetration in condition with external vessel cooling to maintain the reactor vessel integrity in KNGR. In the present experimental program on the integrity estimation of the KNGR penetration, the aluminum oxide melt by thermite reaction and the test section with one penetration of the real size and real material were selected. The melt mass, the pressure of the system, and the vessel geometry were selected as an experimental parameter. (author)

  1. Experimental study on two-phase flow natural circulation in a core catcher cooling channel for EU-APR1400 using air-water system

    Energy Technology Data Exchange (ETDEWEB)

    Song, Ki Won [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); Korea Atomic Energy Research Institute, Daejeon 34057 (Korea, Republic of); Nguyen, Thanh Hung [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47906 (United States); Ha, Kwang Soon; Kim, Hwan Yeol; Song, Jinho [Korea Atomic Energy Research Institute, Daejeon 34057 (Korea, Republic of); Park, Hyun Sun [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); Revankar, Shripad T., E-mail: shripad@postech.ac.kr [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); School of Nuclear Engineering, Purdue University, West Lafayette, IN 47906 (United States); Kim, Moo Hwan [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); Korea Institute of Nuclear Safety, Daejeon 305-338 (Korea, Republic of)

    2017-05-15

    Highlights: • Two-phase flow regimes and transition behavior were observed in the coolant channel. • Test were conducted for natural circulation with air-water. • Data were obtained on flow regime, void fraction, flow rates and re-wetting time. • The data were related to a cooling capability of core catcher system. - Abstract: Ex-vessel core catcher cooling system driven by natural circulation is designed using a full scaled air-water system. A transparent half symmetric section of a core catcher coolant channel of a pressurized water reactor was designed with instrumentations for local void fraction measurement and flow visualization. Two designs of air-water top separator water tanks are studied including one with modified ‘super-step’ design which prevents gas entrainment into down-comer. In the experiment air flow rates are set corresponding to steam generation rate for given corium decay power. Measurements of natural circulation flow rate, spatial local void fraction distribution and re-wetting time near the top wall are carried out for various air flow rates which simulate boiling-induced vapor generation. Since heat transfer and critical heat flux are strongly dependent on the water mass flow rate and development of two-phase flow on the heated wall, knowledge of two-phase flow characteristics in the coolant channel is essential. Results on flow visualization showing two phase flow structure specifically near the high void accumulation regions, local void profiles, rewetting time, and natural circulation flow rate are presented for various air flow rates that simulate corium power levels. The data are useful in assessing the cooling capability of and safety of the core catcher system.

  2. Collagen fibril orientation in ovine and bovine leather affects strength: a small angle X-ray scattering (SAXS) study.

    Science.gov (United States)

    Basil-Jones, Melissa M; Edmonds, Richard L; Cooper, Sue M; Haverkamp, Richard G

    2011-09-28

    There is a large difference in strength between ovine and bovine leather. The structure and arrangement of fibrous collagen in leather and the relationship between collagen structure and leather strength has until now been poorly understood. Synchrotron based SAXS is used to characterize the fibrous collagen structure in a series of ovine and bovine leathers and to relate it to tear strength. SAXS gives quantitative information on the amount of fibrous collagen, the orientation (direction and spread) of the collagen microfibrils, and the d-spacing of the collagen. The amount of collagen varies through the thickness of the leather from the grain to the corium, with a greater concentration of crystalline collagen measured toward the corium side. The orientation index (OI) is correlated strongly with strength in ovine leather and between ovine and bovine leathers. Stronger leather has the fibrils arranged mostly parallel to the plane of the leather surface (high OI), while weaker leather has more out-of-plane fibrils (low OI). With the measurement taken parallel to the animal's backbone, weak (19.9 N/mm) ovine leather has an OI of 0.422 (0.033), stronger (39.5 N/mm) ovine leather has an OI of 0.452 (0.033), and bovine leather with a strength of (61.5 N/mm) has an OI of 0.493 (0.016). The d-spacing profile through leather thickness also varies according to leather strength, with little variation being detected in weak ovine leather (average=64.3 (0.5) nm), but with strong ovine leather and bovine leather (which is even stronger) exhibiting a dip in d-spacing (from 64.5 nm at the edges dropping to 62 nm in the center). This work provides a clear understanding of a nanostructural characteristic of ovine and bovine leather that leads to differences in strength.

  3. Development of Sacrificial Material for the Eu-APR1400 Core Catcher

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Jung Soo; Kim, Mun Soo; Kim, Yong Soo [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    To increase and diversify the export marker of the Korean nuclear reactor design, we developed the Eu- APR1400 reactor design based on the APR1400 reactor design, satisfying the European nuclear design requirements including the European Utility Requirements (EUR) and the Finnish requirements of YVL. As recommended by both requirements, the so called core-catcher molten core ex-vessel cooling facility was developed to manage a severe accident at the Eu-APR1400 reactor involving a core meltdown and to mitigate its consequences. Usually, sacrificial material (SM), which controls the melt properties and modifies melt conditions favorable to corium retention, can be employed to protect the core catcher body from the molten core and increase its cooling capability. The EPR reactor design (by Areva, France) core catcher consists of the initial corium retention space, the transportation channel and the wide spreading room for core melt cooling. The EPR used two kinds of SM to protect the initial core retention space from core melt and to spread the core melt across the wide spreading room using the different compositions. The VVER (Russia) ensures melt localization in a water-cooled vessel located directly beneath the reactor. SM is used to remove the thermal focusing effect by the layer inversion process between metallic and oxidic melts. The functional requirements for the SM determined for the present core catcher are (1) melting spreading improvement, (2) focusing effect prevention, (3) hydrogen explosion prevention, (4) FP (fission product) release decreasing, and (5) melt recriticality exclusion. The rest of the paper is organized as follows. The next section provides detailed descriptions of the composition of the present SM, which satisfies its functional requirements. Following this, the manufacturing process of the SM is presented

  4. Formation and reduction behaviors of zirconium oxide compounds in LiCl-Li2O melt at 923 K

    Science.gov (United States)

    Sakamura, Yoshiharu; Iizuka, Masatoshi; Kitawaki, Shinichi; Nakayoshi, Akira; Kofuji, Hirohide

    2015-11-01

    The reduction behaviors of ZrO2, Li2ZrO3 and (U,Pu,Zr)O2 in a LiCl-Li2O salt bath at 923 K were investigated. This study was conducted as part of a feasibility study on the pyrochemical treatment of damaged fuel debris generated by severe accidents at light water reactors. It was demonstrated in electrolytic reduction tests that the uranium in synthetic corium specimens of (U,Pu,Zr)O2 with various ZrO2 contents could be reduced to the metallic form and that part of the zirconium was converted to Li2ZrO3. Zirconium metal and Li2ZrO3 were obtained by the reduction of ZrO2. The reduction of Li2ZrO3 did not proceed even in LiCl containing no Li2O. Moreover, the stable chemical forms of the ZrO2-Li2O complex oxide were investigated as a function of the Li2O concentration in LiCl. ZrO2 was converted to Li2ZrO3 at a Li2O concentration of 0.018 wt%. As the Li2O concentration was increased, Li2ZrO3 was converted to Li6Zr2O7 and then to Li8ZrO6. It is suggested that the removal of Li2ZrO3 from the reduction product is a key point in the pyrochemical treatment of corium.

  5. A state of the art on penetration failure estimation under external vessel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Min, B. T.; Park, R. J.; Kang, K. H.; Cho, Y. R.; Kim, J. W.; Kim, S. B.; Park, S. Y.; Lee, K. Y

    2000-04-01

    A state of the art on penetration failure was reviewed and analyzed to establish the direction of the experimental program in the KNGR and to decide the test section design. The interaction between the corium and the reactor vessel and the corium behavior in the lower plenum of the reactor vessel were analyzed to investigate the penetration effect on severe accident progression, and the TMI-2 accident was investigated in the point of penetration failure. Theoretical model and experiment results on penetration failure under the severe accident were investigated and reviewed to establish the direction of the experimental program on the estimation of the penetration failure in the KNGR. These results were compared with the TMI-2 results. The existing test facilities on penetration failure were investigated and reviewed to decide the test section design. It can be said from the state of the art review that penetration in the lower plenum of the reactor vessel is a week point in the reactor vessel failure under the severe accident, but the reactor vessel may not be failed by penetration failure in condition with the coolant supply to the penetration. Since the penetration is different with reactor types and there is no study on estimation of the penetration welding, it is necessary to investigate failure or not of the penetration in condition with external vessel cooling to maintain the reactor vessel integrity in KNGR. In the present experimental program on the integrity estimation of the KNGR penetration, the aluminum oxide melt by thermite reaction and the test section with one penetration of the real size and real material were selected. The melt mass, the pressure of the system, and the vessel geometry were selected as an experimental parameter. (author)

  6. Effect of the in- and ex-vessel dual cooling on the retention of an internally heated melt pool in a hemispherical vessel

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, K.I.; Kim, B.S.; Kim, D.H. [Korea Atomic Energy Research Inst., Thermal Hydraulic Safety Research, Taejon (Korea, Republic of)

    2001-07-01

    A concept of in-vessel melt retention (IVMR) by in-vessel reflooding and/or reactor cavity flooding has been considered as one of severe accident management strategies and intensive researches to be performed worldwide. This paper provides some results of analytical investigations on the effect of both in- / ex-vessel cooling on the retention of an internally heated molten pool confined in a hemispherical vessel and the related thermal behavior of the vessel wall. For the present analysis, a scale-down reactor vessel for the KSNP reactor design of 1000 MWe (a large dry PWR) is utilized for a reactor vessel. Aluminum oxide melt simulant is also utilized for a real corium pool. An internal power density in the molten pool is determined by a simple scaling analysis that equates the heat flux on the the scale-down vessel wall to that estimated from KSNP. Well-known temperature-dependent boiling heat transfer curves are applied to the in- and ex-vessel cooling boundaries and radiative heat transfer has been only considered in the case of dry in-vessel. MELTPOOL, which is a computational fluid dynamics (CFD) code developed at KAERI, is applied to obtain the time-varying heat flux distribution from a molten pool and the vessel wall temperature distributions with angular positions along the vessel wall. In order to gain further insights on the effectiveness of in- and ex-vessel dual cooling on the in-vessel corium retention, four different boundary conditions has been considered: no water inside the vessel without ex-vessel cooling, water inside the vessel without ex-vessel cooling, no water inside the vessel with ex-vessel cooling, and water inside the vessel with ex-vessel cooling. (authors)

  7. KROTOS FCI experimental programme at CEA Cadarache: new features and status

    Energy Technology Data Exchange (ETDEWEB)

    Bonnet, J.M.; Bullado, Y.; Journeau, C.; Fouquart, P.; Piluso, P.; Sergeant, C.; Magallon, D. [CEA-Cadarache, DTN/STRI/LMA, Bat 708, BP1, 13108 Saint Paul lez Durance cedex (France)

    2005-07-01

    Full text of publication follows: KROTOS facility has been operated by the European Commission at JRC-Ispra for many years until 1999 as part of the FARO/KROTOS programme. The programme had to be stopped at JRC due to new EC priorities, and an agreement was concluded with CEA to transfer the KROTOS facility and know-how to Cadarache to continue investigating the pending FCI issues. By this agreement, CEA became owner of the facility. The main objective of the KROTOS programme at CEA Cadarache is understanding the role of melt properties in steam explosion propagation and energetics and, in particular, steam explosion behaviour of prototypical corium melts. Possible influence of physicochemical processes on explosiveness will also be explored. Actually, analysis of alumina debris produced at Ispra have shown that formation of metastable phases and chemical reaction with water at high temperature may play a significant role in enhancing heat transfer to water in the explosion phase. In order to reach the objective, steam explosion experiments are performed in well characterised conditions for a large spectrum of conditions and melt compositions of interest for both in- and ex-vessel situations. A trigger is applied as a rule. Advanced technology and instrumentation is used to reduce uncertainties on initial conditions and characterise the various phases of an explosion, with emphasis on high energy X-ray cinematography to qualify pre-mixing. This advanced instrumentation will enable the measurement of detailed variables to consolidate the qualification of the FCI codes. The use of high energy X-rays made it necessary to construct a new building to house the facility and its components. The facility should newly become operative early 2005. Main improvements with respect to Ispra concern melt delivery, hydrogen measurement and X-ray imaging. Preliminary studies have demonstrated that identification of the pre-mixture-water interface and coherent melt jet core, and

  8. Transient refractory material dissolution by a volumetrically-heated melt

    Energy Technology Data Exchange (ETDEWEB)

    Seiler, Jean Marie, E-mail: jean-marie.seiler@cea.fr [CEA, DEN, DTN, 17 Rue des Martyrs, 38054 Grenoble Cedex 9 (France); Ratel, Gilles [CEA, DEN, DTN, 17 Rue des Martyrs, 38054 Grenoble Cedex 9 (France); Combeau, Hervé [Institut Jean Lamour, UMR 7198, Lorraine University, Ecole des Mines de Nancy, Parc de Saurupt, 54042 Nancy Cedex (France); Gaus-Liu, Xiaoyang; Kretzschmar, Frank; Miassoedov, Alexei [Karlsruhe Institut of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2014-12-15

    , the liquid composition at the interface is concentrated in the refractory species. During the transient, the interface temperature is equal to the liquidus temperature corresponding to the local and instantaneous composition of the liquid at the interface. Regarding the design of a protective layer made of refractory materials, we can answer the question of how much ceramic can be dissolved and its impact on melt temperature evolution during the dissolution process. It also impacts on subsequent corium solidification since the additional mass of dissolved ceramic leads to increased volume of the molten material, significantly increasing the time required for complete solidification. For the long term, ceramic material does not offer better confinement than a crust made of solidified corium. This work served as support to a generalisation of the model of transient evolution of interface temperature in various severe accident situations (Seiler and Combeau, 2014)

  9. MELTSPREAD-1 calculations of the transient spreading of core materials in the KNGR

    Energy Technology Data Exchange (ETDEWEB)

    Song, Yong Mann; Park, Soo Yong

    1999-03-01

    Major purpose of the report is to predict whether or not the melt will spread to cover the full floor area under severe accident conditions in KNGR (Korea Next Generation Reactor) cavity and to determine the local distribution of spread material depth as well as concrete attack upon the cavity floor. For this analysis, MELTSPREAD-1 computer code developed at ANL (Argonne National Laboratory) is first applied domestically. This code was originally developed to model the discharge of corium from the vessel and its spreading on the floor of a prototypical BWR Mark 1 containment, but is known to have the flexibility of LWR application via user-specified nodalization scheme. The analysis methodology in this report is assessed to be valid by independent ABB-CE review [ABB-CE 1988]. For the conservative analysis of melt spreading and erosion characteristics, a medium LOCA (loss of coolant accident) as the typical in-vessel low pressure accident, with 100% core mass release into the wet cavity is chosen as the basic sequence. For specified conditions of release from the failed reactor vessel lower head, the core materials are calculated to spread within a very short time and cover the full accessible cavity floor area. The spreading profiles are shown as a scene view according to time with detailed predictions of the extent of local melting-induced erosion of the concrete floor. The MELTSPREAD-1 results are important to the assessment of melt coolability following the transients spreading phase, and the results of the basic LOCA sequence can serve as the bounding calculation in the melt spreading and ablation for the KNGR cavity. In addition to this, sensitivity studies are made for important factors and crust formation and heat transfer models together with initial cavity condition and initial corium mass/temperature are appeared to be significant for the results. For the last, both MELTSPREAD-1 code input deck and calculation note used for the sequence analysis are

  10. Detailed evaluation of two phase natural circulation flow in the cooling channel of the ex-vessel core catcher for EU-APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae-Joon, E-mail: rjpark@kaeri.re.kr; Ha, Kwang-Soon; Rhee, Bo-Wook; Kim, Hwan Yeol

    2016-03-15

    Highlights: • Ex-vessel core catcher of PECS is installed in EU-APR1400. • CE-PECS has been conducted to test a cooling capability of the PECS. • Two phase flow in CE-PECS and PECS was analyzed using RELAP5/MOD3. • RELAP5 results are very similar to the CE-PECS data. • The super-step design is suitable for steam injection into the downcomer in PECS. - Abstract: The ex-vessel core catcher of the PECS (Passive Ex-vessel corium retaining and Cooling System) is installed to retain and cool down the corium in the reactor cavity of the EU (European Union)-APR (Advanced Power Reactor) 1400. A verification experiment on the cooling capability of the PECS has been conducted in the CE (Cooling Experiment)-PECS. Simulations of a two-phase natural circulation flow using the RELAP5/MOD3 computer code in the CE-PECS and PECS have been conducted to predict the two-phase flow characteristics, to determine the natural circulation mass flow rate in the cooling channel, and to evaluate the scaling in the experimental design of the CE-PECS. Particularly from a comparative study of the prototype PECS and the scaled test facility of the CE-PECS, the orifice loss coefficient in the CE-PECS was found to be 6 to maintain the coolant circulation mass flux, which is approximately 273.1 kg/m{sup 2} s. The RELAP5 results on the coolant circulation mass flow rate are very similar to the CE-PECS experimental results. An increase in the coolant injection temperature and the heat flux lead to an increase in the coolant circulation mass flow rate. In the base case simulation, a lot of vapor was injected into the downcomer, which leads to an instability of the two-phase natural circulation flow. A super-step design at a downcomer inlet is suitable to prevent vapor injection into the downcomer piping.

  11. Experimental investigation on melt coolability under bottom flooding with and without decay heat simulation

    Energy Technology Data Exchange (ETDEWEB)

    Singh, Nitendra [Homi Bhabha National Institute, Anushakti Nagar, Mumbai 400094, Maharashtra (India); Kulkarni, Parimal P. [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai 400085, Maharashtra (India); Nayak, Arun K., E-mail: arunths@barc.gov.in [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai 400085, Maharashtra (India)

    2015-04-15

    Highlights: • The effect of decay heat on melt coolability under bottom flooding was studied. • Decay heat of 0.5 MW/m{sup 3} was simulated. • A single simulant material with same mass and initial temperature was used. • Quenching of melt pool does not depend on decay heat. • Comparison of with and without decay heat experiments has been presented. - Abstract: Investigations on severe accident phenomena help us in understanding the realistic accidental phenomena for the assessment of associated risk. The societal impact of radiological leakage to the environment has demanded further robustness in the line of defence of nuclear safety. Thus, to ensure the cooling and stabilization of corium within reactor containment in case of severe accident scenarios, many new reactors have been envisaged with core catcher. In this regard, corium coolability still remains an unresolved issue in spite of several efforts being taken towards its understanding. After studying the various cooling strategies, it has been demonstrated that melt coolability using bottom flooding of water is one of the most efficient techniques so far. To study the effect of decay heat on melt pool coolability under bottom flooding condition, two experiments have been performed in this paper; one without the decay heat and the other with decay heat. The test section used for carrying out these experiments consisted of two parts viz. lower part for retaining the melt from furnace, water inlet and melt quenching, and upper part for steam expansion and its outlet. The total height of the test section was 1400 mm and was made of 33 mm thick carbon steel. Total six stainless steel nozzles of diameter 12 mm were used for injecting water at the bottom of the melt pool. The lower part was surrounded by 10 radiative heaters to simulate decay heat of 10 kW which corresponds to 0.5 MW/m{sup 3}. The experiments showed that quenching of about 25 l of melt at initial temperature of nearly 1200 °C took only a

  12. Status of the FARO/KROTOS melt-coolant interactions tests

    Energy Technology Data Exchange (ETDEWEB)

    Magallon, D.; Huhtiniemi, I.; Annunziato, A.; Yerkess, A.; Hohmann, Y.H. [European Commission, Joint Research Centre Safety Technology Institute, Ispra (Austria)

    1996-03-01

    Results of FARO test L-19 are reported. It involved 155 kg of 80 w% UO{sub 2}+20 w% ZrO{sub 2} at 3073 K quenched in 338-kg, 1-m-depth water at saturation at 5.0 MPa (i.e., 537 K). The test is compared with two former tests (L-06 and L-08) performed in similar conditions (1-m-depth water) but with reduced quantities of test (18 and 44 kg, respectively) and with test L-14, performed with a similar quantity of melt (125 kg) but in 2-m-depth water. No fundamental differences with the former tests have been observed. Particularly, the quenching rate per unit melt mass was of the same order (0.5 MW). On the contrary, the portion of melt which remained as a cake on the bottom was larger (50% against a maximum of 30% in the previous tests). The possible reasons for these discrepancies are discussed. Recalculations of test L-19 by using COMETA and TEXAS are also reported and commented. Results from a new set of KROTOS tests conducted with Al{sub 2}O{sub 2} to investigate further the differences already observed with corium melt are presented and discussed. In these tests the effect of men superheat, water subcooling and ambient pressure on Al{sub 2}O{sub 3}/water system behaviour have been tested. In contrast with corium experiments, the results demonstrated that spontaneous explosions occur in the Al{sub 2}O{sub 3}/water system over all the range of parameters tested in highly subcooled conditions. Some TEXAS results for the latest KROTOS Al{sub 2}O{sub 3} test, in which a violent steam explosion did occur, are presented and compared with experiment. During some KROTOS experiments there are large deformations of the bottom plate and hold-down bolts. Use is made of the 2-D axisymmetric code SEURBNUK-EURDYN to analyse these deformations of the test section and some results are presented.

