core melt retention: Topics by WorldWideScience.org

Sample records for core melt retention

Modeling of melt retention in EU-APR1400 ex-vessel core catcher

Energy Technology Data Exchange (ETDEWEB)

Granovsky, V. S.; Sulatsky, A. A.; Khabensky, V. B.; Sulatskaya, M. B. [Alexandrov Research Inst. of Technology NITI, Sosnovy Bor (Russian Federation); Gusarov, V. V.; Almyashev, V. I.; Komlev, A. A. [Saint Petersburg State Technological Univ. SPbSTU, St.Petersburg (Russian Federation); Bechta, S. [KTH, Stockholm (Sweden); Kim, Y. S. [KHNP, 1312 Gil 70, Yuseongdaero, Yuseong-gu, Daejeon (Korea, Republic of); Park, R. J.; Kim, H. Y.; Song, J. H. [KAERI, 989 Gil 111, Daedeokdaero, Yuseong-gu, Daejeon (Korea, Republic of)

2012-07-01

A core catcher is adopted in the EU-APR1400 reactor design for management and mitigation of severe accidents with reactor core melting. The core catcher concept incorporates a number of engineering solutions used in the catcher designs of European EPR and Russian WER-1000 reactors, such as thin-layer corium spreading for better cooling, retention of the melt in a water-cooled steel vessel, and use of sacrificial material (SM) to control the melt properties. SM is one of the key elements of the catcher design and its performance is critical for melt retention efficiency. This SM consists of oxide components, but the core catcher also includes sacrificial steel which reacts with the metal melt of the molten corium to reduce its temperature. The paper describes the required properties of SM. The melt retention capability of the core catcher can be confirmed by modeling the heat fluxes to the catcher vessel to show that it will not fail. The fulfillment of this requirement is demonstrated on the example of LBLOCA severe accident. Thermal and physicochemical interactions between the oxide and metal melts, interactions of the melts with SM, sacrificial steel and vessel, core catcher external cooling by water and release of non-condensable gases are modeled. (authors)
Survey of melt interactions with core retention material

International Nuclear Information System (INIS)

Powers, D.A.

1979-01-01

A survey of the interactions of up to 220 kg stainless steel melts at 1973 0 K with the candidate core retention materials borax, firebrick, high alumina cement, and magnesia is described. Data collected for the interactions include rates of material erosion, aerosol generation, gas evolution, and upward heat flux. Borax acts as an ablative solid that rapidly quenches the melt. Firebrick is ablated by the steel melt at a rate of 8.2 x 10 -6 m/s. High alumina cement is found to be an attractive melt retention material especially if it can be used in the unhydrated form. Magnesia is also found to be an attractive material though it can be eroded by the molten oxides of steel
Core melt retention and cooling concept of the ERP

Energy Technology Data Exchange (ETDEWEB)

Weisshaeupl, H [SIEMENS/KWU, Erlangen (Germany); Yvon, M [Nuclear Power International, Paris (France)

1996-12-01

For the French/German European Pressurized Water Reactor (EPR) mitigative measures to cope with the event of a severe accident with core melt down are considered already at the design stage. Following the course of a postulated severe accident with reactor pressure vessel melt through one of the most important features of a future design must be to stabilize and cool the melt within the containment by dedicated measures. This measures should - as far as possible - be passive. One very promising solution for core melt retention seems to be a large enough spreading of the melt on a high temperature resistant protection layer with water cooling from above. This is the favorite concept for the EPR. In dealing with the retention of a molten core outside of the RPV several ``steps`` from leaving the RPV to finally stabilize the melt have to gone through. These steps are: collection of the melt; transfer of the melt; distribution of the melt; confining; cooling and stabilization. The technical features for the EPR solution of a large spreading of the melt are: Dedicated spreading chamber outside the reactor pit (area about 150 m{sup 2}); high temperature resistant protection layers (e.g. Zirconia bricks) at the bottom and part of the lateral structures (thus avoiding melt concrete interaction); reactor pit and spreading compartment are connected via a discharge channel which has a slope to the spreading area and is closed by a steel plate, which will resist the core melt for a certain time in order to allow a collection of the melt; the spreading compartments is connected with the In-Containment Refuelling Water Storage Tank (IRWST) with pipes for water flooding after spreading. These pipes are closed and will only be opened by the hot melt itself. It is shown how the course of the different steps mentioned above is processed and how each of these steps is automatically and passively achieved. (Abstract Truncated)
Nuclear reactor melt-retention structure to mitigate direct containment heating

Science.gov (United States)

Tutu, Narinder K.; Ginsberg, Theodore; Klages, John R.

1991-01-01

A light water nuclear reactor melt-retention structure to mitigate the extent of direct containment heating of the reactor containment building. The structure includes a retention chamber for retaining molten core material away from the upper regions of the reactor containment building when a severe accident causes the bottom of the pressure vessel of the reactor to fail and discharge such molten material under high pressure through the reactor cavity into the retention chamber. In combination with the melt-retention chamber there is provided a passageway that includes molten core droplet deflector vanes and has gas vent means in its upper surface, which means are operable to deflect molten core droplets into the retention chamber while allowing high pressure steam and gases to be vented into the upper regions of the containment building. A plurality of platforms are mounted within the passageway and the melt-retention structure to direct the flow of molten core material and help retain it within the melt-retention chamber. In addition, ribs are mounted at spaced positions on the floor of the melt-retention chamber, and grid means are positioned at the entrance side of the retention chamber. The grid means develop gas back pressure that helps separate the molten core droplets from discharged high pressure steam and gases, thereby forcing the steam and gases to vent into the upper regions of the reactor containment building.
Comparison of advanced mid-sized reactors regarding passive features, core damage frequencies and core melt retention features

International Nuclear Information System (INIS)

Wider, H.

2005-01-01

New Light Water Reactors, whose regular safety systems are complemented by passive safety systems, are ready for the market. The special aspect of passive safety features is their actuation and functioning independent of the operator. They add significantly to reduce the core damage frequency (CDF) since the operator continues to play its independent role in actuating the regular safety devices based on modern instrumentation and control (I and C). The latter also has passive features regarding the prevention of accidents. Two reactors with significant passive features that are presently offered on the market are the AP1000 PWR and the SWR 1000 BWR. Their passive features are compared and also their core damage frequencies (CDF). The latter are also compared with those of a VVER-1000. A further discussion about the two passive plants concerns their mitigating features for severe accidents. Regarding core-melt retention both rely on in-vessel cooling of the melt. The new VVER-1000 reactor, on the other hand features a validated ex-vessel concept. (author)
Nuclear reactor melt-retention structure to mitigate direct containment heating

International Nuclear Information System (INIS)

Tutu, N.K.; Ginsberg, T.; Klages, J.R.

1991-01-01

This patent describes a nuclear reactor melt-retention structure that functions to retain molten core material within a melt retention chamber to mitigate the extent of direct containment heating. The structure being adapted to be positioned within or adjacent to a pressurized or boiling water nuclear reactor containment building at a location such that at least a portion of the melt retention structure is lower than and to one side of the nuclear reactor pressure vessel, and such that the structure is adjacent to a gas escape channel means that communicates between the reactor cavity and the containment building of the reactor. It comprises a melt-retention chamber, wall means defining a passageway extending between the reactor cavity underneath the reactor pressure vessel and one side of the chamber, the passageway including vent means extending through an upper wall portion thereof. The vent means being in communication with the upper region of the reactor containment building, whereby gas and steam discharged from the reactor pressure vessel are vented through the passageway and vent means into the gas-escape channel means and the reactor containment building
Melt spreading code assessment, modifications, and initial application to the EPR core catcher design

International Nuclear Information System (INIS)

Farmer, M.T.; Basu, S.

2009-01-01

The Evolutionary Power Reactor (EPR) is a 1,600-MWe Pressurized Water Reactor (PWR) that is undergoing a design certification review by the U.S. Nuclear Regulatory Commission (NRC). The EPR severe accident design philosophy is predicated upon the fact that the projected power rating results in a narrow margin for in-vessel melt retention by external flooding. As a result, the design addresses ex-vessel core melt stabilization using a mitigation strategy that includes: 1) an external core melt retention system to temporarily hold core melt released from the vessel; 2) a layer of 'sacrificial' material that is admixed with the melt while in the core melt retention system; 3) a melt plug that, when failed, provides a pathway for the mixture to spread to a large core spreading chamber; and finally, 4) cooling and stabilization of the spread melt by controlled top and bottom flooding. The melt spreading process relies heavily on inertial flow of a low-viscosity admixed melt to a segmented spreading chamber, and assumes that the melt mass will be distributed to a uniform height in the chamber. The spreading phenomenon thus needs to be modeled properly in order to adequately assess the EPR design. The MELTSPREAD code, developed at Argonne National Laboratory, can model segmented, and both uniform and non-uniform spreading. The NRC is using MELTSPREAD to evaluate melt spreading in the EPR design. The development of MELTSPREAD ceased in the early 1990's, and so the code was first assessed against the more contemporary spreading database and code modifications, as warranted, were carried out before performing confirmatory plant calculations. This paper provides principle findings from the MELTSPREAD assessment activities and resulting code modifications, and also summarizes the results of initial scoping calculations for the EPR plant design and preliminary plant analyses, along with the plan for performing the final set of plant calculations including sensitivity studies
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)
Proceedings of the Workshop on in-vessel core debris retention and coolability

International Nuclear Information System (INIS)

1999-01-01

This conference on in-vessel core debris retention and coolability is composed of 37 papers grouped in three sessions: session 1 (Keynote papers: Key phenomena of late phase core melt progression, accident management strategies and status quo of severe fuel damage codes, In-vessel retention as a severe accident management scheme, GAREC analyses in support of in-vessel retention concept, Latest findings of RASPLAV project); session 2 - Experiments and model development with five sub-sessions: sub-session 1 (Debris bed heat transfer: Debris and Pool Formation/Heat Transfer in FARO-LWR: Experiments and Analyses, Evaporation and Flow of Coolant at the Bottom of a Particle-Bed modelling Relocated Debris, Investigations on the Coolability of Debris in the Lower Head with WABE-2D and MESOCO-2D, Uncertainty and Sensitivity Analysis of the Heat Transfer Mechanisms in the Lower Head, Simulation of the Arrival and Evolution of Debris in a PWR Lower Head with the SFD ICARE2 code), sub-session 2 (Corium properties, molten pool natural convection, and crust formation: Physico-chemistry and corium properties for in-vessel retention, Experimental data on heat flux distribution from volumetrically heated pool with frozen boundaries, Thermal hydraulic phenomena in corium pools - numerical simulation with TOLBIAC and experimental validation with BALI, TOLBIAC code simulations of some molten salt RASPLAV experiments, SIMECO experiments on in-vessel melt pool formation and heat transfer with and without a metallic layer, Numerical investigation of turbulent natural convection heat transfer in an internally-heated melt pool and metallic layer, Current status and validation of CON2D and 3D code, Free convection of heat-generating fluid in a constrained during experimental simulation of heat transfer in slice geometry), sub-session 3 (Gap formation and gap cooling: Quench of molten aluminum oxide associated with in-vessel debris retention by RPV internal water, Experimental investigations
Melt spreading code assessment, modifications, and application to the EPR core catcher design

International Nuclear Information System (INIS)

Farmer, M.T.

2009-01-01

The Evolutionary Power Reactor (EPR) is under consideration by various utilities in the United States to provide base load electrical production, and as a result the design is undergoing a certification review by the U.S. Nuclear Regulatory Commission (NRC). The severe accident design philosophy for this reactor is based upon the fact that the projected power rating results in a narrow margin for in-vessel melt retention by external cooling of the reactor vessel. As a result, the design addresses ex-vessel core melt stabilization using a mitigation strategy that includes: (1) an external core melt retention system to temporarily hold core melt released from the vessel; (2) a layer of 'sacrificial' material that is admixed with the melt while in the core melt retention system; (3) a melt plug in the lower part of the retention system that, when failed, provides a pathway for the mixture to spread to a large core spreading chamber; and finally, (4) cooling and stabilization of the spread melt by controlled top and bottom flooding. The overall concept is illustrated in Figure 1.1. The melt spreading process relies heavily on inertial flow of a low-viscosity admixed melt to a segmented spreading chamber, and assumes that the melt mass will be distributed to a uniform height in the chamber. The spreading phenomenon thus needs to be modeled properly in order to adequately assess the EPR design. The MELTSPREAD code, developed at Argonne National Laboratory, can model segmented, and both uniform and nonuniform spreading. The NRC is thus utilizing MELTSPREAD to evaluate melt spreading in the EPR design. MELTSPREAD was originally developed to support resolution of the Mark I containment shell vulnerability issue. Following closure of this issue, development of MELTSPREAD ceased in the early 1990's, at which time the melt spreading database upon which the code had been validated was rather limited. In particular, the database that was utilized for initial validation consisted
Severe accident mitigation and core melt retention in the European pressurized reactor (EPR)

International Nuclear Information System (INIS)

Fischer, Manfred

2003-01-01

For the mitigation of severe accidents, the FPR has adopted and improved the defense-in-depth approaches of its predecessors, the French 'N4' and the German 'Konvoi' PWR's. Beyond these evolutionary changes, it includes a new, 4-th level of defense aimed at limiting the consequences of a postulated severe accident with core melting. This involves a strengthening of the confinement function and the avoidance of large early releases, by the prevention of scenarios and events with potentially high loads on the containment, incl. RPV failure at high pressure. The remaining low-pressure accidents are mitigated by dedicated design measures. The paper gives an overview and of the measures for H 2 -mitigation and steam explosion and focuses on a detailed description of the precautions and design measures for the stabilization and long-term cooling of the molten core. In the EPR the latter is achieved by melt spreading into a large outside-cooled crucible lateral to the pit, which is passively flooded and cooled with water from the IRWST. The separation of functions between pit and spreading room not only isolates the core catcher from the various loads during RPV failure, but also avoids any risks related to an unintended initiation of flooding during power operation. A stable state of the melt is reached after a few hours. Complete solidification is achieved within days. The core catcher can optionally be cooled actively by the CHRS, which avoids further steaming into the containment and establishes ambient pressure conditions in the long term. (author)
Bayesian estimation of core-melt probability

International Nuclear Information System (INIS)

Lewis, H.W.

1984-01-01

A very simple application of the canonical Bayesian algorithm is made to the problem of estimation of the probability of core melt in a commercial power reactor. An approximation to the results of the Rasmussen study on reactor safety is used as the prior distribution, and the observation that there has been no core melt yet is used as the single experiment. The result is a substantial decrease in the mean probability of core melt--factors of 2 to 4 for reasonable choices of parameters. The purpose is to illustrate the procedure, not to argue for the decrease
Results of out-of-pile experiments to investigate the possibilities of cooling a core melt with internal heat production

International Nuclear Information System (INIS)

Fieg, G.

1976-01-01

After serious hypothetical reactor accidents, melted core materials with internal heat production can occur in large quantities. A retention of these molten core masses within the containment must be ensured. The knowledge of the heat transport from volume-heated layers is necessary to clarify this matter. (orig./LH) [de
Termination of light-water reactor core-melt accidents with a chemical core catcher: the core-melt source reduction system (COMSORS)

International Nuclear Information System (INIS)

Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W.; Kenton, M.A.

1996-09-01

The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate light-water reactor core melt accidents and ensure containment integrity. A special dissolution glass is placed under the reactor vessel. If core debris is released onto the glass, the glass melts and the debris dissolves into the molten glass, thus creating a homogeneous molten glass. The molten glass, with dissolved core debris, spreads into a wide pool, distributing the heat for removal by radiation to the reactor cavity above or by transfer to water on top of the molten glass. Expected equilibrium glass temperatures are approximately 600 degrees C. The creation of a low-temperature, homogeneous molten glass with known geometry permits cooling of the glass without threatening containment integrity. This report describes the technology, initial experiments to measure key glass properties, and modeling of COMSORS operations
Natural Convection Heat Transfer of Oxide Pool During In-Vessel Retention of Core Melts

Energy Technology Data Exchange (ETDEWEB)

Park, Hae-Kyun; Chung, Bum-Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The integrity of reactor vessel may be threatened by the heat generation at the oxide pool and to the natural convection heat transfer to the reactor vessel by those two layers. Therefore, External Reactor Vessel Cooling (ERVC) is performed in order to secure the integrity of the reactor vessel. Whether the IVR(In-Vessel Retention) Strategy can be applicable to a larger reactor is the technical concern, which nourished the research interest for the natural convection heat transfer of metal and oxide pool and ERVC performance. Especially, it is hard to simulate oxide pool by experimentally due to the high level of buoyancy. Moreover, the volumetrically exothermic working fluid should be adopted to simulate the behavior of the core melts. Therefore, the volumetric heat sources that immersed in the working fluid have been adopted to simulate oxide pool by experiment. We investigated oxide pool with two different designs of the volumetric heat sources that adopted previous experiments. The investigation was performed by mass transfer experiment using analogy between heat and mass transfers. The results were compared to previous studies. We simulated the natural convection heat transfer of the oxide pool by mass transfer experiment. The isothermally cooled condition was established by limiting current technique firstly. The results were compared to previous studies under identical design of the volumetric heat sources. The average Nu's of the curvature and the top plate were close to the previous studies.
LACOMERA - large scale experiments on core degradation, melt retention and coolability at the Forschungszentrum Karslruhe

International Nuclear Information System (INIS)

Miassoedov, A.; Alsmeyer, H.; Meyer, L.

2003-01-01

The LACOMERA project at the Forschungszentrum Karlsruhe is a 3 year shared-cost action within the Fifth Framework Programme which started in September 2002. The overall objectives of the LACOMERA project are to provide research institutions from the EU member countries and associated states access to large scale experimental facilities at the Forschungszentrum Karlsruhe which shall be used to increase the knowledge of the quenching of a degraded core and regaining melt coolability in the reactor pressure vessel, of possible melt dispersion to the cavity, of molten core concrete interaction and of ex-vessel melt coolability. One major aspect is to understand how these events affect the safety of European reactors so as to lead to soundly-based accident management procedures. The project will bring together interested partners of different European member states in the area of severe accident analysis and control, with the goal to increase the public confidence in the use of nuclear energy. Moreover, partners from the newly associated states should be included as far as possible, and therefore the needs of Eastern, as well as Western, reactors will be considered in LACOMERA project. The project offers a unique opportunity to get involved in the networks and activities supporting VVER safety, and for Eastern experts to get an access to large scale experimental facilities in a Western research organisation to improve understanding of material properties and core behaviour under severe accident conditions. As a result of the first call for proposals a project on air ingress test in the QUENCH facility has been selected. A second call for proposals is opened with a deadline of 31 December 2003. (author)
The interaction of a core melt with concrete

International Nuclear Information System (INIS)

Reimann, M.; Holleck, H.; Skokan, A.; Perinic, D.

1977-01-01

In its fourth phase, a hypothetic core melt interacts with the concrete of the reactor foundation. This phase may last several days. Experimental laboratory investigations and theoretical models on the basis of model experiments aim at determining the time curve of the temperature of the core melt in order to quantify the processes up to the solidification of the melt and the end of concrete destroyal. Material interactions: 1) The two phases of the core melt, oxidic and metallic, remain separate for a long period of time. In dependence of the degree of oxidation of the system, the elemental distribution and, in particular, the fission products in the melt may be assessed. 2) The changes in the material values of the core melt in dependence of the temperature curve may be qualitatively assessed. 3) The solidification temperature of the oxidic phase of the core melt may be given in dependence of (UO 2 + ZrO 2 ) content. Thermal interactions: 1) The ratio vertical/radial erosion, which determines the cavity shape, is described in the correct order of magnitude by the extended film model. 2) The correct order of magnitude of the erosion rates is described by the concrete destruction model coupled with the film model. 3) The effects of the different concrete destruction enthalpies and concrete compositions (amount of gaseous decomposition products) may be estimated by the model calculations. (orig./HP) [de
Principles of application of mechanical design measures to control severe accident phenomena, applied to the melt retention concept of the EPR

International Nuclear Information System (INIS)

Bittermann, D.

2000-01-01

To retain and stabilize a core melt within the containment, the phenomena which principally have to be dealt with are related to melt discharge, spreading, retention and cooling, plus specific phenomena like melt dispersal and ex-vessel melt water interaction. For the elaboration of mechanical design measures provided to stabilize a melt within the containment, boundary conditions may occur which could pose extremely high thermal and mechanical loads on the structures. This file describes an approach characterized by the idea to influence the course of severe accident scenarios as much as possible in order to generate boundary conditions for mitigation means ''by design'', which enables the development of a mitigation concept with maximum confidence in the effectiveness of the measures provided. (orig.)
In vessel core melt progression phenomena

International Nuclear Information System (INIS)

Courtaud, M.

1993-01-01

For all light water reactor (LWR) accidents, including the so called severe accidents where core melt down can occur, it is necessary to determine the amount and characteristics of fission products released to the environment. For existing reactors this knowledge is used to evaluate the consequences and eventual emergency plans. But for future reactors safety authorities demand decrease risks and reactors designed in such a way that fission products are retained inside the containment, the last protective barrier. This requires improved understanding and knowledge of all accident sequences. In particular it is necessary to be able to describe the very complex phenomena occurring during in vessel core melt progression because they will determine the thermal and mechanical loads on the primary circuit and the timing of its rupture as well as the fission product source term. On the other hand, in case of vessel failure, knowledge of the physical and chemical state of the core melt will provide the initial conditions for analysis of ex-vessel core melt progression and phenomena threatening the containment. Finally a good understanding of in vessel phenomena will help to improve accident management procedures like Emergency Core Cooling System water injection, blowdown and flooding of the vessel well, with their possible adverse effects. Research and Development work on this subject was initiated a long time ago and is still in progress but now it must be intensified in order to meet the safety requirements of the next generation of reactors. Experiments, limited in scale, analysis of the TMI 2 accident which is a unique source of global information and engineering judgment are used to establish and assess physical models that can be implemented in computer codes for reactor accident analysis
A volatile-rich Earth's core inferred from melting temperature of core materials

Science.gov (United States)

Morard, G.; Andrault, D.; Antonangeli, D.; Nakajima, Y.; Auzende, A. L.; Boulard, E.; Clark, A. N.; Lord, O. T.; Cervera, S.; Siebert, J.; Garbarino, G.; Svitlyk, V.; Mezouar, M.

2016-12-01

Planetary cores are mainly constituted of iron and nickel, alloyed with lighter elements (Si, O, C, S or H). Understanding how these elements affect the physical and chemical properties of solid and liquid iron provides stringent constraints on the composition of the Earth's core. In particular, melting curves of iron alloys are key parameter to establish the temperature profile in the Earth's core, and to asses the potential occurrence of partial melting at the Core-Mantle Boundary. Core formation models based on metal-silicate equilibration suggest that Si and O are the major light element components1-4, while the abundance of other elements such as S, C and H is constrained by arguments based on their volatility during planetary accretion5,6. Each compositional model implies a specific thermal state for the core, due to the different effect that light elements have on the melting behaviour of Fe. We recently measured melting temperatures in Fe-C and Fe-O systems at high pressures, which complete the data sets available both for pure Fe7 and other binary alloys8. Compositional models with an O- and Si-rich outer core are suggested to be compatible with seismological constraints on density and sound velocity9. However, their crystallization temperatures of 3650-4050 K at the CMB pressure of 136 GPa are very close to, if not higher than the melting temperature of the silicate mantle and yet mantle melting above the CMB is not a ubiquitous feature. This observation requires significant amounts of volatile elements (S, C or H) in the outer core to further reduce the crystallisation temperature of the core alloy below that of the lower mantle. References 1. Wood, B. J., et al Nature 441, 825-833 (2006). 2. Siebert, J., et al Science 339, 1194-7 (2013). 3. Corgne, A., et al Earth Planet. Sc. Lett. 288, 108-114 (2009). 4. Fischer, R. a. et al. Geochim. Cosmochim. Acta 167, 177-194 (2015). 5. Dreibus, G. & Palme, H. Geochim. Cosmochim. Acta 60, 1125-1130 (1995). 6. Mc

Oxidation effects during corium melt in-vessel retention

Energy Technology Data Exchange (ETDEWEB)

Almyashev, V.I.; Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Sulatsky, A.A.; Vitol, S.A. [Alexandrov Scientific-Research Institute of Technology (NITI), Sosnovy Bor (Russian Federation); Gusarov, V.V. [Ioffe Institute, St. Petersburg (Russian Federation); Bechta, S. [Royal Institute of Technology (KHT), Stockholm (Sweden); Barrachin, M.; Fichot, F. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), St Paul lez Durance (France); Bottomley, P.D., E-mail: paul.bottomley@ec.europa.eu [Joint Research Centre, Institut für Transurane (ITU), Karlsruhe (Germany); Fischer, M. [AREVA GmbH, Erlangen (Germany); Piluso, P. [CEA Cadarache-DEN/DTN/STRI (France)

2016-08-15

Highlights: • Corium–steel interaction tests were re-examined particularly for transient processes. • Oxidation of corium melt was sensitive to oxidant supply and surface characteristics. • Consequences for vessel steel corrosion rates in severe accidents were discussed. - Abstract: In the in-vessel corium retention studies conducted on the Rasplav-3 test facility within the ISTC METCOR-P project and OECD MASCA program, experiments were made to investigate transient processes taking place during the oxidation of prototypic molten corium. Qualitative and quantitative data have been produced on the sensitivity of melt oxidation rate to the type of oxidant, melt composition, molten pool surface characteristics. The oxidation rate is a governing factor for additional heat generation and hydrogen release; also for the time of secondary inversion of oxidic and metallic layers of corium molten pool.
Recent progress in the LACOMERA Project (Large-Scale Experiments on Core Degradation, Melt Retention and Coolability) at the Forschungszentrum Karslruhe

International Nuclear Information System (INIS)

Miassoedov, A.; Alsmeyer, H.; Eppinger, B.; Meyer, L.; Steinbrueck, M.

2004-01-01

The LACOMERA Project at the Forschungszentrum Karlsruhe (FZK) is a 3 year action within the 5 th Framework Programme of the EU. The overall objective of the project is to offer research institutions from the EU member countries and associated states access to four large-scale experimental facilities QUENCH, LIVE, DISCO-H, and COMET which can be used to investigate core melt scenarios from the beginning of core degradation to melt formation and relocation in the vessel, possible melt dispersion to the reactor cavity, and finally corium concrete interaction and corium coolability in the reactor cavity. As a result of two calls for proposals, seven organisations from four countries are expected to profit from the LACOMERA Project participating in preparation, conduct and analysis of the following experiments: QUENCH-L1: Air ingression impact on core degradation. The test has provided unique data for the investigation of air ingress phenomenology in conditions as representative as possible of the reactor case regarding the source term. QUENCH-L2: Boil-off of a flooded bundle. The test will be of a generic interest for all reactor types, providing a link between the severe accident and design basis areas, and would deliver oxidation and thermal hydraulic data at high temperatures. LIVE-L1: Simulation of melt relocation into the Reactor Pressure Vessel (RPV) lower head for VVER conditions. The experiment will provide important information on the melt pool behaviour during the stages of air circulation at the outer RPV surface with a subsequent flooding of the lower head. LIVE-L2: Transient corium spreading and its impact on the heat fluxes to the RPV wall and on the final shape of the melt in the RPV lower head. The test will address the questions of melt stabilisation and the effects of crust formation near the RPV wall for a nonsymmetrical melt pool shape. COMET-L1: Long-term 2D concrete ablation in siliceous concrete cavity at intermediate decay heat power level with
Simulation experiment on the flooding behaviour of core melts: KATS-9

International Nuclear Information System (INIS)

Fieg, G.; Massier, H.; Schuetz, W.; Stegmaier, U.; Stern, G.

2000-11-01

For future Light Water Reactors special devices (core catchers) are being developed to prevent containment failure by basement erosion after reactor pressure vessel meltthrough during a core meltdown accident. Quick freezing of the molten core masses is desirable to reduce release of radioactivity. Several concepts of core catcher devices have been proposed based on the spreading of corium melt onto flat surfaces with subsequent water cooling. A KATS-experiment has been performed to investigate the flooding behaviour of high temperature melts 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 in terms of liquidus and solidus temperatures. Before flooding with water, spreading of the separate oxidic and metallic melts has been done in one-dimensional channels with a silicate concrete as the substrate. The flooding rate was, in relation to the melt surface, identical to the flooding rate in EPR. (orig.) [de
Molten core retention assembly

International Nuclear Information System (INIS)

Lampe, R.F.

1976-01-01

Molten fuel produced in a core overheating accident is caught by a molten core retention assembly consisting of a horizontal baffle plate having a plurality of openings therein, heat exchange tubes having flow holes near the top thereof mounted in the openings, and a cylindrical imperforate baffle attached to the plate and surrounding the tubes. The baffle assembly is supported from the core support plate of the reactor by a plurality of hanger rods which are welded to radial beams passing under the baffle plate and intermittently welded thereto. Preferably the upper end of the cylindrical baffle terminates in an outwardly facing lip to which are welded a plurality of bearings having slots therein adapted to accept the hanger rods
Thermal interaction of core melt debris with the TMI-2 baffle, core-former, and lower head structures

International Nuclear Information System (INIS)

Cronenberg, A.W.; Tolman, E.L.

1987-09-01

Recent inspection of the TMI-2 core-former baffle walls (vertical), former plates (horizontal), and lower plenum has been conducted to assess potential damage to these structures. Video observations show evidence of localized melt failure of the baffle walls, whereas fiberoptics data indicate the presence of resolidified debris on the former plates. Lower plenum inspection also confirms the presence of 20 tons or more of core debris in the lower plenum. These data indicate massive core melt relocation and the potential for melt attack on vessel structural components. This report presents analyses aimed at developing an understanding of melt relocation behavior and damage progression to TMI-2 vessel components. Thermal analysis indicates melt-through of the baffle plates, but maintenance of structural integrity of the former plates and lower head. Differences in the damage of these structures is attributed largely to differences in contact time with melt debris and pressure of water. 29 refs., 17 figs., 9 tabs
BWR core melt progression phenomena: Experimental analyses

International Nuclear Information System (INIS)

Ott, L.J.

1992-01-01

In the BWR Core Melt in Progression Phenomena Program, experimental results concerning severe fuel damage and core melt progression in BWR core geometry are used to evaluate existing models of the governing phenomena. These include control blade eutectic liquefaction and the subsequent relocation and attack on the channel box structure; oxidation heating and hydrogen generation; Zircaloy melting and relocation; and the continuing oxidation of zirconium with metallic blockage formation. Integral data have been obtained from the BWR DF-4 experiment in the ACRR and from BWR tests in the German CORA exreactor fuel-damage test facility. Additional integral data will be obtained from new CORA BWR test, the full-length FLHT-6 BWR test in the NRU test reactor, and the new program of exreactor experiments at Sandia National Laboratories (SNL) on metallic melt relocation and blockage formation. an essential part of this activity is interpretation and use of the results of the BWR tests. The Oak Ridge National Laboratory (ORNL) has developed experiment-specific models for analysis of the BWR experiments; to date, these models have permitted far more precise analyses of the conditions in these experiments than has previously been available. These analyses have provided a basis for more accurate interpretation of the phenomena that the experiments are intended to investigate. The results of posttest analyses of BWR experiments are discussed and significant findings from these analyses are explained. The ORNL control blade/canister models with materials interaction, relocation and blockage models are currently being implemented in SCDAP/RELAP5 as an optional structural component
State of the Art Report for the In-Vessel Late Core Melt Progression

International Nuclear Information System (INIS)

Kim, Hee Dong; Kang, Kyoung Ho; Park, Rae Joon

2009-04-01

The formation of corium pool in the reactor vessel lower head and its behavior is still an important issue. This issue is closely related to understanding of the core melting, its course, critical phases and timing during severe accidents and the influence of these processes on the accident progression, especially the evaluation of in-vessel retention by external reactor vessel cooling (IVR-ERVC) as a severe accident management strategy. The previous researches focused on the quisi-steady state behavior of molten corium pool in the lower head and related in-vessel retention problem. However, questions of the feasibility of the in-vessel retention concept for high power density reactor and uncertainties due to layering effect require further studies. These researches are rather essential to consider the whole evolution of the accident including formation and growth of the molten pool and the characteristic of corium arrival in the lower head and molten pool behavior after the core debris remelting. The general objective of the LIVE program performed at FzK is to study the corium pool formation and behavior with emphasis on the transient behavior through the large scale 3-D experiments. In this report, description of LIVE experimental facility and results of performance test are briefly summarized and the process to select the simulant is depicted. Also, the results of LIVE L1 and L2 tests and analytical models are included. These experimental results are very useful to development and verification of the model of molten corium pool behavior
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.
Risk reduction of core-melt accidents in advaned CAPRA burner cores

International Nuclear Information System (INIS)

Maschek, W.; Struwe, D.; Eigemann, M.

1997-01-01

As part of the CAPRA Program (Consommation Accrue de Plutonium dans les RApides) the feasibility of fast reactors is investigated to burn plutonium and also to destruct minor actinides. The design of CAPRA cores shows significant differences compared to conventional cores. Especially the high Pu-enrichment has an important influence on the core melt-down behavior and the associated recriticality risk. To cope with this risk, inherent design features and special measures/devices are investigated for their potential of early fuel discharge to reduce the criticality of the reactor core. An assessment of such measures/devices is given and experimental needs are formulated. 11 refs., 5 figs
Examination of offsite emergency protective measures for core melt accidents

International Nuclear Information System (INIS)

Aldrich, D.C.; McGrath, P.E.; Ericson, D.M. Jr.; Jones, R.B.; Rasmussen, N.C.

Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to potential nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each protective measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment falure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects
[Characteristics of mercury exchange flux between soil and atmosphere under the snow retention and snow melting control].

Science.gov (United States)

Zhang, Gang; Wang, Ning; Ai, Jian-Chao; Zhang, Lei; Yang, Jing; Liu, Zi-Qi

2013-02-01

Jiapigou gold mine, located in the upper Songhua River, was once the largest mine in China due to gold output, where gold extraction with algamation was widely applied to extract gold resulting in severe mercury pollution to ambient environmental medium. In order to study the characteristics of mercury exchange flux between soil (snow) and atmosphere under the snow retention and snow melting control, sampling sites were selected in equal distances along the slope which is situated in the typical hill-valley terrain unit. Mercury exchange flux between soil (snow) and atmosphere was determined with the method of dynamic flux chamber and in all sampling sites the atmosphere concentration from 0 to 150 cm near to the earth in the vertical direction was measured. Furthermore, the impact factors including synchronous meteorology, the surface characteristics under the snow retention and snow melting control and the mercury concentration in vertical direction were also investigated. The results are as follows: During the period of snow retention and melting the air mercury tends to gather towards valley bottom along the slope and an obvious deposit tendency process was found from air to the earth's surface under the control of thermal inversion due to the underlying surface of cold source (snow surface). However, during the period of snow melting, mercury exchange flux between the soil and atmosphere on the surface of the earth with the snow being melted demonstrates alternative deposit and release processes. As for the earth with snow covered, the deposit level of mercury exchange flux between soil and atmosphere is lower than that during the period of snow retention. The relationship between mercury exchange flux and impact factors shows that in snow retention there is a remarkable negative linear correlation between mercury exchange flux and air mercury concentration as well as between the former and the air temperature. In addition, in snow melting mercury exchange
EPRTM engineered features for core melt mitigation in severe accidents

International Nuclear Information System (INIS)

Fischer, Manfred; Henning, Andreas

2009-01-01

For the prevention of accident conditions, the EPR TM relies on the proven 3-level safety concepts inherited from its predecessors, the French 'N4' and the German 'Konvoi' NPP. In addition, a new, fourth 'beyond safety' level is implemented for the mitigation of postulated severe accidents (SA) with core melting. It is aimed at preserving the integrity of the containment barrier and at significantly reducing the frequency and magnitude of activity releases into the environment under such extreme conditions. Loss of containment integrity is prevented by dedicated design measures that address short- and long-term challenges, like: the melt-through of the reactor pressure vessel under high internal pressure, energetic hydrogen/steam explosions, containment overpressure failure, and basemat melt-through. The EPR TM SA systems and components that address these issues are: - the dedicated SA valves for the depressurization the primary circuit, - the provisions for H 2 recombination, atmospheric mixing, steam dilution, - the core melt stabilization system, - the dedicated SA containment heat removal system. The core melt stabilization system (CMSS) of the EPR TM is based on a two-stage ex-vessel approach. After its release from the RPV the core debris is first accumulated and conditioned in the (dry) reactor pit by the addition of sacrificial concrete. Then the created molten pool is spread into a lateral core catcher to establish favorable conditions for the later flooding, quenching and cooling with water passively drained from the Internal Refueling Water Storage Tank. Long-term heat removal from the containment is achieved by sprays that are supplied with water by the containment heat removal system. Complementing earlier publications focused on the principle function, basic design, and validation background of the EPR TM CMSS, this paper describes the state achieved after detailed design, as well as the technical solutions chosen for its main components, including
Incorporation of Certain Hydrophobic Excipients in the Core of Melt ...

African Journals Online (AJOL)

Patrick Erah

incorporation of hydrophobic materials (talc or magnesium stearate) in the core of such granules may further retard .... (500mg) was filled into a capsule shell and ... of the drug particles. The effect of melt granulation on the release profiles of paracetamol is shown in Fig 1. The melt granulations displayed a retarded release.
Large population center and core melt accident considerations in siting

International Nuclear Information System (INIS)

Camarinopoulos, L.; Yadigaroglu, G.

1983-01-01

The problem of providing suitable demographic siting criteria in the presence of a very large population center in an otherwise sparsely populated region is addressed. Simple calculations were performed making maximum use of pretabulated results of studies where core melt accidents are considered. These show that taking into consideration the air flow patterns in the region can lower the expected population doses from core melt accidents more effectively than distance alone. Expected doses are compared to the annual background radiation dose. A simple siting criterion combining geographical considerations with the probability of a release reaching the large population center is proposed
The WAIS Melt Monitor: An automated ice core melting system for meltwater sample handling and the collection of high resolution microparticle size distribution data

Science.gov (United States)

Breton, D. J.; Koffman, B. G.; Kreutz, K. J.; Hamilton, G. S.

2010-12-01

Paleoclimate data are often extracted from ice cores by careful geochemical analysis of meltwater samples. The analysis of the microparticles found in ice cores can also yield unique clues about atmospheric dust loading and transport, dust provenance and past environmental conditions. Determination of microparticle concentration, size distribution and chemical makeup as a function of depth is especially difficult because the particle size measurement either consumes or contaminates the meltwater, preventing further geochemical analysis. Here we describe a microcontroller-based ice core melting system which allows the collection of separate microparticle and chemistry samples from the same depth intervals in the ice core, while logging and accurately depth-tagging real-time electrical conductivity and particle size distribution data. This system was designed specifically to support microparticle analysis of the WAIS Divide WDC06A deep ice core, but many of the subsystems are applicable to more general ice core melting operations. Major system components include: a rotary encoder to measure ice core melt displacement with 0.1 millimeter accuracy, a meltwater tracking system to assign core depths to conductivity, particle and sample vial data, an optical debubbler level control system to protect the Abakus laser particle counter from damage due to air bubbles, a Rabbit 3700 microcontroller which communicates with a host PC, collects encoder and optical sensor data and autonomously operates Gilson peristaltic pumps and fraction collectors to provide automatic sample handling, melt monitor control software operating on a standard PC allowing the user to control and view the status of the system, data logging software operating on the same PC to collect data from the melting, electrical conductivity and microparticle measurement systems. Because microparticle samples can easily be contaminated, we use optical air bubble sensors and high resolution ice core density
Catalogue of generic plant states leading to core melt in PWRs: includes appendix 1: detailed description of sequences leading to core melt

International Nuclear Information System (INIS)

1996-11-01

The Task Group on thermal-hydraulic system behaviour was given a mandate from PWG 2 on Coolant System-Behaviour with the approval of CSNI to deal with the topic of Accident Management. A writing group was set up to identify generic plant states leading to core melt for pressurized water reactors (PWR) and find 'possible approaches to accident management measures' (AM-Measures) for dealing with them. From a matrix of 15 initiating events and 12 system failures (i.e. from 180 possibilities), 32 event sequences have been identified as leading to core melt. Each sequence has been divided into characteristic plant state intervals according to safety function challenges. For each of the 141 defined characteristic plant state intervals, the members of the Writing Group made proposals for AM-Measures
Praying Mantis Bending Core Breakoff and Retention Mechanism

Science.gov (United States)

Badescu, Mircea; Sherrit, Stewart; Bar-Cohen, Yoseph; Bao, Xiaoqi; Lindermann, Randel A.

2011-01-01

Sampling cores requires the controlled breakoff of the core at a known location with respect to the drill end. An additional problem is designing a mechanism that can be implemented at a small scale, yet is robust and versatile enough to be used for a variety of core samples. The new design consists of a set of tubes (a drill tube, an outer tube, and an inner tube) and means of sliding the inner and outer tubes axially relative to each other. Additionally, a sample tube can be housed inside the inner tube for storing the sample. The inner tube fits inside the outer tube, which fits inside the drill tube. The inner and outer tubes can move axially relative to each other. The inner tube presents two lamellae with two opposing grabbing teeth and one pushing tooth. The pushing tooth is offset axially from the grabbing teeth. The teeth can move radially and their motion is controlled by the outer tube. The outer tube presents two lamellae with radial extrusions to control the inner tube lamellae motion. In breaking the core, the mechanism creates two support points (the grabbing teeth and the bit tip) and one push point. The core is broken in bending. The grabbing teeth can also act as a core retention mechanism. The praying mantis that is disclosed herein is an active core breaking/retention mechanism that requires only one additional actuator other than the drilling actuator. It can break cores that are attached to the borehole bottom as
FE-simulation of the viscoplastic behaviour of different RPV steels in the frame of in-vessel melt retentions scenarios

International Nuclear Information System (INIS)

Altstadt, E.; Willschuetz, H.G.; Mueller, G.

2004-01-01

Assuming the hypothetical scenario of a severe accident with subsequent core meltdown and formation of a melt pool in the reactor pressure vessel (RPV) lower plenum of a Light Water Reactor (LWR) leads to the question about the behavior of the RPV. One accident management strategy could be to stabilize the in-vessel debris configuration in the RPV as one major barrier against uncontrolled release of heat and radio nuclides. To get an improved understanding and knowledge of the melt pool convection and the vessel creep and possible failure processes and modes occurring during the late phase of a core melt down accident the FOREVER-experiments (Failure Of REactor VEssel Retention) have been performed at the Division of Nuclear Power Safety of the Royal Institute of Technology Stockholm. These experiments are simulating the behavior of the lower head of the RPV under the thermal loads of a convecting melt pool with decay heating, and under the pressure loads that the vessel experiences in a depressurization scenario. The geometrical scale of the experiments is 1:10 compared to a common LWR. This paper deals with the experimental, numerical, and metallographical results of the creep failure experiment EC-FOREVER-4, where the American pressure vessel steel SA533B was applied for the lower head. For comparison the results of the experiment EC-FOREVER-3B, build of the French 16MND5 steel, are discussed, too. Emphasis is put on the differences in the viscoplastic behaviour of different heats of the RPV steel. For this purpose, the creep tests in the frame of the LHF/OLHF experiments are reviewed, too. As a hypothesis it is stated that the sulphur content could be responsible for differences in the creep behaviour. (orig.)
OECD/CSNI Workshop on In-Vessel Core Debris Retention and Coolability - Summary and Conclusions

International Nuclear Information System (INIS)

Behbahani, Ali-Reza; Drozd, Andrzej; Kim, Sang-Baik; Micaelli, Jean-Claude; Okkonen, Timo; Sugimoto, Jun; Trambauer, Klaus; Tuomisto, Harri

1999-01-01

understanding and the modelling of several phenomena involved in the domain of interest. They concern: the corium properties, the molten pool convection, the gap formation and cooling, the creep behaviour of the lower head, the ex-vessel critical heat flux, and they are discussed in the session specific conclusions (chapter 2.2). Some conclusions have been reached and some outstanding questions have been identified during the general discussion : Plant specific studies performed on core melt retention indicate with high confidence level that the in-vessel retention concept is possible for low (∼600 MWel) power reactors. Convective corium pool behaviour with all chemical and physical aspects requires still confirmatory research. Such as being carried out in OECD Rasplav project. The pool formation and the initial conditions in the lower plenum are plant and accident sequence dependent and are subject to large uncertainties. Since in-vessel coolability cannot be ensured (e.g., for high power plants), it is highly desirable to understand RPV lower head mechanical behaviour (i.e., size, location and time to failure) and to improve the modelling of creep rupture behaviour. For severe accident analyses different approaches are useful: Obtain physical understanding by well designed experiments and detailed code calculations; Apply integral codes to obtain scenario and system effects such as timing; Complement remaining shortcomings by separate engineering calculations when necessary. One of the most important work is to develop a consistent severe accident management strategy for each individual plant. This ensures the above discussed chemical and physical phenomena do not cause confusion among the reactor safety society
Development of the inductive ring susceptor technique for sustaining oxide melts

International Nuclear Information System (INIS)

Copus, E.R.

1983-09-01

A method for melting and sustaining large volumes of UO 2 has been developed at Sandia. This capability will greatly enhance reactor safety studies in the areas of ex-vessel interactions and degraded core retention by providing out-of-pile simulation for the decay heat process that is inherent to reactor core debris. The method, referred to as the Inductive Ring Susceptor Technique, melts UO 2 powder via inductively heated susceptor rings fashioned from highly conductive refractory metal. These rings are embedded in the non-conductive charge material. Placement of the rings is designed for optimum heat transfer and a controlled pool-type geometry. The technique has been demonstrated by a series of sustained oxide melt experiments

SWR 1000 severe accident control through in-vessel melt retention by external RPV cooling

Energy Technology Data Exchange (ETDEWEB)

Kolev, N.I. [Framatome Advanced Nuclear Power, NDSI, Erlangen (Germany)

2001-07-01

Framatome Advanced Nuclear Power is being designing a new generation NPP with boiling water reactor SWR1000. Besides of various of modern passive and active safety features the system is also designed for controlling of a postulated severe accident with extreme low probability of occurrence. This work presents the rationales behind the decision to select the external cooling as a safety management strategy during severe accident. Bounding scenery are analyzed regarding the core melting, melt-water interaction during relocation of the melt from the core region into the lower head and the external coolability of the lower head. The conclusion is reached that the external cooling for the SWR1000 is a valuable strategy for accident management during postulated severe accidents. (authors)
SWR 1000 severe accident control through in-vessel melt retention by external RPV cooling

International Nuclear Information System (INIS)

Kolev, N.I.

2001-01-01

Framatome Advanced Nuclear Power is being designing a new generation NPP with boiling water reactor SWR1000. Besides of various of modern passive and active safety features the system is also designed for controlling of a postulated severe accident with extreme low probability of occurrence. This work presents the rationales behind the decision to select the external cooling as a safety management strategy during severe accident. Bounding scenery are analyzed regarding the core melting, melt-water interaction during relocation of the melt from the core region into the lower head and the external coolability of the lower head. The conclusion is reached that the external cooling for the SWR1000 is a valuable strategy for accident management during postulated severe accidents. (authors)
An Interconnected Network of Core-Forming Melts Produced by Shear Deformation

Science.gov (United States)

Bruhn, D.; Groebner, N.; Kohlstedt, D. L.

2000-01-01

The formation mechanism of terrestrial planetary is still poorly understood, and has been the subject of numerous experimental studies. Several mechanisms have been proposed by which metal-mainly iron with some nickel-could have been extracted from a silicate mantle to form the core. Most recent models involve gravitational sinking of molten metal or metal sulphide through a partially or fully molten mantle that is often referred to as a'magma ocean. Alternative models invoke percolation of molten metal along an interconnected network (that is, porous flow) through a solid silicate matrix. But experimental studies performed at high pressures have shown that, under hydrostatic conditions, these melts do not form an interconnected network, leading to the widespread assumption that formation of metallic cores requires a magma ocean. In contrast, here we present experiments which demonstrate that shear deformation to large strains can interconnect a significant fraction of initially isolated pockets of metal and metal sulphide melts in a solid matrix of polycrystalline olivine. Therefore, in a dynamic (nonhydrostatic) environment, percolation remains a viable mechanism for the segregation and migration of core-forming melts in a solid silicate mantle.
Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt

International Nuclear Information System (INIS)

Aldrich, D.C.; McGrath, P.E.; Rasmussen, N.C.

1978-06-01

Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects
Nuclear power reactor core melt accidents. Current State of Knowledge

International Nuclear Information System (INIS)

Jacquemain, Didier; Cenerino, Gerard; Corenwinder, Francois; Raimond, Emmanuel IRSN; Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Couturier, Jean; Debaudringhien, Cecile; Duprat, Anna; Dupuy, Patricia; Evrard, Jean-Michel; Nicaise, Gregory; Berthoud, Georges; Studer, Etienne; Boulaud, Denis; Chaumont, Bernard; Clement, Bernard; Gonzalez, Richard; Queniart, Daniel; Peltier, Jean; Goue, Georges; Lefevre, Odile; Marano, Sandrine; Gobin, Jean-Dominique; Schwarz, Michel; Repussard, Jacques; Haste, Tim; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno; Durin, Michel; Andreo, Francois; Atkhen, Kresna; Daguse, Thierry; Dubreuil-Chambardel, Alain; Kappler, Francois; Labadie, Gerard; Schumm, Andreas; Gauntt, Randall O.; Birchley, Jonathan

2015-11-01

For over thirty years, IPSN and subsequently IRSN has played a major international role in the field of nuclear power reactor core melt accidents through the undertaking of important experimental programmes (the most significant being the Phebus-FP programme), the development of validated simulation tools (the ASTEC code that is today the leading European tool for modelling severe accidents), and the coordination of the SARNET (Severe Accident Research Network) international network of excellence. These accidents are described as 'severe accidents' because they can lead to radioactive releases outside the plant concerned, with serious consequences for the general public and for the environment. This book compiles the sum of the knowledge acquired on this subject and summarises the lessons that have been learnt from severe accidents around the world for the prevention and reduction of the consequences of such accidents, without addressing those from the Fukushima accident, where knowledge of events is still evolving. The knowledge accumulated by the Institute on these subjects enabled it to play an active role in informing public authorities, the media and the public when this accident occurred, and continues to do so to this day. Following the introduction, which describes the structure of this book and highlights the objectives of R and D on core melt accidents, this book briefly presents the design and operating principles (Chapter 2) and safety principles (Chapter 3) of the reactors currently in operation in France, as well as the main accident scenarios envisaged and studied (Chapter 4). The objective of these chapters is not to provide exhaustive information on these subjects (the reader should refer to the general reference documents listed in the corresponding chapters), but instead to provide the information needed in order to understand, firstly, the general approach adopted in France for preventing and mitigating the consequences of core melt
Improved Rock Core Sample Break-off, Retention and Ejection System, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — The proposed effort advances the design of an innovative core sampling and acquisition system with improved core break-off, retention and ejection features. The...
Melt propagation in dry core debris beds

International Nuclear Information System (INIS)

Dosanjh, S.S.

1989-01-01

During severe light water reactor accidents like Three Mile Island Unit 2, the fuel rods can fragment and thus convert the reactor core into a large particle bed. The postdryout meltdown of such debris beds is examined. A two-dimensional model that considers the presence of oxidic (UO 2 and ZrO 2 ) as well as metallic (e.g., zirconium) constituents is developed. Key results are that a dense metallic crust is created near the bottom of the bed as molten materials flow downward and freeze; liquid accumulates above the blockage and, if zirconium is present, the pool grows rapidly as molten zirconium dissolved both UO 2 and ZrO 2 particles; if the melt wets the solid, a fraction of the melt flows radially outward under the action of capillary forces and freezes near the radial boundary; in a nonwetting system, all of the melt flows into the bottom of the bed; and when zirconium and iron are in intimate contact and the zirconium metal atomic fraction is > 0.33, these metals can liquefy and flow out of the bed very early in the meltdown sequence
Improved Rock Core Sample Break-off, Retention and Ejection System, Phase II

Data.gov (United States)

National Aeronautics and Space Administration — The proposed effort advances the design of an innovative core sampling and acquisition system with improved core break-off, retention and ejection features. Phase 1...
Universal viscosity growth in metallic melts at megabar pressures: the vitreous state of the Earth's inner core

International Nuclear Information System (INIS)

Brazhkin, Vadim V; Lyapin, A G

2000-01-01

Experimental data on and theoretical models for the viscosity of various types of liquids and melts under pressure are reviewed. Experimentally, the least studied melts are those of metals, whose viscosity is considered to be virtually constant along the melting curve. The authors' new approach to the viscosity of melts involves the measurement of the grain size in solidified samples. Measurements on liquid metals at pressures up to 10 GPa using this method show, contrary to the empirical approach, that the melt viscosity grows considerably along the melting curves. Based on the experimental data and on the critical analysis of current theories, a hypothesis of a universal viscosity behavior is introduced for liquids under pressure. Extrapolating the liquid iron results to the pressures and temperatures at the Earth's core reveals that the Earth's outer core is a very viscous melt with viscosity values ranging from 10 2 Pa s to 10 11 Pa s depending on the depth. The Earth's inner core is presumably an ultraviscous (>10 11 Pa s) glass-like liquid - in disagreement with the current idea of a crystalline inner core. The notion of the highly viscous interior of celestial bodies sheds light on many mysteries of planetary geophysics and astronomy. From the analysis of the pressure variation of the melting and glass-transition temperatures, an entirely new concept of a stable metallic vitreous state arises, calling for further experimental and theoretical study. (reviews of topical problems)
Reaction- and melting behaviour of LWR-core components UO2, Zircaloy and steel during the meltdown period

International Nuclear Information System (INIS)

Hofmann, P.

1976-07-01

The reaction behaviour of the UO 2 , Zircaloy-4 and austenitic steel core components was investigated as a function of temperature (till melting temperatures) under inert and oxidizing conditions. Component concentrations varied between that of Corium-A (65 wt.% UO 2 , 18% Zry, 17% steel) and that of Corium-E (35 wt.% UO 2 , 10% Zry, 55% steel). In addition, Zircaloy and stainless steel were used with different degrees of oxidation. The paper describes systematically the phases that arise during heating and melting. The integral composition of the melts and the qualitative as well as quantitative analysis of the phases present in solidified corium are given. In some cases melting points have been determined. The reaction and melting behaviour of the corium specimens strongly depends on the concentration and on the degree of oxidation of the core components. First liquid phases are formed at the Zry-steel interface at about 1,350 0 C. The maximum temperatures of about 2,500 0 C for the complete melting of the corium-specimens are well below the UO 2 melting point. Depending on the steel content and/or degree of oxidation of Zry and steel, a homogeneous metallic or oxide melt or two immiscible melts - one oxide and the other metallic - are obtained. During the melting experiments performed under inert gas conditions the chemical composition of the molten specimens generally change by evaporation losses of single elements, especially of uranium, zirconium and oxygen. The total weight losses go up to 30%; under oxidizing conditions they are substantially smaller due to the occurrence of different phases. In air or water vapor, the occurrence of the phases and the melting behaviour of the core components are strongly influenced by the oxidation rate and the oxygen supply to the surface of the melt. In the case of the hypothetical core melting accident, a heterogeneous melt (oxide and metallic) is probable after the meltdown period. (orig./RW) [de
In-vessel core melt retention by RPV external cooling for high power PWR. MAAP 4 analysis on a LBLOCA scenario without SI

International Nuclear Information System (INIS)

Cognet, C.; Gandrille, P.

1999-01-01

In-, ex-vessel reflooding or both simultaneously can be envisaged as Accident Management Measures to stop a Severe Accident (SA) in vessel. This paper addresses the possibility of in-vessel core melt retention by RPV external flooding for a high power PWR (4250 MWth). The reactor vessel is assumed to have no lower head penetration and thermal insulation is neglected. The effects of external cooling of high power density debris, where the margin for such a strategy is low, are investigated with the MAAP4 code. MAAP4 code is used to verify the system capability to flood the reactor pit and to predict simultaneously the corium relocation into the lower head with the thermal and mechanical response of the RPV in transient conditions. The corium pool cooling and holding in the RPV lower head is analysed. Attention is paid to the internal heat exchanges between corium components. This paper focuses particularly the heat transfer between oxidic and metallic phases as well as between the molten metallic phase and the RPV wall of utmost importance for challenging the RPV integrity in vicinity of the metallic phase. The metal segregation has a decisive influence upon the attack of the vessel wall due to a very strong peaking of the lateral flux ('focusing effect'). Thus, the dynamics of the formation of the metallic layer characterized by a growing inventory of steel, both from a partial vessel ablation and the degradation of internals steel structures by the radiative heat flux from the debris, is displayed. The analysed sequence is a surge line rupture near the hot leg (LBLOCA) leading to the fastest accident progression
The internal core catcher in Super Phenix 1

International Nuclear Information System (INIS)

Le Rigoleur, C.; Kayser, G.; Maurin, G.; Magnon, B.

1982-07-01

The internal core catcher in SUPER PHENIX 1 is described here in some detail. The fuel retention capabilities are presented for situations of increasing severity. The first situation corresponds to the core catcher design. It relates to a hypothetical subassembly accident that would cause a limited quantity of fuel, corresponding to the mass of seven subassemblies, to be deposited on the core catcher. For this situation and at all levels of the analysis, the most conservative assumptions are made in order to prove the integrity of the core catcher. The second situation corresponds to a hypothetical larger core melt accident. In this case, for some of the parameters, assumptions are made that correspond to the most likely situations based on engineering considerations. Then the maximum retention capabilities are presented
Neutronics simulations on hypothetical power excursion and possible core melt scenarios in CANDU6

International Nuclear Information System (INIS)

Kim, Yonghee

2015-01-01

LOCA (Loss of coolant accident) is an outstanding safety issue in the CANDU reactor system since the coolant void reactivity is strongly positive. To deal with the LOCA, the CANDU systems are equipped with specially designed quickly-acting secondary shutdown system. Nevertheless, the so-called design-extended conditions are requested to be taken into account in the safety analysis for nuclear reactor systems after the Fukushima accident. As a DEC scenario, the worst accident situation in a CANDU reactor system is a unprotected LOCA, which is supposed to lead to a power excursion and possibly a core melt-down. In this work, the hypothetical unprotected LOCA scenario is simulated in view of the power excursion and fuel temperature changes by using a simplified point-kinetics (PK) model accounting for the fuel temperature change. In the PK model, the core reactivity is assumed to be affected by a large break LOCA and the fuel temperature is simulated to account for the Doppler effect. In addition, unlike the conventional PK simulation, we have also considered the Xe-I model to evaluate the impact of Xe during the LOCA. Also, we tried to simulate the fuel and core melt-down scenario in terms of the reactivity through a series of neutronics calculations for hypothetical core conditions. In case of a power excursion and possible fuel melt-down situation, the reactor system behavior is very uncertain. In this work, we tried to understand the impacts of fuel melt and relocation within the pressure vessel on the core reactivity and failure of pressure and calandria tubes. (author)
Simulant melt experiments on performance of the in-vessel core catcher

International Nuclear Information System (INIS)

Kyoung-Ho Kang; Rae-Joon Park; Sang-Baik Kim; Suh, K.Y.; Cheung, F.B.; Rempe, J.L.

2005-01-01

Full text of publication follows: LAVA-GAP experiments are in progress to investigate the performance of the in-vessel core catcher using alumina melt as a corium simulant. The hemispherical in-vessel core catcher made of carbon steel was installed inside the lower head vessel with uniform gap of 5 mm or 10 mm to the inner surface of the lower head vessel. As a performance test of the in-vessel core catcher, the effects of base steel and internal coating materials and gap thickness between the core catcher and the lower head vessel were examined in this study. In the LAVA-GAP-2 and LAVA-GAP-3 tests, the base steel was carbon steel and the gap thickness was 10 mm. On the other hand, in the LAVA-GAP-4 and LAVA-GAP-5 tests, the base steel was stainless steel and the gap thickness was 5 mm. Actual composition of the coating material for the LAVA-GAP-4 test was 92% of ZrO 2 - 8% of Y 2 O 3 including 95% of Ni - 5% of Al bond coat same as the LAVA-GAP-3 test. In these tests, the thickness of ZrO 2 internal coating was 0.5 mm. To examine the effects of the coating material, in-vessel core catcher with a 0.6 mm-thick ZrO 2 coating without bond coat was used in the LAVA-GAP-5 test. This report summarizes the experimental results and the post metallurgical inspection results of the LAVA-GAP-4 and LAVA-GAP- 5 tests. In the LAVA-GAP-4 and LAVA-GAP-5 tests, the core catcher was failed and it was stuck to the inner surface of the lower head vessel. LAVA-GAP-4 and LAVA-GAP-5 test results imply that 5 mm thick gap is rather small for sufficient water ingression and steam venting through the gap. In case of small gap size, water is boiled off and steam increases pressure inside the gap and so water can not ingress into the gap at the initial heat up stage. Metallurgical inspections on the test specimens indicate that the internal coating layer might melt totally and dispersed in the base steel and the solidified iron melt and so the detection frequencies of Zr and O are trivial all
Simulation of core melt spreading with lava: theoretical background and status of validation

International Nuclear Information System (INIS)

Allelein, H.-J.; Breest, A.; Spengler, C.

2000-01-01

The goal of this paper is to present the GRS R and D achievements and perspectives of its approach to simulate ex-vessel core melt spreading. The basic idea followed by GRS is the analogy of core melt spreading to volcanic lava flows. A fact first proposed by Robson (1967) and now widely accepted is that lava rheologically behaves as a Bingham fluid, which is characterized by yield stress and plastic viscosity. Recent experimental investigations by Epstein (1996) reveal that corium-concrete mixtures may be described as Bingham fluids. The GRS code LAVA is based on a successful lava flow model, but is adapted to prototypic corium and corium-simulation spreading. Furthermore some detailed physical models such as a thermal crust model on the free melt surface and a model for heat conduction into the substratum are added. Heat losses of the bulk, which is represented by one mean temperature, are now determined by radiation and by temperature profiles in the upper crust and in the substratum. In order to reduce the weak mesh dependence of the original algorithm, a random space method of cellular automata is integrated, which removes the mesh bias without increasing calculation time. LAVA is successfully validated against a lot of experiments using different materials spread. The validation process has shown that LAVA is a robust and fast running code to simulate corium-type spreading. LAVA provides all integral information of practical interest (spreading length, height of the melt after stabilization) and seems to be an appropriate tool for handling large core melt masses within a plant application. (orig.)
In-core melt progression for the MAAP 4 codes

International Nuclear Information System (INIS)

Wu, C.-D.; Paik, Chan Y.; Henry, Robert E.; Ply, Martin G.

2004-01-01

The MAAP 4 core melt progression model contains provisions for the formation of a molten debris pool surrounded by a crust during late phase core degradation. A predominantly oxidic molten pool with a predominantly metallic lower crust may naturally develop through a combination of models for real material phase diagrams, mechanistic relocation, and rules to recognize extremely low porosity and the liquid fractions of adjacent highly degraded nodes. Pool size and shape thus becomes relatively independent of core nodalization (which only governs the coarseness of the crust location). An upper pool crust is mechanistically allowed during consideration of radiative and convective heat losses from the pool top surface to surrounding core nodes, the core barrel, and upper internals. Circulation within the pool causes mass and energy exchange between participating pool nodes, and determines the heat fluxes to the boundary crusts. Side and bottom node failure is predicted based on the time, temperature, and stress. Calculations demonstrate that this concept allows simulation of the degraded core geometry observed during the TMI-2 accident. (author)
Melting of Fe-Si-O alloys: the Fate of Coexisting Si and O in the Core

Science.gov (United States)

Arveson, S. M.; Lee, K. K. M.

2017-12-01

The light element budget of Earth's core plays an integral role in sustaining outer core convection, which powers the geodynamo. Many experiments have been performed on binary iron compounds, but the results do not robustly agree with seismological observations and geochemical constraints. Earth's core is almost certainly made up of multiple light elements, so the future of core composition studies lies in ternary (or higher order) systems in order to examine interactions between light elements. We perform melting experiments on Fe-Si-O alloys in a laser-heated diamond-anvil cell to 80 GPa and 4000 K. Using 2D multi- wavelength imaging radiometry together with textural and chemical analysis of quenched samples, we measure the high-pressure melting curves and determine partitioning of light elements between the melt and the coexisting solid. Quenched samples are analyzed both in map view and in cross section using scanning electron microscopy (SEM) and electron microprobe analysis (EPMA) to examine the 3D melt structure and composition. Partitioning of light elements between molten and solid alloys dictates (1) the density contrast at the ICB, which drives compositional convection in the outer core and (2) the temperature of the CMB, an integral parameter for understanding the deep Earth. Our experiments suggest silicon and oxygen do not simply coexist in the melt and instead show complex solubility based on temperature. Additionally, we do not find evidence of crystallization of SiO2 at low oxygen content as was recently reported.11 Hirose, K., et al., Crystallization of silicon dioxide and compositional evolution of the Earth's core. Nature, 2017. 543(7643): p. 99-102.
Melting of iron at the Earth's core conditions by molecular dynamics simulation

Directory of Open Access Journals (Sweden)

Y. N. Wu

2011-09-01

Full Text Available By large scale molecular dynamics simulations of solid-liquid coexistence, we have investigated the melting of iron under pressures from 0 to 364 GPa. The temperatures of liquid and solid regions, and the pressure of the system are calculated to estimate the melting point of iron. We obtain the melting temperature of iron is about 6700±200K under the inner-outer core boundary, which is in good agreement with the result of Alfè et al. By the pair analysis technique, the microstructure of liquid iron under higher pressures is obviously different from that of lower pressures and ambient condition, indicating that the pressure-induced liquid-liquid phase transition may take place in iron melts.
Process for retention of iodine and aerosols during containment venting

International Nuclear Information System (INIS)

Eckardt, B.; Betz, R.; Greger, G.U.; Werner, K.D.