  13. Post test calculations of a severe accident experiment for VVER-440 reactors by the ATHLET code

    Energy Technology Data Exchange (ETDEWEB)

    Gyoergy, Hunor [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques (BME NTI); Trosztel, Istvan [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research (MTA EK)

    2013-09-15

    Severe accident - if no mitigation action is taken - leads to core melt. An effective severe accident management strategy can be the external reactor pressure vessel cooling for corium localization and stabilization. For some time discussion was going on, whether the in-vessel retention can be applied for the VVER-440 type reactors. It had to be demonstrated that the available space between the reactor vessel and biological protection allows sufficient cooling to keep the melted core in the vessel, without the reactor pressure vessel losing its integrity. In order to demonstrate the feasibility of the concept an experimental facility was realized in Hungary. The facility called Cooling Effectiveness on the Reactor External Surface (CERES) is modeling the vessel external surface and the biological protection of Paks NPP. A model of the CERES facility for the ATHLET TH system code was developed. The results of the ATHLET calculation agree well with the measurements showing that the vessel cooling can be insured for a long time in a VVER-440 reactor. (orig.)

  14. Effectiveness of Cryogen Tetrfluoroethane on Elimination of Gingival Epithelium and its Clinical Application in Gingival Depigmentation–Histological Findings and Case Series

    Science.gov (United States)

    Kumar, Santhosh; Bhat, G. Subraya; Bhat, K. Mahalinga

    2013-01-01

    Objective: To histologically assess and clinically co-relate the effectiveness of cryogen Tetrafluoroethane (TFE) for gingival depigmentation procedure. Material and Methods: Twelve patients having unaesthetic gingival melanin pigmentation were included in the study. Gingival tissues of eight patients having gingival melanin pigmentation undergoing gingivoplasty or gingivectomy for crownlengthening were exposed to the cryogen and this was used for the histological examination. Gingivectomies were done after 8, 24, 96 hours and after a week of application of tetrafluoroethane. Four fair skinned patients complaining of unaesthetic gingival hyperpigmentation underwent gingival depigmentation using Tetrafluoroethane cryogen. Results: Histologically after 96 hours of application of cryogen there was complete loss of retepegs and epithelial detachment from the corium was evident. Complete re – epithelialisation was noted after a week and was clinically correlated. Conclusion: We therefore, concluded that histologically tetrafluoroethane can effectively destroy gingival epithelium without causing damage to the connective tissue and clinically the color of the gingiva had more pleasing appearance 6 months postoperatively. Hence the cryogen can be used safely for depigmentation procedure. PMID:24551730

  15. Development of Integrated Evaluation System for Severe Accident Management

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Ha; Kim, K. R.; Park, S. H.; Park, S. Y.; Park, J. H.; Song, Y. M.; Ahn, K. I.; Choi, Y

    2007-06-15

    The objective of the project is twofold. One is to develop a severe accident database (DB) for the Korean Standard Nuclear Power plant (OPR-1000) and a DB management system, and the other to develop a localized computer code, MIDAS (Multi-purpose IntegrateD Assessment code for Severe accidents). The MELCOR DB has been constructed for the typical representative sequences to support the previous MAAP DB in the previous phase. The MAAP DB has been updated using the recent version of MAAP 4.0.6. The DB management system, SARD, has been upgraded to manage the MELCOR DB in addition to the MAAP DB and the network environment has been constructed for many users to access the SARD simultaneously. The integrated MIDAS 1.0 has been validated after completion of package-wise validation. As the current version of MIDAS cannot simulate the anticipated transient without scram (ATWS) sequence, point-kinetics model has been implemented. Also the gap cooling phenomena after corium relocation into the RPV can be modeled by the user as an input parameter. In addition, the subsystems of the severe accident graphic simulator are complemented for the efficient severe accident management and the engine of the graphic simulator was replaced by the MIDAS instead of the MELCOR code. For the user's convenience, MIDAS input and output processors are upgraded by enhancing the interfacial programs.

  16. Validation of the THIRMAL-1 melt-water interaction code

    Energy Technology Data Exchange (ETDEWEB)

    Chu, C.C.; Sienicki, J.J.; Spencer, B.W. [Argonne National Lab., IL (United States)

    1995-09-01

    The THIRMAL-1 computer code has been used to calculate nonexplosive LWR melt-water interactions both in-vessel and ex-vessel. To support the application of the code and enhance its acceptability, THIRMAL-1 has been compared with available data from two of the ongoing FARO experiments at Ispra and two of the Corium Coolant Mixing (CCM) experiments performed at Argonne. THIRMAL-1 calculations for the FARO Scoping Test and Quenching Test 2 as well as the CCM-5 and -6 experiments were found to be in excellent agreement with the experiment results. This lends confidence to the modeling that has been incorporated in the code describing melt stream breakup due to the growth of both Kelvin-Helmholtz and large wave instabilities, the sizes of droplets formed, multiphase flow and heat transfer in the mixing zone surrounding and below the melt metallic phase. As part of the analysis of the FARO tests, a mechanistic model was developed to calculate the prefragmentation as it may have occurred when melt relocated from the release vessel to the water surface and the model was compared with the relevant data from FARO.

  17. Chemical Technology Division annual technical report, 1992

    Energy Technology Data Exchange (ETDEWEB)

    Battles, J.E.; Myles, K.M.; Laidler, J.J.; Green, D.W.

    1993-06-01

    In this period, CMT conducted research and development in the following areas: (1) electrochemical technology, including advanced batteries and fuel cells; (2) technology for fluidized-bed combustion and coal-fired magnetohydrodynamics; (3) methods for treatment of hazardous waste, mixed hazardous/radioactive waste, and municipal solid waste; (4) the reaction of nuclear waste glass and spent fuel under conditions expected for an unsaturated repository; (5) processes for separating and recovering transuranic elements from nuclear waste streams, treating water contaminated with volatile organics, and concentrating radioactive waste streams; (6) recovery processes for discharged fuel and the uranium blanket in the Integral Fast Reactor (EFR); (7) processes for removal of actinides in spent fuel from commercial water-cooled nuclear reactors and burnup in IFRs; and (8) physical chemistry of selected materials (corium; Fe-U-Zr, tritium in LiAlO{sub 2} in environments simulating those of fission and fusion energy systems. The Division also conducts basic research in catalytic chemistry associated with molecular energy resources and novel` ceramic precursors; materials chemistry of superconducting oxides, electrified metal/solution interfaces, and molecular sieve structures; and the geochemical processes involved in water-rock interactions occurring in active hydrothermal systems. In addition, the Analytical Chemistry Laboratory in CMT provides a broad range of analytical chemistry support services to the technical programs at Argonne National Laboratory (ANL).

  18. The anatomy and physiology of the suspensory apparatus of the distal phalanx.

    Science.gov (United States)

    Pollitt, Christopher C

    2010-04-01

    The equine hoof capsule protects the softer, more sensitive, structures within. Failure of the connection between hoof and bone (suspensory apparatus of the distal phalanx or SADP) results in the crippling lameness of laminitis. Active basal cell proliferation occurs principally in tubular hoof and proximal and distal lamellae. The remaining lamellae are virtually non-proliferative and the hoof wall moves past the stationary distal phalanx, by controlled activation and inhibition of constituent proteases. The lamellar corium derives most of its blood supply from the branches of the terminal arch which perforate the distal phalanx. Valveless veins within the foot can be exploited clinically for retrograde venous therapy or contrast radiography (venography). The basement membrane (BM) forms the interface between the lamellar epidermis and the adjacent dermis and the plasma membrane of each lamellar basal cell is attached to the BM by numerous electron dense adhesion plaques or hemidesmosomes the ultimate attachment unit of the SADP. Laminitis destroys and dislocates the BM and its components and without an intact, functional BM, the structure and function of the lamellar epidermis is pathologically compromised. Transcription and activation of constituent proteases occurs in normal hoof lamellae but in increased amounts during laminitis.

  19. Experiments to investigate direct containment heating phenomena with scaled models of the Zion Nuclear Power Plant in the Surtsey Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Allen, M.D.; Pilch, M.M.; Blanchat, T.K.; Griffith, R.O. [Sandia National Labs., Albuquerque, NM (United States); Nichols, R.T. [Ktech Corp., Albuquerque, NM (United States)

    1994-05-01

    The Surtsey Facility at Sandia National Laboratories (SNL) is used to perform scaled experiments that simulate hypothetical high-pressure melt ejection (HPME) accidents in a nuclear power plant (NPP). These experiments are designed to investigate the effect of specific phenomena associated with direct containment heating (DCH) on the containment load, such as the effect of physical scale, prototypic subcompartment structures, water in the cavity, and hydrogen generation and combustion. In the Integral Effects Test (IET) series, 1:10 linear scale models of the Zion NPP structures were constructed in the Surtsey vessel. The RPV was modeled with a steel pressure vessel that had a hemispherical bottom head, which had a 4-cm hole in the bottom head that simulated the final ablated hole that would be formed by ejection of an instrument guide tube in a severe NPP accident. Iron/alumina/chromium thermite was used to simulate molten corium that would accumulate on the bottom head of an actual RPV. The chemically reactive melt simulant was ejected by high-pressure steam from the RPV model into the scaled reactor cavity. Debris was then entrained through the instrument tunnel into the subcompartment structures and the upper dome of the simulated reactor containment building. The results of the IET experiments are given in this report.

  20. Flow boiling CHF enhancement in an external reactor vessel cooling (ERVC) channel using graphene oxide nanofluid

    Energy Technology Data Exchange (ETDEWEB)

    Park, Seong Dae; Bang, In Cheol, E-mail: icbang@unist.ac.kr

    2013-12-15

    Highlights: • We investigate CHF limits of graphene oxide nanofluid for IVR-ERVC. • Graphene oxide nanofluid enhanced CHF up to about 20%. • CHF enhancement can be explained by the improved thermal activity. - Abstract: External reactor vessel cooling for in-vessel retention of corium is an important concept to mitigate the consequences of a severe accident by flooding the reactor cavity. Although this system has some merits, it is restricted by the capacity of heat removal through the nucleate boiling on the outer surface of the reactor. In this study, the graphene oxide (GO) nanofluid at 0.0001 vol% was used to enhance the critical heat flux (CHF). The CHF tests were conducted with a closed-loop facility. Test section simulated the reactor vessel of APR-1400 with a small scale. The test results show about ∼20% enhancement of CHF at 50 and 100 kg/m{sup 2} s under a 10 K subcooling condition. It means that the additional thermal margin could be acquired by just adding the GO nanoparticles to the flooding water without severe economic concerns. It is also found that this CHF enhancement is caused by coating the graphene oxide nanoparticles on the heated surface. However, the sessile drop tests on the coated heater surface show that the wettability of GO coated surface is not improved. The results of IR thermography show that one of the promising reasons is the change of thermal activity due to the coated GO nanoparticles on the heated surface.

  1. Therapeutic Effects of Fermented Flax Seed Oil on NC/Nga Mice with Atopic Dermatitis-Like Skin Lesions

    Science.gov (United States)

    Yang, Joonhyoung; Min, Sangyeon

    2017-01-01

    Background. Atopic Dermatitis (AD) is one of the most common chronic inflammatory skin diseases. Objective. This experiment aimed to study the effects of Fermented Flax Seed Oil (FFSO) on symptoms such as redness, eczema, and pruritus induced by AD. Materials and Methods. AD-induced NC/Nga mice were used to observe the immunological and therapeutic effects of FFSO on skin in vivo. Raw 264.7 cells were used to investigate the effects of FFSO in cells. Fc receptor expression and concentration of beta-hexosaminidase were measured. Nitric oxide assay, Western blotting, real-time PCR, image analysis, and statistical analysis were performed in vitro. Results. In the immunohistochemical results, p-ERK 1/2 expression decreased, fibrogenesis strongly increased, and distribution reduction is observed. Distribution of IL-4-positive cells in the corium near the basal portion of the epithelium in the AT group was reduced. FFSO treatment reduced the number of cells showing NF-κB p65 and iNOS expression. The level of LXR in the AT group was higher than that in the AE group, and elevation of PKC expression was significantly reduced by FFSO treatment. Conclusion. FFSO could alleviate symptoms of AD such as epithelial damage, redness, swelling, and pruritus.

  2. Research on the improvement of nuclear safety -The development of a severe accident analysis code-

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Heui Dong; Cho, Sung Won; Park, Jong Hwa; Hong, Sung Wan; Yoo, Dong Han; Hwang, Moon Kyoo; Noh, Kee Man; Song, Yong Man [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    For prevention and mitigation of the containment failure during severe accident, the study is focused on the severe accident phenomena, especially, the ones occurring inside the cavity and is intended to improve existing models and develop analytical tools for the assessment of severe accidents. A correlation equation of the flame velocity of pre mixture gas of H{sub 2}/air/steam has been suggested and combustion flame characteristic was analyzed using a developed computer code. For the analysis of the expansion phase of vapor explosion, the mechanical model has been developed. The development of a debris entrainment model in a reactor cavity with captured volume has been continued to review and examine the limitation and deficiencies of the existing models. Pre-test calculation was performed to support the severe accident experiment for molten corium concrete interaction study and the crust formation process and heat transfer characteristics of the crust have been carried out. A stress analysis code was developed using finite element method for the reactor vessel lower head failure analysis. Through international program of PHEBUS-FP and participation in the software development, the research on the core degradation process and fission products release and transportation are undergoing. CONTAIN and MELCOR codes were continuously updated under the cooperation with USNRC and French developed computer codes such as ICARE2, ESCADRE, SOPHAEROS were also installed into the SUN workstation. 204 figs, 61 tabs, 87 refs. (Author).

  3. Experimental investigation of heat transfer during severe accident of a Pressurized Heavy Water Reactor with simulated decay heat generation in molten pool inside calandria vessel

    Energy Technology Data Exchange (ETDEWEB)

    Prasad, Sumit Vishnu, E-mail: svprasad@barc.gov.in; Nayak, Arun Kumar, E-mail: arunths@barc.gov.in

    2016-07-15

    Highlights: • Scaled test facility simulating the calandria vessel and calandria vault water of PHWR with simulated decay heat was built. • Experiments conducted with simulant material at about 1200 °C. • Experimental result shows that melt coolability and growth rate of crust thickness are affected by presence of decay heat. • No gap was observed between the crust and vessel on opening. • Result shows that vessel integrity is intact with presence of water inside water tank in both cases. - Abstract: The present study focuses on experimental investigation in a scaled facility of an Indian PHWR to investigate the coolability of molten corium with simulated decay heat in the simulated calandria vessel. Molten borosilicate glass was used as the simulant due to its comparable heat transfer characteristics similar to prototypic material. About 60 kg of the molten material was poured into the test section at about 1200 °C. Decay heat in the melt pool was simulated using four high watt heaters cartridges, each having 9.2 kW. The temperature distributions inside the molten pool, across the vessel wall thickness and vault water were measured. Experimental results obtained are compared with the results obtained previously for no decay heat case. The results indicated that presence of decay heat seriously affects the coolability behaviour and formation of crust in the melt pool. The location and magnitude of maximum heat flux and surface temperature of the vessel also are affected in the presence of decay heat.

  4. 1996 scientific report; Rapport scientifique 1996

    Energy Technology Data Exchange (ETDEWEB)

    Berhouet, F.; Delavault, E.; Dupuy, P.; Legre, J.; Comte, M.; Panek, H.; Martin-Daguet, V.; Fleche, J.L.; Meis, C.; Kupecek, Ph.; Mennerat, G.; Mauchien, P.; Chaleard, C.; Kocon, S.; Lacour, J.L.; Gueneau, C.; Dauvois, V.; Gonella, C.; Berhier, C.; Lameille, J.M. [CEA Saclay, 91 - Gif-sur-Yvette (France). Dept. des Procedes d`enrichissement] [and others

    1997-12-31

    This report presents the work realized in 1996 by CEA-DCC (Direction du Cycle du Combustible). The main study fields are: laser beam stabilization, copper vapor laser (LVC) modelling, vapor density measurements, light transport in thick vapor, uranium evaporation, GdR-PRACTIS activities, uranyl organo phosphates, actinide extraction, ultra son in chemistry, oxidation mechanisms, salt free processes, tributyl phosphate (TBP) stability, cation exchanging membranes, filtering mineral membranes, liquid-liquid extraction processes, centrifugal contactor, supercritical phase, iodine reduction, weapon plutonium reprocessing, long term waste packaging, clay behaviour modelling; gas formation in deep storages, water-clay chemical interaction, nuclear glass behaviour, actinide storage dies, radionuclide release in storage, waste storage dies, iodine conditioning, cesium extraction by calix[4]arene molecules, decontamination foam, oxidation by supercritical water, decontamination by electrolysis, microorganism actions on metals, long life radionuclide determination in wastes, active and passive tomography, sulfated wastes, zircon-titanates ceramics, direct storage of irradiated fuel, caramel type fuel dismantling, radioprotection calculation codes, diffraction of backscattered electrons, infrared light coherent source, optical spectrometry, physical and chemical aspect of corium, iron-nickel-chromium alloy corrosion, natural uranium perspective, Atalante installation, interlaboratory comparisons. (A.C.)

  5. Desktop Severe Accident Graphic Simulator Module for CANDU6 : PSAIS

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. Y.; Song, Y. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The ISAAC ((Integrated Severe Accident Analysis Code for CANDU Plant) code is a system level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a Level 2 probabilistic safety assessment or severe accident management strategy developments. The code has the capability to predict a severe accident progression by modeling the CANDU6- specific systems and the expected physical phenomena based on the current understanding of the unique accident progressions. The code models the sequence of accident progressions from a core heatup, pressure tube/calandria tube rupture after an uncovery from inside and outside, a relocation of the damaged fuel to the bottom of the calandria, debris behavior in the calandria, corium quenching after a debris relocation from the calandria to the calandria vault and an erosion of the calandria vault concrete floor, a hydrogen burn, and a reactor building failure. Along with the thermal hydraulics, the fission product behavior is also considered in the primary system as well as in the reactor building.

  6. Noninvasive topographical investigation of functional parameters in the human skin

    Science.gov (United States)

    Kessler, Manfred D.; Krug, Alfons; Hoeper, Jens

    1996-04-01

    A rapid micro-lightguide spectrometer (EMPHO II) coupled to an automatic three axis positioning system enables very precise and fast 2D-scans at the surface of human skin. The positioning accuracy amounts to 1 micrometer. This allows measurements with excellent spatial reproducibility. With this system examinations of local distribution of HbO2 and Hb have been performed in human skin. For this purpose at the back of the hand areas of 5 by 5 mm to 5 by 10 mm were scanned in defined steps of 100 micrometers. Functional images of local hemoglobin concentration and hemoglobin oxygenation of microscopical structures have been resolved by use of 250 micrometer lightguide sensors. Two-dimensional-images of local oxygen supply parameters corresponding directly to morphological structures of human skin have been gained. The local pattern matches the distribution of the papillas of the corium. In the papillas the capillary loops supplying the lower part of the epidermis are situated. The measured parameters describe very exactly the local oxygen supply situation of the area under investigation.

  7. Anatomical mapping of the nasal muscles and application to cosmetic surgery.

    Science.gov (United States)

    Konschake, Marko; Fritsch, Helga

    2014-11-01

    We present an anatomical mapping of the most important muscles influencing the nose, incorporating constant anatomical structures, and their spatial correlations. At our disposal were the midfaces of 18 bodies of both sexes, obtained by informed consent from body donors aged between 60 and 80 years. Macroscopically, we dissected the nasal regions of eight corpses, six midfaces were prepared according to plastination histology, four by creating plastinated slices. On their way from their periosteal origin to the edge of the skin, the muscles of the nose cross the subcutaneous adipose tissue, dividing it into superficial and deep layers. The individual muscle fibers insert into the skin directly at the reticular corium. Sometimes, they reach the border of the epidermis which represents a special arrangement of corial muscle attachments. The course of the anatomical fibers of individual nasal muscles presented macroscopically and microscopically in this study offers surgeons a detailed overview of the anatomically important muscular landmarks of the midface. © 2014 Wiley Periodicals, Inc.

  8. [Epidermal aging and anti-aging strategies].

    Science.gov (United States)

    Wohlrab, J; Hilpert, K; Wolff, L

    2016-02-01

    Epithelial senescence is a complex process depending on intrinsic as well as extrinsic factors (e.g., UV or IR light, tobacco smoke) and must be seen in the context of the aging process especially of the corium and the subcutis. Morphological alterations become apparent in the form of epithelial atrophy, structural changes within the basal membrane, and a decrease in cell count of melanocytes and Langerhans cells. Signs of cellular senescence are reduced proliferation of keratinocytes, cumulation of dysplastic keratinocytes, various mutations (e.g., c-Fos/c-Jun, STAT3, FoxO1), as well as multiple lipid or amino acid metabolic aberrations (e.g., production of advanced glycation endproducts). This causes functional changes within the physical (lipid deficiency, water distribution dysfunction, lack of hygroscopic substances), chemical (pH conditions, oxygen radicals), and immunological barrier. Prophylactically, barrier-protective care products, antioxidant substances (e.g., vitamin C, B3, E, polyphenols, flavonoids), sunscreen products/measurements, and retinoids are used. For correcting alterations in aged epidermis, chemical peelings (fruit acids, β-hydroxy acid, trichloroacetic acid, phenolic compounds), non-ablative (IPL, PDL, Nd:YAG) as well as ablative (CO2, Erbium-YAG) light-assisted methods are used.

  9. Experiments on Sedimentation of Particles in a Water Pool with Gas Inflow

    Directory of Open Access Journals (Sweden)

    Eunho Kim

    2016-04-01

    Full Text Available During the late phase of severe accidents of light water reactors, a porous debris bed is expected to develop on the bottom of the flooded reactor cavity after breakup of the melt in water. The geometrical configuration, i.e., internal and external characteristics, of the debris bed is significant for the adequate assessment of the coolability of the relocated corium. The internal structure of a debris bed was investigated experimentally using the DAVINCI (Debris bed research Apparatus for Validation of the bubble-Induced Natural Convection effect Issue test facility. Particle sedimentation under the influence of a two-phase natural convection flow due to the decay heat in the debris bed was simulated by dropping various sizes of particles into a water vessel with air bubble injection from the bottom. Settled particles were collected and sieved to obtain the particle mass, size distribution in the radial and axial positions, and the bed porosity and permeability. The experimental results showed that the center part of the particle bed tended to have larger particles than the peripheral area. For the axial distribution, the lower layer had a higher fraction of larger particles. As the sedimentation progressed, the size distribution in the upper layers can shift to larger sizes because of the higher vapor generation rate and stronger flow intensity.

  10. Candeocoris bistillatus, new genus and new species of Ochlerini from Ecuador (Hemiptera: Heteroptera: Pentatomidae).

    Science.gov (United States)

    Roell, Talita; Campos, Luiz Alexandre

    2015-09-17

    Recent examination of specimens from Ecuador revealed a series of males and females of an undescribed species clearly belonging to the Discocephalinae. The new species presents characteristics similar to genera of both Discocephalini and Ochlerini, preventing an undoubtful placement of the new species within any genus and tribe. We conducted a cladistic analysis to investigate the possible relationships of the new species within Discocephalinae. The new species was recovered as sister-group to the remaining Ochlerini, supporting the proposition of a new genus, so Candeocoris bistillatus Roell & Campos, gen. n. et sp. n. are described within Ochlerini. The new genus is recognized for its dark glossy aspect, tumid vertex of head, long and sinuous labrum, base of labium placed close to anterior limit of eyes, thick bristles on meso- and metatibiae, pygophore globose, and laterotergites 9 touching each other. The new species is recognized by a large yellow spot on each corium, yellow spots on each segment of the connexivum, bucculae with anterior tooth, laminar projections on superior layer of ventral rim of pygophore, vesica with a single median projection, and broad gonocoxites 8.

  11. Computed tomographic anatomy of the equine foot.

    Science.gov (United States)

    Claerhoudt, S; Bergman, E H J; Saunders, J H

    2014-10-01

    This study describes a detailed computed tomographic reference of the normal equine foot. Ten forefeet of five adult cadavers, without evidence of orthopaedic disease, were used. Computed tomography (CT) was performed on all feet. Two-millimetre thick transverse slices were obtained, and sagittal and dorsal planes were reformatted. The CT images were matched with the corresponding anatomic slices. The phalanges and the distal sesamoid bone showed excellent detail. The extensor and flexor tendons (including their attachments) could be clearly evaluated. The collateral (sesamoidean) ligaments could be readily located, but were difficult to delineate at their proximal attachment. The distal digital annular ligament could only be distinguished from the deep digital flexor tendon proximal to the distal sesamoid bone, and its proximal attachment could be identified, but not its distal insertion. Small ligaments (impar ligament, chondrosesamoidean, chondrocoronal and chondrocompedal ligaments, axial and abaxial palmar ligaments of the proximal inter-phalangeal joint) were seen with difficulty and not at all slices. The joint capsules could not be delineated from the surrounding soft tissue structures. The lateral and medial proprius palmar digital artery and vein could be visualized occasionally on some slices. The ungular cartilages, corium and hoof wall layering were seen. The nerves, the articular and fibrocartilage of the distal sesamoid bone and the chondroungular ligament could not be assessed. Computed tomography of the equine foot can be of great value when results of radiography and ultrasonography are inconclusive. Images obtained in this study may serve as reference for CT of the equine foot.

  12. Collagen fibril alignment and deformation during tensile strain of leather: a small-angle X-ray scattering study.

    Science.gov (United States)

    Basil-Jones, Melissa M; Edmonds, Richard L; Norris, Gillian E; Haverkamp, Richard G

    2012-02-08

    The distribution and effect of applied strain on the collagen fibrils that make up leather may have an important bearing on the ultimate strength and other physical properties of the material. While sections of ovine and bovine leather were being subjected to tensile strain up to rupture, synchrotron-based small-angle X-ray scattering (SAXS) spectra were recorded edge-on to the leather at points from the corium to the grain. Measurements of both fibril orientation and collagen d spacing showed that, initially, the fibers reorient under strain, becoming more aligned. As the strain increases (5-10% strain), further fibril reorientation diminishes until, at 37% strain, the d spacing increases by up to 0.56%, indicating that significant tensile forces are being transmitted to individual fibrils. These changes, however, are not uniform through the cross-section of leather and differ between leathers of different strengths. The stresses are taken up more evenly through the leather cross-section in stronger leathers in comparison to weaker leathers, where stresses tended to be concentrated during strain. These observations contribute to our understanding of the internal strains and structural changes that take place in leather under stress.

  13. Histomorphological evaluation of the digital coronary region at different fetal development stages of Holstein cattle

    Directory of Open Access Journals (Sweden)

    R.E. Rabelo

    2015-02-01

    Full Text Available The scientific literature lacks detailed morphological descriptions of the histological development and cell differentiation of fetal bovine hoof. In this study, 40 extremity members of Holstein bovine fetuses were collected and divided into four groups (G1 to G4 based on the estimated age. Fragments were removed from wall and sole, processed and stained with hematoxylin - eosin (HE for light microscopy observation. In G1, it was found that the epidermis was very thin, including keratinocyte layers and clusters of mesenchymal cells. In group G2 it was observed that the thickness of the epidermis covering the limbs remained variable and laminar corium developed in the germinal layer. In group G3 it was noted that in the germinal epithelium there were papillae in little advanced development and cells of the stratum corneum in the initial process of keratinization. In G4, the epidermis was well developed with layers distributed homogeneously, containing symmetrical and long papillae and intense production of keratin. In this work, the most important cellular events for the formation of the fetal hoof in Holstein cattle were first described in different stages of their formation.

  14. Perilla leaf extract prevents atopic dermatitis induced by an extract of Dermatophagoides farinae in NC/Nga mice.

    Science.gov (United States)

    Komatsu, Ken-Ichi; Takanari, Jun; Maeda, Takahiro; Kitadate, Kentaro; Sato, Takashi; Mihara, Yoshihiro; Uehara, Kaori; Wakame, Koji

    2016-12-01

    Perilla (Perilla frutescens Britton) leaf comprises many types of active components, mainly flavonoids, and acts as an anti-inflammatory agent in in vitro and in vivo atopic dermatitis (AD) models. We investigated the effects of orally administered perilla leaf extract (PLE) on the symptoms of AD induced by Dermatophagoides farinae extract (DFE) in NC/Nga AD model mice. The mice were allowed free intake of 0.5% PLE. Skin lesions were assessed, and blood was sampled from the caudal vein on days 0, 7, 14, 21, and 31. On day 31, all mice were sacrificed to obtain blood, skin, spleen, and intestinal tissue samples. The assessment scores of the skin lesions and total serum IgE levels of PLE-treated mice (PLE group) were significantly lower than DFE-treated mice (DFE group) on days 7, 14, and 21. On day 31, the serum periostin and thymus and activation-regulated chemokine (TARC) levels in the PLE group were significantly lower than those in the DFE group. Histological analysis of the skin revealed that hyperplasia of the epidermal and dermal layers and infiltration of inflammatory cells (cell infiltration in corium tissues) were suppressed by PLE. Periostin deposition was observed in the skin tissue obtained from the DFE group. Moreover, the CD4+/CD8+ ratio of splenic T cells was suppressed in the PLE group but not in the DFE group.

  15. Gastric Ulcers Syndrome in Donkeys

    Directory of Open Access Journals (Sweden)

    Abelardo Morales Briceño

    2015-09-01

    Full Text Available This study aimed to describe gastric ulcer in donkeys. 10 donkeys (Equus asinus were studied in Bodonal de la Sierra, Badajoz-Extremadura, Spain. They were referred for necropsy and dead due to non-digestive causes. 4 males and 6 females were examined. The ages were classified of 4-16 years old. The stomach and gastric mucosa was evaluated for classified Merrit, 2003. Samples of gastric tissue were collected. The samples fixed in formalin were processed by conventional histological techniques and examined by histopathology. None of the donkeys presented clinical signs for gastric ulcers syndrome. Of the 10 donkeys studied, 10% had Grade 0; 30% Grade 1; 40% Grade 2; 10% Grade 3; and 10% Grade 4. In 30% (3/10 parasites such as Gasterophilus sp. were observed. The histological slices revealed severe damage on the gastric mucosa, a loss of continuity of the gastric mucosa with corium exposure, and subchorionic edema with parakeratotic hyperkeratosis, together with a mixed lymphoplasmocytic mononuclear infiltrate. In conclusion, we reported gastric ulcers syndrome in donkeys in Spain.

  16. Evidence for successful acceptance of irradiated free gingival allografts in dogs. [Gamma radiation

    Energy Technology Data Exchange (ETDEWEB)

    Rubenstein, H.S.; Ruben, M.P.; Levy, C.; Peiser, S.

    1975-04-01

    Free graft samples were excised and frozen to -55/sup 0/C. Subsequently the grafts were exposed to 2.5 x 10/sup 6/ rads of /sup 60/Co ..gamma..-radiation. The irradiated allogeneic grafts were later reconstituted and surgically transferred to four recipient subjects. Three autogenous nonirradiated grafts were also placed as controls. The animals were killed so as to furnish healing data at 0, 3, 7, 10, and 40 days postoperatively. Fourteen allografts were evaluated. They were judged to be nonantigenic, immunologically incompetent and nonviable. However, retention of an essentially unaltered connective tissue corium may have been instrumental in supporting subsequent epithelial regeneration from adjacent host tissue, while being passively incorporated into a very dynamic receptor zone. The graft thus served as a scaffolding for connective tissue deposition and attachment. It appears that the experimental regime obviated the immunologic interference usually encountered in allografting procedures. Thus, high intensity irradiation of the graft tissue may have rendered the tissue to be immunologically tolerable. Further studies are required to ascertain the duration of the host's immunologic unresponsiveness to the alien tissue, as well as ultimate structural and biologic fate of the transplanted tissues. Experiments are now in progress which have been designed to test the extent of immunologic sensitization induced by the grafted tissue.

  17. Drain line melt through experiment under water-free condition

    Energy Technology Data Exchange (ETDEWEB)

    Nakada, Kotaro [Nuclear Engineering Lab., Toshiba Corp., Kawasaki, Kanagawa (Japan); Jaeckel, Bernd; Hirschmann, Harald; Patorski, Jacek; Duijvestijn, Guus

    1999-07-01

    In order to investigate the behavior of a BWR drain line attacked by an oxidic melt, the experiment, CORVIS (Corium Reactor Vessel Interaction Studies) 03/2 was performed. The drain line tube was formed according to the design of an existing BWR. Aluminum oxide was used as the core melt substitute. The melt with an initial temperature of 2518 to 2543 K flowed into the water-free drain line and filled it on entire length of 7012 mm. The melt would have penetrated even further it the melt flow was not stopped by a steel plug at the tube end. The drain line did not fail but was distorted at the high temperature and elongated by 50 mm by thermoplastic deformation under its dead weight. Maximum surface temperature of 1323 K were measured near the drain line welding nozzle. It was concluded that the drain was torn off at higher internal pressure under the same thermal conditions. Temperature histories indicate that a crust was formed on the test plate screening temporarily the steel structures against melting. (author)

  18. Experiments on sedimentation of particles in a water pool with gas inflow

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eun Ho; Jung, Woo Hyun; Park, Jin Ho; Park, Hyun Suk; Moriyama, Kiyofumi [Div. of Advanced Nuclear Engineering, Pohang University of Science and Technology, Pohang (Korea, Republic of)

    2016-04-15

    During the late phase of severe accidents of light water reactors, a porous debris bed is expected to develop on the bottom of the flooded reactor cavity after breakup of the melt in water. The geometrical configuration, i.e., internal and external characteristics, of the debris bed is significant for the adequate assessment of the coolability of the relocated corium. The internal structure of a debris bed was investigated experimentally using the DAVINCI (Debris bed research Apparatus for Validation of the bubble-Induced Natural Convection effect Issue) test facility. Particle sedimentation under the influence of a two-phase natural convection flow due to the decay heat in the debris bed was simulated by dropping various sizes of particles into a water vessel with air bubble injection from the bottom. Settled particles were collected and sieved to obtain the particle mass, size distribution in the radial and axial positions, and the bed porosity and permeability. The experimental results showed that the center part of the particle bed tended to have larger particles than the peripheral area. For the axial distribution, the lower layer had a higher fraction of larger particles. As the sedimentation progressed, the size distribution in the upper layers can shift to larger sizes because of the higher vapor generation rate and stronger flow intensity.

  19. Occurrence of hybrids and laboratory evidence of fertility among three species of the Phyllosoma complex (Hemiptera: Reduviidae in Mexico

    Directory of Open Access Journals (Sweden)

    José Alejandro Martínez-Ibarra

    2009-12-01

    Full Text Available In seven studied communities of Western Mexico, triatomine specimens were sympatrically collected, some with atypical morphological characteristics in contrast to pure specimens, which were presumed to be hybrids. More than 200 specimens of Meccus pallidipennis and Meccus longipennis with brown-yellow markings on dorsal connexival segments were collected in Ahuacapán and Quitupan. In La Mesa, more than 60 specimens similar to Meccus picturatus in most morphological characteristics (including size were collected, although they presented a largely yellowish corium like M. pallidipennis. Interfertility was proven between all of the studied wild hybrid specimens, as well as between all the experimental laboratory hybrids. Two different phenotypes (M. picturatus and M. longipennis were obtained from crosses between M. picturatus x M. picturatus and M. longipennis x M. longipennis from the three studied localities in state of Nayarit as from La Mesita. Results support the hypothesis that the subspecific ranking of those triatomines may, therefore, be more appropriate because reproductive isolation has not been developed and complete interbreeding was recorded.

  20. Occurrence of hybrids and laboratory evidence of fertility among three species of the Phyllosoma complex (Hemiptera: Reduviidae) in Mexico.

    Science.gov (United States)

    Martínez-Ibarra, José Alejandro; Salazar-Schettino, Paz María; Nogueda-Torres, Benjamín; Vences, Mauro Omar; Tapia-González, José María; Espinoza-Gutiérrez, Bertha

    2009-12-01

    In seven studied communities of Western Mexico, triatomine specimens were sympatrically collected, some with atypical morphological characteristics in contrast to pure specimens, which were presumed to be hybrids. More than 200 specimens of Meccus pallidipennis and Meccus longipennis with brown-yellow markings on dorsal connexival segments were collected in Ahuacapán and Quitupan. In La Mesa, more than 60 specimens similar to Meccus picturatus in most morphological characteristics (including size) were collected, although they presented a largely yellowish corium like M. pallidipennis. Interfertility was proven between all of the studied wild hybrid specimens, as well as between all the experimental laboratory hybrids. Two different phenotypes (M. picturatus and M. longipennis) were obtained from crosses between M. picturatus x M. picturatus and M. longipennis x M. longipennis from the three studied localities in state of Nayarit as from La Mesita. Results support the hypothesis that the subspecific ranking of those triatomines may, therefore, be more appropriate because reproductive isolation has not been developed and complete interbreeding was recorded.

  1. MELCOR 1.8.2 Assessment: IET direct containment heating tests

    Energy Technology Data Exchange (ETDEWEB)

    Kmetyk, L.N.

    1993-10-01

    MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC, that models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRS. As part of an ongoing assessment program, the MELCOR computer code has been used to analyze several of the IET direct containment heating experiments done at 1:10 linear scale in the Surtsey test facility at Sandia and at 1:40 linear scale in the corium-water thermal interactions (CWTI) COREXIT test facility at Argonne National Laboratory. These MELCOR calculations were done as an open post-test study, with both the experimental data and CONTAIN results available to guide the selection of code input. Basecase MELCOR results are compared to test data in order to evaluate the new HPME DCH model recently added in MELCOR version 1.8.2. The effect of various user-input parameters in the HPME model, which define both the initial debris source and the subsequent debris interaction, were investigated in sensitivity studies. In addition, several other non-default input modelling changes involving other MELCOR code packages were required in our IET assessment analyses in order to reproduce the observed experiment behavior. Several calculations were done to identify whether any numeric effects exist in our DCH IET assessment analyses.

  2. The European Research on Severe Accidents in Generation-II and -III Nuclear Power Plants

    Directory of Open Access Journals (Sweden)

    Jean-Pierre Van Dorsselaere

    2012-01-01

    Full Text Available Forty-three organisations from 22 countries network their capacities of research in SARNET (Severe Accident Research NETwork of excellence to resolve the most important remaining uncertainties and safety issues on severe accidents in existing and future water-cooled nuclear power plants (NPP. After a first project in the 6th Framework Programme (FP6 of the European Commission, the SARNET2 project, coordinated by IRSN, started in April 2009 for 4 years in the FP7 frame. After 2,5 years, some main outcomes of joint research (modelling and experiments by the network members on the highest priority issues are presented: in-vessel degraded core coolability, molten-corium-concrete-interaction, containment phenomena (water spray, hydrogen combustion…, source term issues (mainly iodine behaviour. The ASTEC integral computer code, jointly developed by IRSN and GRS to predict the NPP SA behaviour, capitalizes in terms of models the knowledge produced in the network: a few validation results are presented. For dissemination of knowledge, an educational 1-week course was organized for young researchers or students in January 2011, and a two-day course is planned mid-2012 for senior staff. Mobility of young researchers or students between the European partners is being promoted. The ERMSAR conference is becoming the major worldwide conference on SA research.

  3. Palynology of the Rupelian to Burdigalian (Oligocene to Lower Miocene) interval of the Alma-1X well, Danish North Sea

    Energy Technology Data Exchange (ETDEWEB)

    Schioeler, P.

    2003-07-01

    A palynological study of cuttings samples from the North Sea well Alma-1X documents for the first time the dinoflagellae cyst and acritarch assemblage in the Oligocene to Lower Miocene interval of the Central North Sea. The interval is characterised by a poor foraminifera assemblage yielding a relatively low stratigraphic resolution, whereas the palynomorph assemblage is rich. Two hundred and three microplankton taxa were encountered in the study interval. The distribution of dinoflagellates and acritarchs in the well suggests a subdivision of the Oligocene to Lower Miocene interval into 24 zones based on first downhole occurrence of key taxa. The subdivsion lends support from unpublished consultancy report data from several other wells in the Danish North Sea. However, as the the subdivion builds on published data from one well only, it is considered informal until more documentatin is at hand. Four new species and on e new subspecies are described from the study interval: Amphorosphaeridium almae Schioeler sp. nov., Dalella rota Schioeler sp. nov., Filisphaera pachyderma Schioeler sp. nov., Pentadinium corium Schioeler sp. nov. and Spiniferites pseudofurcatus verrucosus Schioeler ssp., nov. Pseudospiniferites manumii Lund, 2002 is emended and transferred to the genus Spiniferites. (au)

  4. Fukushima Daiichi Unit 1 Ex-Vessel Prediction: Core Concrete Interaction

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Farmer, Mitchell [Argonne National Lab. (ANL), Argonne, IL (United States); Francis, Matthew W [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Lower head failure and corium concrete interaction were predicted to occur at Fukushima Daiichi Unit 1 (1F1) by several different system-level code analyses, including MELCOR v2.1 and MAAP5. Although these codes capture a wide range of accident phenomena, they do not contain detailed models for ex-vessel core melt behavior. However, specialized codes exist for analysis of ex-vessel melt spreading (e.g., MELTSPREAD) and long-term debris coolability (e.g., CORQUENCH). On this basis, an analysis was carried out to further evaluate ex-vessel behavior for 1F1 using MELTSPREAD and CORQUENCH. Best-estimate melt pour conditions predicted by MELCOR v2.1 and MAAP5 were used as input. MELTSPREAD was then used to predict the spatially dependent melt conditions and extent of spreading during relocation from the vessel. The results of the MELTSPREAD analysis are reported in a companion paper. This information was used as input for the long-term debris coolability analysis with CORQUENCH.

  5. Polyanionic collagen membranes for guided tissue regeneration: Effect of progressive glutaraldehyde cross-linking on biocompatibility and degradation.

    Science.gov (United States)

    Veríssimo, D M; Leitão, R F C; Ribeiro, R A; Figueiró, S D; Sombra, A S B; Góes, J C; Brito, G A C

    2010-10-01

    The ultimate goal of periodontal therapy is to control periodontal tissue inflammation and to produce predictable regeneration of that part of the periodontium which has been lost as a result of periodontal disease. In guided tissue regeneration membranes function as mechanical barriers, excluding the epithelium and gingival corium from the root surface and allowing regeneration by periodontal ligament cells. This report aims to study the effect of glutaraldehyde (GA) cross-linking on mineralized polyanionic collagen (PAC) membranes by conducting a histological evaluation of the tissue response (biocompatibility) and by assessing the biodegradation of subcutaneous membrane implants in rats. We studied six different samples: a PAC, a PAC mineralized by alternate soaking processes for either 25 or 75 cycles (PAC 25 and PAC 75, respectively) and these films cross-linked by GA. Inflammatory infiltrate, cytokine dosage, fibrosis capsule thickness, metalloproteinase immunohistochemistry and membrane biodegradation after 1, 7, 15 and 30 days were measured. The inflammatory response was found to be more intense in membranes without cross-linking, while the fibrosis capsules became thicker in cross-linked membranes after 30 days. The membranes without cross-linking suffered intense biodegradation, while the membranes with cross-linking remained intact after 30 days. The cross-linking with GA reduced the inflammatory response and prevented degradation of the membranes over the entire course of the observation period. These membranes are thus an attractive option when the production of new bone depends on the prolonged presence of a mechanical barrier.