1990-05-01

A process for retention of the majority of aerosols and iodine during containment venting was optimized. For this purpose, sections of a two-stage process comprising a venturi scrubber and a metal-fiber filter demister were tested under containment venting conditions assumed to prevail during a hypothetical core - melt accident and optimized with a view to achieving high decontamination factors and loading capacity while minimizing the size of the process. The loading and retention tests performed in a scrubber operating pressure range between 1 and 10 bar, at temperatures from 50 to 200degC (also boiling pools) and in air and steam atmospheres. Under these unfavorable conditions for aerosol retention, the retention efficiencies were determined at various flow rates with soluble and non-soluble aerosols as well as gaseous iodine. The retention efficiencies for BaSO 4 , uranine and SnO 2 aerosols were determined to be 99.95% to 99.99% for venturi scrubbers with metal-fiber filter demister. The retention efficiency for elemental iodine was determined to be ≥99% including revolatization effects over a 24-hour operating period. The high loading capacity of the venturi scrubber unit was verified after process modifications with various aerosols. The use of full-scale process section together with the best possible simulation of containment venting conditions by the test parameters ensured that the results can be transferred to real venting equipment. The aim of ensuring the retention of the majority of the aerosol-borne activity and of elemental iodine activity and minimizing the process size was clearly achieved and verified by means of this optimized venting equipment under an extremely wide range of hypothetical core-melt accident conditions. (orig.) With 17 refs., 3 tabs., 35 annexes [de
An internal core catcher for a pool L.M.F.B.R. and connected studies

International Nuclear Information System (INIS)

Le Rigoleur, C.; Kayser, G.

1979-01-01

This paper describes an internal core catcher for a pool LMFBR. Problems related to retention of debris are studied: downward progression of debris from the core to the core catcher, debris bed formation, heat transfer below the core catcher plate and to the main vessel, mechanical resistance. These results are used to estimate the performances of the internal core catcher for a given core melt-down-accident. It is seen that for a uniform thickness layer on the core catcher the retention capabilities are satisfactory. Then the problem of a heap of debris is approached. Dryout is studied. Uncertainties related to the bed characteristics and problems of extended dryout beds are put forward

A 400-year ice core melt layer record of summertime warming in the Alaska Range

Science.gov (United States)

Winski, D.; Osterberg, E. C.; Kreutz, K. J.; Wake, C. P.; Ferris, D. G.; Campbell, S. W.; Baum, M.; Raudzens Bailey, A.; Birkel, S. D.; Introne, D.; Handley, M.

2017-12-01

Warming in high-elevation regions has socially relevant impacts on glacier mass balance, water resources, and sensitive alpine ecosystems, yet very few high-elevation temperature records exist from the middle or high latitudes. While many terrestrial paleoclimate records provide critical temperature records from low elevations over recent centuries, melt layers preserved in alpine glaciers present an opportunity to develop calibrated, annually-resolved temperature records from high elevations. We present a 400-year temperature record based on the melt-layer stratigraphy in two ice cores collected from Mt. Hunter in the Central Alaska Range. The ice core record shows a 60-fold increase in melt frequency and water equivalent melt thickness between the pre-industrial period (before 1850) and present day. We calibrate the melt record to summer temperatures based on local and regional weather station analyses, and find that the increase in melt production represents a summer warming of at least 2° C, exceeding rates of temperature increase at most low elevation sites in Alaska. The Mt. Hunter melt layer record is significantly (p<0.05) correlated with surface temperatures in the central tropical Pacific through a Rossby-wave like pattern that induces high temperatures over Alaska. Our results show that rapid alpine warming has taken place in the Alaska Range for at least a century, and that conditions in the tropical oceans contribute to this warming.
Contribution of Anticipated Transients Without Scram (ATWS) to core melt at United States nuclear power plants

International Nuclear Information System (INIS)

Giachetti, R.T.

1989-09-01

This report looks at WASH-1400 and several other Probabilistic Risk Assessments (PRAs) and Probabilistic Safety Studies (PSSs) to determine the contribution of Anticipated Transients Without Scram (ATWS) events to the total core melt probability at eight nuclear power plants in the United States. After considering each plant individually, the results are compared from plant to plant to see if any generic conclusions regarding ATWS, or core melt in general, can be made. 8 refs., 34 tabs
Viscosity measurements on metal melts at high pressure and viscosity calculations for the earth's core

International Nuclear Information System (INIS)

Mineev, Vladimir N; Funtikov, Aleksandr I

2004-01-01

A review is given of experimental and calculated data on the viscosity of iron-based melts on the melting curve. The interest in these data originates in the division of opinion on whether viscosity increases rather moderately or considerably in the high-pressure range. This disagreement is especially pronounced in the interpretation of the values of molten iron and its compounds in the environment of the earth's outer core. The conclusion on a substantial rise in viscosity mostly follows from the universal law, proposed by Brazhkin and Lyapin [1], of viscosity changing along the metal melting curve in the high-pressure range. The review analyzes available experimental and computational data, including the most recent ones. Data on viscosity of metals under shock wave compression in the megabar pressure range are also discussed. It is shown that data on viscosity of metal melts point to a small increase of viscosity on the melting curve. Specifics are discussed of the phase diagram of iron made more complex by the presence of several phase transitions and by the uncertainty in the position of the melting curve in the high-pressure range. Inaccuracies that arise in extrapolating the results of viscosity measurements to the pressure range corresponding to the earth's core environment are pointed out. (reviews of topical problems)
Simulant melt experiments on performance of the in-vessel core catcher

International Nuclear Information System (INIS)

Kang, Kyoung-Ho; Park, Rae-Joon; Kim, Sang-Baik; Suh, K.Y.; Cheung, F.B.; Rempe, J.L.

2007-01-01

In order to enhance the feasibility of in-vessel retention (IVR) of molten core material during a severe accident for high-power reactors, an in-vessel core catcher (IVCC) was designed and evaluated as part of a joint United States-Korean International Nuclear Energy Research Initiative (INERI). The proposed IVCC is expected to increase the thermal margin for success of IVR by providing an 'engineered gap' for heat transfer from materials that relocate during a severe accident and potentially serving as a sacrificial material under a severe accident. In this study, LAVA-GAP experiments were performed to investigate the thermal and mechanical performance of the IVCC using the alumina melt as simulant. The LAVA-GAP experiments aim to examine the feasibility and sustainability of the IVCC under the various test conditions using 1/8th scale hemispherical test sections. As a feasibility test of the proposed IVCC in this INERI project, the effects of IVCC base steel materials, internal coating materials, and gap size between the IVCC and the vessel lower head were examined. The test results indicated that the internally coated IVCC has high thermal performance compared with the uncoated IVCC. In terms of integrity of the base steel, carbon steel is superior to stainless steel and the effect of bond coat is found to be trivial for the tests performed in this study. The thermal load is mitigated via boiling heat removal in the gap between the IVCC and the vessel lower head. The current test results imply that gaps less than 10 mm are not enough to guarantee effective cooling induced by water ingression and steam venting there through. Selection of endurable material and pertinent gap size is needed to implement the proposed IVCC concept into advanced reactor designs
Considerations concerning the strategy of corium retention in the reactor vessel

International Nuclear Information System (INIS)

2015-01-01

Third-generation nuclear reactors are characterised by consideration during design of core meltdown accidents. More specifically, dedicated measures or devices must be implemented to avoid basemat melt-through in the reactor building. These devices must have a high level of confidence. The strategy of corium retention in the reactor vessel, if supported by appropriate research and development, makes it possible to achieve this objective. IRSN works alone or in partnerships to address all the issues associated with in-vessel corium retention. This document describes the in-vessel corium retention strategy and its limitations, along with the research programs conducted by IRSN in this area
Shock Compression and Melting of an Fe-Ni-Si Alloy: Implications for the Temperature Profile of the Earth's Core and the Heat Flux Across the Core-Mantle Boundary

Science.gov (United States)

Zhang, Youjun; Sekine, Toshimori; Lin, Jung-Fu; He, Hongliang; Liu, Fusheng; Zhang, Mingjian; Sato, Tomoko; Zhu, Wenjun; Yu, Yin

2018-02-01

Understanding the melting behavior and the thermal equation of state of Fe-Ni alloyed with candidate light elements at conditions of the Earth's core is critical for our knowledge of the region's thermal structure and chemical composition and the heat flow across the liquid outer core into the lowermost mantle. Here we studied the shock equation of state and melting curve of an Fe-8 wt% Ni-10 wt% Si alloy up to 250 GPa by hypervelocity impacts with direct velocity and reliable temperature measurements. Our results show that the addition of 10 wt% Si to Fe-8 wt% Ni alloy slightly depresses the melting temperature of iron by 200-300 (±200) K at the core-mantle boundary ( 136 GPa) and by 600-800 (±500) K at the inner core-outer core boundary ( 330 GPa), respectively. Our results indicate that Si has a relatively mild effect on the melting temperature of iron compared with S and O. Our thermodynamic modeling shows that Fe-5 wt% Ni alloyed with 6 wt% Si and 2 wt% S (which has a density-velocity profile that matches the outer core's seismic profile well) exhibits an adiabatic profile with temperatures of 3900 K and 5300 K at the top and bottom of the outer core, respectively. If Si is a major light element in the core, a geotherm modeled for the outer core indicates a thermal gradient of 5.8-6.8 (±1.6) K/km in the D″ region and a high heat flow of 13-19 TW across the core-mantle boundary.
Melting and solidification behavior of Cu/Al and Ti/Al bimetallic core/shell nanoparticles during additive manufacturing by molecular dynamics simulation

Science.gov (United States)

Rahmani, Farzin; Jeon, Jungmin; Jiang, Shan; Nouranian, Sasan

2018-05-01

Molecular dynamics (MD) simulations were performed to investigate the role of core volume fraction and number of fusing nanoparticles (NPs) on the melting and solidification of Cu/Al and Ti/Al bimetallic core/shell NPs during a superfast heating and slow cooling process, roughly mimicking the conditions of selective laser melting (SLM). One recent trend in the SLM process is the rapid prototyping of nanoscopically heterogeneous alloys, wherein the precious core metal maintains its particulate nature in the final manufactured part. With this potential application in focus, the current work reveals the fundamental role of the interface in the two-stage melting of the core/shell alloy NPs. For a two-NP system, the melting zone gets broader as the core volume fraction increases. This effect is more pronounced for the Ti/Al system than the Cu/Al system because of a larger difference between the melting temperatures of the shell and core metals in the former than the latter. In a larger six-NP system (more nanoscopically heterogeneous), the melting and solidification temperatures of the shell Al roughly coincide, irrespective of the heating or cooling rate, implying that in the SLM process, the part manufacturing time can be reduced due to solidification taking place at higher temperatures. The nanostructure evolution during the cooling of six-NP systems is further investigated. [Figure not available: see fulltext.
Method and device for catching reactor core melt-down masses in hypothetical accidents of nuclear power plants

International Nuclear Information System (INIS)

Morlock, G.; Wiesemes, J.; Bachner, D.

1977-01-01

The device is to receive the afterheat of the molten core and in this way to prevent afterflow of coolant and a new criticality. A tank below the reactor pressure vessel, with the proper diameter, contains a store of salt or a salt mixture suitable to receive the afterheat of a core melt-down as heat of fusion or conversion. Above the salt, there is a layer of thermoplastics or of a material forming a hardening foam. Coolant eventually continuing to flow out is separated from the core melt by this barrier layer, and thus the build-up of high steam pressures is prevented. Neutron-absorbing materials, like boron salts mixed to the salts, as well as a subdivision of the salt surface, e.g. by means of canalizing firebricks, prevent the formation of new criticality. Further installations within the tank, like pipings or channels, permit the introduction of water after cooling down of the core or salt melt-down mass and to wash out the brine with all radioactive and other constituents for transport to reprocessing or ultimate storage. (HP) [de
Effect of periodic melting on geochemical and isotopic signals in an ice core from Lomonosovfonna, Svalbard

NARCIS (Netherlands)

Pohjola, V.A.; Moore, J.C.; Isaksson, E.; Jauhiainen, T.; Wal, R.S.W. van de; Martma, T.; Meijer, H.A.J.; Vaikmäe, R.

2002-01-01

[1] We examine the quality of atmospherically deposited ion and isotope signals in an ice core taken from a periodically melting ice field, Lomonosovfonna in central Spitsbergen, Svalbard. The aim is to determine the degree to which the signals are altered by periodic melting of the ice. We use
Rolling-Tooth Core Breakoff and Retention Mechanism

Science.gov (United States)

Badescu, Mircea; Bickler, Donald B.; Sherrit, Stewart; Bar-Cohen, Yoseph; Bao, Xiaoqi; Hudson, Nicolas H.

2011-01-01

Sampling cores requires the controlled breakoff of the core at a known location with respect to the drill end. An additional problem is designing a mechanism that can be implemented at a small scale that is robust and versatile enough to be used for a variety of core samples. This design consists of a set of tubes (a drill tube and an inner tube) and a rolling element (rolling tooth). An additional tube can be used as a sample tube. The drill tube and the inner tube have longitudinal holes with the axes offset from the axis of each tube. The two eccentricities are equal. The inner tube fits inside the drill tube, and the sample tube fits inside the inner tube. While drilling, the two tubes are positioned relative to each other such that the sample tube is aligned with the drill tube axis and core. The drill tube includes teeth and flutes for cuttings removal. The inner tube includes, at the base, the rolling element implemented as a wheel on a shaft in an eccentric slot. An additional slot in the inner tube and a pin in the drill tube limit the relative motion of the two tubes. While drilling, the drill assembly rotates relative to the core and forces the rolling tooth to stay hidden in the slot along the inner tube wall. When the drilling depth has been reached, the drill bit assembly is rotated in the opposite direction, and the rolling tooth is engaged and penetrates into the core. Depending on the strength of the created core, the rolling tooth can score, lock the inner tube relative to the core, start the eccentric motion of the inner tube, and break the core. The tooth and the relative position of the two tubes can act as a core catcher or core-retention mechanism as well. The design was made to fit the core and hole parameters produced by an existing bit; the parts were fabricated and a series of demonstration tests were performed. This invention is potentially applicable to sample return and in situ missions to planets such as Mars and Venus, to moons such
In-vessel coolability and retention of a core melt. Volume 1

Energy Technology Data Exchange (ETDEWEB)

Theofanous, T.G.; Liu, C.; Additon, S.; Angelini, S.; Kymaelaeinen, O.; Salmassi, T. [California Univ., Santa Barbara, CA (United States). Center for Risk Studies and Safety

1996-10-01

The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is physically unreasonable. Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings. The AP600 is particularly favorable to in-vessel retention. Some ideas to enhance the assessment basis as well as performance in this respect, for applications to larger and/or higher power density reactors are also provided.
In-vessel coolability and retention of a core melt. Volume 2

International Nuclear Information System (INIS)

Theofanous, T.G.; Liu, C.; Additon, S.; Angelini, S.; Kymaelaeinen, O.; Salmassi, T.

1996-10-01

The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is physically unreasonable. Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings. The AP600 is particularly favorable to in-vessel retention. Some ideas to enhance the assessment basis as well as performance in this respect, for applications to larger and/or higher power density reactors are also provided
In-vessel coolability and retention of a core melt. Volume 1

International Nuclear Information System (INIS)

Theofanous, T.G.; Liu, C.; Additon, S.; Angelini, S.; Kymaelaeinen, O.; Salmassi, T.

1996-10-01

The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is physically unreasonable. Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings. The AP600 is particularly favorable to in-vessel retention. Some ideas to enhance the assessment basis as well as performance in this respect, for applications to larger and/or higher power density reactors are also provided
Proposed model for fuel-coolant mixing during a core-melt accident

International Nuclear Information System (INIS)

Corradini, M.L.

1983-01-01

If complete failure of normal and emergency coolant flow occurs in a light water reactor, fission product decay heat would eventually cause melting of the reactor fuel and cladding. The core melt may then slump into the lower plenum and later into the reactor cavity and contact residual liquid water. A model is proposed to describe the fuel-coolant mixing process upon contact. The model is compared to intermediate scale experiments being conducted at Sandia. The modelling of this mixing process will aid in understanding three important processes: (1) fuel debris sizes upon quenching in water, (2) the hydrogen source term during fuel quench, and (3) the rate of steam production. Additional observations of Sandia data indicate that the steam explosion is affected by this mixing process
Phenomena in the interaction among a core melt and protective and sacrificial materials

International Nuclear Information System (INIS)

Steinwarz, W.; Koller, W.; Dyllong, N.; Fischer, M.; Hellmann, S.; Lansmann, V.; Nie, M.; Haefner, W.; Alkan, Z.; Andrae, P.; Rensing, B.

2000-01-01

In a postulated core meltdown accident in a light water reactor there are bound to be interactions, in the ex-vessel phase, among the core melt and the structural materials within and below the reactor cavity. In existing plants, these structural materials normally are structural concrete, while future, evolutionary reactor lines are to have sacrificial and protective materials specially designed for this hypothetical case. To add to the state of knowledge about the phenomena occurring, experiments need to be conducted under conditions as realistic as possible. Within the research programs funded by the European Union, the German Federal Ministry for Economics, and the German nuclear power plant operators, experiments on a laboratory as well as an industrial scale on these problems are being carried out in the two projects called CORESA (COrium on REfractory and SAcrificial materials) and ECOSTAR (Ex-vessel COre melt STAbilization Research). The experiments are accompanied by an extensive analytical theoretical program also serving to advance and validate computer codes on the problems under investigation. The projects, which are carried out with international European participation, are expected to allow a concept to be developed for managing postulated accident scenarios involving core meltdown for innovative nuclear power plants, and to provide findings on risk evaluation of plants now in operation so as to further develop accident management measures. (orig.) [de
Evaluation of Melt Behavior with initial Melt Velocity under SFR Severe Accidents

Energy Technology Data Exchange (ETDEWEB)

Heo, Hyo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Jerng, Dong Wook [Chung-Ang Univ, Seoul (Korea, Republic of)

2015-10-15

In the current Korean sodium-cooled fast reactor (SFR) program, early dispersion of the molten metallic fuel within a subchannel is suggested as one of the inherent safety strategies for the initiating phase of hypothetical core disruptive accident (HCDA). The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. Thus, it could be worthwhile to study the horizontal melt injections at lower temperature as a preliminary study in order to identify the melt dispersion phenomena. For this reason, it is required to clarify whether the coolant vapor pressure is the driving force of the melt dispersion with the core region. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition. As the results, the following results are remarked: 1. The upward melt dispersion did not occur for a given melt and coolant temperature in the nonboiling range. Over current range of conditions
Evaluation of Melt Behavior with initial Melt Velocity under SFR Severe Accidents

International Nuclear Information System (INIS)

Heo, Hyo; Bang, In Cheol; Jerng, Dong Wook

2015-01-01

In the current Korean sodium-cooled fast reactor (SFR) program, early dispersion of the molten metallic fuel within a subchannel is suggested as one of the inherent safety strategies for the initiating phase of hypothetical core disruptive accident (HCDA). The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. Thus, it could be worthwhile to study the horizontal melt injections at lower temperature as a preliminary study in order to identify the melt dispersion phenomena. For this reason, it is required to clarify whether the coolant vapor pressure is the driving force of the melt dispersion with the core region. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition. As the results, the following results are remarked: 1. The upward melt dispersion did not occur for a given melt and coolant temperature in the nonboiling range. Over current range of conditions
Fuel Rod Melt Progression Simulation Using Low-Temperature Melting Metal Alloy

International Nuclear Information System (INIS)

Seung Dong Lee; Suh, Kune Y.; GoonCherl Park; Un Chul Lee

2002-01-01

The TMI-2 accident and various severe fuel damage experiments have shown that core damage is likely to proceed through various states before the core slumps into the lower head. Numerous experiments were conducted to address when and how the core can lose its original geometry, what geometries are formed, and in what processes the core materials are transported to the lower plenum of the reactor pressure vessel. Core degradation progresses along the line of clad ballooning, clad oxidation, material interaction, metallic blockage, molten pool formation, melt progression, and relocation to the lower head. Relocation into the lower plenum may occur from the lateral periphery or from the bottom of the core depending upon the thermal and physical states of the pool. Determining the quantities and rate of molten material transfer to the lower head is important since significant amounts of molten material relocated to the lower head can threaten the vessel integrity by steam explosion and thermal and mechanical attack of the melt. In this paper the focus is placed on the melt flow regime on a cylindrical fuel rod utilizing the LAMDA (Lumped Analysis of Melting in Degrading Assemblies) facility at the Seoul National University. The downward relocation of the molten material is a combination of the external film flow and the internal pipe flow. The heater rods are 0.8 m long and are coated by a low-temperature melting metal alloy. The electrical internal heating method is employed during the test. External heating is adopted to simulate the exothermic Zircaloy-steam reaction. Tests are conducted in several quasi-steady-state conditions. Given the variable boundary conditions including the heat flux and the water level, observation is made for the melting location, progression, and the mass of molten material. Finally, the core melt progression model is developed from the visual inspection and quantitative analysis of the experimental data. As the core material relocates
On the sequence of core-melt accidents: Fission product release, source terms and Chernobyl release

Energy Technology Data Exchange (ETDEWEB)

Albrecht, H

1986-01-01

There is a sketch of our ideas on the course of a core melt-out accident in a PWR. There is then a survey of the most important results on fission product release, which were obtained by experiments on the SASCHA melt-out plant. The 3rd part considers questions which are important for determining source terms for the environment and the last part contains some considerations on radioactivity release from the Chernobyl reactor.
Modeling of heat and mass transfer processes during core melt discharge from a reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Dinh, T.N.; Bui, V.A.; Nourgaliev, R.R. [Royal Institute of Technology, Stockholm (Sweden)] [and others

1995-09-01

The objective of the paper is to study heat and mass transfer processes related to core melt discharge from a reactor vessel is a severe light water reactor accident. The phenomenology of the issue includes (1) melt convection in and heat transfer from the melt pool in contact with the vessel lower head wall; (2) fluid dynamics and heat transfer of the melt flow in the growing discharge hole; and (3) multi-dimensional heat conduction in the ablating lower head wall. A program of model development, validation and application is underway (i) to analyse the dominant physical mechanisms determining characteristics of the lower head ablation process; (ii) to develop and validate efficient analytic/computational methods for estimating heat and mass transfer under phase-change conditions in irregular moving-boundary domains; and (iii) to investigate numerically the melt discharge phenomena in a reactor-scale situation, and, in particular, the sensitivity of the melt discharge transient to structural differences and various in-vessel melt progression scenarios. The paper presents recent results of the analysis and model development work supporting the simulant melt-structure interaction experiments.

Evaluation for In-Vessel Retention Capabilities with In-Vessel Injection and External Reactor Vessel Cooling

International Nuclear Information System (INIS)

Lee, Jeong Seong; Ryu, In Chul; Moon, Young Tae

2016-01-01

If the accident has not progressed to the point of substantial changes in the core geometry, establishing adequate cooling is as straightforward as re-establishing flow through the reactor core. However, if the accident has progressed to the point where the core geometry is substantially altered as a result of material melting and relocation, as was the case in the TMI-2 accident, the means of cooling the debris are not as straightforward. From this time on, the reactor core was either completely or nearly covered by water, with high pressure injection flow initiated shortly after three hours into the accident. However, the core debris was not coolable in this configuration and a substantial quantity of molten core material drained into the bypass region, with approximately twenty metric tons of molten debris draining into the reactor pressure vessel (RPV) lower head. Hence, the core configuration developed at approximately three hours into the accident was not coolable, even submerged in water. The purpose of this paper is to evaluate in-vessel retention capabilities with in-vessel injection (IVI) and external reactor vessel cooling (ERVC) available in a reactor application by using the integrated severe accident analysis code. The MAAP5 models were improved to facilitate evaluation of the in-vessel retention capability of APR1400. In-vessel retention capabilities have been analyzed for the APR1400 using the MAAP5.03 code. The results show that in-vessel retention is feasible when in-vessel injection is initiated within a relatively short time frame under the simulation condition used in the present study
Evaluation for In-Vessel Retention Capabilities with In-Vessel Injection and External Reactor Vessel Cooling

Energy Technology Data Exchange (ETDEWEB)

Lee, Jeong Seong; Ryu, In Chul; Moon, Young Tae [KEPCO Engineering and Construction Co. Ltd., Deajeon (Korea, Republic of)

2016-10-15

If the accident has not progressed to the point of substantial changes in the core geometry, establishing adequate cooling is as straightforward as re-establishing flow through the reactor core. However, if the accident has progressed to the point where the core geometry is substantially altered as a result of material melting and relocation, as was the case in the TMI-2 accident, the means of cooling the debris are not as straightforward. From this time on, the reactor core was either completely or nearly covered by water, with high pressure injection flow initiated shortly after three hours into the accident. However, the core debris was not coolable in this configuration and a substantial quantity of molten core material drained into the bypass region, with approximately twenty metric tons of molten debris draining into the reactor pressure vessel (RPV) lower head. Hence, the core configuration developed at approximately three hours into the accident was not coolable, even submerged in water. The purpose of this paper is to evaluate in-vessel retention capabilities with in-vessel injection (IVI) and external reactor vessel cooling (ERVC) available in a reactor application by using the integrated severe accident analysis code. The MAAP5 models were improved to facilitate evaluation of the in-vessel retention capability of APR1400. In-vessel retention capabilities have been analyzed for the APR1400 using the MAAP5.03 code. The results show that in-vessel retention is feasible when in-vessel injection is initiated within a relatively short time frame under the simulation condition used in the present study.
Analysis of core melt accident in Fukushima Daiichi-Unit 1 nuclear reactor

International Nuclear Information System (INIS)

Tanabe, Fumiya

2011-01-01

In order to obtain a profound understanding of the serious situation in Unit 1 and Unit 2/3 reactors of Fukushima Daiichi Nuclear Power Station (hereafter abbreviated as 1F1 and 1F2/3, respectively), which was directly caused by tsunami due to a huge earthquake on 11 March 2011, analyses of severe core damage are performed. In the present report, the analysis method and 1F1 analysis are described. The analysis is essentially based on the total energy balance in the core. In the analysis, the total energy vs. temperature curve is developed for each reactor, which is based on the estimated core materials inventory and material property data. Temperature and melt fraction are estimated by comparing the total energy curve with the total stored energy in the core material. The heat source is the decay heat of fission products and actinides together with reaction heat from the zirconium steam reaction. (author)
The effect of melt composition on metal-silicate partitioning of siderophile elements and constraints on core formation in the angrite parent body

Science.gov (United States)

Steenstra, E. S.; Sitabi, A. B.; Lin, Y. H.; Rai, N.; Knibbe, J. S.; Berndt, J.; Matveev, S.; van Westrenen, W.

2017-09-01

We present 275 new metal-silicate partition coefficients for P, S, V, Cr, Mn, Co, Ni, Ge, Mo, and W obtained at moderate P (1.5 GPa) and high T (1683-1883 K). We investigate the effect of silicate melt composition using four end member silicate melt compositions. We identify possible silicate melt dependencies of the metal-silicate partitioning of lower valence elements Ni, Ge and V, elements that are usually assumed to remain unaffected by changes in silicate melt composition. Results for the other elements are consistent with the dependence of their metal-silicate partition coefficients on the individual major oxide components of the silicate melt composition suggested by recently reported parameterizations and theoretical considerations. Using multiple linear regression, we parameterize compiled metal-silicate partitioning results including our new data and report revised expressions that predict their metal-silicate partitioning behavior as a function of P-T-X-fO2. We apply these results to constrain the conditions that prevailed during core formation in the angrite parent body (APB). Our results suggest the siderophile element depletions in angrite meteorites are consistent with a CV bulk composition and constrain APB core formation to have occurred at mildly reducing conditions of 1.4 ± 0.5 log units below the iron-wüstite buffer (ΔIW), corresponding to a APB core mass of 18 ± 11%. The core mass range is constrained to 21 ± 8 mass% if light elements (S and/or C) are assumed to reside in the APB core. Incorporation of light elements in the APB core does not yield significantly different redox states for APB core-mantle differentiation. The inferred redox state is in excellent agreement with independent fO2 estimates recorded by pyroxene and olivine in angrites.
Study on severe fuel damage and in-vessel melt progression

International Nuclear Information System (INIS)

Kim, Hee Dong; Kim, Sang Baik; Lee, Gyu Jung

1992-06-01

In-vessel core melt progression describes the progression of the state of a reactor core from core uncovery up to reactor vessel melt through in uncovered accidents or through temperature stabilization in accidents recovered by core reflooding. Melt progression can be thought as two parts; early melt progression and late melt progression. Early phase of core melt progression includes the progression of core material melting and relocation, which mostly consist of metallic materials. On the other hand, the late phase of core melt progression involves ceramic material melt and relocation to the lower plenum and heat-up the reactor vessel lower head. A large number of information are available for the early melt progression through experiments such as SFD, DF, FLHT test and utilized in the severe accident analysis codes. However, understanding of the late phase melt progression phenomenology is based primary on TMI-2 core examinations and not much experimental information is available. Especilally, the great uncertainties exist in vessel failure mode, melt composition, mass, and temperature. Further research is planned to perform to reduce the uncertainties in understanding of core melt down accidents as parts of long term melt progression research program. A study on the core melt progression at KAERI has been being performed through the Severe Accident Research Program with USNRC. KAERI staff had participated in the PBF SFD experiments at INEL and analyses of experiments were performed using SCDAP code. Experiments of core melt program have not been carried out at KAERI yet. It is planned that further research on core melt down accidents will be performed, which is related to design of future generations of nuclear reactors as parts of long-term project for improvement of nuclear reactor safety. (Author)
Nickel and helium evidence for melt above the core-mantle boundary.

Science.gov (United States)

Herzberg, Claude; Asimow, Paul D; Ionov, Dmitri A; Vidito, Chris; Jackson, Matthew G; Geist, Dennis

2013-01-17

High (3)He/(4)He ratios in some basalts have generally been interpreted as originating in an incompletely degassed lower-mantle source. This helium source may have been isolated at the core-mantle boundary region since Earth's accretion. Alternatively, it may have taken part in whole-mantle convection and crust production over the age of the Earth; if so, it is now either a primitive refugium at the core-mantle boundary or is distributed throughout the lower mantle. Here we constrain the problem using lavas from Baffin Island, West Greenland, the Ontong Java Plateau, Isla Gorgona and Fernandina (Galapagos). Olivine phenocryst compositions show that these lavas originated from a peridotite source that was about 20 per cent higher in nickel content than in the modern mid-ocean-ridge basalt source. Where data are available, these lavas also have high (3)He/(4)He. We propose that a less-degassed nickel-rich source formed by core-mantle interaction during the crystallization of a melt-rich layer or basal magma ocean, and that this source continues to be sampled by mantle plumes. The spatial distribution of this source may be constrained by nickel partitioning experiments at the pressures of the core-mantle boundary.
Crust behavior and erosion rate prediction of EPR sacrificial material impinged by core melt jet

Energy Technology Data Exchange (ETDEWEB)

Li, Gen; Liu, Ming, E-mail: ming.liu@mail.xjtu.edu.cn; Wang, Jinshi; Chong, Daotong; Yan, Junjie

2017-04-01

Highlights: • A numerical code was developed to analyze melt jet-concrete interaction in the frame of MPS method. • Crust and ablated concrete layer at UO{sub 2}-ZrO{sub 2} melt and concrete interface periodically developed and collapsed. • Concrete surface temperature fluctuated around a low temperature and ablation temperature. • Concrete erosion by Fe-Zr melt jet was significantly faster than that by UO{sub 2}-ZrO{sub 2} melt jet. - Abstract: Sacrificial material is a special ferro-siliceous concrete, designed in the ex-vessel core melt stabilization system of European Pressurized water Reactor (EPR). Given a localized break of RPV lower head, the melt directly impinges onto the dry concrete in form of compact jet. The concrete erosion behavior influences the failure of melt plug, and further affects melt spreading. In this study, a numerical code was developed in the frame of Moving Particle Semi-implicit (MPS) method, to analyze the crust behavior and erosion rate of sacrificial concrete, impinged by prototypic melt jet. In validation of numerical modeling, the time-dependent erosion depth and erosion configuration matched well with the experimental data. Sensitivity study of sacrificial concrete erosion indicates that the crust and ablated concrete layer presented at UO{sub 2}-ZrO{sub 2} melt and concrete interface, whereas no crust could be found in the interaction of Fe-Zr melt with concrete. The crust went through stabilization-fracture-reformation periodic process, accompanied with accumulating and collapsing of molten concrete layer. The concrete surface temperature fluctuated around a low temperature and ablation temperature. It increased as the concrete surface layer was heated to melting, and dropped down when the cold concrete was revealed. The erosion progression was fast in the conditions of small jet diameter and large concrete inclination angle, and it was significantly faster in the erosion by metallic melt jet than by oxidic melt jet.
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
Sulfur Saturation Limits in Silicate Melts and their Implications for Core Formation Scenarios for Terrestrial Planets

Science.gov (United States)

Holzheid, Astrid; Grove, Timothy L.

2002-01-01

This study explores the controls of temperature, pressure, and silicate melt composition on S solubility in silicate liquids. The solubility of S in FeO-containing silicate melts in equilibrium with metal sulfide increases significantly with increasing temperature but decreases with increasing pressure. The silicate melt structure also exercises a control on S solubility. Increasing the degree of polymerization of the silicate melt structure lowers the S solubility in the silicate liquid. The new set of experimental data is used to expand the model of Mavrogenes and O'Neill(1999) for S solubility in silicate liquids by incorporating the influence of the silicate melt structure. The expected S solubility in the ascending magma is calculated using the expanded model. Because the negative pressure dependence of S solubility is more influential than the positive temperature dependence, decompression and adiabatic ascent of a formerly S-saturated silicate magma will lead to S undersaturation. A primitive magma that is S-saturated in its source region will, therefore, become S-undersaturated as it ascends to shallower depth. In order to precipitate magmatic sulfides, the magma must first cool and undergo fractional crystallization to reach S saturation. The S content in a metallic liquid that is in equilibrium with a magma ocean that contains approx. 200 ppm S (i.e., Earth's bulk mantle S content) ranges from 5.5 to 12 wt% S. This range of S values encompasses the amount of S (9 to 12 wt%) that would be present in the outer core if S is the light element. Thus, the Earth's proto-mantle could be in equilibrium (in terms of the preserved S abundance) with a core-forming metallic phase.
In-situ rock melting applied to lunar base construction and for exploration drilling and coring on the moon

International Nuclear Information System (INIS)

Rowley, J.C.; Neudecker, J.W.