  6. Experiments on melt droplets falling into a water pool

    Energy Technology Data Exchange (ETDEWEB)

    Okkonen, T.; Sehgal, B.R. [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety

    1998-01-01

    This paper presents experimental data and analysis related to melt droplets falling into a water pool. A binary CaO-B{sub 2}O{sub 3} melt mixture is used to study the influence of melt superheat and water subcooling on droplet deformation and fragmentation. For the conditions studied (We {<=} 1000), the surface tension of the melt droplet and the film boiling stability greatly affect the fragmentation behaviour. If the melt temperature is between the liquidus and solidus point (mushy zone) or if the film boiling is stable due to a relatively low subcooling, the droplet deformation and fragmentation are mitigated. This behaviour can be related to the effective Weber number (We) of the melt droplet upon entry into the water pool. Similar phenomena can be expected also for interactions of corium (UO{sub 2}-ZrO{sub 2}) and water, which are characterized by a potentially fast transformation of melt into the mushy zone and by particularly stable film boiling. (author)

  7. On-line measurements of RuO{sub 4} during a PWR severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Reymond-Laruinaz, S.; Doizi, D. [CEA, DEN, Departement de Physico-chimie, CEA/Saclay, 91191 Gif sur Yvette Cedex, (France); Manceron, L. [Societe Civile Synchrotron SOLEIL, L' Orme des Merisiers, St-Aubin BP48, 91192 Gif-sur-Yvette Cedex, (France); MONARIS, UMR 8233, Universite Pierre et Marie Curie, 4 Place Jussieu, case 49, F-75252 Paris Cedex 05, (France); Boudon, V. [Laboratoire Interdisciplinaire Carnot de Bourgogne, UMR 6303 CNRS-Universite de Bourgogne, 9 avenue Alain Savary, BP 47870, F-21078 Dijon Cedex, (France); Ducros, G. [CEA, DEN, Departement d' Etudes des Combustibles, CEA/Cadarache, 13108 Saint-Paul-lez-Durance cedex, (France)

    2015-07-01

    After the Fukushima accident, it became essential to have a way to monitor in real time the evolution of a nuclear reactor during a severe accident, in order to react efficiently and minimize the industrial, ecological and health consequences of the accident. Among gaseous fission products, the tetroxide of ruthenium RuO{sub 4} is of prime importance since it has a significant radiological impact. Ruthenium is a low volatile fission product but in case of the rupture of the vessel lower head by the molten corium, the air entering into the vessel oxidizes Ru into gaseous RuO{sub 4}, which is not trapped by the Filtered Containment Venting Systems. To monitor the presence of RuO{sub 4} allows making a diagnosis of the core degradation and quantifying the release into the atmosphere. To determine the presence of RuO{sub 4}, FTIR spectrometry was selected. To study the feasibility of the monitoring, high-resolution IR measurements were realized at the French synchrotron facility SOLEIL on the infrared beam line AILES. Thereafter, theoretical calculations were done to simulate the FTIR spectrum to describe the specific IR fingerprint of the molecule for each isotope and based on its partial pressure in the air. (authors)

  8. Development of Methodology for Spent Fuel Pool Severe Accident Analysis Using MELCOR Program

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won-Tae; Shin, Jae-Uk [RETech. Co. LTD., Yongin (Korea, Republic of); Ahn, Kwang-Il [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The general reason why SFP severe accident analysis has to be considered is that there is a potential great risk due to the huge number of fuel assemblies and no containment in a SFP building. In most cases, the SFP building is vulnerable to external damage or attack. In contrary, low decay heat of fuel assemblies may make the accident processes slow compared to the accident in reactor core because of a great deal of water. In short, its severity of consequence cannot exclude the consideration of SFP risk management. The U.S. Nuclear Regulatory Commission has performed the consequence studies of postulated spent fuel pool accident. The Fukushima-Daiichi accident has accelerated the needs for the consequence studies of postulated spent fuel pool accidents, causing the nuclear industry and regulatory bodies to reexamine several assumptions concerning beyond-design basis events such as a station blackout. The tsunami brought about the loss of coolant accident, leading to the explosion of hydrogen in the SFP building. Analyses of SFP accident processes in the case of a loss of coolant with no heat removal have studied. Few studies however have focused on a long term process of SFP severe accident under no mitigation action such as a water makeup to SFP. USNRC and OECD have co-worked to examine the behavior of PWR fuel assemblies under severe accident conditions in a spent fuel rack. In support of the investigation, several new features of MELCOR model have been added to simulate both BWR fuel assembly and PWR 17 x 17 assembly in a spent fuel pool rack undergoing severe accident conditions. The purpose of the study in this paper is to develop a methodology of the long-term analysis for the plant level SFP severe accident by using the new-featured MELCOR program in the OPR-1000 Nuclear Power Plant. The study is to investigate the ability of MELCOR in predicting an entire process of SFP severe accident phenomena including the molten corium and concrete reaction. The

  9. Effects of Sulfate on the Community Structure of Phytoplankton in Freshwater%硫酸盐对淡水浮游藻类群落结构的影响研究

    Institute of Scientific and Technical Information of China (English)

    钟远; 樊娟; 刘春光; 庄源益

    2009-01-01

    In order to promote the understanding of sulfate enrichnment on the growth and .sucession of phytoplanktion cotomunities in eutrophic freshwater body laboratory experiment of phytoplankton in a landscapc water body was conducted.Three treatment groups were set up in experiment:cantroal (G0),step by step enrichmenl (G1) and eurichmet at a time (G2). The resolt indicated that the addition of sulfate promoted the increase of Chloropbyta species and biomass, while inhibited the growth of native Cyanobacteria and diatom species. Species and biomass. of G0, G1 group were similar and much higher than those of G2 group. Avcrage biomass of G0, Gland G2 group was 98.46, 96.09 and 81.19 mg·L~(-1) , respectively. C0 group was almost always dominated by Cyanobacteria species such as P. corium and O. amphibia, dominant species in G1 and G2 groups changed from Cyannbacteria species to Chlorophyta species including S. quadricauda, G. radiate and S. obliqus. Compared with that of control group GO (1.49±0.32),the average community diversity indices of G1 (1.70 ±0.1g)and G2 groups ( 1.68±0.40) were elevated.%为了解硫酸盐含量增加对富营养化淡水水体藻类生长及群落演替的影响,对天津市某景观水体藻类进行了室内模拟试验研究.按硫酸盐投加方式分设对照组(G0)、逐步投加组(G1)和一次性投加组(G2).结果表明,向水中投加一定浓度硫酸盐可以抑制原有蓝藻和硅藻生长,而绿藻种类和生物量增加.G0和G1组的种类数和生物量相近但是大于G2组.G0、G1和G2组的平均总生物量分别为98.46、96.09和81.19 mg·L~(-1).对照组几乎始终是皮状席藻(P.corium)和两栖颤藻(0.amphibia)等蓝藻为第一优势种,而投加硫酸盐的G1和G2组优势种从皮状席藻和小颤藻(0.tenuis)等蓝藻向四尾栅藻(S.quadricauda)、放射多芒藻(G.radiate)和斜生栅藻(S.obliqus)等绿藻演替.群落平均多样性指数与对照组G0(1.49±0.32)相比,投加硫酸盐的G1(1.70±0

  10. OECD MCCI project Melt Eruption Test (MET) design report, Rev. 2. April 15, 2003.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Lomperski, S.; Kilsdonk, D. J.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division); (NRC)

    2011-05-23

    The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. The Melt Coolability and Concrete Interaction (MCCI) program is pursuing separate effect tests to examine the viability of the melt coolability mechanisms identified as part of the MACE program. These mechanisms include bulk cooling, water ingression, volcanic eruptions, and crust breach. At the second PRG meeting held at ANL on 22-23 October 2002, a preliminary design1 for a separate effects test to investigate the melt eruption cooling mechanism was presented for PRG review. At this meeting, NUPEC made several recommendations on the experiment approach aimed at optimizing the chances of achieving a floating crust boundary condition in this test. The principal recommendation was to incorporate a mortar sidewall liner into the test design, since data from the COTELS experiment program indicates that corium does not form a strong mechanical bond with this material. Other recommendations included: (i) reduction of the electrode elevation to well below the melt upper surface elevation (since the crust may bond to these solid surfaces), and (ii) favorably taper the mortar liner to facilitate crust detachment and relocation during the experiment. Finally, as a precursor to implementing these modifications, the PRG recommended the development of a design for a small-scale scoping test intended to verify the ability of the mortar liner to preclude formation of an anchored bridge crust under core-concrete interaction conditions. This revised Melt Eruption Test (MET) plan is intended to

  11. Experiment of fIow and quenching for hemisphericaI narrow gaps%半球体窄缝通道流动换热与冷却能力试验

    Institute of Scientific and Technical Information of China (English)

    林波; 田文喜; 秋穗正; 苏光辉

    2015-01-01

    Many molten corium relocated in the lower head of TMI-2 reactor after severe accident.The lower head was not pene-trated during the accident.There are some cooling mechanisms have been suggested and a gap cooling mechanism is considered to be the most plausible one which plays a major role in the cooling of the corium.A narrow gap flow and heat transfer experiment was done to investigate mechanisms of heat transfer and the coolability.Hemispherical copper was heated downward and the flow heat transfer behavior was investigated.The gaps between the copper shell and plastic vessel were 2 mm and 4 mm,respectively. The effect of the gap cooling was investigated.In the experiment,the copper was heated with different powers,and the effects on the heat transfer should be noticed.The experimental results show that the coolability is affected by the wall heat and the gap width.The wall temperature is important for the coolability of the narrow gap.The effect of the forced convection is more markedly than that of the nature convection in single phase convection heat transfer and multi-phase convection heat transfer ex-periments.%在美国 TMI-2发生严重事故时,反应堆下封头内累积了大量的熔融物,经过一段时间观察,下封头未被熔融物熔穿。因此提出了一些可能的冷却机理,其中间隙冷却机理被认为是最有可能的一种。本文进行了窄通道流动换热与冷却能力的试验,旨在探索窄缝的传热机理和冷却能力。试验中半球形采用向下加热,并观察流动换热现象。在分组试验过程中窄缝间隙选取为2 mm 和4 mm 两种,采用不同的加热功率对铜球进行加热。试验结果表明:在窄通道试验中,冷却过程与壁面加热、间隙大小有着不可分割的关系;壁面温度对窄通道冷却有着较大的影响;在窄通道流动换热与冷却过程中,对于单相对流换热,强制对流冷却的效果大于自然对流的冷却效果

  12. Overview of the ACEX project iodine work

    Energy Technology Data Exchange (ETDEWEB)

    Merilo, M.

    1996-12-01

    The ACEX project is an internationally sponsored research program that focuses on several aspects of severe accidents. The areas addressed are iodine behavior in containments, pool scrubbing, molten corium concrete interactions, and ex-vessel core debris coolability. These areas all represent extensions to the previous and current ACE and MACE programs respectively. The ACE-Phase B (iodine) project, and other recent research efforts, have clarified the roles of the important phenomena that influence iodine volatility in reactor containments during severe accidents. The ACE Iodine Chemistry Subcommittee concluded that even though enough data has been generated to support reasonably good quantification of the important phenomena, a few important areas remain where quantification is still uncertain. This is due to a lack of agreement on how to utilize the existing database, as well as the possible absence of critical test and/or property data. Technical resolution of the overall iodine behavior issue is therefore not feasible until these uncertainties are fully assessed and practical solutions have been identified, implemented, and verified. The overall objectives of the ACEX iodine research program are to ensure that the iodine database can be used to predict the airborne concentration of iodine, the conditions for iodine reservoir stability, and to provide a mechanistic understanding for these phenomena. The first phase of this work involves a comprehensive review and interpretation of the existing database in order to formulate practical strategies for dealing with significant uncertainties and/or deficiencies. Several projects are underway involving the effects of organic reactions and structural surface interactions. In addition effort is being expended on standardizing the aqueous iodine kinetics database, specifying useful mass transfer models, and defining methodology for pH prediction. (Abstract Truncated)

  13. Icare/Cathare coupling: three-dimensional thermal hydraulics of severe LWR accidents

    Energy Technology Data Exchange (ETDEWEB)

    Guillard, V.; Fichot, F. [CEA Fontenay aux Roses, Inst. de Protection et de Surete Nucleaire, Dept. de Recherches en Securite, DRS, 92 (France); Boudier, P.; Parent, M. [CEA Grenoble, Dir. des Reacteurs Nucleaires, DRN, 38 (France); Roser, R. [Communication et Systemes Systemes d' Information, CS SI, 38 - Fontaine (France)

    2001-07-01

    In the phenomenology of severe LWR accidents considered in safety studies, the accidental sequences can be divided into three phases: the initial phase, where no severe damage of fuel or control rods and structures occurs; the early core degradation phase, where limited material melting and relocation takes place; and the late core degradation phase during which substantial material relocation happens, molten pools and debris beds can form and corium may fall into the lower plenum and, in case of vessel failure, come into the containment. The CATHARE2 code is a system code which has been developed by CEA for IPSN, EDF and FRAMATOME to describe the thermal-hydraulics behavior of a whole PWR circuit during the first of these three phases, with a core degradation model limited to clad rupture. The ICARE2 code, developed by IPSN, allows the complete description of early and late core degradation phases, with a thermal-hydraulics model limited to the vessel, initial and boundary conditions being provided by a system code. The aim of this paper is to present the main features of the new version of the coupling, ICARE/CATHARE V2. First, the general characteristics of ICARE2 V3mod1 and CATHARE2 V1.5 standard codes, dealing with physical models and numerical aspects, are described. Second, the technical features of the coupling between the two codes are detailed. At last, some results of ICARE/CATHARE V2 calculations are presented which demonstrate the ability of the code to simulate a severe accident in a PWR and notably to describe multi-dimensional effects occurring in the core during the LOCA and degradation phases. (authors)

  14. Advanced in-vessel retention design for next generation risk management

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Kune Y.; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

    1997-12-31

    In the TMI-2 accident, approximately twenty (20) tons of molten core material drained into the lower plenum. Early advanced light water reactor (LWR) designs assumed a lower head failure and incorporated various measures for ex-vessel accident mitigation. However,one of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100 deg C for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel. Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident/risk management strategies. As an advanced in-vessel design concept, this work presents the COrium Attack Syndrome Immunization Structures (COASIS) that are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in-vessel (COASISI) and ex-vessel (COASISO) are demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the Korean Standard Nuclear Power Plant (KSNPP) reactor. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options. 15 refs., 5 figs., 1 tab. (Author)

  15. GROSS AND MICROSCOPIC ANATOMY OF MAMMARY GLAND OF DROMEDARIES UNDER DIFFERENT PHYSIOLOGICAL CONDITIONS

    Directory of Open Access Journals (Sweden)

    R. Kausar, A. Sarwar and C.S. Hayat

    2001-09-01

    Full Text Available Samples of 24 mammary glands from healthy one-humped camel (Camelus dromedarius cows were investigated under different physiological conditions for their gross and light microscopic anatomy. Different groups included equal number of immature heifers, mature heifers, lactating and non-lactating animals. Tissues fixed in 10% NBF were processed as per routine and stained with hematoxylin and eosin (H&E and Massons trichome. Morphometry was done with the help of stage and ocular micrometer. Gross studies revealed that the camel's udder consists of four quarters. The udder and teat showed light brown to solid black tinge in colour. The tips of teats sloped to a point both in immature and mature animals, however, the conformation of teats turned rounded at the tips in lactating camels. Each teat possessed two teat canals. The dimensions of teat and streak canal varied markedly among four different groups. Number of Furestenberg's rosettes ranged from 11.6 to 13.6. Microscopic studies revealed that streak canal was lined by stratified squamous keratinized epithelium that was partially extremely thin in some parts. Cutaneous layer of teat was devoid of hair follicles except at the base of teat. Follicles were associated with sebaceous glands. Sweat glands were less coiled and showed a wide acinous element forming the part of excretory duct. Glomus organs occurred in the stratum profundum of the corium as well as in the subcutis of the skin of mammary gland. They also revealed great variation in structure and size. Epithelial lining of the alveoli varied from flattened to columnar according to physiological state. Number and size of alveoli per lobule decreased and the parenchyma was replaced by loose connective tissue during non-lactating phase. These results suggested that age and lactation considerably influenced gross and microscopic anatomy of mammary gland in camels.

  16. Hydrogen combustion in a flat semi-confined layer with respect to the Fukushima Daiichi accident

    Energy Technology Data Exchange (ETDEWEB)

    Kuznetsov, Mike, E-mail: kuznetsov@kit.edu [Karlsruhe Institute of Technology, 76131 Karlsruhe (Germany); Yanez, Jorge [Karlsruhe Institute of Technology, 76131 Karlsruhe (Germany); Grune, Joachim; Friedrich, Andreas [Pro-Science GmbH, 76275 Ettlingen (Germany); Jordan, Thomas [Karlsruhe Institute of Technology, 76131 Karlsruhe (Germany)

    2015-05-15

    Graphical abstract: - Highlights: • Critical conditions for flame propagation regimes in a layer geometry are analyzed. • Numerical simulation of hydrogen explosion reproduces real strength of shock waves. • From 80 to 200 kg of hydrogen were exploded during Fukushima (Unit I) accident. • A sonic deflagration with TNT equivalent of 800 kg was the most probable regime. - Abstract: Hydrogen accumulations at the top of a containment or reactor building may occur due to the interaction of molten corium and water followed by a severe accident of a nuclear reactor (TMI, Chernobyl, Fukushima Daiichi). The hydrogen that is released from the reactor accumulates usually as a stratified semi-confined layer of hydrogen–air mixture. A series of large scale experiments on hydrogen combustion and explosion in a semi-confined layer of uniform and non-uniform hydrogen–air mixtures in the presence of obstructions or without them was performed at the Karlsruhe Institute of Technology (KIT). Different flame propagation regimes from slow subsonic to relatively fast sonic flames and then to detonations were experimentally investigated in different geometries and then simulated with COM3D code with respect to evaluate the amount of hydrogen that was involved in the Fukushima Daiichi Accident (FDA). The experiments were performed in a horizontal semi-confined layer with the dimensions 9 × 3 × 0.6 m with/without obstacles opened from below. The hydrogen concentration in the mixtures with air was varied in the range of 10–34 vol.% without or with a gradient of 20–60 vol.%H{sub 2}/m. Effects of hydrogen concentration gradient, layer thickness, obstruction geometry, average and maximum hydrogen concentration on the flame propagation regimes were investigated with respect to evaluate the maximum pressure loads on internal structures. Blast wave strength and dynamics of propagation after the explosion of the hydrogen–air mixture layer were numerically simulated to reproduce

  17. Investigation of debris bed formation, spreading and coolability

    Energy Technology Data Exchange (ETDEWEB)

    Kudinov, P.; Konovalenko, A.; Grishchenko, D.; Yakush, S.; Basso, S.; Lubchenko, N.; Karbojian, A. [Royal Institute of Technology, KTH. Div. of Nuclear Power Safety, Stockholm (Sweden)

    2013-08-15

    The work is motivated by the severe accident management strategy adopted in Nordic type BWRs. It is assumed that core melt ejected from the vessel will fragment, quench and form a coolable debris bed in a deep water pool below the vessel. In this work we consider phenomena relevant to the debris bed formation and coolability. Several DEFOR-A (Debris Bed Formation - Agglomeration) tests have been carried out with new corium melt material and a melt releasing nozzle mockup. The influence of the melt material, melt superheat, jet free fall height on the (i) faction of agglomerated debris, (ii) particle size distribution, (iii) ablation/plugging of the nozzle mockup has been addressed. Results of the DECOSIM (Debris Coolability Simulator) code validation against available COOLOCE data are presented in the report. The dependence of DHF on system pressure from COOLOCE experiments can be reproduced quite accurately if either the effective particle diameter or debris bed porosity is increased. For a cylindrical debris bed, good agreement is achieved in DECOSIM simulations for the particle diameter 0.89 mm and porosity 0.4. The results obtained are consistent with MEWA simulation where larger particle diameters and porosities were found to be necessary to reproduce the experimental data on DHF. It is instructive to note that results of DHF prediction are in better agreement with POMECO-HT data obtained for the same particles. It is concluded that further clarification of the discrepancies between different experiments and model predictions. In total 13 exploratory tests were carried out in PDS (particulate debris spreading) facility to clarify potential influence of the COOLOCE (VTT) facility heaters and TCs on particle self-leveling process. Results of the preliminary analysis suggest that there is no significant influence of the pins on self-leveling, at least for the air superficial velocities ranging from 0.17 up to 0.52 m/s. Further confirmatory tests might be needed

  18. Numerical simulation of 2D ablation profile in CCI-2 experiment by moving particle semi-implicit method

    Energy Technology Data Exchange (ETDEWEB)

    Chai, Penghui, E-mail: phchai@vis.t.u-tokyo.ac.jp; Kondo, Masahiro; Erkan, Nejdet; Okamoto, Koji

    2016-05-15

    Highlights: • Multiphysics models were developed based on Moving Particle Semi-implicit method. • Mixing process, chemical reaction can be simulated in MCCI calculation. • CCI-2 experiment was simulated to validate the models. • Simulation and experimental results for sidewall ablation agree well. • Simulation results confirm the rapid erosion phenomenon observed in the experiment. - Abstract: Numerous experiments have been performed to explore the mechanisms of molten core-concrete interaction (MCCI) phenomena since the 1980s. However, previous experimental results show that uncertainties pertaining to several aspects such as the mixing process and crust behavior remain. To explore the mechanism governing such aspects, as well as to predict MCCI behavior in real severe accident events, a number of simulation codes have been developed for process calculations. However, uncertainties exist among the codes because of the use of different empirical models. In this study, a new computational code is developed using multiphysics models to simulate MCCI phenomena based on the moving particle semi-implicit (MPS) method. Momentum and energy equations are used to solve the velocity and temperature fields, and multiphysics models are developed on the basis of the basic MPS method. The CCI-2 experiment is simulated by applying the developed code. With respect to sidewall ablation, good agreement is observed between the simulation and experimental results. However, axial ablation is slower in the simulation, which is probably due to the underestimation of the enhancement effect of heat transfer provided by the moving bubbles at the bottom. In addition, the simulation results confirm the rapid erosion phenomenon observed in the experiment, which in the numerical simulation is explained by solutal convection provided by the liquid concrete at the corium/concrete interface. The results of the comparison of different model combinations show the effect of each

  19. Containment loads due to direct containment heating and associated hydrogen behavior: Analysis and calculations with the CONTAIN code

    Energy Technology Data Exchange (ETDEWEB)

    Williams, D C; Bergeron, K D; Carroll, D E; Gasser, R D; Tills, J L; Washington, K E

    1987-05-01

    One of the most important unresolved issues governing risk in many nuclear power plants involves the phenomenon called direct containment heating (DCH), in which it is postulated that molten corium ejected under high pressure from the reactor vessel is dispersed into the containment atmosphere, thereby causing sufficient heating and pressurization to threaten containment integrity. Models for the calculation of potential DCH loads have been developed and incorporated into the CONTAIN code for severe accident analysis. Using CONTAIN, DCH scenarios in PWR plants having three different representative containment types have been analyzed: Surry (subatmospheric large dry containment), Sequoyah (ice condenser containment), and Bellefonte (atmospheric large dry containment). A large number of parameter variation and phenomenological uncertainty studies were performed. Response of DCH loads to these variations was found to be quite complex; often the results differ substantially from what has been previously assumed concerning DCH. Containment compartmentalization offers the potential of greatly mitigating DCH loads relative to what might be calculated using single-cell representations of containments, but the actual degree of mitigation to be expected is sensitive to many uncertainties. Dominant uncertainties include hydrogen combustion phenomena in the extreme environments produced by DCH scenarios, and factors which affect the rate of transport of DCH energy to the upper containment. In addition, DCH loads can be aggravated by rapid blowdown of the primary system, co-dispersal of moderate quantities of water with the debris, and quenching of de-entrained debris in water; these factors act by increasing steam flows which, in turn, accelerates energy transport. It may be noted that containment-threatening loads were calculated for a substantial portion of the scenarios treated for some of the plants considered.