1984-01-01

An excavation technology based upon melting of rock and soil has been extensively developed at the prototype hardware and conceptual design levels for terrestrial conditions. Laboratory and field tests of rock-melting penetration have conclusively indicated that this excavation method is insensitive to rock, soil types, and conditions. Especially significant is the ability to form in-place glass linings or casings on the walls of boreholes, tunnels, and shafts. These factors indicate the unique potential for in situ construction of primary lunar base facilities. Drilling and coring equipment for resource exploration on the moon can also be devised that are largely automated and remotely operated. It is also very likely that lunar melt-glasses will have changed mechanical properties when formed in anhydrous and hard vacuum conditions. Rock melting experiments and prototype hardware designs for lunar rock-melting excavation applications are suggested
Planetesimal core formation with partial silicate melting using in-situ high P, high T, deformation x-ray microtomography

Science.gov (United States)

Anzures, B. A.; Watson, H. C.; Yu, T.; Wang, Y.

2017-12-01

Differentiation is a defining moment in formation of terrestrial planets and asteroids. Smaller planetesimals likely didn't reach high enough temperatures for widescale melting. However, we infer that core formation must have occurred within a few million years from Hf-W dating. In lieu of a global magma ocean, planetesimals likely formed through inefficient percolation. Here, we used in-situ high temperature, high pressure, x-ray microtomography to track the 3-D evolution of the sample at mantle conditions as it underwent shear deformation. Lattice-Boltzmann simulations for permeability were used to characterize the efficiency of melt percolation. Mixtures of KLB1 peridotite plus 6.0 to 12.0 vol% FeS were pre-sintered to achieve an initial equilibrium microstructure, and then imaged through several consecutive cycles of heating and deformation. The maximum calculated melt segregation velocity was found to be 0.37 cm/yr for 6 vol.% FeS and 0.61 cm/year for 12 vol.% FeS, both below the minimum velocity of 3.3 cm/year required for a 100km planetesimal to fully differentiate within 3 million years. However, permeability is also a function of grain size and thus the samples having smaller grains than predicted for small planetesimals could have contributed to low permeability and also low migration velocity. The two-phase (sulfide melt and silicate melt) flow at higher melt fractions (6 vol.% and 12 vol.% FeS) was an extension of a similar study1 containing only sulfide melt at lower melt fraction (4.5 vol.% FeS). Contrary to the previous study, deformation did result in increased permeability until the sample was sheared by twisting the opposing Drickamer anvils by 360 degrees. Also, the presence of silicate melt caused the FeS melt to coalesce into less connected pathways as the experiment with 6 vol.% FeS was found to be less permeable than the one with 4.5 vol.% FeS but without any partial melt. The preliminary data from this study suggests that impacts as well as
On heat transfer characteristics of real and simulant melt pool experiments

Energy Technology Data Exchange (ETDEWEB)

Dinh, T.N.; Nourgaliev R.R.; Sehgal, B.R. [Royal Institute of Technology, Stockholm (Sweden)

1995-09-01

The paper presents results of analytical studies of natural convection heat transfer in scaled and/or simulant melt pool experiments related to the PWR in-vessel melt retention issue. Specific reactor-scale effects of a large decay-heated core melt pool in the reactor pressure vessel lower plenum are first reviewed, and then the current analytical capability of describing physical processes under prototypical situations is examined. Experiments and experimental approaches are analysed by focusing on their ability to represent prototypical situations. Calculations are carried out in order to assess the significance of some selected effects, including variations in melt properties, pool geometry and heating conditions. Rayleigh numbers in the present analysis are limited to 10{sup 12}, where uncertainties in turbulence modeling are not overriding other uncertainties. The effects of fluid Prandtl number on heat transfer to the lowermost part of cooled pool walls are examined for square and semicircular cavities. Calculations are performed also to explore limitations of using side-wall heating and direct electrical heating in reproducing the physical picture of interest. Needs for further experimental and analytical efforts are discussed as well.
External cooling: The SWR 1000 severe accident management strategy. Part 1: motivation, strategy, analysis: melt phase, vessel integrity during melt-water interaction

International Nuclear Information System (INIS)

Kolev, Nikolay Ivanov

2004-01-01

This paper provides the description of the basics behind design features for the severe accident management strategy of the SWR 1000. The hydrogen detonation/deflagration problem is avoided by containment inertization. In-vessel retention of molten core debris via water cooling of the external surface of the reactor vessel is the severe accident management concept of the SWR 1000 passive plant. During postulated bounding severe accidents, the accident management strategy is to flood the reactor cavity with Core Flooding Pool water and to submerge the reactor vessel, thus preventing vessel failure in the SWR 1000. Considerable safety margins have determined by using state of the art experiment and analysis: regarding (a) strength of the vessel during the melt relocation and its interaction with water; (b) the heat flux at the external vessel wall; (c) the structural resistance of the hot structures during the long term period. Ex-vessel events are prevented by preserving the integrity of the vessel and its penetrations and by assuring positive external pressure at the predominant part of the external vessel in the region of the molten corium pool. Part 1 describes the motivation for selecting this strategy, the general description of the strategy and the part of the analysis associated with the vessel integrity during the melt-water interaction. (author)
Assessment of uncertainties in core melt phenomenology and their impact on risk at the Z/IP facilities

International Nuclear Information System (INIS)

Pratt, W.T.; Ludewig, H.; Bari, R.A.; Meyer, J.F.

1983-01-01

An evaluation of core meltdown accidents in the Z/IP facilities has been performed. Containment event trees have been developed to relate the progression of a given accident to various potential containment building failure modes. An extensive uncertainty analysis related to core melt phenomenology has been performed. A major conclusion of the study is that large variations in parameters associated with major phenomenological uncertainties have a relatively minor impact on risk when external initiators are considered. This is due to the inherent capability fo the Z/IP containment buildings to contain a wide range of core meltdown accidents. 12 references, 2 tables
The influence of chemistry on core melt accidents

International Nuclear Information System (INIS)

Liljenzin, J.O.

1990-01-01

Chemical reactions play an important role in assessing the safety of nuclear power plants. The main source of heat in the early stage of an accident is due to a chemical reaction between steam and the circonium encapsulating the nuclear fuel. The heating and melting of fuel leads to a release of fission products which rapidly condense to form particles suspended in the surrounding gas. These aerosols are the main carriers of radioactivity as they may transport active material from the reactor vessel into the reactor containment building where it is deposited. The content of fission products in the aerosol particles and their chemical form determine their interaction with water molecules. Chemical forces laed to an absorption of water in the particles which transforms them into droplets with increased mass. The particles become spherical and hence deposit more rapidly on surrounding surfaces. There is a rapid reaction between boron carbide and stainless steel in the control blades of boiling water reactors. There is only a small formation of boric acid. This leads to a smaller formation of volatile iodine compounds. But the alloying process is likely to cause melting of the control blades so the are removed from the reactor core, a process which may have negative secondary effects. It has been found that a series of materials that are present in the reactor containment are likely to participate in various chemical reactions during an accident. Among these are electric cables, motors, thermal insulation, surface coatings and sheet metal. Metallic surface coatings and sheet metal can be some of the main sources of hydrogen. Effects from chemical reactions can be more accurately predicted by the new SHMAPP code, developed within this project, combining thermal, hydraulic and chemical phenomena. (AB)
Examination of off-site emergency protective measures for core melt accidents

International Nuclear Information System (INIS)

Aldrich, D.C.; Ericson, D.M. Jr.; Jones, R.B.

1978-01-01

Results from the Reactor Safety Study (RSS) have shown that to cause significant impacts off-site, i.e., sufficient quantities of biologically important radionuclides released, it is necessary to have a core melt accident. To mitigate the impact of such potential accidents, the design of appropriate emergency response actions requires information as to the relative merit of publicly available protective measures. In order to provide this information, a study using the consequence model developed for the RSS is being conducted to evaluate (in terms of reduced public health effects and dose exposure) potential off-site protective strategies. The paper describes the methods being used in the study as well as the results and conclusions obtained
Ex-vessel boiling experiments: laboratory- and reactor-scale testing of the flooded cavity concept for in-vessel core retention. Pt. II. Reactor-scale boiling experiments of the flooded cavity concept for in-vessel core retention

International Nuclear Information System (INIS)

Chu, T.Y.; Bentz, J.H.; Slezak, S.E.; Pasedag, W.F.

1997-01-01

For pt.I see ibid., p.77-88 (1997). This paper summarizes the results of a reactor-scale ex-vessel boiling experiment for assessing the flooded cavity design of the heavy water new production reactor. The simulated reactor vessel has a cylindrical diameter of 3.7 m and a torispherical bottom head. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling mainly results from the gravity head, which in turn results from flooding the side of the reactor vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid-solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion. The results show that, under prototypic heat load and heat flux distributions, the flooded cavity will be effective for in-vessel core retention in the heavy water new production reactor. The results also demonstrate that the heat dissipation requirement for in-vessel core retention, for the central region of the lower head of an AP-600 advanced light water reactor, can be met with the flooded cavity design. (orig.)
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.
Melting in super-earths.

Science.gov (United States)

Stixrude, Lars

2014-04-28

We examine the possible extent of melting in rock-iron super-earths, focusing on those in the habitable zone. We consider the energetics of accretion and core formation, the timescale of cooling and its dependence on viscosity and partial melting, thermal regulation via the temperature dependence of viscosity, and the melting curves of rock and iron components at the ultra-high pressures characteristic of super-earths. We find that the efficiency of kinetic energy deposition during accretion increases with planetary mass; considering the likely role of giant impacts and core formation, we find that super-earths probably complete their accretionary phase in an entirely molten state. Considerations of thermal regulation lead us to propose model temperature profiles of super-earths that are controlled by silicate melting. We estimate melting curves of iron and rock components up to the extreme pressures characteristic of super-earth interiors based on existing experimental and ab initio results and scaling laws. We construct super-earth thermal models by solving the equations of mass conservation and hydrostatic equilibrium, together with equations of state of rock and iron components. We set the potential temperature at the core-mantle boundary and at the surface to the local silicate melting temperature. We find that ancient (∼4 Gyr) super-earths may be partially molten at the top and bottom of their mantles, and that mantle convection is sufficiently vigorous to sustain dynamo action over the whole range of super-earth masses.
Studies of Behavior Melting Temperature Characteristics for Multi Thermocouple In-Core Instrument Assembly

Energy Technology Data Exchange (ETDEWEB)

Shin, Donghyup; Chae, Myoungeun; Kim, Sungjin; Lee, Kyulim [Woojin inc, Hwasung (Korea, Republic of)

2015-05-15

Bottom-up type in-core instruments (ICIs) are used for the pressurized water reactors of OPR-1000, APR- 1400 in order to measure neutron flux and temperature in the reactor. It is a well-known technique and a proven design using years in the nuclear field. ICI consists of one pair of K-type thermocouple, five self-powered neutron detectors (SPNDs) and one back ground detector. K-type thermocouple's purpose is to measure the core exit temperature (CET) in the reactor. The CET is a very important factor for operating nuclear power plants and it is 327 .deg. C when generally operating the reactor in the nuclear power plant(NPP) in case of OPR- 1000. If the CET will exceed 650 .deg. C, Operators in the main control room should be considered to be an accident situation in accordance with a severe accident management guidance(SAMG). The Multi Thermocouple ICI is a new designed ICI assuming severe accident conditions. It consists of four more thermocouples than the existing design, so it has five Ktype thermocouples besides the thermocouple measuring CET is located in the same elevation as the ICI. Each thermocouple is able to be located in the desired location as required. The Multi Thermocouple ICI helps to measure the temperature distribution of the entire reactor. In addition, it will measure certain point of melted core because of the in-vessel debris of nuclear fuel when an accident occurs more seriously. In this paper, to simulate a circumstance such as a nuclear reactor severe accident was examined. In this study, the K-type thermocouples of Multi Thermocouple ICI was confirmed experimentally to be able to measure up to 1370 .deg. C before the thermocouples have been melted. And after the thermocouples were melted by debris, it was able to be monitored that the signal of EMF directed the infinite value of voltage. Therefore through the results of the test, it can be assumed that if any EMF data among the Multi Thermocouple ICI will direct the infinite value

Studies of Behavior Melting Temperature Characteristics for Multi Thermocouple In-Core Instrument Assembly

International Nuclear Information System (INIS)

Shin, Donghyup; Chae, Myoungeun; Kim, Sungjin; Lee, Kyulim

2015-01-01

Bottom-up type in-core instruments (ICIs) are used for the pressurized water reactors of OPR-1000, APR- 1400 in order to measure neutron flux and temperature in the reactor. It is a well-known technique and a proven design using years in the nuclear field. ICI consists of one pair of K-type thermocouple, five self-powered neutron detectors (SPNDs) and one back ground detector. K-type thermocouple's purpose is to measure the core exit temperature (CET) in the reactor. The CET is a very important factor for operating nuclear power plants and it is 327 .deg. C when generally operating the reactor in the nuclear power plant(NPP) in case of OPR- 1000. If the CET will exceed 650 .deg. C, Operators in the main control room should be considered to be an accident situation in accordance with a severe accident management guidance(SAMG). The Multi Thermocouple ICI is a new designed ICI assuming severe accident conditions. It consists of four more thermocouples than the existing design, so it has five Ktype thermocouples besides the thermocouple measuring CET is located in the same elevation as the ICI. Each thermocouple is able to be located in the desired location as required. The Multi Thermocouple ICI helps to measure the temperature distribution of the entire reactor. In addition, it will measure certain point of melted core because of the in-vessel debris of nuclear fuel when an accident occurs more seriously. In this paper, to simulate a circumstance such as a nuclear reactor severe accident was examined. In this study, the K-type thermocouples of Multi Thermocouple ICI was confirmed experimentally to be able to measure up to 1370 .deg. C before the thermocouples have been melted. And after the thermocouples were melted by debris, it was able to be monitored that the signal of EMF directed the infinite value of voltage. Therefore through the results of the test, it can be assumed that if any EMF data among the Multi Thermocouple ICI will direct the infinite value
Core degradation and fission product release

International Nuclear Information System (INIS)

Wright, R.W.; Hagen, S.J.L.

1992-01-01

Experiments on core degradation and melt progression in severe LWR accidents have provided reasonable understanding of the principal processes involved in the early phase of melt progression that extends through core degradation and metallic material melting and relocation. A general but not a quantitative understanding of late phase melt progression that involves ceramic material melting and relocation has also been obtained, primarily from the TMI-2 core examination. A summary is given of the current state of knowledge on core degradation and melt progression obtained from these integral experiments and of the principal remaining significant uncertainties. A summary is also given of the principal results on in-vessel fission product release obtained from these experiments. (author). 8 refs, 5 figs, 3 tabs
Experiments and analyses on melt-structure-water interactions during severe accidents

International Nuclear Information System (INIS)

Seghal, B.R.; Dinh, T.N.; Bui, V.A.; Green, J.A.; Nourgaliev, R.R.; Okkonen, T.O.; Dinh, A.T.

1998-04-01

This report is the final report for the research project Melt Structure Water Interactions (MSWI). It describes results of analytical and experimental studies concerning MSWI during the course of a hypothetical core meltdown accident in a LWR. Emphasis has been placed on phenomena which govern vessel failure mode and timing and the mechanisms and properties which govern the fragmentation and breakup of melt jets and droplets. It was found that: 2-D effects significantly diminished the focusing effect of an overlying metallic layer on top of an oxide melt pool. This result improves the feasibility of in-vessel retention of a melt pool through external cooling of the lower head; phenomena related to hole ablation and melt discharge, in the event of vessel failure, are affected significantly by crust formation; the jet fragmentation process is a function of many related phenomena. The fragmentation rate depends not only on the traditional parameters but also on the melt physical properties, which change as the melt cools down from liquid to solid temperature; film boiling was investigated by developing a two-phase flow model and inserting it in a multi-D fluid dynamics code. It was concluded that the thickness of the film on the surface of a melt jet would be small and that the effects of the film on the process should not be large. This conclusion is contrary to the modeling employed in some other codes. The computer codes were developed and validated against the data obtained in the MSWI Project. The melt vessel interaction thermal analysis code describes the process of melt pool formation and convection and the resulting vessel thermal loadings. In addition, several innovative models were developed to describe the melt-water interaction process. The code MELT-3D treats the melt jet as a collection of particles whose movement is described with a three-dimensional Eulerian formulation. The model (SIPHRA) tracks the melt jet with an additional equation, using the
Melt cooling by bottom flooding: The experiment CometPC-H3. Ex-vessel core melt stabilization research

International Nuclear Information System (INIS)

Alsmeyer, H.; Cron, T.; Merkel, G.; Schmidt-Stiefel, S.; Tromm, W.; Wenz, T.

2003-03-01

The CometPC-H3 experiment was performed to investigate melt cooling by water addition to the bottom of the melt. The experiment was performed with a melt mass of 800 kg, 50% metal and 50% oxide, and 300 kW typical decay heat were simulated in the melt. As this was the first experiment after repair of the induction coil, attention was given to avoid overload of the induction coil and to keep the inductor voltage below critical values. Therefore, the height of the sacrificial concrete layer was reduced to 5 cm only, and the height of the porous concrete layers was also minimized to have a small distance and good coupling between heated melt and induction coil. After quite homogeneous erosion of the upper sacrificial concrete layer, passive bottom flooding started from the porous concrete after 220 s with 1.3 liter water/s. The melt was safely stopped, arrested and cooled. The porous, water filled concrete was only slightly attacked by the hot melt in the upper 25 mm of one sector of the coolant device. The peak cooling rate in the early contact phase of coolant water and melt was 4 MW/m 2 , and exceeded the decay heat by one order of magnitude. The cooling rate remarkably dropped, when the melt was covered by the penetrating water and a surface crust was formed. Volcanic eruptions from the melt during the solidification process were observed from 360 - 510 s and created a volcanic dome some 25 cm high, but had only minor effect on the generation of a porous structure, as the expelled melt solidified mostly with low porosity. Unfortunately, decay heat simulation in the melt was interrupted at 720 s by an incorrect safety signal, which excluded further investigation of the long term cooling processes. At that time, the melt was massively flooded by a layer of water, about 80 cm thick, and coolant water inflow was still 1 l/s. The melt had reached a stable situation: Downward erosion was stopped by the cooling process from the water filled, porous concrete layer. Top
The matter of probability controlling melting of nuclear ship reactor

International Nuclear Information System (INIS)

Pihowicz, W.; Sobczyk, S.

2008-01-01

In the first part of this work beside description of split power, power of radioactivity disintegration and afterpower and its ability to extinguish, the genera condition of melting nuclear reactor core and its detailed versions were described. This paper also include the description of consequences melting nuclear reactor core both in case of stationary and mobile (ship) reactor and underline substantial differences. Next, fulfilled with succeed, control under melting of stationary nuclear reactor core was characterized.The middle part describe author's idea of controlling melting of nuclear ship reactor core. It is based on: - the suggestion of prevention pressure's untightness in safety tank of nuclear ship reactor by '' corium '' - and the suggestion of preventing walls of this tank from melting by '' corium ''. In the end the technological and construction barriers of the prevention from melting nuclear ship reactor and draw conclusions was presented. (author)
A fast running method for predicting the efficiency of core melt spreading for application in ASTEC

International Nuclear Information System (INIS)

Spengler, C.

2010-01-01

The integral Accident Source Term Evaluation Code (ASTEC) is jointly developed by the French Institut de Radioprotection et de Surete Nucleaire (IRSN) and the German Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH to simulate the complete scenario of a hypothetical severe accident in a nuclear light water reactor, from the initial event until the possible radiological release of fission products out of the containment. In the frame of the new series of ASTEC V2 versions appropriate model extensions to the European Pressurised Water Reactor (EPR) are under development. With view to assessing with ASTEC the proper operation of the ex-vessel melt retention and coolability concept of the EPR with regard to melt spreading an approximation of the area finally covered by the corium and of the distance run by the corium front before freezing is required. A necessary capability of ASTEC is in a first step to identify such boundary cases, for which there is a potential that the melt will freeze before the spreading area is completely filled. This paper presents a fast running method for estimating the final extent of the area covered with melt on which a simplified criterion in ASTEC for detecting such boundary cases will be based. If a boundary case is detected the application of a more-detailed method might be necessary to assess further the consequences for the accident sequence. The major objective here is to provide a reliable method for estimating the final result of the spreading and not to provide highly detailed methods to simulate the dynamics of the transient process. (orig.)
Melt cooling by bottom flooding. The COMET core-catcher concept

International Nuclear Information System (INIS)

Foit, Jerzy Jan; Alsmeyer, Hans; Tromm, Walter; Buerger, Manfred; Journeau, Christophe

2009-01-01

The COMET concept has been developed to cool an ex-vessel corium melt in case of a hypothetical severe accident leading to vessel melt-through. After erosion of a sacrificial concrete layer the melt is passively flooded by bottom injection of coolant water. The open porosities and large surface that are generated during melt solidification form a porous permeable structure that is permanently filled with the evaporating water and thus allows an efficient short-term as well as long-term removal of the decay heat. The advantages of this concept are the fast cool-down and complete solidification of the melt within less than one hour typically. This stops further release of fission products from the corium. A drawback may be the fast release of steam during the quenching process. Several experimental series have been performed by FZK (Germany) to test and optimise the functionality of the different variants of the COMET concept. Thermite generated melts of iron and aluminium oxide were used. The large scale COMET-H test series with sustained inductive heating includes nine experiments performed with an array of water injection channels embedded in a sacrificial concrete layer. Variation of the water inlet pressure and melt height showed that melts up to 50 cm height can be safely cooled with an overpressure of the coolant water of 0.2 bar. The CometPC concept is based on cooling by flooding the melt from the bottom through layers of porous, water filled concrete. The third variant of the COMET design, CometPCA, uses a layer of porous, water filled concrete CometPCA from which flow channels protrude into the layer of sacrificial concrete. This modified concept combines the advantages of the original COMET concept with flow channels and the high resistance of a water-filled porous concrete layer against downward melt attack. Four large scale CometPCA experiments (FZK, Germany) have demonstrated an efficient cooling of melts up to 50 cm height using the recommended water
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)
Effect of the in- and ex-vessel dual cooling on the retention of an internally heated melt pool in a hemispherical vessel

International Nuclear Information System (INIS)

Ahn, K.I.; Kim, B.S.; Kim, D.H.

2001-01-01

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)
Simulant - water experiments to characterize the debris bed formed in severe core melt accidents

International Nuclear Information System (INIS)

Mathai, Amala M.; Anandan, J.; Sharma, Anil Kumar; Murthy, S.S.; Malarvizhi, B.; Lydia, G.; Das, Sanjay Kumar; Nashine, B.K.; Selvaraj, P.

2015-01-01

Molten Fuel Coolant Interaction (WO) and debris bed configuration on the core catcher plate assumes importance in assessing the Post Accident Heat Removal (PARR) of a heat generating debris bed. The key factors affecting the coolability of the debris bed are the bed porosity, morphology of the fragmented particles, degree of spreading/heaping of the debris on the core catcher and the fraction of lump formed. Experiments are conducted to understand the fragmentation kinetics and subsequent debris bed formation of molten woods metal in water at interface temperatures near the spontaneous nucleation temperature of water. Morphology of the debris particles is investigated to understand the fragmentation mechanisms involved. The spreading behavior of the debris on the catcher plate and the particle size distribution are presented for 5 kg and 10 kg melt inventories. Porosity of the undisturbed bed on the catcher plate is evaluated using a LASER sensor technique. (author)
Corium melt researches at VESTA test facility

Directory of Open Access Journals (Sweden)

Hwan Yeol Kim

2017-10-01

Full Text Available VESTA (Verification of Ex-vessel corium STAbilization and VESTA-S (-small test facilities were constructed at the Korea Atomic Energy Research Institute in 2010 to perform various corium melt experiments. Since then, several tests have been performed for the verification of an ex-vessel core catcher design for the EU-APR1400. Ablation tests of an impinging ZrO2 melt jet on a sacrificial material were performed to investigate the ablation characteristics. ZrO2 melt in an amount of 65–70 kg was discharged onto a sacrificial material through a well-designed nozzle, after which the ablation depths were measured. Interaction tests between the metallic melt and sacrificial material were performed to investigate the interaction kinetics of the sacrificial material. Two types of melt were used: one is a metallic corium melt with Fe 46%, U 31%, Zr 16%, and Cr 7% (maximum possible content of U and Zr for C-40, and the other is a stainless steel (SUS304 melt. Metallic melt in an amount of 1.5–2.0 kg was delivered onto the sacrificial material, and the ablation depths were measured. Penetration tube failure tests were performed for an APR1400 equipped with 61 in-core instrumentation penetration nozzles and extended tubes at the reactor lower vessel. ZrO2 melt was generated in a melting crucible and delivered down into an interaction crucible where the test specimen is installed. To evaluate the tube ejection mechanism, temperature distributions of the reactor bottom head and in-core instrumentation penetration were measured by a series of thermocouples embedded along the specimen. In addition, lower vessel failure tests for the Fukushima Daiichi nuclear power plant are being performed. As a first step, the configuration of the molten core in the plant was investigated by a melting and solidification experiment. Approximately 5 kg of a mixture, whose composition in terms of weight is UO2 60%, Zr 10%, ZrO2 15%, SUS304 14%, and B4C 1%, was melted in a
Radioactive contamination of Danish territory after core-melt accidents at the Barsebaeck power plant

International Nuclear Information System (INIS)

Gjoerup, H.L.; Jensen, N.O.; Hedemann Jensen, P.; Kristensen, L.; Nielsen, O.J.; Petersen, E.L.; Petersen, T.; Roed, J.; Thykier-Nielsen, S.; Heikel Vinter, F.; Warming, L.; Aarkrog, A.

1982-03-01

An assessment is made of the radioactive contamination of Danish territory in the event of a core-melt accident at the Barsebaeck nuclear power plant in Sweden. Accidents including both core melt-down and containment failure are considered. Consequences are calculated for a BWR-3 release under common meteorological conditions and for a BWR-2 release under extreme meteorological conditions. Calculations are based on experiments and theoretical work relating to deposition velocities for different types of surface, shielding effect of structures, and weathering. The effects are described of different dose-reducing measures, e.g., decontamination, relocation, destruction of contaminated foodstuffs. The collective effective dose equivalent from external gamma radiation from deposited activity integrated over a time period of 30 years, is calculated to be 3.6 Megamanrem in the BWR-3 case without dose-reducing measures. For the BWR-2 case, the corresponding dose is approx. 41 Megamanrem. A combination of temporary relocation, hosing of roads etc. and digging of gardens is estimated to reduce these doses to approx. 2.5 Megamanrem and approx. 15 Megamanrem, respectively. The collective committed effective dose equivalent from the consumption of contaminated foodstuffs is calculated to 23 Megamanrem in the BWR-3 case without dose-reducing measures. This dose could be reduced to 0.2 Megamanrem if contaminated crops are destroyed during the first year after the accident and if changes are made in agricultural production in the contaminated area. The corresponding doses in the BWR-2 case would be 197 Megamanrem and 1.4 Megmanrem, respectively. (author)
Comparison of the effect of hazard and response/fragility uncertainties on core melt probability uncertainty

International Nuclear Information System (INIS)

Mensing, R.W.

1985-01-01

This report proposes a method for comparing the effects of the uncertainty in probabilistic risk analysis (PRA) input parameters on the uncertainty in the predicted risks. The proposed method is applied to compare the effect of uncertainties in the descriptions of (1) the seismic hazard at a nuclear power plant site and (2) random variations in plant subsystem responses and component fragility on the uncertainty in the predicted probability of core melt. The PRA used is that developed by the Seismic Safety Margins Research Program
Rapid, dynamic segregation of core forming melts: Results from in-situ High Pressure- High Temperature X-ray Tomography

Science.gov (United States)

Watson, H. C.; Yu, T.; Wang, Y.

2011-12-01

The timing and mechanisms of core formation in the Earth, as well as in Earth-forming planetesimals is a problem of significant importance in our understanding of the early evolution of terrestrial planets . W-Hf isotopic signatures in meteorites indicate that core formation in small pre-differentiated planetesimals was relatively rapid, and occurred over the span of a few million years. This time scale is difficult to achieve by percolative flow of the metallic phase through a silicate matrix in textural equilibrium. It has been suggested that during this active time in the early solar system, dynamic processes such as impacts may have caused significant deformation in the differentiating planetesimals, which could lead to much higher permeability of the core forming melts. Here, we have measured the change in permeability of core forming melts in a silicate matrix due to deformation. Mixtures of San Carlos olivine and FeS close to the equilibrium percolation threshold (~5 vol%FeS) were pre-synthesized to achieve an equilibrium microstructure, and then loaded into the rotational Drickamer apparatus at GSE-CARS, sector 13-BMD, at the Advanced Photon Source (Argonne National Laboratory). The samples were subsequently pressed to ~2GPa, and heated to 1100°C. Alternating cycles of rotation to collect X-ray tomography images, and twisting to deform the sample were conducted until the sample had been twisted by 1080°. Qualitative and quantitative analyses were performed on the resulting 3-dimensional x-ray tomographic images to evaluate the effect of shear deformation on permeability and migration velocity. Lattice-Boltzmann simulations were conducted, and show a marked increase in the permeability with increasing deformation, which would allow for much more rapid core formation in planetesimals.
Prediction of thermoplastic failure of a reactor pressure vessel under a postulated core melt accident

International Nuclear Information System (INIS)

Duijvestijn, G.; Birchley, J.; Reichlin, K.

1997-01-01

This paper presents the lower head failure calculations performed for a postulated accident scenario in a commercial nuclear power plant. A postulated one inch break in the primary coolant circuit leads to dryout and subsequent meltdown of the core. The reference plant is a pressurized water reactor without penetrations in the reactor vessel lower head. The molten core material accumulates in the lower head, eventually causing failure of the vessel. The analysis investigates flow conditions in the melt pool, temperature evolution in the reactor vessel wall, and structure mechanical evaluation of the vessel under strong thermal loads and a range of internal pressures. The calculations were performed using the ADINA finite element codes. The analysis focusses on the failure processes, time and mode of failure. The most likely mode of failure at low pressure is global rupture due to gradual accumulation of creep strain over a large part of the heated area. In contrast, thermoplasticity becomes important at high pressure or following a pressure spike and can lead to earlier local failure. In situations in which part of the heat load is concentrated over a small area, resulting in a hot spot, local failure occurs, but not until the temperatures are close to the melting point. At low pressure, in particular, the hot spot area remains intact until the structure is molten across more than half of the thickness. (author) 14 figs., 16 refs
Penetration of a heated pool into a melting miscible substrate

International Nuclear Information System (INIS)

Eck, G.; Werle, H.

1986-01-01

Core-catchers have been proposed, which, after a core disruptive accident in a nuclear reactor, prevent containment failure caused by contact of the molten debris with the underlying ex-vessel structural materials. Most of these core-catchers are provided with sacrificial layers which on melting consume some fraction of the decay heat and dilute the heat sources and the fissionable material as the core masses are dissolved by the molten sacrificial material. Dilution of the core masses results in relatively low heat fluxes and temperatures at the wall of the core-catcher and, in addition, reduces the probability of recriticality. An experimental study was conducted on melting systems consisting of a liquid over-lying a solid substrate, which after melting of the solid, are mutually miscible. To initiate melting, the liquid was heated either by a planar heater from above or internally by an ac current. The density of the liquid was varied systematically, and it was found that downward heat transfer increases strongly with this parameter. In addition to heat transfer, mass transfer was studied by measuring the local concentration of the molten material in the liquid. A few experiments were performed in which sideward melting and two-dimensional pool growth were investigated
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.
Assessment of mass fraction and melting temperature for the application of limestone concrete and siliceous concrete to nuclear reactor basemat considering molten core-concrete interaction

Energy Technology Data Exchange (ETDEWEB)

Lee, Ho Jae; Kim, Do Gyeum [Korea Institute of Civil Engineering and Building Technology, Goyang (Korea, Republic of); Cho, Jae Leon [Korea Hydro and Nuclear Power Co., Ulsan (Korea, Republic of); Yoon, Eui Sik [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Cho, Myung Suk [Korea Hydro and Nuclear Power Co., Central Research Institute, Daejeon (Korea, Republic of)

2016-04-15

Severe accident scenarios in nuclear reactors, such as nuclear meltdown, reveal that an extremely hot molten core may fall into the nuclear reactor cavity and seriously affect the safety of the nuclear containment vessel due to the chain reaction caused by the reaction between the molten core and concrete. This paper reports on research focused on the type and amount of vapor produced during the reaction between a high-temperature molten core and concrete, as well as on the erosion rate of concrete and the heat transfer characteristics at its vicinity. This study identifies the mass fraction and melting temperature as the most influential properties of concrete necessary for a safety analysis conducted in relation to the thermal interaction between the molten core and the basemat concrete. The types of concrete that are actually used in nuclear reactor cavities were investigated. The H2O content in concrete required for the computation of the relative amount of gases generated by the chemical reaction of the vapor, the quantity of CO2 necessary for computing the cooling speed of the molten core, and the melting temperature of concrete are evaluated experimentally for the molten core-concrete interaction analysis.
Modelling of RPV lower head under core melt severe accident condition using OpenFOAM

International Nuclear Information System (INIS)

Madokoro, Hiroshi; Kretzschmar, Frank; Miassoedov, Alexei

2017-01-01

Although six years have been passed since the tragic severe accident at Fukushima Daiichi, still large uncertainties exist in modeling of core degradation and reactor pressure vessel (RPV) failure. It is extremely important to obtain a better understanding of complex phenomena in the lower head in order to improve accident management measures. The possible failure mode of reactor pressure vessel and its failure time are especially a matter of importance. Thermal behavior of the molten pool can be simulated by the Phase-change Effective Convectivity Model (PECM), which is a distributed-parameter model developed in the Royal Institute of Technology (KTH), Sweden. The model calculates convective currents not using a pure CFD approach but based on so called “characteristic velocities” that are determined by empirical correlations depending on the geometry and physical properties of the molten pool. At the Karlsruhe Institute of Technology (KIT), the PECM has been implemented in the open-source CFD software OpenFOAM in order to receive detailed predictions of a core melt behavior in the RPV lower head under severe accident conditions. An advantage of using OpenFOAM is that it is very flexible to add and modify models and physical properties. In the current work, the solver is extended to couple PECM with a structure analysis model of the vessel wall. The model considers thermal expansion, plasticity, creep and damage. The model and physical properties are based on those implemented in ANSYS. Although the previous implementation had restriction that the amount of and geometry of the melt cannot be changed, our coupled model allows flexibility of the melt amount and geometry. The extended solver was used to simulate the LIVE-L1 and -L7V experiments and has demonstrated good prediction of the temperature distribution in the molten pool and heat flux distribution through the vessel wall. Regarding the vessel failure the model was applied to one of the FOREVER tests
Correlation for downward melt penetration into a miscible low-density substrate

International Nuclear Information System (INIS)

Fang, L.J.; Cheung, F.B.; Pedersen, D.R.; Linehan, J.H.

1984-01-01

Downward penetration of a sacrificial bed material or a concrete basemat structure by an overlying layer of core melt resulting from a hypothetical core disruptive accident has been a major issue in post accident heat removal studies. One characteristic feature of this problem is that the solid substrate, when molten, is miscible with and lighter than the core melt so that the rate of penetration is strongly dependent upon the motion of natural convection in the melt layer driven by the density difference between the core melt and the molten substrate. This fundamentally interesting and technologically important problem has been investigated by a number of researchers. Significantly different melting rates, however, were observed in these studies. Questions concerning the occurrence of flow transition and its effect on melt penetration remain to be answered. To promote the understanding of the phenomena and to strengthen the data base of melt penetration, simulation experiments were conducted using various kinds of salt solutions (KI, NaCl, CaCl 2 , and MgCl 2 solutions) as the working fluid and an air-bubble-free ice slab as the solid substrate

Generalized Thermohydraulics Module GENFLO for Combining With the PWR Core Melting Model, BWR Recriticality Neutronics Model and Fuel Performance Model

International Nuclear Information System (INIS)

Miettinen, Jaakko; Hamalainen, Anitta; Pekkarinen, Esko

2002-01-01

Thermal hydraulic simulation capability for accident conditions is needed at present in VTT in several programs. Traditional thermal hydraulic models are too heavy for simulation in the analysis tasks, where the main emphasis is the rapid neutron dynamics or the core melting. The GENFLO thermal hydraulic model has been developed at VTT for special applications in the combined codes. The basic field equations in GENFLO are for the phase mass, the mixture momentum and phase energy conservation equations. The phase separation is solved with the drift flux model. The basic variables to be solved are the pressure, void fraction, mixture velocity, gas enthalpy, liquid enthalpy, and concentration of non-condensable gas fractions. The validation of the thermohydraulic solution alone includes large break LOCA reflooding experiments and in specific for the severe accident conditions QUENCH tests. In the recriticality analysis the core neutronics is simulated with a two-dimensional transient neutronics code TWODIM. The recriticality with one rapid prompt peak is expected during a severe accident scenario, where the control rods have been melted and ECCS reflooding is started after the depressurization. The GENFLO module simulates the BWR thermohydraulics in this application. The core melting module has been developed for the real time operator training by using the APROS engineering simulators. The core heatup, oxidation, metal and fuel pellet relocation and corium pool formation into the lower plenum are calculated. In this application the GENFLO model simulates the PWR vessel thermohydraulics. In the fuel performance analysis the fuel rod transient behavior is simulated with the FRAPTRAN code. GENFLO simulates the subchannel around a single fuel rod and delivers the heat transfer on the cladding surface for the FRAPTRAN. The transient boundary conditions for the subchannel are transmitted from the system code for operational transient, loss of coolant accidents and
XPS and EPXMA investigation and chemical speciation of aerosol samples formed in LWR core melting experiments

International Nuclear Information System (INIS)

Moers, H.; Jenett, H.; Kaufmann, R.; Klewe-Nebenius, H.; Pfennig, G.; Ache, H.J.

1985-09-01

Aerosol samples consisting of fission products and elements of light water reactor structural materials were collected during simulating in a laboratory scale the heat-up phase of a core melt accident. The aerosol particles were formed in a steam atmosphere at temperatures between 1200 and 1900 0 C of the melting charge. The investigation of the samples by use of X-ray photoelectron spectroscopy (XPS) permitted the chemical speciation of the detected aerosol constituents silver, cadmium, indium, tellurium, iodine, and cesium. A comparison of the elemental analysis results obtained from XPS with those achieved from electron probe X-ray micro analysis (EPXMA) revealed that aerosol particle surface and aerosol particle bulk are principally composed of the same elements and that these compositions vary with release temperature. In addition, quantitative differences between the composition of surface and bulk have only been observed for those aerosol samples which were collected at higher melting charge temperatures. In order to obtain direct information on chemical species below the surface selected samples were argon ion bombarded. Changes in composition and chemistry were monitored by XPS, and the results were interpreted in light of the effects, which were observed when appropriate standard samples were sputtered. (orig.) [de
Simulation of melt spreading in consideration of phase transitions

Energy Technology Data Exchange (ETDEWEB)

Spengler, C. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Koeln (Germany)

2002-07-01

The analysis of melt spreading and relocation phenomena in the containment of LWR power plants in case of hypothetical severe accidents leading to core melting is an important issue for reactor safety investigations. For the simulation of melt spreading the code LAVA has been developed on the basis of a method from the related subject of volcanology by adding more detailed models for heat transfer phenomena and flow rheology. The development is supported by basic analysis of the spreading of gravity currents as well as experimental investigations of the rheology of solidifying melts. These exhibit strong non-Newtonian effects in case of a high content of solids in the freezing melt. The basic model assumption in LAVA is the ideal Bingham plastic approach to the non-Newtonian, shear-thinning characteristic of solidifying melts. For the recalculation of melt spreading experiments, the temperature-dependent material properties for solidifying melt mixtures have been calculated using correlations from the literature. With the parameters and correlations for the rheological material properties approached by results from literature, it was possible to recalculate successfully recent spreading experiments with simulant materials and prototypic reactor core materials. An application to the behaviour of core melt in the reactor cavity assumed a borderline case for the issue of spreading. This limit is represented by melt conditions (large solid fraction, low volume flux), under which the melt is hardly spreadable. Due to the persistent volume flux the reactor cavity is completely, but inhomogeneously filled with melt. The degree of inhomogeneity is rather small, so it is concluded, that for the long-term coolability of a melt pool in narrow cavities the spreading of melt will probably have only negligible influence. (orig.)
Retention Management Of Critical (Core) Employees A Challenging Issue Confronting Organisations In The 21st Century

OpenAIRE

Janet Chew; Lanny Entrekin

2011-01-01

Employee retention is one of the challenges facing many business organisations today. Many industries are afflicted with high demand for specialised employees and are also suffering high levels of turnover. We have moved into a knowledge-based society where human capital is considered a key resource and a competitive business advantage. The high attrition rate of critical (core) employees is costly to corporations. Loss of these high talent employees results in the stripping of valuable human...
Review of the Technical Status on the Debris Bed Cooling Model

Energy Technology Data Exchange (ETDEWEB)

Kim, Eui Kwang; Cho, Chung Ho; Lee, Yong Bum

2007-09-15

Preliminary safety analyses of the KALIMER-600 design have shown that the design has inherent safety characteristics and is capable of accommodating double-fault initiators such as ATWS events without coolant boiling or fuel melting. However, for the future design of sodium cooled fast reactor, the evaluation of the safety performance and the determination of containment requirements may be worth due consideration of triple-fault accident sequences of extremely low probability of occurrence that leads to core melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will be required as a design requirement for the future design of sodium cooled fast reactor. Also, proof of the capacity of the debris bed cooling is an essential condition to solve the problem of in-vessel retention of the core debris. In this study, review of the technical status on the debris bed cooling model was carried out for in-vessel retention of the core debris0.
Review of the Technical Status on the Debris Bed Cooling Model

International Nuclear Information System (INIS)

Kim, Eui Kwang; Cho, Chung Ho; Lee, Yong Bum

2007-09-01

Preliminary safety analyses of the KALIMER-600 design have shown that the design has inherent safety characteristics and is capable of accommodating double-fault initiators such as ATWS events without coolant boiling or fuel melting. However, for the future design of sodium cooled fast reactor, the evaluation of the safety performance and the determination of containment requirements may be worth due consideration of triple-fault accident sequences of extremely low probability of occurrence that leads to core melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will be required as a design requirement for the future design of sodium cooled fast reactor. Also, proof of the capacity of the debris bed cooling is an essential condition to solve the problem of in-vessel retention of the core debris. In this study, review of the technical status on the debris bed cooling model was carried out for in-vessel retention of the core debris
Visualization Study of Melt Dispersion Behavior for SFR with a Metallic Fuel under Severe Accidents

Energy Technology Data Exchange (ETDEWEB)

Heo, Hyo Heo; Park, Seong Dae; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Jerng, Dong Wook [Jungang Univ., Seoul (Korea, Republic of)

2015-05-15

The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition.
LFR core design for prevention & mitigation of severe accidents

International Nuclear Information System (INIS)

Grasso, Giacomo

2012-01-01

Conclusions: • Aiming at fully complying Gen-IV safety requirements – even in case of Fukushima-like events –, prevention and mitigation strategies must be stressed in FR design. • The safety of Lead-cooled Fast Reactors can rely on intrinsic features due to the coolant, such as: • the practical impossibility of Lead boiling, hence the unreliability of core (only) voiding for wide safety margins, and the retention of corium; • the high density of lead, for the buoyancy of Control Rods (allowing their safe positioning below the core), and the dispersion of molten core up to the setting up of a “cold melting pot”. • the possibility to adopt wide coolant channels for encouraging natural circulation, without affecting the hardness of the neutron spectrum; • the hard neutron spectrum allows the adiabatic operation of LFRs (which implies minimal criticality swings even through long cycles) with small amounts of Mas (hence with a negligible detriment to the safety features); • an effective reduction of the coolant density effect simply through the shortening of the active height
Nuclear Power Reactor Core Melt Accidents. Current State of Knowledge

International Nuclear Information System (INIS)

Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

2013-01-01

For over thirty years, IPSN and subsequently IRSN has played a major international role in the field of nuclear power reactor core melt accidents through the undertaking of important experimental programmes (the most significant being the Phebus- FP programme), the development of validated simulation tools (the ASTEC code that is today the leading European tool for modelling severe accidents), and the coordination of the SARNET (Severe Accident Research Network) international network of excellence. These accidents are described as 'severe accidents' because they can lead to radioactive releases outside the plant concerned, with serious consequences for the general public and for the environment. This book compiles the sum of the knowledge acquired on this subject and summarises the lessons that have been learnt from severe accidents around the world for the prevention and reduction of the consequences of such accidents, without addressing those from the Fukushima accident, where knowledge of events is still evolving. The knowledge accumulated by the Institute on these subjects enabled it to play an active role in informing public authorities, the media and the public when this accident occurred, and continues to do so to this day
Core-melting accidents in Chernobyl and Harrisburg

International Nuclear Information System (INIS)

Loon, A.J. van; Vonderen, A.C.M. van

1987-01-01

This publication deals with the essences of the reactor accident in Chernobylsk and the conclusions to be drawn from these with regard to reactor safety. Therein the technical differences between the reactor types in the West and the East play an important role. Also attention is spent to the now generally accepted philosophy that by simplification and making use of proven technologies, a further deminishing of the risks can be achieved step by step. In ch.'s 2 and 4 the origin and course of the accidents in respectively Chernobylsk and Harrisburg are analyzed; in the analysis of the Chernobylsk accident also date have been used which were provided by the Sovjet-Union, supplied with results of studies of the U.S. Department of Energy (DOE). In ch. 3 this information is compared with the insights which have grown at KEMA about these on the base of reactor physical and thermohydraulic considerations and of computer calculations reproducing the course of the accident. An important question is if, and if so: to which extent, an accident such as the one in Chernobylsk also can take place in the West. In order to answer that question as accurate as possible the consequences of core meltings accidents and the risk for such an accident taking place are pursued. In ch. 6 the legal frameworks are indicated by which the risk may be limited and by which eventually yet occurring damage may be arranged. Ch. 7 finally deals with the lessons which the accidents in Chernobylsk and Harrisburg have learnt us and with the possible consequences of these for the further application of nuclear power in the Netherlands. (H.W.). 105 refs.; 42 figs.; 17 refs
Core catcher concepts future PWR-Plants

International Nuclear Information System (INIS)

Alsmeyer, H.; Werle, H.

1994-01-01

Light water reactors of the next generation should have still greater passive safety, even in the most serious accidents. This includes the long term safe inclusion of the core inventory in the case of core meltdown accidents. The three concepts for cooling the liquefied core outside the reactor pressure vessel examined by KfK should remove the post-shutdown heat by direct contact of the melt with water. The geometric distribution of the melt increases its surface area, so that favourable conditions for heat removal from the poorly thermally-conducting melt are created and complete quick solidification occurs. The experiments examine both the relocation and distribution mechanisms of the melt and the reactions occurring when water enters. As strong interaction is possible on direct contact of the melt with water, an important aim is experimental determination and limitation of any resulting mechanical stresses. (orig./HP) [de
Final results of the XR2-1 BWR metallic melt relocation experiment

International Nuclear Information System (INIS)

Gauntt, R.O.; Humphries, L.L.

1997-08-01

This report documents the final results of the XR2-1 boiling water reactor (BWR) metallic melt relocation experiment, conducted at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. The objective of this experiment was to investigate the material relocation processes and relocation pathways in a dry BWR core following a severe nuclear reactor accident such as an unrecovered station blackout accident. The imposed test conditions (initial thermal state and the melt generation rates) simulated the conditions for the postulated accident scenario and the prototypic design of the lower core test section (in composition and in geometry) ensured that thermal masses and physical flow barriers were modeled adequately. The experiment has shown that, under dry core conditions, the metallic core materials that melt and drain from the upper core regions can drain from the core region entirely without formation of robust coherent blockages in the lower core. Temporary blockages that suspended pools of molten metal later melted, allowing the metals to continue draining downward. The test facility and instrumentation are described in detail. The test progression and results are presented and compared to MERIS code analyses. 6 refs., 55 figs., 4 tabs
Melt migration modeling in partially molten upper mantle

Science.gov (United States)

Ghods, Abdolreza

The objective of this thesis is to investigate the importance of melt migration in shaping major characteristics of geological features associated with the partial melting of the upper mantle, such as sea-floor spreading, continental flood basalts and rifting. The partial melting produces permeable partially molten rocks and a buoyant low viscosity melt. Melt migrates through the partially molten rocks, and transfers mass and heat. Due to its much faster velocity and appreciable buoyancy, melt migration has the potential to modify dynamics of the upwelling partially molten plumes. I develop a 2-D, two-phase flow model and apply it to investigate effects of melt migration on the dynamics and melt generation of upwelling mantle plumes and focusing of melt migration beneath mid-ocean ridges. Melt migration changes distribution of the melt-retention buoyancy force and therefore affects the dynamics of the upwelling plume. This is investigated by modeling a plume with a constant initial melt of 10% where no further melting is considered. Melt migration polarizes melt-retention buoyancy force into high and low melt fraction regions at the top and bottom portions of the plume and therefore results in formation of a more slender and faster upwelling plume. Allowing the plume to melt as it ascends through the upper mantle also produces a slender and faster plume. It is shown that melt produced by decompressional melting of the plume migrates to the upper horizons of the plume, increases the upwelling velocity and thus, the volume of melt generated by the plume. Melt migration produces a plume which lacks the mushroom shape observed for the plume models without melt migration. Melt migration forms a high melt fraction layer beneath the sloping base of the impermeable oceanic lithosphere. Using realistic conditions of melting, freezing and melt extraction, I examine whether the high melt fraction layer is able to focus melt from a wide partial melting zone to a narrow region
Melt Fragmentation Characteristics of Metal Fuel with Melt Injection Mass during Initiating Phase of SFR Severe Accidents

Energy Technology Data Exchange (ETDEWEB)

Heo, Hyo; Lee, Min Ho; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Jerng, Dong Wook [Chung-Ang Univ., Seoul (Korea, Republic of)

2016-05-15

The PGSFR has adopted the metal fuel for its inherent safety under severe accident conditions. However, this fuel type is not demonstrated clearly yet under the such severe accident conditions. Additional experiments for examining these issues should be performed to support its licensing activities. Under initiating phase of hypothetic core disruptive accident (HCDA) conditions, the molten metal could be better dispersed and fragmented into the coolant channel than in the case of using oxide fuel. This safety strategy provides negative reactivity driven by a good dispersion of melt. If the coolant channel does not sufficient coolability, the severe recriticality would occur within the core region. Thus, it is important to examine the extent of melt fragmentation. The fragmentation behaviors of melt are closely related to a formation of debris shape. Once the debris shape is formed through the fragmentation process, its coolability is determined by the porosity or thermal conductivity of the melt. There were very limited studies for transient irradiation experiments of the metal fuel. These studies were performed by Transient Reactor Test Facility (TREAT) M series tests in U.S. The TREAT M series tests provided basic information of metal fuel performance under transient conditions. The effect of melt injection mass was evaluated in terms of the fragmentation behaviors of melt. These behaviors seemed to be similar between single-pin and multi-pins failure condition. However, the more melt was agglomerated in case of multi-pins failure.
Methods to prevent the source term of methyl lodide during a core melt accident

Energy Technology Data Exchange (ETDEWEB)

Karhu, A. [VTT Energy (Finland)

1999-11-01

The purpose of this literature review is to gather available information of the methods to prevent a source term of methyl iodide during a core melt accident. The most widely studied methods for nuclear power plants include the impregnated carbon filters and alkaline additives and sprays. It is indicated that some deficiencies of these methods may emerge. More reactive impregnants and additives could make a great improvement. As a new method in the field of nuclear applications, the potential of transition metals to decompose methyl iodide, is introduced in this review. This area would require an additional research, which could elucidate the remaining questions of the reactions. The ionization of the gaseous methyl iodide by corona-discharge reactors is also shortly described. (au)
Numerical analysis of the induction melting process of oxide fuel material

International Nuclear Information System (INIS)

Kondala Rao, R.; Mangarjuna Rao, P.; Nashine, B.K.; Selvaraj, P.

2015-01-01

For the experimental simulation of Molten Fuel-Coolant Interaction (MFCI) phenomenon under hypothetical core meltdown accident scenario in a nuclear reactor, it is required to generate the molten pool of core materials. For this purpose, a laboratory scale Cold wall Crucible induction melting system has been developed. To optimize the system for efficient and reliable melting process, it is required to have comprehensive knowledge on the heat and mass transfer processes along with electromagnetic process that occur during the melting of core materials. Hence, a 2D axi-symmetric numerical model has been developed using a multiphysics software to simulate the induction melting process. The phase change phenomenon is taken into account by using enthalpy formulation. The experimental data available in literature for magnetic field and flow field are used for model validation. The model predicted temperatures are also in good agreement with experimentally measured values. The validated model has been used to study the induction melting behavior of UO_2 fuel material. (author)
Effect of controlling recrystallization from the melt on the residual stress and structural properties of the Silica-clad Ge core fiber

Science.gov (United States)

Zhao, Ziwen; Cheng, Xueli; He, Ting; Xue, Fei; Zhang, Wei; Chen, Na; Wen, Jianxiang; Zeng, Xianglong; Wang, Tingyun

2017-09-01

Effect of controlling recrystallization from the melt (1000 °C) on the residual stress and structural properties of a Ge core fiber via molten core drawing (MCD) method is investigated. Ge core fibers is investigated using Raman spectroscopy, scanning electron microscope (SEM), and X-ray diffraction (XRD). Compared with the as-drawn Ge fiber, the Raman peak of the recrystallized Ge fiber shift from 300 cm-1 to 300.6 cm-1 and full width at half maximum (FWHM) decreased from 5.36 cm-1 to 4.48 cm-1. The Ge crystal grains which sizes are of 200-600 nm were formed during the process of recrystallization; the XRD peak of (1 1 1) plane is observed after recrystallization. These results show that controlling recrystallization allows the release of the thermal stress, and improvement of the crystal quality of Ge core.
In-calandria retention of corium in Indian PHWR - experimental simulations with decay heat

International Nuclear Information System (INIS)

Nayak, A.K.

2015-01-01

The severe accident at Fukushima has compelled the nuclear community to relook at the safety of existing nuclear power plants (NPP) against natural origin events of beyond design basis and prolonged station black out (SBO). A major lesson learned is to assess the capability of the safety systems to cool the reactor core and spent fuel storage facilities in the event of a prolonged station black out (SBO). Similar safety review is planned for the Indian Pressurized Heavy Water Reactors (PHWRs) considering a prolonged SBO. The Indian PHWR is a heavy water-moderated and cooled, natural uranium-fuelled reactor in which the horizontal fuel channels are submerged in a pool of heavy water moderator located inside the calandria vessel. The calandria vessel is surrounded by a calandria vault having large volume of light water. Concerns are raised that in the event of an unmitigated SBO, it may result into a low probable severe accident leading to core melt down. The core melt may further fail the calandria vessel in case the melt is not quenched. If the calandria vessel fails, the corium shall interact with the cold calandria vault water and concrete resulting in generation of large amount of non-condensable gases and steam which will lead to over pressurization of containment and may cause its failure. Therefore, in-calandria corium retention via external cooling using vault water can be considered as an important accident management program in PHWR. In this strategy, the core melt retains inside the calandria vessel by continually removing the stored heat and decay heat through outer surface of the vessel by cooling water and maintaining the integrity of the vessel. 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 by using the calandria vault water. Molten borosilicate glass was used as the simulant due to its comparable heat transfer characteristics
An assessment of Class-9 (core-melt) accidents for PWR dry-containment systems

International Nuclear Information System (INIS)

Theofanous, T.G.; Saito, M.

1981-01-01

The phenomenology of core-melt accidents in dry containments was examined for the purpose of identifying the margins of safety in such Class-9 situations. The scale (geometry) effects appear to crucially limit the extent (severity) of steam explosions. This together with the established reduced explosivity of the corium-A/water system, and the inherently high capability of dry containments (redinforced concrete, and shields in some cases, seismic design etc.) lead to the conclusion that failure due to steam explosions may be considered essentially incredible. These premixture scaling considerations also impact ultimate debris disposition and coolability and need additional development. A water-flooded reactor cavity would have beneficial effects in limiting (but not necessarily eliminating) melt-concrete interactions. Independently of the initial degree of quenching and/or scale of fragmentation, mechanisms exist that drive the system towards ultimate stability (coolability). Additional studies, with intermediate-scale prototypic materials are recommended to better explore these mechanisms. Containment heat removal systems must provide the crucial capability of mitigating such accidents. Passive systems should be explored and assessed against currently available and/or improved active systems taking into account the rather loose time constraints required for activation. It appears that containment margins for accommodating the hydrogen problem are limited. This problem appears to stand out not only in terms of potential consequences but also in terms of lack of any readily available and clear cut solutions at this time. (orig.)
Investigation of the Potential for In-Vessel Melt Retention in the Lower Head of a BWR by Cooling through the Control Rod Guide Tubes. APRl 4, Stage 2 Report

International Nuclear Information System (INIS)

Sehgal, B.R.; Jasiulevicius, A.; Konovalikhin, M.

2004-01-01

This report describes the experiments performed at the Division, investigating the coolability potential offered by the Control Rod Guide Tubes (CRGTs), which are present in large numbers in the lower head of a BWR and there is a water flow circuit in each one of them. This investigation is related to the overall goal of retaining the core melt in the lower head of a BWR during a postulated severe accident, through accident management procedures, or strategy. The experiments were performed in two facilities, i.e. POMECO (Porous MEdium COolability) and COMECO (Core MElt COolability), respectively, for investigating the coolability when the core material is in the form of a particulate debris bed and when it is in the form of a melt. The POMECO facility employed a sand bed heated electrically to heating levels of up to 1 MW/m 3 and experiments performed in that facility obtained the enhancement in the dryout heat flux and in the quench velocity due to presence of a CRGT, with, and without, water flow in it. The COMECO facility employed a simulant material melt pool heated electrically to power levels of = 1.3 MW/m 3 and the experiments in it also determined the enhancement in the heat removal from the melt pool that could be obtained by the presence of a CRGT, with, or without water flow in it. In each of the experiments in these facilities, the scaling employed was of a unit cell of core material around a prototypic geometry CRGT with the prototypic decay heat input. The experimental results showed that a CRGT is able to offer a substantial additional potential for coolability of particulate and melt material in the lower head of a BWR. Analysis of the data obtained in the set of experiments performed lead to the following results for the heat flux through the CRGT: - for a water filled particulate debris bed: - 40 kW/m 2 ; - for a day hot particulate debris bed: - 150 kW/m 2 ; - for a melt pool with a crust formed on the CRGT surface: -350 kW/m 2 . It is

Earth's core formation due to the Rayleigh-Taylor instability

International Nuclear Information System (INIS)

Ida, S.; Nakagawa, Y.; Nakazawa, K.

1987-01-01

A protoearth accretion stage configuration consisting of an undifferentiated solid core, an intermediate metal-melt layer, and an outer silicate-melt layer, is presently taken as the initial state in an investigation of Rayleigh-Taylor instability-induced core formation. The Ida et al. (to be published) quantitative results on the instability in a self-gravitating fluid sphere are used. The instability is found to occur through the translational mode on a time-scale of about 10 hr, in the case where the metal-melt layer is greater than about 1 km; this implies that the earth's core formed due to the undifferentiated solid core's translation upon the outer layer's melting. Differentiation would then have occurred in the late accretion stage. 17 references
Experimental studies on melt spreading, bubbling heat transfer, and coolant layer boiling

International Nuclear Information System (INIS)

Greene, G.A.; Finfrock, C.; Klages, J.; Schwarz, C.E.; Burson, S.B.

1988-01-01

Melt spreading studies have been undertaken to investigate the extent to which molten core debris may be expected to spread under gravity forces in a BWR drywell geometry. The objectives are to determine the extent of melt spreading as a function of melt mass,melt superheat, and water depth. These studies will enable an objective determination of whether or not core debris can spread up to and contact containment structures or boundaries upon vessel failure. Results indicate that the most important variables are the melt superheat and the water depth. Studies have revealed five distinct regimes of melt spreading ranging from hydrodynamically-limited to heat transfer-limited. A single parameter dimensionless correlation is presented which identified the spreading regime and allows for mechanistic calculation of the average thickness to which the melt will spread. 7 refs., 12 figs
Thermal-hydraulic studies on molten core-concrete interactions

International Nuclear Information System (INIS)

Greene, G.A.

1986-10-01

This report discusses studies carried out in connection with light water power reactor accidents. Recent assessments have indicated that the consequences of molten-core concrete interactions dominate the considerations of severe accidents. The two areas of interest that have been investigated are interlayer heat and mass transfer and liquid-liquid boiling. Interlayer heat and mass transfer refers to processes that occur within a core melt between the stratified, immiscible phases of core oxides and metals. Liquid-liquid boiling refers to processes that occur at the melt-concrete on melt-coolant interface
Sodalite as a vehicle to increase Re retention in waste glass simulant during vitrification

Energy Technology Data Exchange (ETDEWEB)

Luksic, Steven A., E-mail: steven.luksic@pnnl.gov; Riley, Brian J.; Parker, Kent E.; Hrma, Pavel

2016-10-15

Technetium (Tc) retention during Hanford waste vitrification can be increased if the volatility can be controlled. Incorporating Tc into a thermally stable mineral phase, such as sodalite, is one way to achieve increased retention. Here, rhenium (Re)-bearing sodalite was tested as a vehicle to transport perrhenate (ReO{sub 4}{sup −}), a nonradioactive surrogate for pertechnetate (TcO{sub 4}{sup −}), into high-level (HLW) and low-activity waste (LAW) glass simulants. After melting HLW and LAW simulant feeds, the retention of Re in the glass was measured and compared with the Re retention in glass prepared from a feed containing Re{sub 2}O{sub 7}. Phase analysis of sodalite in both these glasses across a profile of temperatures describes the durability of Re-sodalite during the feed-to-glass transition. The use of Re sodalite improved the Re retention by 21% for HLW glass and 85% for LAW glass, demonstrating the potential improvement in Tc-retention if TcO{sub 4}{sup −} were to be encapsulated in a Tc-sodalite prior to vitrification. - Highlights: • Re retention is improved by incorporation into sodalite structure. • LAW-type glass shows lower retention but larger improvement with Re-sodalite. • Sodalite is stable to higher temperatures in high-alumina glass melts.
Effect of Technetium-99 sources on its retention in low activity waste glass

Science.gov (United States)

Luksic, Steven A.; Kim, Dong-Sang; Um, Wooyong; Wang, Guohui; Schweiger, Michael J.; Soderquist, Chuck Z.; Lukens, Wayne; Kruger, Albert A.

2018-05-01

Small-scale crucible melting tests on simulated waste glass were performed with technetium-99 (Tc-99) introduced as different species in a representative low activity waste simulant. The glass saw an increase in Tc-99 retention when TcO2•2H2O and various Tc-minerals containing reduced tetravalent Tc were used compared to tests in which pertechnetate with heptavalent Tc was used. We postulate that the increase of Tc retention is likely caused by different reaction paths for Tc incorporation into glass during early stages of melting, rather than the low volatility of reduced tetravalent Tc compounds, which has been a generally accepted idea. Additional studies are needed to clarify the exact mechanisms relevant to the effect of reduced Tc compounds on Tc incorporation into or volatilization from the glass melt.
Effect of Technetium-99 Sources on Its Retention in Low Activity Waste Glass

Energy Technology Data Exchange (ETDEWEB)

Luksic, Steven A.; Kim, Dong-Sang; Um, Wooyong; Wang, Guohui; Schweiger, Michael J.; Soderquist, Chuck Z.; Lukens, Wayne W.; Kruger, Albert A.

2018-05-01

Small-scale crucible melting tests on simulated waste glass were performed with technetium-99 (Tc-99) introduced as different species in a representative low activity waste simulant. The glass saw an increase in Tc-99 retention when TcO2∙2H2O and various Tc-minerals containing reduced tetravalent Tc were used compared to tests in which pertechnetate with hexavalent Tc was used. We postulate that the increase of Tc retention is likely caused by different reaction paths for Tc incorporation into glass during early stages of melting, rather than the low volatility of reduced tetravalent Tc compounds, which has been a generally accepted idea. Additional studies are needed to clarify the exact mechanisms relevant to the effect of reduced Tc compounds on Tc incorporation into or volatilization from glass melt.
Eutectic melting temperature of the lowermost Earth's mantle

Science.gov (United States)

Andrault, D.; Lo Nigro, G.; Bolfan-Casanova, N.; Bouhifd, M.; Garbarino, G.; Mezouar, M.

2009-12-01

Partial melting of the Earth's deep mantle probably occurred at different stages of its formation as a consequence of meteoritic impacts and seismology suggests that it even continues today at the core-mantle boundary. Melts are important because they dominate the chemical evolution of the different Earth's reservoirs and more generally the dynamics of the whole planet. Unfortunately, the most critical parameter, that is the temperature profile inside the deep Earth, remains poorly constrained accross the planet history. Experimental investigations of the melting properties of materials representative of the deep Earth at relevant P-T conditions can provide anchor points to refine past and present temperature profiles and consequently determine the degree of melting at the different geological periods. Previous works report melting relations in the uppermost lower mantle region, using the multi-anvil press [1,2]. On the other hand, the pyrolite solidus was determined up to 65 GPa using optical observations in the laser-heated diamond anvil cell (LH-DAC) [3]. Finally, the melting temperature of (Mg,Fe)2SiO4 olivine is documented at core-mantle boundary (CMB) conditions by shock wave experiments [4]. Solely based on these reports, experimental data remain too sparse to draw a definite melting curve for the lower mantle in the relevant 25-135 GPa pressure range. We reinvestigated melting properties of lower mantle materials by means of in-situ angle dispersive X-ray diffraction measurements in the LH-DAC at the ESRF [5]. Experiments were performed in an extended P-T range for two starting materials: forsterite and a glass with chondrite composition. In both cases, the aim was to determine the onset of melting, and thus the eutectic melting temperatures as a function of pressure. Melting was evidenced from drastic changes of diffraction peak shape on the image plate, major changes in diffraction intensities in the integrated pattern, disappearance of diffraction rings
KATS experiments to simulate corium spreading in the EPR core catcher concept

International Nuclear Information System (INIS)

Eppinger, B.; Fieg, G.; Schuetz, W.; Stegmaier, U.