  20. HIGH-FIDELITY SIMULATION-DRIVEN MODEL DEVELOPMENT FOR COARSE-GRAINED COMPUTATIONAL FLUID DYNAMICS

    Energy Technology Data Exchange (ETDEWEB)

    Hanna, Botros N.; Dinh, Nam T.; Bolotnov, Igor A.

    2016-06-01

    Nuclear reactor safety analysis requires identifying various credible accident scenarios and determining their consequences. For a full-scale nuclear power plant system behavior, it is impossible to obtain sufficient experimental data for a broad range of risk-significant accident scenarios. In single-phase flow convective problems, Direct Numerical Simulation (DNS) and Large Eddy Simulation (LES) can provide us with high fidelity results when physical data are unavailable. However, these methods are computationally expensive and cannot be afforded for simulation of long transient scenarios in nuclear accidents despite extraordinary advances in high performance scientific computing over the past decades. The major issue is the inability to make the transient computation parallel, thus making number of time steps required in high-fidelity methods unaffordable for long transients. In this work, we propose to apply a high fidelity simulation-driven approach to model sub-grid scale (SGS) effect in Coarse Grained Computational Fluid Dynamics CG-CFD. This approach aims to develop a statistical surrogate model instead of the deterministic SGS model. We chose to start with a turbulent natural convection case with volumetric heating in a horizontal fluid layer with a rigid, insulated lower boundary and isothermal (cold) upper boundary. This scenario of unstable stratification is relevant to turbulent natural convection in a molten corium pool during a severe nuclear reactor accident, as well as in containment mixing and passive cooling. The presented approach demonstrates how to create a correction for the CG-CFD solution by modifying the energy balance equation. A global correction for the temperature equation proves to achieve a significant improvement to the prediction of steady state temperature distribution through the fluid layer.

  1. Ex-vessel coolability and energetics of steam explosions in nordic light water reactors - EXCOOLSE project report 2004

    Energy Technology Data Exchange (ETDEWEB)

    Park, H.S.; Nayak, A.K.; Hansson, R.C.; Sehgal, B.R. [Royal Inst. of Technology, Div. of Nuclear Power Safety (Sweden)

    2005-10-01

    Beyond-the-design-basis accidents, i.e. severe accidents, involve melting of the nuclear reactor core and release of radioactivity. Intensive research has been performed for years to evaluate the consequence of the postulated severe accidents. Severe accidents posed, to the reactor researchers, a most interesting and most difficult set of phenomena to understand, and to predict the consequences, for the various scenarios that could be contemplated. The complexity of the interactions, occurring at such high temperatures ({approx} 2500 deg. C), between different materials, which are changing phases and undergoing chemical reactions, is simply indescribable with the accuracy that one may desire. Thus, it is a wise approach to pursue research on SA phenomena until the remaining uncertainty in the predicted consequence, or the residual risk, can be tolerated. In the PRE-DELI-MELT project at NKS, several critical issues on the core melt loadings in the BWR and PWR reactor containments were identified. Many of Nordic nuclear power plants, particularly in boiling water reactors, adopted the Severe Accident Management Strategy (SAMS) which employed the deep subcooled water pool in lower dry-well. The success of this SAMS largely depends on the issues of steam explosions and formation of debris bed and its coolability. From the suggestions of the PRE-DELI-MELT project, a series of research plan was proposed to investigate the remaining issues specifically on the ex-vessel coolability of corium during severe accidents; (a) ex-vessel coolability of the melt or particulate debris, and (b) energetics and debris characteristics of fuel-coolant interactions endangering the integrity of the reactor containments. (au)

  2. Digital Cushion Fatty Acid Composition and Lipid Metabolism Gene Network Expression in Holstein Dairy Cows Fed a High-Energy Diet.

    Directory of Open Access Journals (Sweden)

    Zeeshan Muhammad Iqbal

    Full Text Available The hoof digital cushion is a complex structure composed of adipose tissue beneath the distal phalanx, i.e. axial, middle and abaxial fat pad. The major role of these fat depots is dampening compression of the corium underneath the cushion. The study aimed to determine expression of target genes and fatty acid profiles in the hoof of non-pregnant dry Holstein cows fed low (CON or high-energy (OVE diets. The middle fat pad of the hoof digital cushion was collected soon after slaughter. Despite the lack of effect on expression of the transcription regulators SREBF1 and PPARG, the expression of the lipogenic enzymes ACACA, FASN, SCD, and DGAT2 was upregulated with OVE. Along with the upregulation of G6PD and IDH1, important for NADPH synthesis during lipogenesis, and the basal glucose transporter SLC2A1, these data indicated a pro-lipogenic response in the digital cushion with OVE. The expression of the lipid droplet-associated protein PLIN2 was upregulated while expression of lipolytic enzymes (ATGL, ABDH5, and LIPE only tended to be upregulated with OVE. Therefore, OVE induced lipogenesis, lipid droplet formation, and lipolysis, albeit to different extents. Although concentration of monounsaturated fatty acids (MUFA did not differ, among the polyunsaturated fatty acids (PUFA, the concentration of 20:5n3 was lower with OVE. Among the saturated fatty acids, 20:0 concentration was greater with OVE. Although data indicated that the hoof digital cushion metabolic transcriptome is responsive to higher-energy diets, this did not translate into marked differences in the fatty acid composition. The decrease in concentration of PUFA, which could contribute to synthesis of inflammatory molecules, in OVE-fed cows indicated that feeding higher-energy diets might be detrimental for the mediation of inflammation in digital cushion. This effect could be further exacerbated by physiologic and endocrine changes during the peripartal period that favor inflammation.

  3. Optimization of the Severe Accident Management Strategy for Domestic Plants and Validation Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. B.; Kim, H. D.; Koo, K. M.; Park, R. J.; Hong, S. H.; Cho, Y. R.; Kim, J. T.; Ha, K. S.; Kang, K. H

    2007-04-15

    nuclear power plants, a technical basis report and computational aid tools were developed in parallel with the experimental and analytical works for the resolution of the uncertain safety issues. ELIAS experiments were carried out to quantify the boiling heat removal rate at the upper surface of a metallic layer for precise evaluations on the effect of a late in-vessel coolant injection. T-HERMES experiments were performed to examine the two-phase natural circulation phenomena through the gap between the reactor vessel and the insulator in the APR1400. Detailed analyses on the hydrogen control in the APR1400 containment were performed focused on the effect of spray system actuation on the hydrogen burning and the evaluation of the hydrogen behavior in the IRWST. To develop the technical basis report for the severe accident management, analyses using SCDAP/RELAP5 code were performed for the accident sequences of the OPR1000. Based on the experimental and analytical results performed in this study, the computational aids for the evaluations of hydrogen flammability in the containment, criteria of the in-vessel corium cooling, criteria of the external reactor vessel cooling were developed. An ASSA code was developed to validate the signal from the instrumentations during the severe accidents and to process the abnormal signal. Since ASSA can perform the signal processing from the direct input of the nuclear power plant during the severe accident, it can be platform of the computational aids. In this study, the ASSA was linked with the computaional aids for the hydrogen flammability.

  4. Development of pyro-processing technology at CRIEPI for carving out the future of nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Iizuka, M.; Koyama, T.; Sakamura, Y.; Uozumi, K.; Fujihata, K.; Kato, T.; Murakami, T.; Tsukada, T. [Central Research Institute of Electric Power Industry, Komae-shi, Tokyo 201-8511 (Japan); Glatz, J.P. [European Commission, JRC, Institute for Transuranium Elements (Germany)

    2013-07-01

    Pyro-processing has been attracting increasing attention as a promising candidate as an advanced nuclear fuel cycle technology. It provides economic advantage as well as reduction in proliferation risk and burden of long live radioactive waste, especially when it is combined with advanced fuels such as metallic or nitride fuel which gives excellent burning efficiency of minor actinides (MA). CRIEPI has been developing pyro-processing technology since late eighties with both domestic and international collaborations. In the early stage, electrochemical and thermodynamic properties in LiCl-KCl eutectic melt, and fundamental feasibility of core technology like electrorefining were chiefly investigated. Currently, stress in the process chemistry development is also placed on supporting technologies, such as treatment of anode residue and high temperature distillation for cathode product from electrorefining, and so on. Waste treatment process development, such as studies on adsorption behavior of various FP elements into zeolite and conditions for the fabrication of glass-bonded sodalite waste form, are steadily improved as well. In parallel, dedicated pyro-processing equipment such as zeolite column for treatment of spent electro-refiner salt is currently in progress. Recently, an integrated engineering-scale fuel cycle tests were performed funded by Japanese government (MEXT) as an important step before proceeding to large scale hot demonstration of pyro-processing. Oxide fuels can be readily introduced into the pyro-processing by reducing them to metals by adoption of electrochemical reduction technique. Making use of this advantage, the pyro-processing is currently under preliminary evaluation for its applicability to the treatment of the corium, mainly consisting of (U,Zr)O{sub 2}, formed in different composition during the accident of the Fukushima Daiichi nuclear power plant. (authors)

  5. Malignant atrophic papulosis (Köhlmeier-Degos disease - A review

    Directory of Open Access Journals (Sweden)

    Theodoridis Athanasios

    2013-01-01

    Full Text Available Abstract Definition of the disease Malignant atrophic papulosis (MAP, described independently by Köhlmeier and Degos et al., is a rare, chronic, thrombo-obliterative vasculopathy characterized by papular skin lesions with central porcelain-white atrophy and surrounding teleangiectatic rim. Epidemiology Less than 200 cases have been described in the literature. The first manifestation of MAP usually occurs between the 20th and 50th year of life. Clinical description The cutaneous clinical picture is almost pathognomonic. The histology is not consistent but in most cases it shows a wedge-shaped connective tissue necrosis in the deep corium due to a thrombotic occlusion of the small arteries. In the systemic variant, manifestations mostly occur at the intestine and central nervous system. Etiology The etiopathogenesis of the disease remains unknown, a genetic predisposition may occur. Vasculitis, coagulopathy or primary dysfunction of the endothelial cells have been implicated. Diagnostic methods Diagnosis is only based on the characteristic skin lesions. Differrential diagnosis It depends on the clinical presentation of MAP, but systemic lupus erythematosus and other connective tissue diseases need to be considered. Management No effective treatment exists for the systemic manifestations, while compounds that facilitate blood perfusion have achieved a partial regression of the skin lesions in single cases. Prognosis An apparently idiopathic, monosymptomatic, cutaneous, benign variant and a progressive, visceral one with approx. 50% lethality within 2–3 years have been reported. Systemic manifestations can develop years after the occurrence of skin lesions leading to bowel perforation and peritonitis, thrombosis of the cerebral arteries or massive intracerebral hemorrhage, meningitis, encephalitis, radiculopathy, myelitis.

  6. Latest findings from the OECD Rasplav Project

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.

    1997-01-01

    During the late phase of a severe accident in a light water reactor (current and future designs of BWRs, PWRs and VVERs), a significant amount of core material may relocate downward to the lower head of the reactor vessel. If molten core materials were to relocate to the lower head of the reactor pressure vessel (RPV), a molten pool consisting primarily of a mixture of ZrO{sub 2} and UO{sub 2} and some combination of a metal would form on the lower head. A solid crust of material would form around the boundaries of the pool, but internal heat generation resulting from radioactive decay of fission products would assure that most of the pool remains molten. In fact, the molten pool would undergo significant internal natural convection which would reach steady state conditions in about a few hours. Detailed understanding of all aspects of this natural convection process, in conjunction with the thermal boundary conditions imposed on the outer surface, determines the fraction of the total heat dissipation that is transferred through the upper crust to the inside of the reactor vessel by radiative heat exchange and the fraction which must be conducted through the wall of the reactor vessel lower head. This distribution is critical in determining whether and under what conditions the molten material can be cooled and retained in the reactor pressure vessel. The OECD Rasplav Project was established in 1994 as a three year program to study molten pool behavior and its interactions with structural materials in the lower head. This paper reviews the establishment of the project, its initial studies and proposed experimental testing, and the construction, preparation, and actual testing of a chamber of corium heated to well above liquid temperature.

  7. Validation of NEPTUNE-CFD on ULPU-V experiments

    Energy Technology Data Exchange (ETDEWEB)

    Jamet, Mathieu, E-mail: mathieu.jamet@edf.fr; Lavieville, Jerome; Atkhen, Kresna; Mechitoua, Namane

    2015-11-15

    In-vessel retention (IVR) of molten corium through external cooling of the reactor pressure vessel is one possible means of severe accident mitigation for a class of nuclear power plants. The aim is to successfully terminate the progression of a core melt within the reactor vessel. The probability of success depends on the efficacy of the cooling strategy; hence one of the key aspects of an IVR demonstration relates to the heat removal capability through the vessel wall by convection and boiling in the external water flow. This is only possible if the in-vessel thermal loading is lower than the local critical heat flux expected along the outer wall of the vessel, which is in turn highly dependent on the flow characteristics between the vessel and the insulator. The NEPTUNE-CFD multiphase flow solver is used to obtain a better understanding at local scale of the thermal hydraulics involved in this situation. The validation of the NEPTUNE-CFD code on the ULPU-V facility experiments carried out at the University of California Santa Barbara is presented as a first attempt of using CFD codes at EDF to address such an issue. Two types of computation are performed. On the one hand, a steady state algorithm is used to compute natural circulation flow rates and differential pressures and, on the other, a transient algorithm computation reveals the oscillatory nature of the pressure data recorded in the ULPU facility. Several dominant frequencies are highlighted. In both cases, the CFD simulations reproduce reasonably well the experimental data for these quantities.

  8. Cavity structural integrity evaluation of steam explosion using LS-DYNA

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dae-Young; Park, Chang-Hwan [FNC Technology Co. Ltd., Yongin (Korea, Republic of); Kim, Kap-sun [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    For investigating the mechanical response of the newly-designed NPP against an steam explosion, the cavity structural integrity evaluation was performed, in which the mechanical load resulted from a steam explosion in the reactor cavity was calculated. In the evaluation, two kinds of approach were considered, one of which is a deterministic manner and the other is a probabilistic one. In this report, the procedure and the results of the deterministic analysis are presented When entering the severe accident, the core is relocated to the lower head. In this case, an Ex-Vessel Steam Explosion(EVSE) can occur. It can threaten the structural integrity of the cavity due to the load applied to the walls or slabs of the cavity. The large amount of the energy transmitted from interaction between the molten corium and the water causes a dynamic loading onto the concrete walls resulting not only to affect the survivability of the various equipment but also to threaten the integrity of the containment. In this report, the response of the cavity wall structure is analyzed using the nonlinear finite element analysis (FEA) code. The resulting stress and strain of the structure were evaluated by the criteria in NEI07-13. Until now, deterministic analysis was performed via finite element analysis for the dynamic load generated by the steam explosion to investigate the effect on the cavity structure. A deterministic method was used in this study using the specific values of material properties and clearly defined steam explosion pressure curve. The results showed that the rebar and the liner are kept intact even at the high pressure pulse given by the steam explosion. The liner integrity is more critical to judge the preservation of the lean-tightness. In the meantime, there were found cracks in concrete media.

  9. Contribution to the study of thermal-hydraulic problems in nuclear reactors; Contribution a l`etude de problemes de thermohydraulique dans les reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Cognet, G

    1998-07-07

    In nuclear reactors, whatever the type considered, Pressurized Water Water Reactors (PWRs), Fast Breeder reactors (FBRs)..., thermal-hydraulics, the science of fluid mechanics and thermal behaviour, plays an essential role, both in nominal operating and accidental conditions. Fluid can either be the primary fluid (liquid or gas) or a very specific fluid called corium, which, in case of severe accident, could result from core and environning structure melting. The work reported here represents a 20-year contribution to thermal-hydraulic issues which could occur in FBRs and PWRs. Working on these two types of reactors, both in nominal and severe accident situations, has allowed me to compare the problems and to realize the importance of communication between research teams. The evolution in the complexity of studied problems, unavoidable in order to reduce costs and significantly improve safety, has led me from numerical modelling of single-phase flow turbulence to high temperature real melt experiments. The difficulties encountered in understanding the observed phenomena and in increasing experimental databases for computer code qualification have often entailed my participation in specific measurement device developments or adaptations, in particular non-intrusive devices generally based on optical techniques. Being concerned about the end-use of this research work, I actively participated in `in-situ` thermalhydraulic experiments in the FBRs: Phenix and Super-Phenix, of which I appreciated their undeniable scientific contribution. In my opinion, the thermal-hydraulic questions related to severe accidents are the most complex as they are at the cross-roads of several scientific specialities. Consequently, they require a multi-disciplinary approach and a continuous see-saw motion between experimentalists and modelling teams. After a brief description of the various problems encountered, the main ones are reported. Finally, the importance for research teams to

  10. A computer code for analysis of severe accidents in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)

  11. [Skin slice used as a model for investigating acupuncture effects].

    Science.gov (United States)

    Wang, Li-Na; Schwarz, Wolfgang; Gu, Quan-Bao; Ding, Guang-Hong

    2009-10-01

    To establish an acupoint-connective tissue model for studying the mechanism of acupuncture by using in vitro patch clamp technique in the rat skin slices. The local connective tissue under the corium of "Housanli" (ST 36) area from SD rat was acutely and bluntly separated and fixed in a chamber filled with artificial incubation solution. Mast cells in the prepared connective tissue slice were labeled by toluidine blue (TB) or neutral red (NR). The whole-cell current of mast cells responding to pressure stimulation applied through a patch pipette was recorded in rat slices derived from acupoint ST 36 area by using in vitro patch-clamp technique. 1) After staining with TB and NR, the labeled mast cells were found to distribute in the extracellular matrix of the connective tissue samples, and their degranulation phenomenon could be seen clearly. 2) The whole-cell current of mast cells in response to mechanical stress stimulation was successfully recorded in the connective tissue slices of the rat acupoint ST 36 area. The cellular membrane currents increased evidently when pressure gradients of -30, -60 or -90 cmH2O were applied to the recorded mast cells. The connective tissue slice from the rat ST 36 area may be used as a model for investigating the peripheral mechanism of acupuncture by combining the microtechniques and electrophysiological techniques. The results obtained in this model prove for the first time by electrophysiology that the mast cells in the connective tissue are probably involved in the transduction process of the mechanical signal from acupuncture stimulation. This new model provides a base for investigating the characters of the cells, collagen fibers, proteoglycans, etc. and their interactions in the acupoint connective tissue in the future.