2001-01-01

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)
In-vessel coolability and retention of a core melt

International Nuclear Information System (INIS)

Theofanous, T.G.; Liu, C.; Additon, S.

1997-01-01

The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. The technical treatment in this assessment includes: (a) new data on energy flow from either volumetrically heated pools or non-heated layers on top, boiling and critical heat flux in inverted, curved geometries, emissivity of molten (superheated) samples of steel, and chemical reactivity proof tests, (b) a simple but accurate mathematical formulation that allows prediction of thermal loads by means of convenient hand calculations, (c) a detailed model programmed on the computer to sample input parameters over the uncertainty ranges, and to produce probability distributions of thermal loads and margins for departure from nucleate boiling at each angular position on the lower head, and (d) detailed structural evaluations that demonstrate that departure from nucleate boiling is a necessary and sufficient criterion for failure. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is open-quotes physically unreasonable.close quotes Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings
The melting curve of iron to 250 gigapascals - A constraint on the temperature at earth's center

Science.gov (United States)

Williams, Quentin; Jeanloz, Raymond; Bass, Jay; Svendsen, Bob; Ahrens, Thomas J.

1987-01-01

The melting curve of iron, the primary constituent of earth's core, has been measured to pressures of 250 gigapascals with a combination of static and dynamic techniques. The melting temperature of iron at the pressure of the core-mantle boundary (136 GPa) is 4800 + or - 200 K, whereas at the inner core-outer core boundary (330 GPa), it is 7600 + or - 500 K. A melting temperature for iron-rich alloy of 6600 K at the inner core-outer core boundary and a maximum temperature of 6900 K at earth's center are inferred. This latter value is the first experimental upper bound on the temperature at earth's center, and these results imply that the temperature of the lower mantle is significantly less than that of the outer core.
Laboratory studies of the meltfront propagation in a borax core-catcher

International Nuclear Information System (INIS)

Dalle Donne, M.; Werle, H.

1980-08-01

A series of seven laboratory experiments concerning the meltdown of a borax core catcher have been performed. By the selection of the simulant materials the most important thermophysical properties of the core catcher materials were taken into account. Fission product heating of the molten core masses was simulated by electrolytically heating of the molten region. The experiments reveal interesting details of the phenomena to be expected during melt-down of a borax core catcher, especially on the flow pattern, the mixing processes of molten materials and the layer formation the melt. The most interesting result is that the ratio of downward to sideward melting rate is heavily reduced by high melting barriers and that a cubic structure of barriers will not equalize downward and sideward melting rates. A super 8 film is available as additional information. (orig.) [de
Load histories from steam explosions during core melt accidents

International Nuclear Information System (INIS)

Jacobs, H.; Kolev, N.I.

1992-01-01

For the analysis of steam explosions a multicomponent multiphase thermohydraulic code is required which describes at least the motions of melt, water, and steam by separate velocity fields. One example of these very rare codes is the IVA3 code the development of which was brought to an interim close in 1991. As an example of a typical application of this code, precalculations of the FARO LWR Scoping Test 2 performed at Ispra are discussed. Unfortunately, the calculation results cannot be compared directly to the test results because of important differences between planned and achieved test parameters. Above all, only about one third of the planned melt mass actually entered the water. The test was performed in a closed vessel at an initial pressure of 50 bar. The water was saturated at this temperature and its level was at 1 m height. The simulation starts with the release of 50 kg of simulated corium from an intermediate catcher at about 3.2 m height. The calculation predicts a gradual pressure rise without fast transients worth mentioning from 50 to about 76 bar within roughly one second and stabilizes slightly below the maximum. Also described are the material distributions predicted during the process and the 'mixed' masses according to two different criteria. The former indicate that the melt jet penetrates the water without desintegrating while being surrounded by a thick vapor layer. Subsequently the melt collects at the level bottom and much of the liquid water is blown upwards by the steam being produced. The amounts of mass being 'mixed' with liquid water (and thus are thought to potentially participate in a steam explosion) remain below 10% for the known Theofanous criterion and below 30% for a more conservative criterion. It is however more important that the calculation demonstrates that further mixing could be the result of the onset of a steam explosion. This may strongly limit the usefulness of local mixing criteria. (orig./DG)
An assessment of the radiological consequences of releases to groundwater following a core-melt accident at the Sizewell PWR

International Nuclear Information System (INIS)

Maul, P.R.

1984-03-01

In the extremely unlikely event of a degraded core accident at the proposed Sizewell PWR it is theoretically possible for the core to melt through the containment, after which activity could enter groundwater directly or as a result of subsequent leaching of the core in the ground. The radiological consequences of such an event are analysed and compared with the analysis undertaken by the NRPB for the corresponding releases to atmosphere. It is concluded that the risks associated with the groundwater route are much less important than those associated with the atmospheric route. The much longer transport times in the ground compared with those in the atmosphere enable countermeasures to be taken, if necessary, to restrict doses to members of the public to very low levels in the first few years following the accident. The entry of long-lived radionuclides into the sea over very long timescales results in the largest contribution to population doses, but these are delivered at extremely low dose rates which would be negligible compared with background exposure. (author)
Event course analysis of core disruptive accidents

International Nuclear Information System (INIS)

Hering, W.; Homann, C.; Sengpiel, W.; Struwe, D.; Messainguiral, C.

1995-01-01

The theortical studies of the behavior of a PWR core in a meltdown accident are focused on hydrogen release, materials redistribution in the core area including forming of an oxide melt pool, quantity of melt and its composition, and temperatures attained by the RPV internals (esp. in the upper plenum) during the accident up to the time of melt relocation into the lower plenum. The calculations are done by the SCDAP/RELAP5 code. For its validation selected CORA results and Phebus FPTO results have been used. (orig.)
Using just-in-time teaching and peer instruction in a residency program's core curriculum: enhancing satisfaction, engagement, and retention.

Science.gov (United States)

Schuller, Mary C; DaRosa, Debra A; Crandall, Marie L

2015-03-01

To assess use of the combined just-in-time teaching (JiTT) and peer instruction (PI) instructional strategy in a residency program's core curriculum. In 2010-2011, JiTT/PI was piloted in 31 core curriculum sessions taught by 22 faculty in the Northwestern University Feinberg School of Medicine's general surgery residency program. JiTT/PI required preliminary and categorical residents (n=31) to complete Web-based study questions before weekly specialty topic sessions. Responses were examined by faculty members "just in time" to tailor session content to residents' learning needs. In the sessions, residents answered multiple-choice questions (MCQs) using clickers and engaged in PI. Participants completed surveys assessing their perceptions of JiTT/PI. Videos were coded to assess resident engagement time in JiTT/PI sessions versus prior lecture-based sessions. Responses to topic session MCQs repeated in review sessions were evaluated to study retention. More than 70% of resident survey respondents indicated that JiTT/PI aided in the learning of key points. At least 90% of faculty survey respondents reported positive perceptions of aspects of the JiTT/PI strategy. Resident engagement time for JiTT/PI sessions was significantly greater than for prior lecture-based sessions (z=-2.4, P=.016). Significantly more review session MCQ responses were correct for residents who had attended corresponding JiTT/PI sessions than for residents who had not (chi-square=13.7; df=1; P<.001). JiTT/PI increased learner participation, learner retention, and the amount of learner-centered time. JiTT/PI represents an effective approach for meaningful and active learning in core curriculum sessions.
Fundamentals of Melt-Water Interfacial Transport Phenomena: Improved Understanding for Innovative Safety Technologies in ALWRs

Energy Technology Data Exchange (ETDEWEB)

M. Anderson; M. Corradini; K.Y. Bank; R. Bonazza; D. Cho

2005-04-26

The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The goal of this work is to provide the fundamental understanding needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability. The work considers the ex-vessel coolability phenomena in two stages. The first stage is the melt quenching process and is being addressed by Argonne National Lab and University of Wisconsin in modified test facilities. Given a quenched melt in the form of solidified debris, the second stage is to characterize the long-term debris cooling process and is being addressed by Korean Maritime University in via test and analyses. We then address the appropriate scaling and design methodologies for reactor applications.
Fundamentals of Melt-Water Interfacial Transport Phenomena: Improved Understanding for Innovative Safety Technologies in ALWRs

International Nuclear Information System (INIS)

Anderson, M.; Corradini, M.; Bank, K.Y.; Bonazza, R.; Cho, D.

2005-01-01

The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The goal of this work is to provide the fundamental understanding needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability. The work considers the ex-vessel coolability phenomena in two stages. The first stage is the melt quenching process and is being addressed by Argonne National Lab and University of Wisconsin in modified test facilities. Given a quenched melt in the form of solidified debris, the second stage is to characterize the long-term debris cooling process and is being addressed by Korean Maritime University in via test and analyses. We then address the appropriate scaling and design methodologies for reactor applications
Acoustic detection of melt particles

International Nuclear Information System (INIS)

Costley, R.D. Jr.

1988-01-01

The Reactor Safety Research Department at Sandia National Laboratories is investigating a type of Loss of Coolant Accident (LOCA). In this particular type of accident, core meltdown occurs while the pressure within the reactor pressure vessel (RPV) is high. If one of the instrument tube penetrations in the lower head fails, melt particles stream through the cavity and into the containment vessel. This experiment, which simulates this type accident, was performed in the Surtsev Direct Heating Test Facility which is approximately a 1:10 linear scaling of a large dry containment volume. A 1:10 linear scale model of the reactor cavity was placed near the bottom of the Surtsey vessel so that the exit of the cavity was at the vertical centerline of the vessel. A pressure vessel used to create the simulated molten core debris was located at the scaled height of the RPV. In order to better understand how the melt leaves the cavity and streams into the containment an array of five acoustic sensors was placed directly in the path of the melt particles about 30 feet from the exit of the sealed cavity. Highly damped, broadband sensors were chosen to minimize ringing so that individual particle hits could be detected. The goal was to count the signals produced by the individual particle hits to get some idea of how the melt particles left the cavity. This document presents some of the results of the experiment. 9 figs
Nitrogen partitioning during core-mantle differentiation

Science.gov (United States)

Speelmanns, I. M.; Schmidt, M. W.; Liebske, C.

2016-12-01

This study investiagtes nitrogen partitioing between metal and silicate melts as relevant for core segregation during the accretion of planetesimals into the Earth. On present day Earth, N belongs to the most important elements, as it is one of the key constituents of our atmosphere and forms the basis of life. However, the geochemistry of N, i.e. its distribution and isotopic fractionation between Earth's deep reservoirs is not well constrained. In order to determine the partitioning behaviour of N, a centrifuging piston cylinder was used to euqilibrate and then gravitationally separate metal-silicate melt pairs at 1250 °C, 1 GPa over the range of oxygen fugacities thought to have prevailied druing core segreagtion (IW-4 to IW). Complete segregation of the two melts was reached within 3 hours at 1000 g, the interface showing a nice meniscus The applied double capsule technique, using an outer metallic and inner non-metallic (mostly graphite) capsule, minimizes volatile loss over the course of the experiment compared to single non-metallic capsules. The two quenched melts were cut apart, cleaned at the outside and N concentrations of the melts were analysed on bulk samples by an elemental analyser. Nevertheless, the low amount of sample material and the N yield in the high pressure experiments required the developement of new analytical routines. Despite these experimental and analytical difficulties, we were able to determine a DNmetal/silicateof 13±0.25 at IW-1, N partitioning into the core froming metal. The few availible literature data [1],[2] suggest that N changes its compatibility favoring the silicate melt or magma ocean at around IW-2.5. In order to asses how much N may effectively be contained in the core and the silicate Earth, experiments characterizing N behaviour over the entire range of core formation condtitions are well under way. [1] Kadik et al., (2011) Geochemistry International 49.5: 429-438. [2] Roskosz et al., (2013) GCA 121: 15-28.
Luna 24 ferrobasalt as a low-Mg primary melt

International Nuclear Information System (INIS)

Norman, M.; Ryder, G.

1980-01-01

Luna 24 very-low titanium (VLT) ferrobasalts, metabasalts, brown glasses and impact melts form a tight compositional cluster with no gradation to other groupings postulated for the Luna 24 core components. This suggests that the Luna 24 VLT ferrobasalt was extruded as a liquid of its own composition and was not derived by fractional crystallization from a more magnesian parent in a surface flow. Furthermore, the characteristics of the core lithologies are not easily visualized as components of such a differential flow, e.g. brown glasses. Gravitative settling models purporting to demonstrate the validity of the flow differentiation model are merely permissive. Subsurface fractionation requires that plagioclase, not olivine, be the liquidus phase. The high-Mg component in the Luna 24 core can be constrained, though not identified, chemically, and it has neither the major element, trace element, isotopic, nor mineralogical characteristics required of a possible parent to the Luna 24 VLT ferrobasalt. Thus models of fractionation lack a physical expression of the less differentiated compositions, contrary to the belief that the high-Mg component in the core is the parent material. The Luna 24 VLT ferrobasalt is probably a primary low-Mg melt from a plagioclase-bearing source region, and may have undergone little or no fractionation prior to eruption. Such a model is compatible with, and suggested by, chemical and experimental data. Caution against posulating that all Mg-poor melts are fractionated products, based on terrestrial models, is advised. The terrestrial oceanic situation of 'primary melts' with similar Mg/Fe is probably not valid for the Moon. (Auth.)

Investigating the translation of Earth's inner core

DEFF Research Database (Denmark)

Day, Elizabeth A; Cormier, Vernon F; Geballe, Zachary M

2012-01-01

The Earth’s inner core provides unique insights into processes that are occurring deep within our Earth today, as well as processes that occurred in the past. The seismic structure of the inner core is complex, and is dominated by anisotropic and isotropic differences between the Eastern...... for models of a translating inner core. Additionally, we investigate the structure at the base of the outer core and the inner core boundary by analyzing PKP-Cdiff waves. The search for observable PKP-Cdiff is particularly concentrated in regions that are predicted to be actively freezing and melting...... and Western ‘hemispheres’ of the inner core. Recent geodynamical models suggest that this hemispherical dichotomy can be explained by a fast translation of the inner core. In these models one side of the inner core is freezing, while the other side is melting, leading to the development of different seismic...
Melt quenching and coolability by water injection from below: Co-injection of water and non-condensable gas

International Nuclear Information System (INIS)

Cho, Dae H.; Page, Richard J.; Abdulla, Sherif H.; Anderson, Mark H.; Klockow, Helge B.; Corradini, Michael L.

2006-01-01

The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The goal of our work is to provide the fundamental understanding needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability. The work considers the ex-vessel coolability phenomena in two stages. The first stage is the melt quenching process and is being addressed by Argonne National Lab and University of Wisconsin in modified test facilities. Given a quenched melt in the form of solidified debris, the second stage is to characterize the long-term debris cooling process and is being addressed by Korean Maritime University via test and analyses. In this paper, experiments on melt quenching by the injection of water from below are addressed. The test section represented one-dimensional flow-channel simulation of the bottom injection of water into a core melt in the reactor cavity. The melt simulant was molten lead or a lead alloy (Pb-Bi). For the experimental conditions employed (i.e., melt depth and water flow rates), it was found that: (1) the volumetric heat removal rate increased with increasing water mass flow rate and (2) the non-condensable gas mixed with the injected water had no impairing effect on the overall heat removal rate. Implications of these current experimental findings for ALWR ex-vessel coolability are discussed
Materials problems related to the core catcher of sodium cooled reactors

International Nuclear Information System (INIS)

Goetzmann, O.

1975-05-01

There are in principal two possible solutions for the external core catcher as far as materials are concerned. 1) A barrier consisting of a material with a high melting point, 2) a tray of comparatively low melting material with a high solubility for the fuel. In case of the first concept one has to look for materials whose melting temperatures are above the temperature of the molten core. Based on metallurgical reasons it seems very likely that the molten core does not exceed a temperature in the range between 2,500 and 2,800 0 C. Due to the compatibility situation with the molten core only a few high melting oxides will be suitable as liner materials for a core catcher. In the second case basalt or concrete, if free of water and lime, are suitable materials. Graphite is a high melting material, however, due to its behaviour with the molten core it should be listed under the second group. By the reaction of graphite with the core materials the melt can be kept liquid down to temperatures of around 1,100 0 C. The evolution of CO by this reaction should be supportable. It is an endothermal reaction. Experiments on the behaviour of core catcher materials have shown that sodium is capable of penetrating into sintered bodies of UO 2 with densities of 90% TD at temperatures higher than 200 0 C. This may lead to the desintegration of these bodies. The exposure to moist air has not done much harm to UO 2 pellets of densities from 80 to 90% TD. Even after one year of exposure, swelling or desintegration could not be observed. Sodium is also capable of penetrating into bodies of synthetic carbon and graphite. Only well graphitized material will not be destroyed. (orig.) [de
Possible Role of Hydrogen in the Earth Core

Science.gov (United States)

Takahashi, E.; Imai, T.

2011-12-01

Possible role of hydrogen in the Earth core has been discussed by Stevenson (1977) and demonstrated experimentally by Fukai (1984), Okuchi (1997) and others. Planetary theory proposes a possibility of hydrogen incorporation in Earth's magma ocean from ambient solar nebula gas (Ikoma & Genda 2005, Genda & Ikoma 2008). More recently, migration of snow line during planet formation was examined (Min et al., 2010; Oka et al, 2011) and it was proposed that the Earth building material originally contained abundant water as ice and hydrous minerals. Therefore, it is very important to investigate the fate of water in the planet building process and clarify the role of hydrogen in the planetary core. Using SPring-8 synchrotron (NaCl capsule, LiAlH4 as hydrogen source), we determined the melting curve of FeH up to 20 GPa under hydrogen saturated conditions (Sakamaki, Takahashi et al, 2009). Observed melting point is below 1300C and has a very small dT/dP slope. By extrapolating the melting curve using Lindeman's law, we proposed that hydrogen could lower the melting temperature of the Earth core by more than 1500K than current estimate. Here we report our new experiments using SPring-8 synchrotron (single crystal diamond capsule, water as hydrogen source). Hydrogen concentration and melting temperature of FeHx that coexists with hydrous mantle minerals were determined at 15-20GPa and 1000-1600C. We show that 1) hydrogen concentration in FeHx at 1000C, coexisting with hydrous-B and ringwoodite is approximately X=0.6. 2) Upon heating, hydrous-B decomposes and hydrogen strongly partitions into FeHx (X=0.8~1.0) than ringwoodite. 3) FeHx that coexists with ringwoodite melts between ~1300C (solidus) and ~1600C (liquidus). Combined our new experiments with those by Sakamaki et al (2009) and Shibazaki et al (2009), partitioning of hydrogen between proto-core and primitive mantle is discussed. We propose that >90% of water in the source material may have entered the Earth core. Given
Enhanced 99 Tc retention in glass waste form using Tc(IV)-incorporated Fe minerals

Energy Technology Data Exchange (ETDEWEB)

Um, Wooyong; Luksic, Steven A.; Wang, Guohui; Saslow, Sarah; Kim, Dong-Sang; Schweiger, Michael J.; Soderquist, Chuck Z.; Bowden, Mark E.; Lukens, Wayne W.; Kruger, Albert A.

2017-11-01

Technetium (99Tc) immobilization by doping into iron oxide mineral phases may alleviate the problems with Tc volatility during vitrification of nuclear waste. Reduced Tc, Tc(IV), substitutes for Fe(III) in the crystal structure by a process of Tc reduction from Tc(VII) to Tc(IV) followed by co-precipitation of Fe oxide minerals. Two Tc-incorporated Fe minerals (Tc-goethite and Tc-magnetite/maghemite) were prepared and tested for Tc retention in glass melt samples at temperatures between 600 – 1,000 oC. After being cooled, the solid glass specimens prepared at different temperatures were analyzed for Tc oxidation state using Tc K-edge XANES. In most samples, Tc was partially oxidized from Tc(IV) to Tc(VII) as the melt temperature increased. However, Tc retention in glass melt samples prepared using Tc-incorporated Fe minerals were moderately higher than in glass prepared using KTcO4 because of limited and delayed Tc volatilization.
Calculation of individual and population doses on Danish territory resulting from hypothetical core-melt accidents at the Barsebaeck reactor

International Nuclear Information System (INIS)

1977-01-01

Individual and population doses within Danish territory are calculated from hypothetical, severe core-melt accidents at the Swedish nuclear plant at Barsebaeck. The fission product inventory of the Barsebaeck reactor is calculated. The release fractions for the accidents are taken from WASH-1400. Based on parametric studies, doses are calculated for very unfavourable, but not incredible weather conditions. The probability of such conditions in combination with wind direction towards Danish territory is estimated. Doses to bone marrow, lungs, GI-tract and thyroid are calculated based on dose models developed at Risoe. These doses are found to be consistent with doses calculated with the models used in WASH-1400. (author)
Status of degraded core issues. Synthesis paper prepared by G. Bandini in collaboration with the NEA task group on degraded core cooling

International Nuclear Information System (INIS)

2001-02-01

The in-vessel evolution of a severe accident in a nuclear reactor is characterised, generally, by core uncover and heat-up, core material oxidation and melting, molten material relocation and debris behaviour in the lower plenum up to vessel failure. The in-vessel core melt progression involves a large number of physical and chemical phenomena that may depend on the severe accident sequence and the reactor type under consideration. Core melt progression has been studied in the last twenty years through many experimental works. Since then, computer codes are being developed and validated to analyse different reactor accident sequences. The experience gained from the TMI-2 accident also constitutes an important source of data. The understanding of core degradation process is necessary to evaluate initial conditions for subsequent phases of the accident (ex-vessel and within the containment), and define accident management strategies and mitigative actions for operating and advanced reactors. This synthesis paper, prepared within the Task Group on Degraded Core Cooling (TG-DCC) of PWG2, contains a brief summary of current views on the status of degraded core issues regarding light water reactors. The in-vessel fission product release and transport issue is not addressed in this paper. The areas with remaining uncertainties and the needs for further experimental investigation and model development have been identified. The early phase of core melt progression is reasonably well understood. Remaining uncertainties may be addressed on the basis of ongoing experimental activities, e.g. on core quenching, and research programs foreseen in the near future. The late phase of core melt progression is less understood. Ongoing research programs are providing additional valuable information on corium molten pool behaviour. Confirmatory research is still required. The pool crust behaviour and material relocation into the lower plenum are the areas where additional research should
CFD to modeling molten core behavior simultaneously with chemical phenomena

International Nuclear Information System (INIS)

Vladimir V Chudanov; Anna E Aksenova; Valerii A Pervichko

2005-01-01

Full text of publication follows: This paper deals with the basic features of a computing procedure, which can be used for modeling of destruction and melting of a core with subsequent corium retaining into the reactor vessel. The destruction and melting of core mean the account of the following phenomena: a melting, draining (moving of the melt through a porous layer of core debris), freezing with release of an energy, change of geometry, formation of the molten pool, whose convective intermixing and distribution influence on a mechanism of borders destruction. It is necessary to take into account that during of heating molten pool and development in it of convective fluxes a stratification of a multi-component melt on two layers of metal light and of oxide heavy components is observed. These layers are in interaction, they can exchange by the separate components as result of diffusion or oxidizing reactions. It can have an effect considerably on compositions, on a specific weight, and on properties of molten interacting phases, and on a structure of the molten stratified pool. In turn, the retaining of the formed molten masses in reactor vessel requires the solution of a matched heat exchange problem, namely, of a natural convection in a heat generating fluid in partially or completely molten corium and of heat exchange problem with taking into account of a melting of the reactor vessel. In addition, it is necessary to take into account phase segregation, caused by influence of local and of global natural convective flows and thermal lag of heated up boundaries. The mathematical model for simulation of the specified phenomena is based on the Navier-Stokes equations with variable properties together with the heat transfer equation. For modeling of a corium moving through a porous layer of core debris, the special computing algorithm to take into account density jump on interface between a melt and a porous layer of core debris is designed. The model was
Preliminary model for core/concrete interactions

International Nuclear Information System (INIS)

Murfin, W.B.

1977-08-01

A preliminary model is described for computing the rate of penetration of concrete by a molten LWR core. Among the phenomena included are convective stirring of the melt by evolved gases, admixture of concrete decomposition products to the melt, chemical reactions, radiative heat loss, and variation of heat transfer coefficients with local pressure. The model is most applicable to a two-phase melt (metallic plus oxidic) having a fairly high metallic content
Simulation experiments concerning core meltdown

International Nuclear Information System (INIS)

Werle, H.

1979-01-01

A gas stream causes a remarkable increase in the interfacial heat flux (by a factor of 8 for v = 0.63 cm/s, v = gas volume flux/horizontal area). The most important characteristics of the system investigated (silicon oil/wood metal) are relatively similar to those of a core melt, Therefore a remarkable increase of the interfacial heat transfer by the gas release may be expected also for a core melt, compared with earlier investigations at the system silicon oil/water the influence of a gas stream is nevertheless remarkably lower for silicon oil/wood metal. This shows that the density ratio plays an important role. (orig./RW) [de
Analysis of core degradation and relocation phenomena and scenarios in a Nordic-type BWR

Energy Technology Data Exchange (ETDEWEB)

Galushin, Sergey, E-mail: galushin@kth.se; Kudinov, Pavel, E-mail: pkudinov@kth.se

2016-12-15

Highlights: • A data base of the debris properties in lower plenum generated using MELCOR code. • The timing of safety systems has significant effect on the relocated debris properties. • Loose coupling between core relocation and vessel failure analyses was established. - Abstract: Severe Accident Management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel cooling of core melt debris. The melt is released from the failed vessel and poured into a deep pool of water located under the reactor. The melt is expected to fragment, quench, and form a debris bed, coolable by a natural circulation and evaporation of water. Success of the strategy is contingent upon melt release conditions from the vessel and melt-coolant interaction that determine (i) properties of the debris bed and its coolability (ii) potential for energetic melt-coolant interactions (steam explosions). Risk Oriented Accident Analysis Methodology (ROAAM+) framework is currently under development for quantification of the risks associated with formation of non-coolable debris bed and occurrence of steam explosions, both presenting a credible threats to containment integrity. The ROAAM+ framework consist of loosely coupled models that describe each stage of the accident progression. Core relocation analysis framework provides initial conditions for melt vessel interaction, vessel failure and melt release frameworks. The properties of relocated debris and melt release conditions, including in-vessel and ex-vessel pressure, lower drywell pool depth and temperature, are sensitive to the accident scenarios and timing of safety systems recovery and operator actions. This paper illustrates a methodological approach and relevant data for establishing a connection between core relocation and vessel failure analysis in ROAAM+ approach. MELCOR code is used for analysis of core degradation and relocation phenomena. Properties of relocated debris are obtained as functions of the accident scenario
SAXS study of transient pre-melting in chain-folded alkanes

International Nuclear Information System (INIS)

Ungar, G.; Wills, H.H.

1990-01-01

A pronounced pre-melting effect is observed in chain-folded crystals of pure monodisperse n-alkane C 246 H 494 . The effect is reversible on a short time scale, but at longer times the once-folded chain crystals are irreversibly lost as slow chain extension proceeds by solid diffusion well below the melting point. The melting process is thus monitored by rapid time-resolved small-angle X-ray (SAXS) measurements, using synchrotron radiation. The results show that the observed pronounced broadening of the DSC melting endotherm for chain-folded crystals is entirely due to genuine pre-melting of lamellar surfaces. Although a significant portion of material is already molten below the final melting point of chain-folded crystals T F , no recrystallization in the chain-extended form can occur until the cores of the crystalline lamellae melt at T F . Pre-melting of extended chain crystals is significantly less pronounced than that of folded chain crystals
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.
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

This report is the final report for the work performed in 1998 in the research project Melt Structure Water Interactions (MSWI), under the auspices of the APRI Project, jointly funded by SKI, HSK, USNRC and the Swedish and Finnish power companies. The present report describes results of advanced analytical and experimental studies concerning melt-water-structure interactions during the course of a hypothetical severe core meltdown accident in a light water reactor (LWR). Emphasis has been placed on phenomena and properties which govern the fragmentation and breakup of melt jets and droplets, melt spreading and coolability, and thermal and mechanical loadings of a pressure vessel during melt-vessel interaction. Many of the investigations performed in support of this project have produced papers which have been published in the proceedings of technical meetings. A short summary of the results achieved in these papers is provided in this overview. Both experimental and analytical studies were performed to improve knowledge about phenomena of melt-structure-water interactions. We believe that significant technical advances have been achieved during the course of these studies. It was found that: the solidification has a strong effect on the drop deformation and breakup. Initially appearing at the drop surface and, later, thickening inwards, the solid crust layer dampens the instability waves on the drop surface and, therefore, hinders drop deformation and breakup. The drop thermal properties also affect the thermal behavior of the drop and, therefore, have impact on its deformation behavior. The jet fragmentation process is a function of many related phenomena. The fragmentation rate depends not only on the traditional parameters, e.g. the Weber number, but also on the melt physical properties, which change as the melt cools down from the liquidus to the solidus temperature. Additionally, the crust formed on the surface of the melt jet will also reduce the propensity
Investigation on Melt-Structure-Water Interactions (MSWI) during severe accidents

International Nuclear Information System (INIS)

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

1999-08-01

This report is the final report for the work performed in 1998 in the research project Melt Structure Water Interactions (MSWI), under the auspices of the APRI Project, jointly funded by SKI, HSK, USNRC and the Swedish and Finnish power companies. The present report describes results of advanced analytical and experimental studies concerning melt-water-structure interactions during the course of a hypothetical severe core meltdown accident in a light water reactor (LWR). Emphasis has been placed on phenomena and properties which govern the fragmentation and breakup of melt jets and droplets, melt spreading and coolability, and thermal and mechanical loadings of a pressure vessel during melt-vessel interaction. Many of the investigations performed in support of this project have produced papers which have been published in the proceedings of technical meetings. A short summary of the results achieved in these papers is provided in this overview. Both experimental and analytical studies were performed to improve knowledge about phenomena of melt-structure-water interactions. We believe that significant technical advances have been achieved during the course of these studies. It was found that: the solidification has a strong effect on the drop deformation and breakup. Initially appearing at the drop surface and, later, thickening inwards, the solid crust layer dampens the instability waves on the drop surface and, therefore, hinders drop deformation and breakup. The drop thermal properties also affect the thermal behavior of the drop and, therefore, have impact on its deformation behavior. The jet fragmentation process is a function of many related phenomena. The fragmentation rate depends not only on the traditional parameters, e.g. the Weber number, but also on the melt physical properties, which change as the melt cools down from the liquidus to the solidus temperature. Additionally, the crust formed on the surface of the melt jet will also reduce the propensity
Characterisation of Ceramic-Coated 316LN Stainless Steel Exposed to High-Temperature Thermite Melt and Molten Sodium

Science.gov (United States)

Ravi Shankar, A.; Vetrivendan, E.; Shukla, Prabhat Kumar; Das, Sanjay Kumar; Hemanth Rao, E.; Murthy, S. S.; Lydia, G.; Nashine, B. K.; Mallika, C.; Selvaraj, P.; Kamachi Mudali, U.

2017-11-01

Currently, stainless steel grade 316LN is the material of construction widely used for core catcher of sodium-cooled fast reactors. Design philosophy for core catcher demands its capability to withstand corium loading from whole core melt accidents. Towards this, two ceramic coatings were investigated for its application as a layer of sacrificial material on the top of core catcher to enhance its capability. Plasma-sprayed thermal barrier layer of alumina and partially stabilised zirconia (PSZ) with an intermediate bond coat of NiCrAlY are selected as candidate material and deposited over 316LN SS substrates and were tested for their suitability as thermal barrier layer for core catcher. Coated specimens were exposed to high-temperature thermite melt to simulate impingement of molten corium. Sodium compatibility of alumina and PSZ coatings were also investigated by exposing samples to molten sodium at 400 °C for 500 h. The surface morphology of high-temperature thermite melt-exposed samples and sodium-exposed samples was examined using scanning electron microscope. Phase identification of the exposed samples was carried out by x-ray diffraction technique. Observation from sodium exposure tests indicated that alumina coating offers better protection compared to PSZ coating. However, PSZ coating provided better protection against high-temperature melt exposure, as confirmed during thermite melt exposure test.
Event course analysis of core disruptive accidents; Ereignisablaufanalyse kernzerstoerender Unfaelle

Energy Technology Data Exchange (ETDEWEB)

Hering, W.; Homann, C.; Sengpiel, W.; Struwe, D.; Messainguiral, C.

1995-08-01

The theortical studies of the behavior of a PWR core in a meltdown accident are focused on hydrogen release, materials redistribution in the core area including forming of an oxide melt pool, quantity of melt and its composition, and temperatures attained by the RPV internals (esp. in the upper plenum) during the accident up to the time of melt relocation into the lower plenum. The calculations are done by the SCDAP/RELAP5 code. For its validation selected CORA results and Phebus FPTO results have been used. (orig.)
Climate change and forest fires synergistically drive widespread melt events of the Greenland Ice Sheet.

Science.gov (United States)

Keegan, Kaitlin M; Albert, Mary R; McConnell, Joseph R; Baker, Ian

2014-06-03

In July 2012, over 97% of the Greenland Ice Sheet experienced surface melt, the first widespread melt during the era of satellite remote sensing. Analysis of six Greenland shallow firn cores from the dry snow region confirms that the most recent prior widespread melt occurred in 1889. A firn core from the center of the ice sheet demonstrated that exceptionally warm temperatures combined with black carbon sediments from Northern Hemisphere forest fires reduced albedo below a critical threshold in the dry snow region, and caused the melting events in both 1889 and 2012. We use these data to project the frequency of widespread melt into the year 2100. Since Arctic temperatures and the frequency of forest fires are both expected to rise with climate change, our results suggest that widespread melt events on the Greenland Ice Sheet may begin to occur almost annually by the end of century. These events are likely to alter the surface mass balance of the ice sheet, leaving the surface susceptible to further melting.
Simulation of Two-Phase Natural Circulation Loop for Core Cather Cooling Using Air Water

International Nuclear Information System (INIS)

Revankar, S. T.; Huang, S. F.; Song, K. W.; Rhee, B. W.; Park, R. J.; Song, J. H.

2012-01-01

A closed loop natural circulation system employs thermally induced density gradients in single phase or two-phase liquid form to induce circulation of the working fluid thereby obviating the need for any mechanical moving parts such as pumps and pump controls. This increases the reliability and safety of the cooling system and reduces installation, operation and maintenance costs. That is the reason natural circulation cooling has been considered in advanced reactor core cooling and in engineered safety systems. Natural circulation cooling has been proposed to remove reactor decay heat by external vessel cooling for in-vessel core retention during sever accident scenario. Recently in APR1400 reactor core catcher design natural circulation cooling is proposed to stabilize and cool the corium ejected from the reactor vessel following core melt and breach of reactor vessel. The natural circulation flow is similar to external vessel cooling where water flows through an inclined narrow gap below hot surface and is heated to produce boiling. The two-phase natural circulation enables cooling of the corium pool collected on core catcher. Due to importance of this problem this paper focuses simulation of the two-phase natural circulation through inclined gap using air-water system. Scaling criteria for air-water loop are derived that enable simulation of the flow regimes and natural circulation flow rates in such systems using air-water system
How to demonstrate adequacy of protection against a core melt

International Nuclear Information System (INIS)

Hock, R.

1996-01-01

After the Chernobyl accident the public - and consequently the politicians - in Western countries requested improvements in safety for future reactors even in those designs where the type of accident which had destroyed the Chernobyl plant is excluded by fundamental physics. When the major German and French suppliers of nuclear power stations, Siemens and Framatome, decided to develop jointly a next generation reactor type, this political 'request' had to be taken into account. It was decided to include safety features to mitigate the consequences of a core melt - the severest type of accident in a western light water reactor - should it occur despite the many other safety features which are included in this design in order to reduce the probability of occurrence of this type of event to extremely low values. The question arose: How to demonstrate the adequacy of his additional protection? It was evident that the methodology proposed by ICRP namely to demonstrate that the risk of individual members of the public caused by 'probabilistic events' is sufficiently low, could not be used: Due to the low probability of occurrence the contribution of this kind of accident to the risk of any average member of a critical group would already be sufficiently low even if there were no additional countermeasures. In addition, this approach would not cover severe societal effects potentially caused by such an accident. We therefore introduced a different methodology in order to demonstrate the adequacy of additional design features which are only required 'just in case': The consequences of such an unlikely but nevertheless very severe event shall be restricted to the plant itself. Severe consequences outside the immediate vicinity of the plant shall be excluded by the design. (author)

Electrical resistivity discontinuity of iron along the melting curve

Science.gov (United States)

Wagle, Fabian; Steinle-Neumann, Gerd

2018-04-01

Discontinuous changes of electrical resistivity ρel (increase), density ϱ and isothermal compressibility βT (decrease) occur across the melting temperature of metals and can be directly related by Ziman's theory in the long-wavelength approximation. By evaluating experimental data at ambient pressure, we show that Ziman's approximation holds for iron and other simple and transition metals. Using a thermodynamic model to determine βT for γ-, ɛ- and liquid Fe and a previously published model for ρel of liquid Fe, we apply Ziman's approximation to calculate ρel of solid Fe along the melting curve. For pure Fe, we find the discontinuity in ρel to decrease with pressure and to be negligibly small at inner core boundary conditions. However, if we account for light element enrichment in the liquid outer core, the electrical resistivity decrease across the inner core boundary is predicted to be as large as 36 per cent.
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

Highlights: • We describe a test investigating ceramic dissolution by a molten non-eutectic melt. • The evolution of the interface temperature between melt and refractory is measured. • A theoretical model describing dissolution kinetics is proposed. • When dissolution stops, interface temperature is the liquidus temperature of the melt. - Abstract: The present work addresses the question of corium–ceramic interaction in a core catcher during a core-melt accident in a nuclear power plant. It provides an original insight into transient aspects concerning dissolution of refractory material by a volumetrically heated pool. An experiment with simulant material (LIVECERAM) is presented. Test results clearly show that dissolution of solid refractory material can occur in a non-eutectic melt at a temperature which is lower than the melting temperature of the refractory material. During the dissolution transient, the interface temperature rises above the liquidus temperature, corresponding to the instantaneous average composition of the melt pool. With constant power dissipation in the melt and external cooling of the core-catcher, a final steady-state situation is reached. Dissolution stops when the heat flux (delivered by the melt to the refractory) can be removed by conduction through the residual thickness of the ceramic, with T{sub interface} = T{sub liquidus} (calculated for the average composition of the final liquid pool). The final steady state corresponds to a uniform pool composition and uniform interface temperature distribution. Convection in the pool is governed by natural thermal convection and the heat flux distribution is therefore similar to what would be obtained for a single component pool. An interpretation of the experiment with two model-based approaches (0D and 1D) is presented. The mass transfer kinetics between the interface and the bulk is controlled by a diffusion sublayer within the boundary layer. During the dissolution transient
Structure of a mushy layer at the inner core boundary

Science.gov (United States)

Deguen, R.; Huguet, L.; Bergman, M. I.; Labrosse, S.; Alboussiere, T.

2015-12-01

We present experimental results on the solidification of ammonium chloride from an aqueous solution, yielding a mushy zone, under hyper-gravity. A commercial centrifuge has been equipped with a slip-ring so that electric power, temperature and ultrasonic signals could be transmitted between the experimental setup and the laboratory. A Peltier element provides cooling at the bottom of the cell. Probes monitor the temperature along the height of the cell. Ultrasound measurements (2 to 6 MHz) is used to detect the position of the front of the mushy zone and to determine attenuation in the mush. A significant increase of solid fraction (or decrease of mushy layer thickness) and attenuation in the mush is observed as gravity is increased. Kinetic undercooling is significant in our experiments and has been included in a macroscopic mush model. The other ingredients of the model are conservation of energy and chemical species, along with heat/species transfer between the mush and the liquid phase: boundary-layer exchanges at the top of the mush and bulk convection within the mush (formation of chimneys). The outputs of the model compare well with our experiments. We have then run the model in a range of parameters suitable for the Earth's inner core, which has shown the role of bulk mush convection for the inner core and the reason why a solid fraction very close to unity should be expected. We have also run melting experiments: after crystallization of a mush, the liquid has been heated from above until the mush started to melt, while the bottom cold temperature was maintained. These melting experiments were motivated by the possible local melting at the inner core boundary that has been invoked to explain the formation of the anomalously slow F-layer at the bottom of the outer core or inner core hemispherical asymmetry. Oddly, the consequences of melting are an increase in solid fraction and a decrease in attenuation. It is hence possible that surface seismic velocity
Post-accident core coolability of light water reactors

International Nuclear Information System (INIS)

Michio, I.; Teruo, I.; Tomio, Y.; Tsutao, H.

1983-01-01

A study on post-accident core coolability of LWR is discussed based on the practical fuel failure behavior experienced in NSRR, PBF, PNS and others. The fuel failure behavior at LOCA, RIA and PCM conditions are reviewed, and seven types of fuel failure modes are extracted as the basic failure mechanism at accident conditions. These are: cladding melt or brittle failure, molten UO 2 failure, high temperature cladding burst, low temperature cladding burst, failure due to swelling of molten UO 2 , failure due to cracks of embrittled cladding for irradiated fuel rods, and TMI-2 core failure. The post-accident core coolability at each failure mode is discussed. The fuel failures caused actual flow blockage problems. A characteristic which is common among these types is that the fuel rods are in the conditions violating the present safety criteria for accidents, and UO 2 pellets are in melting or near melting hot conditions when the fuel rods failed
A core city problem: recruitment and retention of salaried physicians.

Science.gov (United States)

Paxton, G S; Sbarbaro, J A; Nossaman, N

1975-03-01

The professional and personal characteristics of all physicians recruited into a large urban governmentally sponsored health system were evaluated and correlated to staff retention and loss. The results were tabulated for 84 physicians, approximately 90 per cent of the physician work force, over a three-year period. Eighty per cent resided in either Denver or the state of Colorado prior to entry. This is further reflected in a significant percentage being enrolled in the local medical school or training programs prior to entry. These facts suggest a possible source of manpower for beginning programs. Twenty-six per cent came from private practice, 32 per cent from the military and 14 per cent from the Public Health Service. The turnover rate averaged 6.2 per cent per year, with 4.4 per cent being initiated by the physician and 1.8 per cent leaving because of administrative pressure. Data from other studies are reviewed. Factors which appear to influence retention positively were residency training (pediatricians), sex (females), age (over 38) and those with team experience. These factors suggest directions as to the type of physician who, if recruited, tend to reduce turnover. The establishment of a group practice atmosphere with rewards for clinical skills and the offering of unusual specialty opportunities are proposed as positive factors in the retention of staff.
A review of the core catcher design in LMR

International Nuclear Information System (INIS)

Lee, Yong Bum; Hahn, Do Hee

2001-08-01

The overwhelming emphasis in reactor safety is on the prevention of core meltdown. Moreover, although there have been several accidents that have resulted in some fuel melting, to date there have been no accidents severe enough to cause the syndrome of core collapse, reactor vessel melt-through, containment penetration, and dispersal into the ground. Nevertheless, a number of proposals have been made for the design of core catcher systems to control or stop the motion of the molten core mass should such an accident take place. Core catchers may differ in both their location within the reactor system and in the mechanism that is used to cool and control the motion of the core debris. In this report the classification, configuration and main features of the core catcher are described. And also, The core catcher design technologies and processes are presented. Finally the core catcher provisions in constructed and planned LMRs (Liquid Metal Reactors) are summarized and the preliminary assessment on the core catcher installation in KALIMER is presented
Core Formation on Asteroid 4 Vesta: Iron Rain in a Silicate Magma Ocean

Science.gov (United States)

Kiefer, Walter S.; Mittlefehldt, David W.

2017-01-01

Geochemical observations of the eucrite and diogenite meteorites, together with observations made by NASA's Dawn spacecraft, suggest that Vesta resembles H chondrites in bulk chemical composition, possibly with about 25% of a CM-chondrite like composition added in. For this model, the core is 15% by mass (or 8 volume %) of the asteroid. The abundances of moderately siderophile elements (Ni, Co, Mo, W, and P) in eucrites require that essentially all of the metallic phase in Vesta segregated to form a core prior to eucrite solidification. Melting in the Fe-Ni-S system begins at a cotectic temperature of 940 deg. C. Only about 40% of the total metal phase, or 3-4 volume % of Vesta, melts prior to the onset of silicate melting. Liquid iron in solid silicate initially forms isolated pockets of melt; connected melt channels, which are necessary if the metal is to segregate from the silicate, are only possible when the metal phase exceeds about 5 volume %. Thus, metal segregation to form a core does not occur prior to the onset of silicate melting.
A condensed review of the core catcher in the LMR

International Nuclear Information System (INIS)

Lee, Yong Bum; Hahn, Do hee

2001-03-01

The overwhelming emphasis in reactor safety is on the prevention of core meltdown. Moreover, although there have been several accidents that have resulted in some fuel melting, to date there have been no accidents severe enough to cause the syndrome of core collapse, reactor vessel melt-through, containment penetration, and dispersal into the ground. Nevertheless, a number of proposals have been made for the design of core catcher systems to control or stop the motion of the molten core mass should such an accident take place. Core catchers may differ in both their location within the reactor system and in the mechanism that is used to cool and control the motion of the core debris. In this report the classification, configuration and main features of the core catcher are described. And also, the core catcher provisions in constructed and planned LMRs (Liquid Metal Reactors) are summarized
Simulation of heat and mass transfer processes in molten core debris-concrete systems. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Felde, D K

1979-01-01

The heat and mass transport phenomena taking place in volumetrically-heated fluids have become of interest in recent years due to their significance in assessments of fast reactor safety and post-accident heat removal (PAHR). Following a hypothetical core disruptive accident (HCDA), the core and reactor internals may melt down. The core debis melting through the reactor vessel and guard vessel may eventually contact the concrete of the reactor cell floor. The interaction of the core debris with the concrete as well as the melting of the debris pool into the concrete will significantly affect efforts to prevent breaching of the containment and the resultant release of radioactive effluents to the environment.
Structural Investigation of Fe-Ni-S and Fe-Ni-Si Melts by High-temperature Fluorescence XAFS Measurements

International Nuclear Information System (INIS)

Manghnani, Murli H.; Balogh, John; Hong Xinguo; Newville, Matthew; Amulele, G.

2007-01-01

Iron-nickel (Fe-Ni) alloy is regarded as the most abundant constituent of Earth's core, with an amount of 5.5 wt% Ni in the core based on geochemical and cosmochemical models. The structural role of nickel in liquid Fe-Ni alloys with light elements such as S or Si is poorly understood, largely because of the experimental difficulties of high-temperature melts. Recently, we have succeeded in acquiring Ni K-edge fluorescence x-ray absorption fine structure (XAFS) spectra of Fe-Ni-S and Fe-Ni-Si melts and alloys. Different structural environment of Ni atoms in Fe-Ni-S and Fe-Ni-Si melts is observed, supporting the effect of light elements in Fe-Ni melts
Chiral hide-and-seek: retention of enantiomorphism in laser-induced nucleation of molten sodium chlorate.

Science.gov (United States)

Ward, Martin R; Copeland, Gary W; Alexander, Andrew J

2011-09-21

We report the observation of non-photochemical laser-induced nucleation (NPLIN) of sodium chlorate from its melt using nanosecond pulses of light at 1064 nm. The fraction of samples that nucleate is shown to depend linearly on the peak power density of the laser pulses. Remarkably, we observe that most samples are nucleated by the laser back into the enantiomorph (dextrorotatory or levorotatory) of the solid prior to melting. We do not observe a significant dependence on polarization of the light, and we put forward symmetry arguments that rule out an optical Kerr effect mechanism. Our observations of retention of chirality can be explained by decomposition of small amounts of the sodium chlorate to form sodium chloride, which provide cavities for retention of clusters of sodium chlorate even 18 °C above the melting point. These clusters remain sub-critical on cooling, but can be activated by NPLIN via an isotropic polarizability mechanism. We have developed a heterogeneous model of NPLIN in cavities, which reproduces the experimental data using simple physical data available for sodium chlorate.
Evaluation of an experiment modelling heat transfer from the melt pool for use in VVER 440/213 reactors

International Nuclear Information System (INIS)

Skop, J.

2003-12-01

The strategy of confining core melt within the reactor vessel is among promising strategies to mitigate severe accidents of VVER 440/213 reactors. This strategy consists in residual heat removal from the melt by external vessel cooling from the outside, using water from the flooded reactor downcomer. This approach can only be successful if the critical heat flux on the external vessel surface is not exceeded. This can be assessed based on the parameters of heat transfer from the core melt pool in the conditions of natural circulation within the pool. Those parameters are the subject of the report. A basic description of the terms and physical basis of the strategy of confining core melt inside the vessel is given in Chapter 2, which also briefly explains similarity theory, based on which the results obtained on experimental facilities, using simulation materials, can be related to the actual situation inside a real reactor. Chapter 3 presents an overview of experimental work addressing the characteristics of heat transfer from the core melt pool in natural circulation conditions and a description of the experimental facilities. An overview of the results emerging from the experiments and their evaluation with respect to their applicability to reactors in Czech nuclear power plants are given in Chapter 4
Development of coring, consolidating, subterrene penetrators

International Nuclear Information System (INIS)

Murphy, H.D.; Neudecker, J.W.; Cort, G.E.; Turner, W.C.; McFarland, R.D.; Griggs, J.E.

1976-02-01

Coring penetrators offer two advantages over full face-melting penetrators, i.e., formation of larger boreholes with no increase in power and the production of glass-lined, structurally undisturbed cores which can be recovered with conventional core-retrieval systems. These cores are of significant value in geological exploratory drilling programs. The initial design details and fabrication features of a 114-mm-diam coring penetrator are discussed; significant factors for design optimization are also presented. Results of laboratory testing are reported and compared with performance predictions, and an initial field trial is described
Study on coolability of melt pool with different strategies

International Nuclear Information System (INIS)

Kulkarni, P.P.; Nayak, A.K.

2014-01-01

Highlights: • Experiments have been performed to test quenching of molten pool with different schemes. • Top flooding, bottom flooding and indirect cooling schemes were used. • A single simulant material with same mass and initial temperature was used. • Bottom flooding technique is found to be the most effective technique. • A comparison of all the three techniques has been presented. - Abstract: After the Fukushima accident, there have been a lot of concerns regarding long term core melt stabilization following a severe accident in nuclear reactors. Several strategies have been contemplated for quenching and stabilization of core melt like top flooding, bottom flooding, indirect cooling, etc. However, the effectiveness of these schemes is yet to be determined properly, for which, lot of experiments are needed. Several experiments have been performed for coolability of molten pool under top flooding condition. A few experiments have been performed for study of coolability of melt pool under bottom flooding as well as for indirect cooling. Besides, these tests are very scattered because they involve different simulant materials, initial temperatures and masses of melt, which makes it very difficult to judge the effectiveness of a particular technique and advantage over the other. In the present paper we have carried out different experiments wherein a single simulant material with same mass was cooled with different techniques starting from the same initial temperature. The result showed that, while top flooding and indirect cooling took several hours to cool, bottom flooding took a few minutes to cool the melt which makes it the most effective technique
Devon island ice cap: core stratigraphy and paleoclimate.

Science.gov (United States)

Koerner, R M

1977-04-01

Valuable paleoclimatic information can be gained by studying the distribution of melt layers in deep ice cores. A profile representing the percentage of ice in melt layers in a core drilled from the Devon Island ice cap plotted against both time and depth shows that the ice cap has experienced a period of very warm summers since 1925, following a period of colder summers between about 1600 and 1925. The earlier period was coldest between 1680 and 1730. There is a high correlation between the melt-layer ice percentage and the mass balance of the ice cap. The relation between them suggests that the ice cap mass balance was zero (accumulation equaled ablation) during the colder period but is negative in the present warmer one. There is no firm evidence of a present cooling trend in the summer conditions on the ice cap. A comparison with the melt-layer ice percentage in cores from the other major Canadian Arctic ice caps shows that the variation of summer conditions found for the Devon Island ice cap is representative for all the large ice caps for about 90 percent of the time. There is also a good correlation between melt-layer percentage and summer sea-ice conditions in the archipelago. This suggests that the search for the northwest passage was influenced by changing climate, with the 19th-century peak of the often tragic exploration coinciding with a period of very cold summers.
SCDAP/RELAP5 modeling of movement of melted material through porous debris in lower head

International Nuclear Information System (INIS)

Siefken, L. J.; Harvego, E. A.

2000-01-01

A model is described for the movement of melted metallic material through a ceramic porous debris bed. The model is designed for the analysis of severe accidents in LWRs, wherein melted core plate material may slump onto the top of a porous bed of relocated core material supported by the lower head. The permeation of the melted core plate material into the porous debris bed influences the heatup of the debris bed and the heatup of the lower head supporting the debris. A model for mass transport of melted metallic material is applied that includes terms for viscosity and turbulence but neglects inertial and capillary terms because of their small value relative to gravity and viscous terms in the momentum equation. The relative permeability and passability of the porous debris are calculated as functions of debris porosity, particle size, and effective saturation. An iterative numerical solution is used to solve the set of nonlinear equations for mass transport. The effective thermal conductivity of the debris is calculated as a function of porosity, particle size, and saturation. The model integrates the equations for mass transport with a model for the two-dimensional conduction of heat through porous debris. The integrated model has been implemented into the SCDAP/RELAP5 code for the analysis of the integrity of LWR lower heads during severe accidents. The results of the model indicate that melted core plate material may permeate to near the bottom of a 1m deep hot porous debris bed supported by the lower head. The presence of the relocated core plate material was calculated to cause a 12% increase in the heat flux on the external surface of the lower head
Method of reducing the hazard which may occur as a consequence of a reactor core meltdown

International Nuclear Information System (INIS)

Donne, M.D.; Dorner, S.; Schumacher, G.

1978-01-01

The core melt resulting from a meltdown accident of a GFB, LWR or LMFRR is collected by a core catcher from graphite placed below the core. The core melt is penetrating step by step into a borate store in the collecting vessel and is dissolving in it. Therefore the borate at the same time will absorb the decay heat. In order to remove the solidified and cooled down melted mass water is applied eliminating the borate. The remaining oxide state of the powdery core is sucked off again from the core catcher together with the water. The borate store (e.g. alkali borate) itself consists of separate layers with shaped parts, the coverings of which are made of steel, iron, cast iron, nickel, iron or nickel alloys, ceramic material or glass. (DG) [de
Method of reducing the hazard which may occur as a consequence of a reactor core meltdown

International Nuclear Information System (INIS)

Donne, M.D.; Dorner, S.; Schumacher, G.

1985-01-01

The core melt resulting from a meltdown accident of a GFB, LWR or LMFRR is collected by a core catcher from graphite placed below the core. The core melt is penetrating step by step into a borate store in the collecting vessel and is dissolving in it. Therefore the borate at the same time will absorb the decay heat. In order to remove the solidified and cooled down melted mass water is applied eliminating the borate. The remaining oxide states of the powdery core is sucked off again from the core catcher together with the water. The borate store (e.g. alkali borate) itself consists of separate layers with shaped parts, the coverings of which are made of steel, iron, cast iron, nickel, iron or nickel alloys, ceramic material or glass. (orig./PW)
Frictional melt generated by the 2008 Mw 7.9 Wenchuan earthquake and its faulting mechanisms

Science.gov (United States)

Wang, H.; Li, H.; Si, J.; Sun, Z.; Zhang, L.; He, X.

2017-12-01

Fault-related pseudotachylytes are considered as fossil earthquakes, conveying significant information that provide improved insight into fault behaviors and their mechanical properties. The WFSD project was carried out right after the 2008 Wenchuan earthquake, detailed research was conducted in the drilling cores. 2 mm rigid black layer with fresh slickenlines was observed at 732.6 m in WFSD-1 cores drilled at the southern Yingxiu-Beichuan fault (YBF). Evidence of optical microscopy, FESEM and FIB-TEM show it's frictional melt (pseudotachylyte). In the northern part of YBF, 4 mm fresh melt was found at 1084 m with similar structures in WFSD-4S cores. The melts contain numerous microcracks. Considering that (1) the highly unstable property of the frictional melt (easily be altered or devitrified) under geological conditions; (2) the unfilled microcracks; (3) fresh slickenlines and (4) recent large earthquake in this area, we believe that 2-4 mm melt was produced by the 2008 Wenchuan earthquake. This is the first report of fresh pseudotachylyte with slickenlines in natural fault that generated by modern earthquake. Geochemical analyses show that fault rocks at 732.6 m are enriched in CaO, Fe2O3, FeO, H2O+ and LOI, whereas depleted in SiO2. XRF results show that Ca and Fe are enriched obviously in the 2.5 cm fine-grained fault rocks and Ba enriched in the slip surface. The melt has a higher magnetic susceptibility value, which may due to neoformed magnetite and metallic iron formed in fault frictional melt. Frictional melt visible in both southern and northern part of YBF reveals that frictional melt lubrication played a major role in the Wenchuan earthquake. Instead of vesicles and microlites, numerous randomly oriented microcracks in the melt, exhibiting a quenching texture. The quenching texture suggests the frictional melt was generated under rapid heat-dissipation condition, implying vigorous fluid circulation during the earthquake. We surmise that during
Analysis of natural convection in volumetrically-heated melt pools

International Nuclear Information System (INIS)

Sehgal, B.R.; Dinh, T.N.; Nourgaliev, R.R.

1996-12-01

Results of series of studies on natural convection heat transfer in decay-heated core melt pools which form in a reactor lower plenum during the progression of a core meltdown accident are described. The emphasis is on modelling and prediction of turbulent heat transfer characteristics of natural convection in a liquid pool with an internal energy source. Methods of computational fluid dynamics, including direct numerical simulation, were applied for investigation

Induced tungsten melting events in the divertor of ASDEX Upgrade and their influence on plasma performance

International Nuclear Information System (INIS)

Krieger, K.; Lunt, T.; Dux, R.; Janzer, A.; Kallenbach, A.; Mueller, H.W.; Neu, R.; Puetterich, T.; Rohde, V.

2011-01-01

Tungsten rods of 1 x 1 x 3 mm were exposed at the outer divertor plate of ASDEX Upgrade using a manipulator system. Controlled melting of the W-rod in H-mode discharges was induced by moving the outer strike point towards the W-rod position. Visible light emission of ejected W droplets was recorded by fast camera systems. The resulting increase of tungsten concentration in the confined plasma was compared to that induced by W laser ablation into the outer main chamber boundary plasma. The resulting divertor retention expressed as ratio of tungsten core penetration probability from a divertor source to that of a main chamber source is ∼100. Ejected droplets are found to move always in general direction of the plasma flow. The measured magnitude of droplet acceleration shows that droplets are mainly subject to rocket forces and friction forces. Typical initial droplet size can be inferred from the time evolution of the droplet light emission to be ≥100μm.
The extreme melt across the Greenland ice sheet in 2012

Science.gov (United States)

Nghiem, S. V.; Hall, D. K.; Mote, T. L.; Tedesco, M.; Albert, M. R.; Keegan, K.; Shuman, C. A.; DiGirolamo, N. E.; Neumann, G.

2012-10-01

The discovery of the 2012 extreme melt event across almost the entire surface of the Greenland ice sheet is presented. Data from three different satellite sensors - including the Oceansat-2 scatterometer, the Moderate-resolution Imaging Spectroradiometer, and the Special Sensor Microwave Imager/Sounder - are combined to obtain composite melt maps, representing the most complete melt conditions detectable across the ice sheet. Satellite observations reveal that melt occurred at or near the surface of the Greenland ice sheet across 98.6% of its entire extent on 12 July 2012, including the usually cold polar areas at high altitudes like Summit in the dry snow facies of the ice sheet. This melt event coincided with an anomalous ridge of warm air that became stagnant over Greenland. As seen in melt occurrences from multiple ice core records at Summit reported in the published literature, such a melt event is rare with the last significant one occurring in 1889 and the next previous one around seven centuries earlier in the Medieval Warm Period. Given its rarity, the 2012 extreme melt across Greenland provides an exceptional opportunity for new studies in broad interdisciplinary geophysical research.
PLUGM: a coupled thermal-hydraulic computer model for freezing melt flow in a channel

International Nuclear Information System (INIS)

Pilch, M.

1982-01-01

PLUGM is a coupled thermal-hydraulic computer model for freezing liquid flow and plugging in a cold channel. PLUGM is being developed at Sandia National Laboratories for applications in Sandia's ex-vessel Core Retention Concept Assessment Program and in Sandia's LMFBR Transition Phase Program. The purpose of this paper is to introduce PLUGM and demonstrate how it can be used in the analysis of two of the core retention concepts under investigation at Sandia: refractory brick crucibles and particle beds
Nuclear reactor core safety device

International Nuclear Information System (INIS)

Colgate, S.A.

1977-01-01

The danger of a steam explosion from a nuclear reactor core melt-down can be greatly reduced by adding a gasifying agent to the fuel that releases a large amount of gas at a predetermined pre-melt-down temperature that ruptures the bottom end of the fuel rod and blows the finely divided fuel into a residual coolant bath at the bottom of the reactor. This residual bath should be equipped with a secondary cooling loop
Applications of liquid state physics to the earth's core

Science.gov (United States)

Stevenson, D. J.

1980-01-01

New results derived for application to the earth's outer core using the modern theory of liquids and the hard-sphere model of liquid structure are presented. An expression derived in terms of the incompressibility and pressure is valid for a high-pressure liquid near its melting point, provided that the pressure is derived from a strongly repulsive pair potential; a relation derived between the melting point and density leads to a melting curve law of essentially the same form as Lindemann's law. Finally, it is shown that the 'core paradox' of Higgins and Kennedy (1971) can occur only if the Gruneisen parameter is smaller than 2/3, and this constant is larger than this value in any liquid for which the pair potential is strongly repulsive.
Comparison of SAS3A and MELT-III predictions for a transient overpower hypothetical accident

International Nuclear Information System (INIS)

Wilburn, N.P.

1976-01-01

A comparison is made of the predictions of the two major codes SAS3A and MELT-III for the hypothetical unprotected transient overpower accident in the FFTF. The predictions of temperatures, fuel restructuring, fuel melting, reactivity feedbacks, and core power are compared
Drag Moderation by the Melting of an Ice Surface in Contact with Water

KAUST Repository

Vakarelski, Ivan Uriev; Chan, Derek Y. C.; Thoroddsen, Sigurdur T

2015-01-01

We report measurements of the effects of a melting ice surface on the hydrodynamic drag of ice-shell-metal-core spheres free falling in water at a Reynolds of number Re∼2×104–3×105 and demonstrate that the melting surface induces the early onset of the drag crisis, thus reducing the hydrodynamic drag by more than 50%. Direct visualization of the flow pattern demonstrates the key role of surface melting. Our observations support the hypothesis that the drag reduction is due to the disturbance of the viscous boundary layer by the mass transfer from the melting ice surface.
Drag Moderation by the Melting of an Ice Surface in Contact with Water

KAUST Repository

Vakarelski, Ivan Uriev

2015-07-24

We report measurements of the effects of a melting ice surface on the hydrodynamic drag of ice-shell-metal-core spheres free falling in water at a Reynolds of number Re∼2×104–3×105 and demonstrate that the melting surface induces the early onset of the drag crisis, thus reducing the hydrodynamic drag by more than 50%. Direct visualization of the flow pattern demonstrates the key role of surface melting. Our observations support the hypothesis that the drag reduction is due to the disturbance of the viscous boundary layer by the mass transfer from the melting ice surface.
COMSORS: A light water reactor chemical core catcher

International Nuclear Information System (INIS)

Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W.

1997-01-01

The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate lightwater reactor (LWR) core-melt accidents and ensure containment integrity. A special dissolution glass made of lead oxide (PbO) and boron oxide (B 2 O 3 ) is placed under the reactor vessel. If molten core debris is released onto the glass, the following sequence happens: (1) the glass absorbs decay heat as its temperature increases and the glass softens; (2) the core debris dissolves into the molten glass; (3) molten glass convective currents create a homogeneous high-level waste (HLW) glass; (4) the molten glass spreads into a wider pool, distributing the heat for removal by radiation to the reactor cavity above or transfer to water on top of the molten glass; and (5) the glass solidifies as increased surface cooling area and decreasing radioactive decay heat generation allows heat removal to exceed heat generation
Material properties influence on steam explosion efficiency. Prototypic versus simulant melts, eutectic versus non-eutectic melts

International Nuclear Information System (INIS)

Leskovar, M.; Mavko, B.

2006-01-01

A steam explosion may occur during a severe nuclear reactor accident if the molten core comes into contact with the coolant water. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. Details of processes taking place prior and during the steam explosion have been experimentally studied for a number of years with adjunct efforts in modelling these processes to address the scaling of these experiments. Steam explosion experiments have shown that there are important differences of behaviour between simulant and prototypical melts, and that also at prototypical melts the fuel coolant interactions depend on the composition of the corium. In experiments with prototypic materials no spontaneous steam explosions occurred (except with an eutectic composition), whereas with simulant materials the steam explosions were triggered spontaneously. The energy conversion ratio of steam explosions with prototypic melts is at least one order of magnitude lower than the energy conversion ratio of steam explosions with simulant melts. Although the different behaviour of prototypic and simulant melts has been known for a number of years, there is no reliable explanation for these differences. Consequently it is not possible to reliably estimate whether corium would behave so non-explosive also in reactor conditions, where the mass of poured melt is nearly three orders of magnitude larger than in experimental conditions. An even more fascinating material effect was observed recently at corium experiments with eutectic and non-eutectic compositions. It turned out that eutectic corium always exploded spontaneously, whereas non-eutectic corium never exploded spontaneously. In the paper, a possible explanation of both material effects (prototypic/simulant melts, eutectic/non-eutectic corium) on the steam explosion is provided. A model for the calculation of the
Analysis of natural convection in volumetrically-heated melt pools

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-01

Results of series of studies on natural convection heat transfer in decay-heated core melt pools which form in a reactor lower plenum during the progression of a core meltdown accident are described. The emphasis is on modelling and prediction of turbulent heat transfer characteristics of natural convection in a liquid pool with an internal energy source. Methods of computational fluid dynamics, including direct numerical simulation, were applied for investigation. Refs, figs, tabs.
PWR degraded core analysis

International Nuclear Information System (INIS)

Gittus, J.H.