  12. Non-Invasive Delivery of dsRNA into De-Waxed Tick Eggs by Electroporation

    Science.gov (United States)

    Ruiz, Newton; de Abreu, Leonardo Araujo; Parizi, Luís Fernando; Kim, Tae Kwon; Mulenga, Albert; Braz, Gloria Regina Cardoso; Vaz, Itabajara da Silva; Logullo, Carlos

    2015-01-01

    RNA interference-mediated gene silencing was shown to be an efficient tool for validation of targets that may become anti-tick vaccine components. Here, we demonstrate the application of this approach in the validation of components of molecular signaling cascades, such as the Protein Kinase B (AKT) / Glycogen Synthase Kinase (GSK) axis during tick embryogenesis. It was shown that heptane and hypochlorite treatment of tick eggs can remove wax, affecting corium integrity and but not embryo development. Evidence of AKT and GSK dsRNA delivery into de-waxed eggs of via electroporation is provided. Primers designed to amplify part of the dsRNA delivered into the electroporated eggs dsRNA confirmed its entry in eggs. In addition, it was shown that electroporation is able to deliver the fluorescent stain, 4',6-diamidino-2-phenylindole (DAPI). To confirm gene silencing, a second set of primers was designed outside the dsRNA sequence of target gene. In this assay, the suppression of AKT and GSK transcripts (approximately 50% reduction in both genes) was demonstrated in 7-day-old eggs. Interestingly, silencing of GSK in 7-day-old eggs caused 25% reduction in hatching. Additionally, the effect of silencing AKT and GSK on embryo energy metabolism was evaluated. As expected, knockdown of AKT, which down regulates GSK, the suppressor of glycogen synthesis, decreased glycogen content in electroporated eggs. These data demonstrate that electroporation of de-waxed R. microplus eggs could be used for gene silencing in tick embryos, and improve the knowledge about arthropod embryogenesis. PMID:26091260

  13. Phenomenological Studies on Melt-Structure-Water Interactions (MSWI) during Postulated Severe Accidents: Year 2004 Activity. APRI 5 report

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R.; Park, H.S.; Nayak, A.K.; Hansson, R.C.; Chiferaw, D.; Stepanyan, A.; Rao, R.S.; Karbojian, A. [Royal Inst. of Technology, Stockholm (Sweden). Div. of Nuclear Power Safety

    2005-04-01

    This report presents descriptions of the major results obtained in the research program 'Melt-Structure-Water Interaction (MSWI)' at NPS/RIT during the year 2004. The primary objectives of the MSWI Project in year 2004 were to study (1) the in-vessel and exvessel melt/debris bed coolability process when melt is flooded with water, and (2) the energetics and characteristics of steam explosions. Our general approaches are to establish scaling relationships so that the data obtained in the experiments could be extended to prototypical accident geometries and conditions, develop phenomenological or computational models for the processes under investigation and validate the existing and newly developed models against data obtained at RIT and at other laboratories. In 2004, several experimental programs, such as the COMECO (Corium MElt COolability), POMECO (POrous MEdia COolability) and MISTEE (Micro-Interactions in STeam Explosion Experiments) programs were continued. The SIMECO (Simulation of MElt Coolability) program was restarted in 2004. The construction of the POMECO-GRAND (POrous MEdia COolability) facility was delayed due to lack of finances. However, existing POMECO facility was modified to study 3-D effects on debris coolability. In this report, the results from the COMECO experiment with high temperature oxidic melt, from the POMECO experiments for the multi-dimensional effects on debris bed coolability, from the SIMECO experiment for three-layer pool configuration and from the MISTEE experiments for steam explosion characteristics and loads are described. For analytical efforts, results from the COMETA code for the entire process of the steam explosions are discussed.

  14. Investigation of the structure of debris beds formed from fuel rods fragmentation

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, Duc-Hanh; Fichot, Florian; Topin, Vincent, E-mail: vincent.topin@irsn.fr

    2017-03-15

    This paper is a study of debris beds that can form in the core of a nuclear power plant under severe accident conditions. Such beds are formed of fragments of pellets and cladding remnants, as observed in the TMI-2 core. Many important issues are related with the morphology of those debris beds: are they coolable in case of water injection and how does molten corium progress through them if they are not coolable? The answers to those questions depend on the structure of the debris bed: porosity, number and arrangement of particles. In order to obtain relevant information, a numerical simulation of the formation of the debris bed is proposed. It relies on a granular approach of the type called “Contact Dynamics” to simulate the collapse of debris and their accumulation. Two different schemes of fuel pellet fragmentation are considered and simulations for different degrees of fragmentation of the pellets are performed. The results show that the number of axial cracks on fuel pellets strongly influences the final porosity of the debris bed. Porosities vary between 31% (less coolable cases) and 45% (similar to TMI-2 observations), with a most probable configuration around 41%. The specific surface of the bed is also evaluated. In the last part, a simple model is used to estimate the impact of the variation in geometry of the numeric debris beds on their flow properties. We show that the permeability and passability can vary respectively with a range of 30% and 15% depending on the number of fragment per pellet. The other benefits of the approach are finally discussed. Among them, the possibility to print 3D samples from the calculated images of debris beds appears as a promising perspective to perform experiments with realistic debris beds.

  15. A wound healing model with sonographic monitoring.

    Science.gov (United States)

    Hoffmann, K; Winkler, K; el-Gammal, S; Altmeyer, P

    1993-05-01

    The methods used hitherto for quantification of skin repair processes only allow an examiner a two-dimensional assessment of superficial wound healing. With the recent advent of high frequency B-scan ultrasonography in dermatology it has become possible to follow the course of healing and evaluate the healing processes in deeper layers of the skin. In this investigation 80 patients received cryosurgery for treatment of basal cell carcinomas on the face or neck region. As the size of cryosurgical defects can be precisely controlled they are potentially useful as standardized wound healing models. The course of wound healing after cryosurgery using a digital ultrasound scanner (DUB 20, Taberna pro medicum, Lüneburg, Germany) was monitored. The usable depth of penetration of the echo signal is approximately 7 mm. The lateral resolution is approximately 200 microns, the axial resolution approximately 80 microns. The cryolesion and the repair processes were examined ultrasonographically and clinically over a period of at least 3 weeks or until the wound had completely healed. The depth of invasion and lateral extent of the basal cell carcinoma as well as the size of the induced cryolesion can be determined by ultrasound. The exudative phase after cryosurgery, with developing oedema and necrosis, can be quantified on the basis of the reduced reflectivity in the corium. The repair processes taking place in the region of necrosis can be visualized in the ultrasound scan. The ultrasonically monitored wound healing model which we have demonstrated is particularly suitable for investigating the efficacy of drugs which promote healing.

  16. Fission-product releases from a PHWR terminal debris bed

    Energy Technology Data Exchange (ETDEWEB)

    Brown, M.J.; Bailey, D.G., E-mail: morgan.brown@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    During an unmitigated severe accident in a pressurized heavy water reactor (PHWR) with horizontal fuel channels, the core may disassemble and relocate to the bottom of the calandria vessel. The resulting heterogeneous in-vessel terminal debris bed (TDB) would likely be quenched by any remaining moderator, and some of the decay heat would be conducted through the calandria vessel shell to the surrounding reactor vault or shield tank water. As the moderator boiled off, the solid debris bed would transform into a more homogeneous molten corium pool located between top and bottom crusts. Until recently, the severe accident code MAAP-CANDU assumed that unreleased volatile and semi-volatile fission products remained in the TDB until after calandria vessel failure, due to low diffusivity through the top crust and the lack of gases or steam to flush released fission products from the debris. However, national and international experimental results indicate this assumption is unlikely; instead, high- and medium-volatility fission products would be released from a molten debris pool, and their volatility and transport should be taken into account in TDB modelling. The resulting change in the distribution of fission products within the reactor and containment, and the associated decay heat, can have significant effects upon the progression of the accident and fission-product releases to the environment. This article describes a postulated PHWR severe accident progression to generate a TDB and the effects of fission-product releases from the terminal debris, using the simple release model in the MAAP-CANDU severe accident code. It also provides insights from various experimental programs related to fission-product releases from core debris, and their applicability to the MAAP-CANDU TDB model. (author)

  17. Ablation characteristics of special concrete due to an impinging zirconium-dioxide melt jet

    Energy Technology Data Exchange (ETDEWEB)

    An, S.M., E-mail: sangmoan@kaeri.re.kr; Ha, K.S.; Min, B.T.; Kim, H.Y.; Song, J.H.

    2015-04-01

    Highlights: • The jet impingement tests were performed for a special concrete of core-catcher. • The ablation rate and depth were measured 1.59 mm/s and 4.33 mm, respectively. • The experimental results were estimated well between the model prediction bounds. • The material ablation was described reasonably by a convective heat transfer model. - Abstract: Jet impingement experiments were performed to investigate the ablation characteristics of special concrete, which has been developed as one of the candidate protecting materials for the EU-APR1400 ex-vessel core catcher. In order to simulate the jet impingement phenomenon owing to the reactor vessel failure during a severe core meltdown accident, the experimental facility was established and the experimental conditions were determined based on parametric studies. The special concrete specimen was manufactured in accordance with the standard procedures, and its microstructures and physicochemical properties were analyzed to verify the requirements for the qualification. An induction melting technique in a cold crucible was employed to generate the zirconium-dioxide melt as a simulant of the corium melt. The special concrete was ablated uniformly over the impact area by jet impingement, and the average ablation depth was measured to be 4.33 mm. The average ablation rate in depth was evaluated as 1.59 mm/s using the temperature measurements of the specimen. As compared with the predictions by the models based on the convective and radiative heat transfer analysis, both the measured ablation rate and depth were estimated appropriately within the bounds of their limits. However, the convective heat transfer model turned out to predict the ablation characteristics of the special concrete more reasonably during the jet impingement even though some water content within the special concrete could lead to a sudden generation of the steam layer through which the material ablation is attenuated substantially by the

  18. Comparison of MCCI Analysis Results using the Newest MAAP5

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Mi Ro; Kim, Hyeong Taek [KHNP-CRI, Daejeon (Korea, Republic of)

    2013-10-15

    In this study, we try to find the improvement in the newest MAAP5 MCCI model by comparing the results using the same sequence and the same condition those used in the previous paper. In this study, we can find that the problem which overestimated the concrete ablation for Basaltic concrete is somewhat resolved. And, in case that the cavity is flooded, it is confirmed that the debris coolability for Limestone and Limestone Common Sand concrete is maintained. But, in case of Basaltic concrete, though the interaction is somewhat inactive, but the debris Coolability is not maintained. We are planning to report these results to EPRI in order to confirm whether this phenomenon is appropriate. After the Fukushima accident, EPRI has developed the MAAP5 that is expected to make up the limitation of MAAP4. The newest version of MAAP5 is known as the Ver.5.0.2 (Build 5.01.2182, simply called 502D) which is published in July this year for final beta testing. In this, it is expected that so many models should be upgraded such as the Lower head plenum model, Debris Coolability model, Molten Core Concrete Interaction, Spent Fuel Pool model and Containment Heat Sink model, etc. During the severe accident progression, the molten corium ejected into the cavity reacted with the concrete in the cavity floor and the phenomenon is called MCCI (Molten Core Concrete Interaction). In the last KNS Spring Meeting, KHNP presented the MCCI analysis results according to the concrete composition using MAAP Ver. 5.0.2 (Build 5.01.1100, simply called 502B) published in April, 2012. In that report, we pointed out that the results of MCCI for Basaltic concrete was too much conservative, so we raised the issues related the MCCI model in the MUG meeting.

  19. A lepra tuberculoide, ou melhor, a lesão tuberculoide na lepra, representa uma fase de transição desta dermatose e não uma forma clínica autônoma: novos achados bacterioscópicos na linfa subcutânea de leprosos

    Directory of Open Access Journals (Sweden)

    H. C. de Souza-Araujo

    1943-08-01

    Full Text Available The A., after an intorductory history of his experience in leprosy, discusses the more convinient routine method of classification of leprosy cases, basing it in the facte that every case is mixt, i. e. when the skin shows any lesion the nerves of that region are also affected by the bacilli. Studying by a new thecnics, which he baptised before as "Lleras' method", the scarching of the agent of leprosy in tuberculoid cases, by examination of sub-corium lymph obtained from the lesion, he discovered new forms of the Hansen bacillus, which describes briefly, arriving at the following conclusions: 1. The A., after discussing about the evolution and clinical classification of leprosy, describes new forms of the HANSIN bacillus, discoverd in the lymph extracted from subcutis of leprosy lesion. 2. In 100 % of tuberculoid cases (total studied 29 the A. found, in the subcutis lymph, bacilli, granules, clubs or other forms of HANSEN bacillus. 3. Such bacteriological findigs and the proved mutation of tuberculoid leprosy into lepromatous type, demolished the basis of the so-called "polar" classification of leprosy. 4. Considering the proved facts already referred to, the A. arrived at the conclusion that 50 % of all papers published about tuberculoid leprosy, within the last ten years, are fanciful. 5. The presence, in the subcutis of lepers, of metamorphosic forms of HANSEN bacillus, is the cause of common relapses of negativated cases by treatment, which fact suggests a new therapeutics method to destroy such elements in loco, and exiges more strict examination before release of interned patients.

  20. Experimental study of in-and-ex-vessel melt cooling during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Baik; Yoo, K. J.; Park, C. K.; Seok, S. D.; Park, R. J.; Yi, S. J.; Kang, K. H.; Ham, Y. S.; Cho, Y. R.; Kim, J. H.; Jeong, J. H.; Shin, K. Y.; Cho, J. S.; Kim, D. H.

    1997-07-01

    After code damage during a severe accident in a nuclear reactor, the degraded core has to be cooled down and the decay heat should be removed in order to cease the accident progression and maintain a stable state. The cooling of core melt is divided into in-vessel and ex-vessel cooling depending on the location of molten core which is dependent on the timing of vessel failure. Since the cooling mechanism varies with the conditions of molten core and surroundings and related phenomena, it contains many phenomenological uncertainties so far. In this study, an experimental study for verification of in-vessel corium cooling and several separate effect experiments for ex-vessel cooling are carried out to verify in- and ex-vessel cooling phenomena and finally to develop the accident management strategy and improve engineered reactor design for the severe accidents. SONATA-IV (Simulation of Naturally Arrested Thermal Attack in Vessel) program is set up for in-vessel cooling and a progression of the verification experiment has been done, and an integral verification experiment of the containment integrity for ex-vessel cooling is planned to be carried out based on the separate effect experiments performed in the first phase. First phase study of SONATA-IV is proof of principle experiment and it is composed of LALA (Lower-plenum Arrested Vessel Attack) experiment to find the gap between melt and the lower plenum during melt relocation and to certify melt quenching and CHFG (Critical Heat Flux in Gap) experiment to certify heat transfer mechanism in an artificial gap. As separate effect experiments for ex-vessel cooling, high pressure melt ejection experiment related to the initial condition for debris layer formation in the reactor cavity, crust formation and heat transfer experiment in the molten pool and molten core concrete interaction experiment are performed. (author). 150 refs., 24 tabs., 127 figs.

  1. An experimental study of hypervapotron structure in external reactor vessel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Yufeng; Zhang, Ming [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Hou, Fangxin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China); Gao, Tianfang [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Chen, Peipei, E-mail: chenpeipei@snptc.com.cn [State Power Investment Group Corporation, Beijing (China)

    2016-07-15

    Highlights: • Experiments are performed to study the application of hypervapotron in ERVC design. • CHF experiments on two surfaces are conducted under different flow conditions. • Hypervapotron improves CHF performance by 40–60% compared with smooth surface. • Visualization shows fin structure removes vapor mushroom for better liquid supply. - Abstract: In vessel retention (IVR) is one of the key strategies for many advanced LWR designs to mitigate postulated severe accidents. The success of IVR substantially relies on external reactor vessel cooling (ERVC) by which the decay heat is removed from the melt core in the reactor vessel lower head. The main challenge of IVR is to provide an adequate safety margin of ERVC against critical heat flux (CHF) of subcooled flow boiling in the reactor lower head flow channel. Due to uncertainties in corium melt pool configuration, large CHF margin of ERVC is usually required by regulatory authorities to demonstrate reliability of severe accident mitigation methods. Various CHF enhancement designs have been proposed and studied in literature. In this paper, an experimental study of hypervapotron structure as a novel design to improve CHF performance of ERVC is conducted. Hypervapotron is chosen as one of the potential engineering options for International Thermonuclear Experimental Reactor (ITER) program as a divertor structure to remove highly intense heat from fusion chamber. This study is to conduct CHF experiments at typical PWR ERVC working conditions. The CHF experiments are performed in a 30 mm by 61 mm rectangular flow channel with a 200 mm long heated surface along the flow direction. Both smooth and hypervapotron surface are tested at various inclination angles of the test section to simulate various positions of the reactor lower head. The hypervapotron is found to have a 40–60% CHF improvement compared with the smooth surface. The high speed visualization indicates that hypervapotron is able to

  2. Ex-vessel coolability and energetics of steam explosions in nordic light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Park, H.S.; Dinh, T.N. [Royal Institute of Technology (Sweden)

    2007-04-15

    The report summarizes activities conducted at the Division of Nuclear Power Safety, Royal Institute of Technology-Sweden (KTH-NPS) within the ExCoolSe project during the year 2005, which is a transition year for the KTH-NPS program. The ExCoolSe project supported by NKS contributes to the severe accident research at KTH-NPS concurrently supported by APRI, HSK and EU SARNET. The main objective in ExCoolSe project is to scrutinize research on risk-significant safety issues related to severe accident management (SAM) strategy adopted for Nordic BWR plants, namely the Ex-vessel Coolability and Energetic Steam explosion. The work aims to pave way toward building a tangible research framework to tackle these long-standing safety issues. Chapter 1 describes the project objectives and work description. Chapter 2 provides a critical assessment of research results obtained from several past programs at KTH. This includes review of key data, insights and implications from POMECO (Porous Media Coolability) program, COMECO (Corium Melt Coolability) program, SIMECO (Study of In-Vessel Melt Coolability) program, and MISTEE (Micro-Interactions in Steam Explosion Experiments) program. Chapter 3 discusses the rationale of the new research program focusing on the SAM issue resolution. The program emphasizes identification and qualification of physics-based limiting mechanisms for both in-vessel phenomena (melt progression and debris coolability in the lower head, vessel failure), and ex-vessel phenomena. Chapter 4 introduces research results from the newly established DEFOR (Debris Formation) program and the ongoing MISTEE program. The focus of DEFOR is fulfill an apparent gap in the contemporary knowledge of severe accidents, namely mechanisms which govern the debris bed formation and bed characteristics. The later control the debris bed coolability. In the MISTEE program, methods for image synchronization and data processing were developed and tested, which enable processing of

  3. OECD MCCI 2-D Core Concrete Interaction (CCI) tests : CCI-2 test data report-thermalhydraulic results, Rev. 0 October 15, 2004.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Lomperski, S.; Kilsdonk, D. J.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division); (NRC)

    2011-05-23

    The Melt Attack and Coolability Experiments (MACE) program addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two program objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of satisfying these objectives, the Management Board (MB) approved the conduct of two long-term 2-D Core-Concrete Interaction (CCI) experiments designed to provide information in several areas, including: (i) lateral vs. axial power split during dry core-concrete interaction, (ii) integral debris coolability data following late phase flooding, and (iii) data regarding the nature and extent of the cooling transient following breach of the crust formed at the melt-water interface. This data report provides thermal hydraulic test results from the CCI-2 experiment, which was conducted on August 24, 2004. Test specifications for CCI-2 are provided in Table 1-1. This experiment investigated the interaction of a fully oxidized 400 kg

  4. 改良微小切口腋臭根治术%Improved radical treatment for bromidrosis with a tiny incision

    Institute of Scientific and Technical Information of China (English)

    王自谦; 石成方; 常冬青; 谷廷敏

    2012-01-01

    Objective To present a improved radical treatment for bromidrosis with a tiny incision. Methods To observe clinical efficiency of 82 bromidrosis patients from Feb 2010 to Jun 2011. Under tumescent anaesthesia.one tiny incisions about 1.0cm in armpit creases were perfomed.Then segregated between corium and adipose layer to border of the designed region.trim the skin flap to Thick skin grafts thickness.AII patients were followed for six months to two years.Results Among the 82 patients of this group,81 patients had primary healing and 1 case of smell residue,and cured after revision surgery. 1 case of subcutaneous extravasated blood,was cured by compression bandaging after ruling out the blood.AII patients healed cosmetically with scar concealed.Conclusion The surgery technique designed in this trial is a safe and facile treatment of bromidros.Therefore, it should be extensively used in clinic.%目的:介绍一种改良微小切口腋臭根治术.方法:对2010年2月~ 2011年6月我科门诊腋臭患者82例,在肿胀麻醉下沿腋窝皱襞作一1.0cm微小切口,盲视下钝性分离皮瓣至设计范围,修剪皮瓣至中厚皮片厚度,随访患者6~ 24个月.结果:81例患者腋臭完全消失,异味残留1例,再次行此术后治愈;术后发生皮下瘀血1例,排除淤血后加压包扎,愈合良好.所有患者术后瘢痕隐蔽,外形美观.结论:本术式是一种微创、简便、安全、疗效确切的腋臭根治技术,值得临床推广应用.

  5. COOLOCE debris bed experiments and simulations investigating the coolability of cylindrical beds with different materials and flow modes

    Energy Technology Data Exchange (ETDEWEB)

    Takasuo, E.; Kinnunen, T.; Holmstroem, S.; Lehtikuusi, T. [VTT Technical Research Centre of Finland (Finland)

    2013-07-15

    The COOLOCE experiments aim at investigating the coolability of debris beds of different geometries, flow modes and materials. A debris bed may be formed of solidified corium as a result of a severe accident in a nuclear power reactor. The COOLOCE-8 test series consisted of experiments with a top-flooded test bed with irregular gravel as the simulant material. The objective was to produce comparison data useful in estimating the effects of different particle materials and the possible effect of the test arrangement on the results. It was found that the dryout heat flux (DHF) measured for the gravel was lower compared to previous experiments with spherical beads, and somewhat lower compared to the early STYX experiments. The difference between the beads and gravel is at least partially explained by the smaller average size of the gravel particles. The COOLOCE-9 test series included scoping experiments examining the effect of subcooling of the water pool in which the debris bed is immersed. The experiments with initially subcooled pool suggest that the subcooling may increase DHF and increase coolability. The aim of the COOLOCE-10 experiments was to investigate the effect of lateral flooding on the DHF a cylindrical test bed. The top of the test cylinder and its sidewall were open to water infiltration. It was found that the DHF is increased compared to a top-flooded cylinder by more than 50%. This suggests that coolability is notably improved. 2D simulations of the top-flooded test beds have been run with the MEWA code. Prior to the simulations, the effective particle diameter for the spherical beads and the irregular gravel was estimated by single-phase pressure loss measurements performed at KTH in Sweden. Parameter variations were done for particle size and porosity used as input in the models. It was found that with the measured effective particle diameter and porosity, the simulation models predict DHF with a relatively good accuracy in the case of spherical

  6. Analysis of KROTOS KS-2 and KS-4 steam explosion experiments with TEXAS-VI

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Ronghua, E-mail: rhchen@mail.xjtu.edu.cn [State Key Laboratory of Multiphase Flow in Power Engineering, School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Wang, Jun [Nuclear Engineering and Engineering Physics, College of Engineering, University of Wisconsin Madison, WI 53706 (United States); Su, G.H.; Qiu, Suizheng [State Key Laboratory of Multiphase Flow in Power Engineering, School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Corradini, M.L., E-mail: Corradini@engr.wisc.edu [Nuclear Engineering and Engineering Physics, College of Engineering, University of Wisconsin Madison, WI 53706 (United States)

    2016-12-01

    Highlights: • The KS-2 and KS-4 steam explosion experiments were analyzed by TEXAS-VI. • The coarse mixing status up to the explosion triggering time was well predicted by TEXAS-VI. • The predicted dynamic explosion pressure was in good agreement with the experimental results. - Abstract: TEXAS-VI is a transient, three-field, one-dimensional mechanistic model for the steam explosion phenomena. A fuel solidification model and associated fragmentation criteria of the solidifying particle for both the mixing phase and explosion phase were developed and incorporated into TEXAS-VI to account for solidification. In the present study, TEXAS-VI was used to analyze the KS-2 and KS-4 steam explosion experiments, which were performed in the KROTOS facility as part of the OECD-SERENA-2 program. In the simulation, the KROTOS experimental facility was modeled as Eulerian control volumes based on the facility geometry. The molten corium jet was divided up into a series of LaGrangian master particles equal to the initial jet diameter. Both the mixing phase and the explosion phase of the experiments were simulated by TEXAS-VI. Comparison to test data indicates that the fuel jet kinematics and the vapor volume during the mixing phase were well predicted by TEXAS-VI. The TEXAS-VI prediction of the dynamic explosion pressure at different axial locations in the test was also in good agreement with the experimental results. The maximum pressure of KS-2 and KS-4 predicted by TEXAS-VI were 16.7 MPa and 41.9 MPa, respectively. The KS-4 maximum steam explosion pressure predicted by TEXAS-VI was higher than that of KS-2, which was consistent with experiment observation. The observed differences of the dynamic explosion pressure between the KS-2 and KS-4 experiments were also successfully simulated by TEXAS-VI. This suggests that TEXAS-VI is able to analyze the effect of prototypic melt compositions on the steam explosion phenomena. Additional benchmarking and evaluations are ongoing.

  7. OECD MCCI project 2-D Core Concrete Interaction (CCI) tests : CCI-3 test data report-thermalhydraulic results. Rev. 0 October 15, 2005.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Lomperski, S.; Kilsdonk, D. J.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division); (NRC)

    2011-05-23

    The Melt Attack and Coolability Experiments (MACE) program addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two program objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of satisfying these objectives, the Management Board (MB) approved the conduct of a third long-term 2-D Core-Concrete Interaction (CCI) experiment designed to provide information in several areas, including: (i) lateral vs. axial power split during dry core-concrete interaction, (ii) integral debris coolability data following late phase flooding, and (iii) data regarding the nature and extent of the cooling transient following breach of the crust formed at the melt-water interface. This data report provides thermal hydraulic test results from the CCI-3 experiment, which was conducted on September 22, 2005. Test specifications for CCI-3 are provided in Table 1-1. This experiment investigated the interaction of a fully oxidized 375

  8. Ranking of severe accident research priorities

    Energy Technology Data Exchange (ETDEWEB)

    Schwinges, B. [Gesell Anlagen and Reaktorsicherheit GRS mbH, D-50667 Cologne (Germany); Journeau, C. [CEA Cadarache, DEN STRI LMA, F-13115 St Paul Les Durance (France); Haste, T. [Paul Scherrer Inst, NES LTH, OVGA 312, CH-5232 Villigen (Switzerland); Meyer, L.; Tromm, W. [Forschungszentrum Karlsruhe, D-76021 Karlsruhe (Germany); Trambauer, K. [GRS mbH, Forschungsgelande, D-85748 Garching (Germany)

    2010-07-01

    The objectives of the SARNET network are to define common research programmes in the field of severe accidents and to develop common computer tools and methodologies for safety assessment in this field. To reach these objectives, one of the work packages, named 'Severe Accident Research Priorities' (SARP), aimed at reviewing and reassessing the priorities of research issues as a basis to harmonize and to re-orient research programmes, to define new ones, and to close - if possible - resolved issues on a common basis. The work was performed in close collaboration with 8 participating institutions, led by GRS, representing technical safety organisations, industry and utilities (IRSN, CEA, EDF, FZK, GRS, KTH, TUS, VTT). This action made use notably of (1) the outcomes of the EURSAFE project in the 5. Framework Programme, i. e. the Phenomena Identification and Ranking Tables (PIRT) on severe accidents, (2) the results of the validation and benchmarking activities on ASTEC, (3) the results of reactor calculations carried out in the other SARNET tasks, and (4) the outcome of the research performed in the three thematic sub-domains of SARNET (corium, containment and source term). The main outcome of EURSAFE was a list of 21 topics which included recommendations for experimental programmes and code developments. This list formed the basis of the work in SARP. Also the methodology applied in EURSAFE to consider both the risk potential and the severe accident issues where large uncertainties still subsist was adopted. The analyses of the progress of research and development activities considered whether (1) any research issue was resolved due to reduction of uncertainties or gain of scientific insights, (2) any new issue had to be added to the list of needed research, (3) any new process or phenomenon had to be included in the general PIRT list taking into account the safety relevance and the lack of knowledge, and (4) any new accident management program has to be

  9. Improved solidification influence modelling for Eulerian fuel-coolant interaction codes

    Energy Technology Data Exchange (ETDEWEB)

    Ursic, Mitja, E-mail: mitja.ursic@ijs.s [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Leskovar, Matjaz; Mavko, Borut [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

    2011-04-15

    Steam explosion experiments revealed important differences in the efficiency between simulant alumina and oxidic corium melts. The experimentally observed differences are importantly attributed to the differences in the melt droplets solidification and void production, which are limiting phenomena in the steam explosion process and have to be adequately modelled in fuel-coolant interaction codes. This article focuses on the modelling of the solidification effect. An improved solidification influence modelling approach for Eulerian fuel-coolant interaction codes was developed and is presented herein. The solidification influence modelling in fuel-coolant interaction codes is strongly related to the modelling of the temperature profile and the mechanical effect of the crust on the fragmentation process. Therefore the first objective was to introduce an improved temperature profile modelling and a fragmentation criterion for partly solidified droplets. The fragmentation criterion was based on the established modified Weber number, which considers the crust stiffness as a stabilizing force acting to retain the crust under presence of the hydrodynamic forces. The modified Weber number was validated on experimental data. The application of the developed improved solidification influence modelling enables an improved determination of the melt droplet mass, which can be efficiently involved in the fine fragmentation during the steam explosion process. Additionally, also the void production modelling is improved, because it is strongly related to the temperature profile modelling in the frame of the solidification influence modelling. Therefore the second objective was to enable an improved solidification influence modelling in codes with an Eulerian formulation of the droplet field. Two additional transported model parameters based on the most important droplets features regarding the fuel-coolant interaction behaviour, were derived. First, the crust stiffness was

  10. Analysis of flammability in the attached buildings to containment under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Rosa, J.C. de la, E-mail: juan-carlos.de-la-rosa-blul@ec.europa.eu [European Commission Joint Research Centre (Netherlands); Fornós, Joan, E-mail: jfornosh@anacnv.com [Asociación Nuclear Ascó-Vandellós (Spain)

    2016-11-15

    analyzed exercise provided three conditions are met: H{sub 2} and CO recombiner devices are found inside the containment; corium is submerged and cooled down to quenching by flooding the reactor cavity; and the containment remains isolated along the accident evolution so that gases flowing into attached buildings to containment are limited to the so-called allowable leakage.

  11. In-vessel coolability and steam explosion in Nordic BWRs

    Energy Technology Data Exchange (ETDEWEB)

    Ma, W.; Hansson, R.; Li, L.; Kudinov, P.; Cadinu, F.; Tran, C-.T. (Royal Institute of Technology (KTH), Stockholm (Sweden))

    2010-05-15

    The INCOSE project is to reduce the uncertainty in quantification of steam explosion risk and in-vessel coolability in Nordic BWR plants with the cavity flooding as a severe accident management (SAM) measure. During 2009 substantial advances and new insights into physical mechanisms were gained for studies of: (i) in-vessel corium coolability - development of the methodologies to assess the efficiency of the control rod guide tube (CRGT) cooling as a potential SAM measure; (ii) debris bed coolability - characterization of the effective particle diameter of multi-size particles and qualification of friction law for two-phase flow in the beds packed with multi-size particles; and (iii) steam explosion - investigation of the effect of binary oxides mixture's properties on steam explosion. An approach for coupling of ECM/PECM models with RELAP5 was developed to enhance predictive fidelity for melt pool heat transfer. MELCOR was employed to examine the CRGT cooling efficiency by considering an entire accident scenario, and the simulation results show that the nominal flowrate (approx10kg/s) of CRGT cooling is sufficient to maintain the integrity of the vessel in a BWR of 3900 MWth, if the water injection is activated no later than 1 hour after scram. The POMECO-FL experimental data suggest that for a particulate bed packed with multi-size particles, the effective particle diameter can be represented by the area mean diameter of the particles, while at high velocity (Re>7) the effective particle diameter is closer to the length mean diameter. The pressure drop of two-phase flow through the particulate bed can be predicted by Reed's model. The steam explosion experiments performed at high melt superheat (>200oC) using oxidic mixture of WO3-CaO didn't detect an apparent difference in steam explosion energetics and preconditioning between the eutectic and noneutectic melts. This points out that the next step of MISTEE experiment will be conducted at lower

  12. OECD 2-D Core Concrete Interaction (CCI) tests : CCI-2 test plan, Rev. 0 January 31, 2004.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Kilsdonk, D. J.; Lomperski, S.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division)

    2011-05-23

    The Melt Attack and Coolability Experiments (MACE) program addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two program objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of satisfying these objectives, the Management Board (MB) approved the conduct of two long-term 2-D Core-Concrete Interaction (CCI) experiments designed to provide information in several areas, including: (i) lateral vs. axial power split during dry core-concrete interaction, (ii) integral debris coolability data following late phase flooding, and (iii) data regarding the nature and extent of the cooling transient following breach of the crust formed at the melt-water interface. The first of these two tests, CCI-1, was conducted on December 19, 2003. This test investigated the interaction of a fully oxidized 400 kg PWR core melt, initially containing 8 wt % calcined siliceous concrete, with a specially designed two

  13. OECD MCCI project long-term 2-D molten core concrete interaction test design report, Rev. 0. September 30, 2002.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Kilsdonk, D. J.; Lomperski, S.; Aeschliman, R. W.; Basu, S. (Nuclear Engineering Division); (NRC)

    2011-05-23

    The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following two technical objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of the first program objective, the Small-Scale Water Ingression and Crust Strength (SSWICS) test series has been initiated to provide fundamental information on the ability of water to ingress into cracks and fissures that form in the debris during quench, thereby augmenting the otherwise conduction-limited heat transfer process. A test plan for Melt Eruption Separate Effects Tests (MESET) has also been developed to provide information on the extent of crust growth and melt eruptions as a function of gas sparging rate under well-controlled experiment conditions. In terms of the second program objective, the project Management Board (MB) has approved startup activities required to carry out

  14. OECD MMCI 2-D Core Concrete Interaction (CCI) tests : CCCI-1 test data report-thermalhydraulic results. Rev 0 January 31, 2004.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Lomperski, S.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division)

    2011-05-23

    The Melt Attack and Coolability Experiments (MACE) program addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten coreconcrete interactions under both wet and dry cavity conditions. Achievement of these two program objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of satisfying these objectives, the Management Board (MB) approved the conduct of two long-term 2-D Core-Concrete Interaction (CCI) experiments designed to provide information in several areas, including: (i) lateral vs. axial power split during dry core-concrete interaction, (ii) integral debris coolability data following late phase flooding, and (iii) data regarding the nature and extent of the cooling transient following breach of the crust formed at the melt-water interface. This data report provides thermal hydraulic test results from the CCI-1 experiment, which was conducted on December 19, 2003. Test specifications for CCI-1 are provided in Table 1-1. This experiment investigated the interaction of a fully oxidized 400 kg

  15. OECD MCCI 2-D Core Concrete Interaction (CCI) tests : CCI-2 test data report-thermalhydraulic results, Rev. 0 October 15, 2004.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Lomperski, S.; Kilsdonk, D. J.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division); (NRC)

    2011-05-23

    The Melt Attack and Coolability Experiments (MACE) program addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two program objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of satisfying these objectives, the Management Board (MB) approved the conduct of two long-term 2-D Core-Concrete Interaction (CCI) experiments designed to provide information in several areas, including: (i) lateral vs. axial power split during dry core-concrete interaction, (ii) integral debris coolability data following late phase flooding, and (iii) data regarding the nature and extent of the cooling transient following breach of the crust formed at the melt-water interface. This data report provides thermal hydraulic test results from the CCI-2 experiment, which was conducted on August 24, 2004. Test specifications for CCI-2 are provided in Table 1-1. This experiment investigated the interaction of a fully oxidized 400 kg

  16. Melt dispersion and direct containment heating (DCH) experiments in the DISCO-H test facility

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, L.; Albrecht, G.; Kirstahler, M.; Schwall, M.; Wachter, E.; Woerner, G.

    2004-05-01

    The DISCO-H Test Facility at Forschungszentrum Karlsruhe was set up to perform scaled experiments that simulate melt ejection scenarios under low system pressure in Severe Accidents in Pressurized Water Reactors (PWR). These experiments are designed to investigate the fluid-dynamic, thermal and chemical processes during melt ejection out of a breach in the lower head of a PWR pressure vessel at pressures below 2 MPa with an iron-alumina melt and steam. In the past, a detailed study of pressure and geometry effects on the fluid dynamics of the melt dispersion process had been performed with cold model fluids in the facility DISCO-C. The main components of the facility are scaled about 1:18 linearly to a large European pressurized water reactor. Standard test results are: pressure and temperature history in the RPV, the cavity, the reactor compartment and the containment, post test melt fractions in all locations with size distribution of the debris, video film in reactor compartment and containment (timing of melt flow and hydrogen burning), and pre- and post test gas analysis in the cavity and the containment. The results of six experiments are presented here. All experiments were done with 10.6 kg of iron-alumina melt (scaling to 16 m{sup 3} corium), and a hole of 56 mm diameter (1 m scaled) or 28 mm at the center of the lower head. For comparison with a similar experiment conducted in a larger scale (1:10), the basis experiment was performed with an open path from the reactor pit to the containment (open pit), with prototypical conditions concerning the steam driven ejection out of the RPV, and a containment atmosphere, that was part air and part steam at an elevated pressure, with 3 mole-% hydrogen. In this and other tests, hydrogen production and combustion occurred. In one experiment the hydrogen effect was excluded by using only nitrogen as driving gas and a pure air atmosphere in the containment. In some tests the direct path to the containment was closed

  17. Influence of Glutamine on Superoxidedismutase, Creatine Phosphokinase and Lactate Dehydrogenase in Elderly Rats with Occult Pressure Sore%谷氨酰胺对隐匿性压疮老年大鼠SOD、CPK及LDH的影响

    Institute of Scientific and Technical Information of China (English)

    王艳; 郑宁; 刁波; 高金华; 陈慧敏

    2013-01-01

    sacrificed The skin color and shape of the areas under pressure were observed with naked eye The pathological changes of the skin muscle tissue were observed under the light microscope.The activity and content of serum superoxidedismutase (SOD), creatine phosphokinase (CPK) and lactate dehydrogenase (LDH) transudate changes were determined Results Model group and Gin group were all normal in skin, nevertheless parts of their skin under pressure appeared sustainable aglow.Under the microscope, the multicoat scalelike epidermis of model group was thinner than before, the structure was blurred, and the epidermis was partly peeled from the coriura The muscle fiber was ruptured with a light dropsy and wider gap.The corium and the muscle fiber were infiltrated by inflammatory cells.The epithelial tissue of skin in Gin group was still multicoat scalelike epidermis with clearer epithelial structure and less infiltrated corium inflammatory cells than model group.Their muscle fiber was arranged more closer than model group with light fracture, edema and infiltration by inflammatory cells.The levels of SOD, CPK and LDH in Gin group and model group wse significantly different (P<0.05).Conclusion Gin can improve the oxidation resistance and reduce the vascular endothelial injury, which can reduce the incidence of pressure sore in elderly rats.

  18. Alterações microscópicas na erupção do Sarampo Histology of measles eruption

    Directory of Open Access Journals (Sweden)

    C. Magarinos Torres

    1952-03-01

    soon in the corium and could be demonstrated twelve hours after the onset of the eruption. The early lesions (twelve to thirty-six hours in the epidermis show usually different stages in a single slide examined. They are described as minute vesicles and pustules; in older lesions the pustules have dried up forming thickened plaques in and beneath cornified layer. Parakeratotic cells with intranuclear bodies first described by TORRES & TEIXIERA (1932 b while inconstant are regarded as a pathognomonic change in measles eruption. Edema of the papillary layer and perivascular infiltrations in the reticular layer by large mononuclears some of them containing small irregular deeply stained granules (MALLORY-MEDLAR-LIPSCHÜTZ' cells are well known changes largely referred in the literature. Evidence is here submitted in support of the opinion that such cells correspond to macrophages with keratohyaline granules phagocited as a consequence of changes in cornification determined by the virus itself. Microscopic examination is necessary for the demonstration of the minute vesicles and pustules which are such an important detail in the histology of the measles eruption as it establishes connections between measles usually considered in the group of exanthematous diseases with chicken-pox, zoster, small-pox and alastrim (pustulous diseases. Epidermal changes are no more found seventy-two hours after the onset of the eruption while well-defined mantles of cells about the vessels and a moderate proliferation of fibrocytes is noticed in the corium.

  19. Effects of environmental factors on phytoplankton abundance in park water bodies of Shanghai%公园水体浮游植物与环境因子的关系

    Institute of Scientific and Technical Information of China (English)

    蒋嫣红; 程婧蕾; 王丽卿

    2012-01-01

    于2006年10月-2007年9月,对上海市10个公园景观水体水质环境因子及浮游植物群落结构进行逐月监测,应用主成分分析(PCA)和典范对应分析(CCA)探讨了浮游植物数量与水质环境因子之间的关系,评价了城市水环境状况,以期为公园水体的水质管理提供科学依据.结果表明:共鉴定公园水体浮游植物8门167种,浮游植物丰度范围为2.16×106~7.87×106 cells·L-1,主要以蓝藻、硅藻、绿藻为主,优势种由皮状席藻(Phormidium corium)、窝形席藻(Ph.fovedarum)、微小平裂藻(Merismopedia tenuissima)、尖针杆藻(Synedra acus)、银灰平裂藻(M.glauca)、啮蚀隐藻(Cryptomonas erosa)等组成;公园景观水体水温变幅为7.9~29℃,水深0.79~1.05 m,透明度0.5~0.70 m,总氮0.896~3.9mg·L-1,铵氮0.224~1.979 mg·L-1,硝酸盐0.126~0.346 mg·L-1,亚硝酸盐0.015~0.140 mg·L-1,总磷0.063~0.372 mg·L-1,活性磷0.007~0.194 mg·L-1,化学需氧量为5.418~10.685 mg·L-1.PCA分析表明,水温、透明度、氮磷营养因子以及化学需氧量是影响浮游植物密度变化的主要因素.CCA分析表明,总氮、总磷、透明度和水温是影响浮游植物群落结构季节变化的主要驱动因子.%From October 2006 to September 2007, a monthly monitoring was conducted on the phytoplankton community structure and water parameters in the landscape water bodies of 10 parks in Shanghai. Principal components analysis (PCA) and canonical correspondence analysis (CCA) were applied to analyze the relationships between the phytoplankton community structure and water parameters and to assess the status of city water environment, aimed to provide scientific basis to manage the water quality of park water body. A total of 167 phytoplankton spe-cies were identified, belonging to eight phyla. The phytoplankton abundance ranged from 2.16× 106 to 7.87×106 cells · L-1, and the phytoplankton community was dominated by Cyanophyta

  20. Relação entre a quantidade de AgNORS, atividade proliferativa e o estágio de desenvolvimento placentário em equinos Relationship between the amount of AgNORs, proliferative activity and stage of placental development in horses

    Directory of Open Access Journals (Sweden)

    Ana C.F. Mançanares

    2012-12-01

    silver nitrate and are related to the activity of rRNA synthesis and to the agility and speed of cell proliferation in the tissues studied. The objective of this study was to relate the amount of AgNORs, proliferative activity and stage of pregnancy in horses, using the coloring of Silver Nitrate. The embryonic attachments were collected, fixed in 10% buffered formaldehyde, embedded in paraplast and stained by silver nitrate. The groups were determined according to the gestational age. The amount of the corium NOR found in early pregnancy indicates the onset of cell activity, and in that the pregnancy progresses, the amount of NOR increases, suggesting higher activity and increased synthesis of their importance in maintaining the fetus. Contrary to what occurs in the corium, the quantification of NORs was higher in late pregnancy than in the beginning, suggesting the stabilization of these membranes in late pregnancy. The chorionic girdle and the yolk sac were found in early pregnancy and had lots of NORs, suggesting synthesis function and proliferation in early pregnancy, since their functions is maintenance of the embryo until the complete formation of the true placenta (chorio-allantoic membranes. We conclude that the membranes that develop in a progressive manner in accordance with the growing embryo/fetal (chorion, amnion and allantoic membranes have an increased number of NORs and the membranes that involute after the formation of the embryo/fetus (yolk sac and chorionic girdle have a decrease in number, suggesting a reduction in proliferative activity in these membranes.

  1. An Experimental Study on the Dynamics of a Single Droplet Vapor Explosion

    Energy Technology Data Exchange (ETDEWEB)

    Concilio Hansson, Roberta

    2010-07-01

    The present study aims to develop a mechanistic understanding of the thermal-hydraulic processes in a vapor explosion, which may occur in nuclear power plants during a hypothetical severe accident involving interactions of high-temperature corium melt and volatile coolant. Over the past several decades, a large body of literature has been accumulated on vapor explosion phenomenology and methods for assessment of the related risk. Vapor explosion is driven by a rapid fragmentation of high temperature melt droplets, leading to a substantial increase of heat transfer areas and subsequent explosive evaporation of the volatile coolant. Constrained by the liquid-phase coolant, the rapid vapor production in the interaction zone causes pressurization and dynamic loading on surrounding structures. While such a general understanding has been established, the triggering mechanism and subsequent dynamic fine fragmentation have yet not been clearly understood. A few mechanistic fragmentation models have been proposed, however, computational efforts to simulate the phenomena generated a large scatter of results. Dynamics of the hot liquid (melt) droplet and the volatile liquid (coolant) are investigated in the MISTEE (Micro-Interactions in Steam Explosion Experiments) facility by performing well-controlled, externally triggered, single-droplet experiments, using a high-speed visualization system with synchronized digital cinematography and continuous X-ray radiography, called SHARP (Simultaneous High-speed Acquisition of X-ray Radiography and Photography). After an elaborate image processing, the SHARP images depict the evolution of both melt material (dispersal) and coolant (bubble dynamics), and their microscale interactions, i.e. the triggering phenomenology. The images point to coolant entrainment into the droplet surface as the mechanism for direct contact/mixing ultimately responsible for energetic interactions. Most importantly, the MISTEE data reveals an inverse

  2. Investigation on Melt-Structure-Water Interactions (MSWI) during severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R.; Yang, Z.L.; Dinh, T.N.; Nourgaliev, R.R.; Bui, V.A.; Haraldsson, H.O.; Li, H.X.; Konovakhin, M.; Paladino, D.; Leung, W.H [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety

    1999-08-01

    to fragment. The molten fuel-coolant mixtures with a binary oxidic core melt (UO{sub 2}-ZrO{sub 2}) may feature low triggerability and low explosivity. The mechanical-to-thermal conversion ratio may be very low. The presence of metallic component in the melt may significantly enhance triggerability and explosivity of molten corium. The characteristics of melt spreading into 2-D channel are much different from those into 1-D channel. The evaluations of spreading parameters in reactor accident situations would be strongly affected by the uncertainties in the boundary conditions of the melt spreading process, such as downward heat removal and of upward heat removal in case of core melt spreading under water. Computer codes were developed and validated against the data obtained in the MSWI Project. The Multiphase Eulerian Lagrangian Transport code (MELT-313) was validated to investigate the hydrodynamics during the premixing process of jet penetrating into a water pool. The FlowLab code employing the multi-fluid multi-phase Lattice-Boltzmann method was developed and validated. In the Lattice-Boltzmann approach, no special treatment is needed to track the interface. A scaling technology was developed to describe the spreading efficiency of melt into I -D channel. The melt vessel interaction thermal analysis (MVITA) code describes the process of melt pool formation, melt pool convection and the resulting vessel thermal loadings. A gap cooling model and model representing penetrations were incorporated into this code. The code was coupled with ANSYS code, which performs structure analysis. Thus a coupled thermal and mechanical analysis of the loadings of a pressure vessel during the melt-vessel interaction, could be performed.

  3. OECD/MCCI 2-D Core Concrete Interaction (CCI) tests : final report February 28, 2006.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Lomperski, S.; Kilsdonk, D. J.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division); (NRC)

    2011-05-23

    Although extensive research has been conducted over the last several years in the areas of Core-Concrete Interaction (CCI) and debris coolability, two important issues warrant further investigation. The first issue concerns the effectiveness of water in terminating a CCI by flooding the interacting masses from above, thereby quenching the molten core debris and rendering it permanently coolable. This safety issue was investigated in the EPRI-sponsored Melt Attack and Coolability Experiments (MACE) program. The approach was to conduct large scale, integral-type reactor materials experiments with core melt masses ranging up to two metric tons. These experiments provided unique, and for the most part repeatable, indications of heat transfer mechanism(s) that could provide long term debris cooling. However, the results did not demonstrate definitively that a melt would always be completely quenched. This was due to the fact that the crust anchored to the test section sidewalls in every test, which led to melt/crust separation, even at the largest test section lateral span of 1.20 m. This decoupling is not expected for a typical reactor cavity, which has a span of 5-6 m. Even though the crust may mechanically bond to the reactor cavity walls, the weight of the coolant and the crust itself is expected to periodically fracture the crust and restore contact with the melt. Although crust fracturing does not ensure that coolability will be achieved, it nonetheless provides a pathway for water to recontact the underlying melt, thereby allowing other debris cooling mechanisms to proceed. A related task of the current program, which is not addressed in this particular report, is to measure crust strength to check the hypothesis that a corium crust would not be strong enough to sustain melt/crust separation in a plant accident. The second important issue concerns long-term, two-dimensional concrete ablation by a prototypic core oxide melt. As discussed by Foit the existing

  4. Experimental Investigation on the APR1400 In-Core Instrumentation Penetration Failure

    Energy Technology Data Exchange (ETDEWEB)

    An, Sang Mo; Jung, Jaehoon; Kim, Hwan Yeol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    wall where a pressure rupture can occur, thus breaching the pressure boundary. This paper focuses on the experimental investigation on the tube ejection failure during a severe accident for the APR1400 ICI penetration, as it is exposed to high temperature of melt and the force exerted by the RCS pressure. An APR1400 ICI penetration specimen was provided by Doosan Heavy Industry and ZrO{sub 2} was used as a prototypic melt. During the interaction of the penetration specimen with the melt, the temperature distributions at the reactor vessel, reactor vessel hole and penetration nozzle were monitored to estimate the weld failure and subsequent tube ejection possibilities. For the verification experiment during a severe accident, the APR1400 ICI penetration specimen was manufactured according to the real manufacturing process with the same materials and dimensions. Zirconium dioxide was used as a simulant of corium melt and interacted with the penetration specimen. The specimen was pressurized up to 2.5 bar during the interaction with melt to induce the tube ejection. The penetration weld was heated up to its melting temperature, and the penetration tube and weld above the reactor vessel surface were eroded severely by the melt. However, even though it was estimated that the weld failure occurred and the reactor vessel with stainless cladding was ablated severely by the melt, the tube ejection phenomenon was not observed and consequently the integrity for the APR1400 ICI penetration was confirmed in the present experimental conditions.

  5. Investigation on species and community ecology of Acaroid mites breeding in stored traditional Chinese animal medicinal materials%芜湖地区储藏动物性中药材孳生粉螨种类及其多样性研究

    Institute of Scientific and Technical Information of China (English)

    陶宁; 段彬彬; 王少圣; 郭伟; 李朝品

    2016-01-01

    Objective To investigate the species of Acaroid mites breeding in the stored traditional Chinese animal medici⁃nal materials and the relationship between its community and habitats. Methods A total of 30 samples of traditional Chinese animal medicinal herbs were collected from Wuhu City,Anhui Province,China. The mites were isolated by the directly micro⁃scopic and floatation microscopic examinations,and then identified and counted under a light microscope. Results Acaroid mites was represented in 28 of the 30 samples,and the breeding rate accounted for as high as 93.3%(28/30). Totally,13 spe⁃cies of Acaroid mites were identified,which belonged to 4 families and 9 genera. The densities of Acaroid mites were top in 6 Chinese herbal medicines,such as corium erinacei,aspongopus,hirudo,pheretima aspergillum,Apostichopus and huechys. The diversity parameters of these six traditional Chinese animal medicinal herbs were calculated. The highest richness indexes were in aspongopus and hirudo,the highest diversity index was in hirudo,and the highest evenness index was in Apostichopus. Conclusions There are Acaroid mites breeding in parts of the traditional Chinese animal medicinal herbs stored in Wuhu. In the storage and processing of Chinese herbal medicines,we should pay attention to the prevention and control of mites.%目的:了解芜湖地区储藏动物性中药材中孳生的粉螨种类及其群落与生境相互关系。方法从安徽省芜湖市中药房采集30种储藏动物性中药材样本,采用直接镜检法和水膜镜检法分离其中的孳生粉螨,光镜下鉴定粉螨种类并计数,计算粉螨生态参数。结果30种储藏动物性中药材样本中有28种孳生粉螨,粉螨孳生率为93.3%(28/30);共检出粉螨13种,隶属于4科9属。粉螨孳生密度最高的前6种中药材依次为刺猬皮、九香虫、水蛭、地龙、海参和红娘子,九香虫和水蛭中的粉螨丰富度指数最高,水蛭中的粉

  6. OECD MCCI project final report, February 28, 2006.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Lomperski, S.; Kilsdonk, D. J.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division); (NRC)

    2011-05-23

    Although extensive research has been conducted over the last several years in the areas of Core-Concrete Interaction (CCI) and debris coolability, two important issues warrant further investigation. The first issue concerns the effectiveness of water in terminating a CCI by flooding the interacting masses from above, thereby quenching the molten core debris and rendering it permanently coolable. This safety issue was investigated in the Melt Attack and Coolability Experiments (MACE) program. The approach was to conduct large scale, integral-type reactor materials experiments with core melt masses ranging up to two metric tons. These experiments provided unique, and for the most part repeatable, indications of heat transfer mechanism(s) that could provide long term debris cooling. However, the results did not demonstrate definitively that a melt would always be completely quenched. This was due to the fact that the crust anchored to the test section sidewalls in every test, which led to melt/crust separation, even at the largest test section lateral span of 1.20 m. This decoupling is not expected for a typical reactor cavity, which has a span of 5-6 m. Even though the crust may mechanically bond to the reactor cavity walls, the weight of the coolant and the crust itself is expected to periodically fracture the crust and restore contact with the melt. The fractured crust will provide a pathway for water to recontact the underlying melt, thereby allowing other debris cooling mechanisms to proceed and contribute to terminating the core-concrete interaction. Thus, one of the key aims of the current program was to measure crust strength to check the hypothesis that a corium crust would not be strong enough to sustain melt/crust separation in a plant accident. The second important issue concerns long-term, two-dimensional concrete ablation by a prototypic core oxide melt. As discussed by Foit, the existing reactor material database for dry cavity conditions is solely one

  7. OECD MCCI project final report, February 28, 2006.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Lomperski, S.; Kilsdonk, D. J.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division); (NRC)

    2011-05-23

    Although extensive research has been conducted over the last several years in the areas of Core-Concrete Interaction (CCI) and debris coolability, two important issues warrant further investigation. The first issue concerns the effectiveness of water in terminating a CCI by flooding the interacting masses from above, thereby quenching the molten core debris and rendering it permanently coolable. This safety issue was investigated in the Melt Attack and Coolability Experiments (MACE) program. The approach was to conduct large scale, integral-type reactor materials experiments with core melt masses ranging up to two metric tons. These experiments provided unique, and for the most part repeatable, indications of heat transfer mechanism(s) that could provide long term debris cooling. However, the results did not demonstrate definitively that a melt would always be completely quenched. This was due to the fact that the crust anchored to the test section sidewalls in every test, which led to melt/crust separation, even at the largest test section lateral span of 1.20 m. This decoupling is not expected for a typical reactor cavity, which has a span of 5-6 m. Even though the crust may mechanically bond to the reactor cavity walls, the weight of the coolant and the crust itself is expected to periodically fracture the crust and restore contact with the melt. The fractured crust will provide a pathway for water to recontact the underlying melt, thereby allowing other debris cooling mechanisms to proceed and contribute to terminating the core-concrete interaction. Thus, one of the key aims of the current program was to measure crust strength to check the hypothesis that a corium crust would not be strong enough to sustain melt/crust separation in a plant accident. The second important issue concerns long-term, two-dimensional concrete ablation by a prototypic core oxide melt. As discussed by Foit, the existing reactor material database for dry cavity conditions is solely one

  8. OECD/MCCI 2-D Core Concrete Interaction (CCI) tests : final report February 28, 2006.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Lomperski, S.; Kilsdonk, D. J.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division); (NRC)

    2011-05-23

    Although extensive research has been conducted over the last several years in the areas of Core-Concrete Interaction (CCI) and debris coolability, two important issues warrant further investigation. The first issue concerns the effectiveness of water in terminating a CCI by flooding the interacting masses from above, thereby quenching the molten core debris and rendering it permanently coolable. This safety issue was investigated in the EPRI-sponsored Melt Attack and Coolability Experiments (MACE) program. The approach was to conduct large scale, integral-type reactor materials experiments with core melt masses ranging up to two metric tons. These experiments provided unique, and for the most part repeatable, indications of heat transfer mechanism(s) that could provide long term debris cooling. However, the results did not demonstrate definitively that a melt would always be completely quenched. This was due to the fact that the crust anchored to the test section sidewalls in every test, which led to melt/crust separation, even at the largest test section lateral span of 1.20 m. This decoupling is not expected for a typical reactor cavity, which has a span of 5-6 m. Even though the crust may mechanically bond to the reactor cavity walls, the weight of the coolant and the crust itself is expected to periodically fracture the crust and restore contact with the melt. Although crust fracturing does not ensure that coolability will be achieved, it nonetheless provides a pathway for water to recontact the underlying melt, thereby allowing other debris cooling mechanisms to proceed. A related task of the current program, which is not addressed in this particular report, is to measure crust strength to check the hypothesis that a corium crust would not be strong enough to sustain melt/crust separation in a plant accident. The second important issue concerns long-term, two-dimensional concrete ablation by a prototypic core oxide melt. As discussed by Foit the existing

  9. [Effects of local transplantation of autologous adipose-derived mesenchymal stem cells on the formation of hyperplastic scar on rabbit ears].

    Science.gov (United States)

    Chen, L; Wang, D L; Wei, Z R; Wang, B; Qi, J P; Sun, G F

    2016-10-20

    test. Results: (1) The complete epithelization time of wounds of rabbits' ears was (20.0±2.0) d post injury, and hyperplastic scars were formed on post injury day 35.0±2.2. On post injury day 40, hyperplastic scars of rabbits of control group were still obvious, while those of group ADSCs became smaller, flat, soft, and light colored. (2) Compared with those in control group, epithelial cell layers and the number of nucleated cells in corium layer of hyperplastic scars of rabbits of group ADSCs were increased, and epithelium foot like and dermal papilla like structures were observed. The collagen density of hyperplastic scars of rabbits of control group was tight and arranged disorderly, while that of group ADSCs were decreased significantly and arranged regularly as compared with that of control group. (3) On post injury day 40, BrdU-labeled ADSCs were still observed in the hyperplastic scars of rabbits of group ADSCs. (4) The protein content of type Ⅰ collagen, type Ⅲ collagen, TGF-β1, and decorin in hyperplastic scars of rabbits of group ADSCs were respectively (1.40±0.04) and (8.18±0.23) μg/L, (25.1±0.7) ng/L, and (4.872±0.101) ng/mL, and those in hyperplastic scars of rabbits of control group were respectively (2.29±0.05) and (12.20±0.38) μg/L, (37.2±1.1) ng/L, and (4.143±0.024) ng/mL. Compared with those in control group, the protein content of type Ⅰ collagen, type Ⅲ collagen, and TGF-β1 in hyperplastic scars of rabbit of group ADSCs were significantly decreased (with t values from -33.66 to -22.84, P values below 0.001), while the protein content of decorin were significantly increased (t=10.41, Prabbits of group ADSCs was significantly decreased (t=4.45, Prabbit at the early stage can inhibit the formation of hyperplastic scar, promote the quality of wound healing, and the mechanism may relate to the down-regulation of TGF-β1, type Ⅰ collagen, and type Ⅲ collagen and the up-regulation of decorin induced by ADSCs.

  10. The Effective Convectivity Model for Simulation and Analysis of Melt Pool Heat Transfer in a Light Water Reactor Pressure Vessel Lower Head

    Energy Technology Data Exchange (ETDEWEB)

    Tran, Chi Thanh

    2009-09-15

    Severe accidents in a Light Water Reactor (LWR) have been a subject of intense research for the last three decades. The research in this area aims to reach understanding of the inherent physical phenomena and reduce the uncertainties in their quantification, with the ultimate goal of developing models that can be applied to safety analysis of nuclear reactors, and to evaluation of the proposed accident management schemes for mitigating the consequences of severe accidents. In a hypothetical severe accident there is likelihood that the core materials will be relocated to the lower plenum and form a decay-heated debris bed (debris cake) or a melt pool. Interactions of core debris or melt with the reactor structures depend to a large extent on the debris bed or melt pool thermal hydraulics. In case of inadequate cooling, the excessive heat would drive the structures' overheating and ablation, and hence govern the vessel failure mode and timing. In turn, threats to containment integrity associated with potential ex-vessel steam explosions and ex-vessel debris uncoolability depend on the composition, superheat, and amount of molten corium available for discharge upon the vessel failure. That is why predictions of transient melt pool heat transfer in the reactor lower head, subsequent vessel failure modes and melt characteristics upon the discharge are of paramount importance for plant safety assessment. The main purpose of the present study is to develop a method for reliable prediction of melt pool thermal hydraulics, namely to establish a computational platform for cost-effective, sufficiently-accurate numerical simulations and analyses of core Melt-Structure-Water Interactions in the LWR lower head during a postulated severe core-melting accident. To achieve the goal, an approach to efficient use of Computational Fluid Dynamics (CFD) has been proposed to guide and support the development of models suitable for accident analysis. The CFD method, on the one hand

  11. 自体脂肪源性间充质干细胞局部移植对兔耳增生性瘢痕形成的影响%Effects of local transplantation of autologous adipose-derived mesenchymal stem cells on the formation of hyperplastic scar on rabbit ears

    Institute of Scientific and Technical Information of China (English)

    陈璐; 王达利; 魏在荣; 王波; 祁建平; 孙广峰

    2016-01-01

    epithelization time of wounds of rabbits' ears was (20.0 ± 2.0) d post injury,and hyperplastic scars were formed on post injury day 35.0 ± 2.2.On post injury day 40,hyperplastic scars of rabbits of control group were still obvious,while those of group ADSCs became smaller,flat,soft,and light colored.(2) Compared with those in control group,epithelial cell layers and the number of nucleated cells in corium layer of hyperplastic scars of rabbits of group ADSCs were increased,and epithelium foot like and dermal papilla like structures were observed.The collagen density of hyperplastic scars of rabbits of control group was tight and arranged disorderly,while that of group ADSCs were decreased significantly and arranged regularly as compared with that of control group.(3) On post injury day 40,BrdU-labeled ADSCs were still observed in the hyperplastic scars of rabbits of group ADSCs.(4) The protein content of type Ⅰ collagen,type Ⅲ collagen,TGF-β1,and decorin in hyperplastic scars of rabbits of group ADSCs were respectively (1.40 ±0.04) and (8.18 ±0.23) μg/L,(25.1 ±0.7) ng/L,and (4.872 ±0.101) ng/mL,and those in hyperplastic scars of rabbits of control group were respectively (2.29 ± 0.05) and (12.20 ±0.38) μg/L,(37.2 ±1.1) ng/L,and (4.143 ±0.024) ng/mL.Compared with those in control group,the protein content of type Ⅰ collagen,type Ⅲ collagen,and TGF-β1 in hyperplastic scars of rabbit of group ADSCs were significantly decreased (with t values from-33.66 to-22.84,P values below 0.001),while the protein content of decorin were significantly increased (t =10.41,P < 0.001).(5) Compared with those in control group,the mRNA expression of TGF-β1 in hyperplastic scars of rabbits of group ADSCs was significantly decreased (t =4.45,P < 0.01),while the mRNA expression of decorin was significantly increased (t =5.61,P < 0.01).Conclusions Autologous transplantation of ADSCs into scar of rabbit at the early stage can inhibit the formation of hyperplastic scar

  12. Alterações cutâneas do cão no Kala-Azar sul-americano

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    C. Magarinos Torres

    1941-01-01

    affords serial observations of the cutaneous changes in Kala Azar as most of the blocks of skin were taken every fortnight. The following conclusions were drawn after a careful microscopic study. (1 Skin changes directly induced in the dog by the parasites of South-American Kala Azar may b described as an infiltration of the corium (pars papillaris and upper portion of the reticular layer by histocytes. Parasites are scanty, at first, latter becoming very numerous in the cytoplasm of such cells. Sometimes the histocytes either embedding or not leishman bodies appear as distinct nodes of infiltration or cell aggregations (histocytic granuloma, Figs. 8 and 22 having a perivascular distribution. The capillary loops in the papillae, the vessels of the sweat glands, the subpapillary plexus, the vertical twigs connecting the superficial and deep plexuses are the ordinary seats of the histocytic Kala Azar granulomata. (2 Some of the cutaneous changes are transient, and show spontaneous tendency to heal. A gradual transformation of the histocytes either containing or not leishman bodies into fixed connective tissue cells or fibroblasts occut and accounts for the natural regression just mentioned. Figs. 3, 5, 18, 19 and 20 are good illustrations of such fibroblastic transformation of the histocytic Kala Azar granulomata. (3 Skin changes induced by the causative organism of South-American Kala Azar are neither uniform nor simultaneous. The same stage may be found in the same dog in different periods of the disease, and not the same changes take place when pieces from several regions are examined in the same moment. The fibroblastic transformation of the histocytic granulomata marking the beginning of the process of repair, e. g., was recognised in dog C, in the 196th as well as in the 213rd (Fig. 18 and 231st (Fig. 19 days after the inoculation. (4 The connective tissue of the skin in dogs experimentally infected with South-American Kala Azar is overflowed by blood cells (monocytes