1982-04-01

A review is presented of the various phenomena involved in degraded core accidents and the ensuing transport of fission products from the fuel to the primary circuit and the containment. The dominant accident sequences found in the PWR risk studies published to date are briefly described. Then chapters deal with the following topics: the condition and behaviour of water reactor fuel during normal operation and at the commencement of degraded core accidents; the generation of hydrogen from the Zircaloy-steam and the steel-steam reactions; the way in which the core deforms and finally melts following loss of coolant; debris relocation analysis; containment integrity; fission product behaviour during a degraded core accident. (U.K.)
Structural integrity investigation for RPV with various cooling water levels under pressurized melting pool

Directory of Open Access Journals (Sweden)

J. Mao

2018-03-01

Full Text Available The strategy denoted as in-vessel retention (IVR is widely used in severe accident (SA management by most advanced nuclear power plants. The essence of IVR mitigation is to provide long-term external water cooling in maintaining the reactor pressure vessel (RPV integrity. Actually, the traditional IVR concept assumed that RPV was fully submerged into the water flooding, and the melting pool was depressurized during the SA. The above assumptions weren't seriously challenged until the occurrence of Fukushima accident on 2011, suggesting the structural behavior had not been appropriately assessed. Therefore, the paper tries to address the structure-related issue on determining whether RPV safety can be maintained or not with the effect of various water levels and internal pressures created from core meltdown accident. In achieving it, the RPV structural behaviors are numerically investigated in terms of several field parameters, such as temperature, deformation, stress, plastic strain, creep strain, and total damage. Due to the presence of high temperature melt on the inside and water cooling on the outside, the RPV failure is governed by the failure mechanisms of creep, thermal-plasticity and plasticity. The creep and plastic damages are interacted with each other, which further accelerate the failure process. Through detailed investigation, it is found that the internal pressure as well as water levels plays an important role in determining the RPV failure time, mode and site.
Physico-Chemistry and Corium Properties for In-Vessel Retention

International Nuclear Information System (INIS)

Froment, K.; Seiler, J.M.; Gueneau, C.; Dauvois, V.; Barbier, F.; Bellon, M.; Tourasse, M.; Ducros, G.; Cognet, G.; Sudreau, F.

1999-01-01

This paper focuses on some important aspects of consequences of material behaviour and interactions on in-vessel retention capabilities. It discusses the behaviour of corium oxide mixtures at elevated temperatures (miscibility gap and density effects, separation due to density effects in the solid-liquid mixture according to the analysis of the Rasplav experiment results), and then the interaction between metallic layer and vessel wall (physical-chemical interaction of corium with the carbon steel vessel wall, migration of low melting point metallic elements in the solid vessel wall). It proposes a mode for the calculation of melt viscosity (liquid phase viscosity and viscosity in the solidification range), addresses the issue of barium release and residual power and of distribution of the residual power in an oxidic corium
A descriptive and factor analytical study of salient retention ...

African Journals Online (AJOL)

Targeting, acquiring and retaining the right customers are at the core of many ... Future research on a comprehensive examination of satisfaction, loyalty, ... relationship marketing, customer retention, customer defection, complaints handling.
Tomographic location of potential melt-bearing phenocrysts in lunar glass spherules

International Nuclear Information System (INIS)

Ebel, D.S.; Fogel, R.A.; Rivers, M.L.

2005-01-01

Apollo 17 orange glass spherules contain olivine phenocrysts with melt inclusions from depth. Tomography ( 200 spherules located 1 phenocryst. We will try to find melt inclusions and obtain original magma volatiles and compositions. In 1971, Apollo 17 astronauts collected a 10 cm soil sample (74220) comprised almost entirely of orange glass spherules. Below this, a double drive-tube core sampled a 68 cm thick horizon comprised of orange glass and black beads (crystallized equivalents of orange glass). Primitive lunar glass spherules (e.g.-A17 orange glasses) are thought to represent ejecta from lunar mare fire fountains. The fire-fountains were apparently driven by a combination of C-O gas exsolution from orange glass melt and the oxidation of graphite. Upon eruption, magmas lost their volatiles (e.g., S, CO, CO 2 ) to space. Evidence for volatile escape remains as volatile-rich coatings on the exteriors of many spherules. Moreover, it showed that Type I and II Fe-Ni-rich metal particles found within orange glass olivine phenocrysts, or free-floating in the glass itself, are powerful evidence for the volatile driving force for lunar fire fountains. More direct evidence for the volatile mechanism has yet to be uncovered. Issues remaining include: the exact composition of magmatic volatiles; the hypothesized existence of graphite in the magma; the oxygen fugacity of the magma and of the lunar interior. In 1996 reported a single ∼450 micron, equant olivine phenocryst, containing four glassy melt inclusions (or inclusion cores), the largest ∼30micron in size, in a thin section of the 74001/2 drill core. The melt is assumed to sample the parent magma of the lunar basalts at depth, evidenced by the S content of the inclusion (600 ppm) which is 400 ppm greater than that of the orange glass host. Such melts potentially contain a full complement of the volatile components of the parent magma, which can be analyzed by infrared spectroscopy. Although the A17 orange glass
The role of noble metals in electric melting of nuclear waste glass

International Nuclear Information System (INIS)

Roth, G.; Weisenburger, S.

1990-01-01

Electrical melting of nuclear waste glass in ceramic melters applies Joule heating, with the molten glass acting as the conductive medium. The local energy release inside the melt relieves from the restriction of external heat addition, allowing to scale up the melter to industrial units. Certainly, that principle makes the melter operation susceptible for changes of the electrical properties of the glass melt. Hence, the melt properties are required to be locally uniform and constant with time. Temporary fluctuations in the feed composition, however, are usually attenuated by the high retention times being in the order of a day and more. More essential for the melter operation are segregation effects occurring systematically. This behaviour can be observed in the case of the so-called noble metal elements Ruthenium, Palladium and Rhodium, belonging to the Platinum metal group. The subject of this paper is to describe the behaviour of the noble metals in electric melting and the problems they can contribute to. The discussion is based on detailed knowledge gained from PAMELA's LEWC processing and from large-scale vitrification of commercial-like waste simulate at INE/KfK. Finally, ways are indicated to solve the noble metal problem technically
Melting of metallic intermediate level waste

Energy Technology Data Exchange (ETDEWEB)

Huutoniemi, Tommi; Larsson, Arne; Blank, Eva [Studsvik Nuclear AB, Nykoeping (Sweden)

2013-08-15

This report presents a feasibility study of a melting facility for core components and reactor internals. An overview is given of how such a facility for treatment of intermediate level waste might be designed, constructed and operated and highlights both the possibilities and challenges. A cost estimate and a risk analysis are presented in order to make a conclusion of the technical feasibility of such a facility. Based on the authors' experience in operating a low level waste melting facility, their conclusion is that without technical improvements such a facility is not feasible today. This is based on the cost of constructing and operating such a facility, in conjunction with the radiological risks associated with operation and the uncertain benefits to disposal and long term safety.
Melting of metallic intermediate level waste

International Nuclear Information System (INIS)

Huutoniemi, Tommi; Larsson, Arne; Blank, Eva

2013-08-01

This report presents a feasibility study of a melting facility for core components and reactor internals. An overview is given of how such a facility for treatment of intermediate level waste might be designed, constructed and operated and highlights both the possibilities and challenges. A cost estimate and a risk analysis are presented in order to make a conclusion of the technical feasibility of such a facility. Based on the authors' experience in operating a low level waste melting facility, their conclusion is that without technical improvements such a facility is not feasible today. This is based on the cost of constructing and operating such a facility, in conjunction with the radiological risks associated with operation and the uncertain benefits to disposal and long term safety
Ice core melt features in relation to Antarctic coastal climate

NARCIS (Netherlands)

Kaczmarska, M.; Isaksson, E.; Karlöf, L.; Brandt, O.; Winther, J.G.; van de Wal, R.S.W.; van den Broeke, M.R.; Johnsen, S.J.

2006-01-01

Measurement of light intensity transmission was carried out on an ice core S100 from coastal Dronning Maud Land (DML). Ice lenses were observed in digital pictures of the core and recorded as peaks in the light transmittance record. The frequency of ice layer occurrence was compared with climate

Experiments on the behaviour of thermite melt injected into sodium: Final report on the THINA test results

International Nuclear Information System (INIS)

Huber, F.; Kaiser, A.; Peppler, W.

1994-01-01

During hypothetical accidents of fast breeder reactors the core melts and part of the core material inventory is ejected into the upper coolant plenum. As a consequence, a fuel to coolant thermal interaction occurs between the melt and the sodium. A series of simulating experiments was carried out in KfK/IRS to improve the knowledge about the phenomenology of molten fuel/coolant interactions and to support theoretical work on the safety of fast breeder reactors. In the tests, a thermite melt of up to 3270 K is injected from below into a sodium pool the temperature of which is between 770 and 820 K. The masses of the melt and the sodium are about five and 150 kg, respectively. Thermal interactions have been observed to occur as a sequence of small local pressure events mainly during the melt injection. Large-scale vapour explosions have not been observed. Generally, the conversion ratios of thermal to mechanical energy have been low. (author)
State-of-the-Art Report on Molten Corium Concrete Interaction and Ex-Vessel Molten Core Coolability

International Nuclear Information System (INIS)

Bonnet, Jean-Michel; Cranga, Michel; Vola, Didier; Marchetto, Cathy; Kissane, Martin; ); Robledo, Fernando; Farmer, Mitchel T.; Spengler, Claus; Basu, Sudhamay; Atkhen, Kresna; Fargette, Andre; Fisher, Manfred; Foit, Jerzi; Hotta, Akitoshi; Morita, Akinobu; Journeau, Christophe; Moiseenko, Evgeny; Polidoro, Franco; Zhou, Quan

2017-01-01

Activities carried out over the last three decades in relation to core-concrete interactions and melt coolability, as well as related containment failure modes, have significantly increased the level of understanding in this area. In a severe accident with little or no cooling of the reactor core, the residual decay heat in the fuel can cause the core materials to melt. One of the challenges in such cases is to determine the consequences of molten core materials causing a failure of the reactor pressure vessel. Molten corium will interact, for example, with structural concrete below the vessel. The reaction between corium and concrete, commonly referred to as MCCI (molten core concrete interaction), can be extensive and can release combustible gases. The cooling behaviour of ex-vessel melts through sprays or flooding is also complex. This report summarises the current state of the art on MCCI and melt coolability, and thus should be useful to specialists seeking to predict the consequences of severe accidents, to model developers for severe-accident computer codes and to designers of mitigation measures
Studies on melt-water-structure interaction during severe accidents

International Nuclear Information System (INIS)

Sehgal, B.R.; Dinh, T.N.; Okkonen, T.J.; Bui, V.A.; Nourgaliev, R.R.; Andersson, J.

1996-10-01

Results of a series of studies, on melt-water-structure interactions which occur during the progression of a core melt-down accident, are described. The emphasis is on the in-vessel interactions and the studies are both experimental and analytical. Since, the studies performed resulted in papers published in proceedings of the technical meetings, and in journals, copies of a set of selected papers are attached to provide details. A summary of the results obtained is provided for the reader who does not, or cannot, venture into the perusal of the attached papers. (au)
Studies on melt-water-structure interaction during severe accidents

Energy Technology Data Exchange (ETDEWEB)

Sehgal, B.R.; Dinh, T.N.; Okkonen, T.J.; Bui, V.A.; Nourgaliev, R.R.; Andersson, J. [Royal Inst. of Technology, Div. of Nucl. Power Safety, Stockholm (Sweden)

1996-10-01

Results of a series of studies, on melt-water-structure interactions which occur during the progression of a core melt-down accident, are described. The emphasis is on the in-vessel interactions and the studies are both experimental and analytical. Since, the studies performed resulted in papers published in proceedings of the technical meetings, and in journals, copies of a set of selected papers are attached to provide details. A summary of the results obtained is provided for the reader who does not, or cannot, venture into the perusal of the attached papers. (au).
Penetration of molten core materials into basaltic and limestone concrete

International Nuclear Information System (INIS)

Sutherland, H.J.

1978-01-01

In conjunction with the small-scale, melt-concrete interaction tests being conducted at Sandia Laboratories, an acoustic technique has been used to monitor the penetration of molten core materials into basaltic and limestone concrete. Real time plots of the position of the melt/concrete interface have been obtained, and they illustrate that the initial penetration rate of the melt may be of the order of 80 mm/min. Phenomena deduced by the technique include a non-wetted melt/concrete interface
Zircon (Hf, O isotopes) as melt indicator: Melt infiltration and abundant new zircon growth within melt rich layers of granulite-facies lenses versus solid-state recrystallization in hosting amphibolite-facies gneisses (central Erzgebirge, Bohemian Massif)

Science.gov (United States)

Tichomirowa, Marion; Whitehouse, Martin; Gerdes, Axel; Schulz, Bernhard

2018-03-01

In the central Erzgebirge within the Bohemian Massif, lenses of high pressure and ultrahigh pressure felsic granulites occur within meta-sedimentary and meta-igneous amphibolite-facies felsic rocks. In the felsic granulite, melt rich parts and restite form alternating layers, and were identified by petrology and bulk rock geochemistry. Mineral assemblages representing the peak P-T conditions were best preserved in melanocratic restite layers. In contrast, in the melt rich leucocratic layers, garnet and related HP minerals as kyanite are almost completely resorbed. Both layers display differences in accessory minerals: melanosomes have frequent and large monazite and Fe-Ti-minerals but lack xenotime and apatite; leucosomes have abundant apatite and xenotime while monazite is rare. Here we present a detailed petrographic study of zircon grains (abundance, size, morphology, inclusions) in granulite-facies and amphibolite-facies felsic gneisses, along with their oxygen and hafnium isotope compositions. Our data complement earlier Usbnd Pb ages and trace element data (REE, Y, Hf, U) on zircons from the same rocks (Tichomirowa et al., 2005). Our results show that the degree of melting determines the behaviour of zircon in different layers of the granulites and associated amphibolite-facies rocks. In restite layers of the granulite lenses, small, inherited, and resorbed zircon grains are preserved and new zircon formation is very limited. In contrast, new zircons abundantly grew in the melt rich leucocratic layers. In these layers, the new zircons (Usbnd Pb age, trace elements, Hf, O isotopes) best preserve the information on peak metamorphic conditions due to intense corrosion of other metamorphic minerals. The new zircons often contain inherited cores. Compared to cores, the new zircons and rims show similar or slightly lower Hf isotope values, slightly higher Hf model ages, and decreased oxygen isotope ratios. The isotope compositions (Hf, O) of new zircons indicate
Fe-based nanocrystalline powder cores with ultra-low core loss

Energy Technology Data Exchange (ETDEWEB)

Wang, Xiangyue, E-mail: wangxiangyue1986@163.com [China Iron and Steel Research Institute Group, Beijing 100081 (China); Center of Advanced Technology and Materials Co., Ltd., Beijing 100081 (China); Lu, Zhichao; Lu, Caowei; Li, Deren [China Iron and Steel Research Institute Group, Beijing 100081 (China); Center of Advanced Technology and Materials Co., Ltd., Beijing 100081 (China)

2013-12-15

Melt-spun amorphous Fe{sub 73.5}Cu{sub 1}Nb{sub 3}Si{sub 15.5}B{sub 7} alloy strip was crushed to make flake-shaped fine powders. The passivated powders by phosphoric acid were mixed with organic and inorganic binder, followed by cold compaction to form toroid-shaped bonded powder-metallurgical magnets. The powder cores were heat-treated to crystallize the amorphous structure and to control the nano-grain structure. Well-coated phosphate-oxide insulation layer on the powder surface decreased the the core loss with the insulation of each powder. FeCuNbSiB nanocrystalline alloy powder core prepared from the powder having phosphate-oxide layer exhibits a stable permeability up to high frequency range over 2 MHz. Especially, the core loss could be reduced remarkably. At the other hand, the softened inorganic binder in the annealing process could effectively improve the intensity of powder cores. - Highlights: • Fe-based nanocrystalline powder cores were prepared with low core loss. • Well-coated phosphate-oxide insulation layer on the powder surface decreased the core loss. • Fe-based nanocrystalline powder cores exhibited a stable permeability up to high frequency range over 2 MHz. • The softened inorganic binder in the annealing process could effectively improve the intensity of powder cores.
Continuous greenhouse gas measurements from ice cores

DEFF Research Database (Denmark)

Stowasser, Christopher

Ice cores offer the unique possibility to study the history of past atmospheric greenhouse gases over the last 800,000 years, since past atmospheric air is trapped in bubbles in the ice. Since the 1950s, paleo-scientists have developed a variety of techniques to extract the trapped air from...... individual ice core samples, and to measure the mixing ratio of greenhouse gases such as carbon dioxide, methane and nitrous oxide in the extracted air. The discrete measurements have become highly accurate and reproducible, but require relatively large amounts of ice per measured species and are both time......-consuming and labor-intensive. This PhD thesis presents the development of a new method for measurements of greenhouse gas mixing ratios from ice cores based on a melting device of a continuous flow analysis (CFA) system. The coupling to a CFA melting device enables time-efficient measurements of high resolution...
Translation and convection of Earth's inner core

Science.gov (United States)

Monnereau, M.; Calvet, M.; Margerin, L.; Mizzon, H.; Souriau, A.

2012-12-01

The image of the inner core growing slowly at the center of the Earth by gradual cooling and solidification of the surrounding liquid outer core is being replaced by the more vigorous image of a ``deep foundry'', where melting and crystallization rates exceed by many times the net growth rate. Recently, a particular mode of convection, called translation, has been put forward as an important mode of inner core dynamics because this mechanism is able to explain the observed East-West asymmetry of P-wave velocity and attenuation (Monnereau et al. 2010). Translation is a pure solid displacement of the inner core material (solid iron) within its envelop, implying crystallization of entering iron on one side of the inner core and melting on the opposite side. Translation is consistent with multiple scattering models of wave propagation. If they do not experience deformation, iron crystals grow as they transit from one hemisphere to the other. Larger crystals constituting a faster and more attenuating medium, a translation velocity of some cm/yr (about ten times the growth rate) is enough to account for the superficial asymmetry observed for P-wave velocity and attenuation, with grains of a few hundred meters on the crystallizing side (West) growing up to a few kilometers before melting on the East side, and a drift direction located in the equatorial plane. Among all hypotheses that have been proposed to account for the seismic asymmetry, translation is the only one based on a demonstrated link between the seismic data and the proposed dynamics, notably through a model of seismic wave propagation. This mechanism was also proposed to be responsible for the formation of a dense layer at the bottom of the outer core, since the high rate of melting and crystallization would release a liquid depleted in light elements at the surface of the inner core (Alboussiere et al 2010). This would explain the anomalously low gradient of P wave velocity in the lowermost 200 km of the
A study on transient heat transfer of the EU-ABWR external core catcher using the phase-change effective convectivity model

International Nuclear Information System (INIS)

Tran Chi Thanh; Nguyen Viet Hung; Tahara, Mika; Kojima, Yoshihiro; Hamazaki, Ryoichi; Kudinov, Pavel

2015-01-01

In advanced designs of Nuclear Power Plants (NPPs), for mitigation of severe accident consequences, on the one hand, the In-Vessel Retention (IVR) concept has been implemented. On the other hand in other new NPP designs (Generation III and III+) with large power reactors, the External Core Catcher (ECC) has been widely adopted. Assessment of ECC design robustness is largely based on analysis of heat transfer of a melt pool formed in the ECC. Transient heat transfer analysis of an ECC is challenging due to (i) uncertainty in the in-vessel accident progression and subsequent vessel failure modes; (ii) long transient, (iii) high Rayleigh number and complex flows involving phase change of the melt pool formed in an ECC. The present paper is concerned with analysis of transient melt pool heat transfer in the ECC of new Advanced Boiling Water Reactor (ABWR) designed by Toshiba Corporation (Japan). According to the ABWR severe accident management strategy, the ECC is initially dry. In order to prevent steam explosion flooding is initiated after termination of melt relocation from the vessel. The ECC full of melt is cooled from the top directly by water and from the bottom through the ECC walls. In order to assess sustainability of the ECC, heat transfer simulation of a stratified melt pool formed in the ECC is carried out. The problem addressed in this work is heat flux distribution at ECC boundaries when cooling is applied (i) from the bottom, (ii) from the top and from the bottom. To perform melt pool heat transfer simulation, we employ Phase-change Effective Convectivity Model (PECM) which was originally developed as a computationally efficient, sufficiently accurate, 2D/3D accident analysis tools for simulation of transient melt pool heat transfer in the reactor lower plenum. Thermal loads from the melt pool to ECC boundaries are determined for selected ex-vessel accident scenarios. Performance of the ECC, efficiency of severe accident management (SAM) measures and
A study of retention characteristics and quality control of nutraceuticals containing resveratrol and polydatin using fused-core column chromatography.

Science.gov (United States)

Fibigr, Jakub; Šatínský, Dalibor; Solich, Petr

2016-02-20

A new high-performance liquid chromatography method using fused-core column for fast separation of resveratrol and polydatin has been developed and used for quality control of nutraceuticals with resveratrol and polydatin content. Retention characteristics (log k) were studied under different conditions on C-18, RP-Amide C-18, Phenyl-hexyl, Pentafluorophenyl (F5) and Cyano stationary phases for both compounds. The effect of the volume fraction of acetonitrile on a retention factors log k of resveratrol and polydatin were evaluated. The optimal separation conditions for resveratrol, polydatin and internal standard p-nitrophenol were found on the fused-core column Ascentis Express ES-Cyano (100×3.0mm), particle size 2.7μm, with mobile phase acetonitrile/water solution with 0.5% acetic acid pH 3 (20:80, v/v) at a flow rate of 1.0mL/min and at 60°C. The detection wavelength was set at 305nm. Under the optimal chromatographic conditions, good linearity with regression coefficients in the range (r=0.9992-0.9998; n=10) for both compounds was achieved. Commercial samples of nutraceuticals were extracted with methanol using ultrasound bath for 15min. A 5μL sample volume of the filtered solution was directly injected into the HPLC system. Accuracy of the method defined as a mean recovery was in the range 83.2-107.3% for both nutraceuticals. The intraday method precision was found satisfactory and relative standard deviations of sample analysis were in the range 0.8-4.7%. The developed method has shown high sample throughput during sample preparation process, modern separation approach, and short time (3min) of analysis. The results of study showed that the declared content of resveratrol and polydatin varied widely in different nutraceuticals according the producers (71.50-115.00% of declared content). Copyright © 2015 Elsevier B.V. All rights reserved.
Model for melt blockage (slug) relocation and physico-chemical interactions during core degradation under severe accident conditions

International Nuclear Information System (INIS)

Veshchunov, M.S.; Shestak, V.E.

2008-01-01

The model describing massive melt blockage (slug) relocation and physico-chemical interactions with steam and surrounding fuel rods of a bundle is developed on the base of the observations in the CORA tests. Mass exchange owing to slug oxidation and fuel rods dissolution is described by the previously developed 2D model for the molten pool oxidation. Heat fluxes in oxidising melt along with the oxidation heat effect at the melt relocation front are counterbalanced by the heat losses in the surrounding media and the fusion heat effect of the Zr claddings attacked by the melt. As a result, the slug relocation velocity is calculated from the heat flux matches at the melt propagation front (Stefan problem). A numerical module simulating the slug behaviour is developed by tight coupling of the heat and mass exchange modules. The new model demonstrates a reasonable capability to simulate the main features of the massive slug behaviour observed in the CORA-W1 test
Calculations with ANSYS/FLOTRAN to a core catcher benchmark

International Nuclear Information System (INIS)

Willschuetz, H.G.

1999-01-01

There are numerous experiments for the exploration of the corium spreading behaviour, but comparable data have not been available up to now in the field of the long-term behaviour of a corium expanded in a core catcher. For the calculations a pure liquid oxidic melt with a homogeneous internal heat source was assumed. The melt was distributed uniformly over the spreading area of the EPR core catcher. All codes applied the well known k-ε-turbulence-model to simulate the turbulent flow regime of this melt configuration. While the FVM-code calculations were performed with three dimensional models using a simple symmetry, the problem was modelled two-dimensionally with ANSYS due to limited CPU performance. In addition, the 2D results of ANSYS should allow a comparison for the planned second stage of the calculations. In this second stage, the behaviour of a segregated metal oxide melt should be examined. However, first estimates and pre-calculations showed that a 3D simulation of the problem is not possible with any of the codes due to lacking computer performance. (orig.)
Petrological Geodynamics of Mantle Melting II. AlphaMELTS + Multiphase Flow: Dynamic Fractional Melting

Science.gov (United States)

Tirone, Massimiliano

2018-03-01

In this second installment of a series that aims to investigate the dynamic interaction between the composition and abundance of the solid mantle and its melt products, the classic interpretation of fractional melting is extended to account for the dynamic nature of the process. A multiphase numerical flow model is coupled with the program AlphaMELTS, which provides at the moment possibly the most accurate petrological description of melting based on thermodynamic principles. The conceptual idea of this study is based on a description of the melting process taking place along a 1-D vertical ideal column where chemical equilibrium is assumed to apply in two local sub-systems separately on some spatial and temporal scale. The solid mantle belongs to a local sub-system (ss1) that does not interact chemically with the melt reservoir which forms a second sub-system (ss2). The local melt products are transferred in the melt sub-system ss2 where the melt phase eventually can also crystallize into a different solid assemblage and will evolve dynamically. The main difference with the usual interpretation of fractional melting is that melt is not arbitrarily and instantaneously extracted from the mantle, but instead remains a dynamic component of the model, hence the process is named dynamic fractional melting (DFM). Some of the conditions that may affect the DFM model are investigated in this study, in particular the effect of temperature, mantle velocity at the boundary of the mantle column. A comparison is made with the dynamic equilibrium melting (DEM) model discussed in the first installment. The implications of assuming passive flow or active flow are also considered to some extent. Complete data files of most of the DFM simulations, four animations and two new DEM simulations (passive/active flow) are available following the instructions in the supplementary material.
Coolability of severely degraded CANDU cores

International Nuclear Information System (INIS)

Meneley, D.A.; Blahnik, C.; Rogers, J.T.; Snell, V.G.; Mijhawan, S.

1995-07-01

Analytical and experimental studies have shown that the separately cooled moderator in a CANDU reactor provides an effective heat sink in the event of a loss-of-coolant accident (LOCA) accompanied by total failure of the emergency core cooling system (ECCS). The moderator heat sink prevents fuel melting and maintains the integrity of the fuel channels, therefore terminating this severe accident short of severe core damage. Nevertheless, there is a probability, however low, that the moderator heat sink could fail in such an accident. The pioneering work of Rogers (1984) for such a severe accident using simplified models showed that the fuel channels would fail and a bed of dry, solid debris would be formed at the bottom of the calandria which would heat up and eventually melt. However, the molten pool of core material would be retained in the calandria vessel, cooled by the independently cooled shield-tank water, and would eventually re solidify. Thus, the calandria vessel would act inherently as a core-catcher as long as the shield tank integrity is maintained. The present paper reviews subsequent work on the damage to a CANDU core under severe accident conditions and describes an empirically based mechanistic model of this process. It is shown that, for such severe accident sequences in a CANDU reactor, the end state following core disassembly consists of a porous bed of dry solid, coarse debris, irrespective of the initiating event and the core disassembly process. (author). 48 refs., 3 tabs., 18 figs
Nitrogen partitioning during Earth's accretion and core-mantle differentiation

Science.gov (United States)

Speelmanns, I. M.; Schmidt, M. W.; Liebske, C.

2017-12-01

On present day Earth, N is one of the key constituents of our atmosphere and forms the basis of life. However, the deep Earth geochemistry of N, i.e. its distribution and isotopic fractionation between Earth's deep reservoirs is not well constrained. This study investigates nitrogen partitioning between metal and silicate melts as relevant for core segregation during the accretion of planetesimals into the Earth. We have determined N-partitioning coefficients over a wide range of temperatures (1250-2000 °C), pressures (15-35 kbar) and oxygen fugacity's, the latter in the relevant range of core segregation (IW-5 to IW). Centrifuging piston cylinders were used to equilibrate and then gravitationally separate metal-silicate melt pairs. Separation of the two melts is necessary to avoid micro nugget contamination in the silicate melt at reducing conditions double capsule technique in all experiments, using an outer metallic (Pt) and inner non-metallic capsule (graphite or Al2O3), minimizes N-loss over the course of the experiments compared to single non-metallic capsules. The two quenched melts were cut apart mechanically, cleaned at the outside, their N concentrations were then analysed on bulk samples by an elemental analyser, the low abslute masses requiring careful development of analytical routines. Despite these difficulties, we were able to determine a DNmetal/silicate of 13±0.3 at IW-1 decreasing to 2.0±0.2 at IW-5.5, at 1250°C and 15 kbar, N partitioning into the core forming metal. Increasing temperature dramatically lowers the DNmetal/silicate to e.g. 0.5±0.15 at IW-4, during early core formation N was hence mildly incompatible in the metal. The results suggest that under magma ocean conditions (> 2000 oC and fO2 IW-2.5), N-partition coefficents were within a factor of 2 of unity. Hence, N did not partition into the core, which should contain negliligible quantities of N. The few available literature data [1],[2],[3] support N changing compatibility with
Coolability of severely degraded CANDU cores. Revised

International Nuclear Information System (INIS)

Meneley, D.A.; Blahnik, C.; Rogers, J.T.; Snell, V.G.; Nijhawan, S.

1996-01-01

Analytical and experimental studies have shown that the separately cooled moderator in a CANDU reactor provides an effective heat sink in the event of a loss-of-coolant accident (LOCA) accompanied by total failure of the emergency core cooling system (ECCS). The moderator heat sink prevents fuel melting and maintains the integrity of the fuel channels, therefore terminating this severe accident short of severe core damage. Nevertheless, there is a probability, however low, that the moderator heat sink could fail in such an accident. The pioneering work of Rogers (1984) for such a severe accident using simplified models showed that the fuel channels would fail and a bed of dry, solid debris would be formed at the bottom of the calandria which would heat up and eventually melt. However, the molten pool of core material would be retained in the calandria vessel, cooled by the independently cooled shield-tank water, and would eventually resolidify. Thus, the calandria vessel would act inherently as a 'core-catcher' as long as the shield tank integrity is maintained. The present paper reviews subsequent work on the damage to a CANDU core under severe accident conditions and describes an empirically based mechanistic model of this process. It is shown that, for such severe accident sequences in a CANDU reactor, the end state following core disassembly consists of a porous bed of dry solid, coarse debris, irrespective of the initiating event and the core disassembly process. (author)
Free energy and structure of dislocation cores in two-dimensional crystals

NARCIS (Netherlands)

Bladon, P.B.; Frenkel, D.

2004-01-01

The nature of the melting transition in two dimensions is critically dependent on the core energy of dislocations. In this paper, we report calculations of the core free energy and the core size of dislocations in two-dimensional solids of systems interacting via square well, hard disk, and r-12
Interpretation of the results of the CORA-33 dry core BWR test

International Nuclear Information System (INIS)

Ott, L.J.; Hagen, S.

1993-01-01

All BWR degraded core experiments performed prior to CORA-33 were conducted under ''wet'' core degradation conditions for which water remains within the core and continuous steaming feeds metal/steam oxidation reactions on the in-core metallic surfaces. However, one dominant set of accident scenarios would occur with reduced metal oxidation under ''dry'' core degradation conditions and, prior to CORA-33, this set had been neglected experimentally. The CORA-33 experiment was designed specifically to address this dominant set of BWR ''dry'' core severe accident scenarios and to partially resolve phenomenological uncertainties concerning the behavior of relocating metallic melts draining into the lower regions of a ''dry'' BWR core. CORA-33 was conducted on October 1, 1992, in the CORA tests facility at KfK. Review of the CORA-33 data indicates that the test objectives were achieved; that is, core degradation occurred at a core heatup rate and a test section axial temperature profile that are prototypic of full-core nuclear power plant (NPP) simulations at ''dry'' core conditions. Simulations of the CORA-33 test at ORNL have required modification of existing control blade/canister materials interaction models to include the eutectic melting of the stainless steel/Zircaloy interaction products and the heat of mixing of stainless steel and Zircaloy. The timing and location of canister failure and melt intrusion into the fuel assembly appear to be adequately simulated by the ORNL models. This paper will present the results of the posttest analyses carried out at ORNL based upon the experimental data and the posttest examination of the test bundle at KfK. The implications of these results with respect to degraded core modeling and the associated safety issues are also discussed
Influence of gas generation on high-temperature melt/concrete interactions

International Nuclear Information System (INIS)

Powers, D.A.

1979-01-01

Accidents involving fuel melting and eventual contact between the high temperature melt and structural concrete may be hypothesized for both light water thermal reactors and liquid metal cooled breeder reactors. Though these hypothesized accidents have a quite low probability of occurring, it is necessary to investigate the probable natures of the accidents if an adequate assessment of the risks associated with the use of nuclear reactors is to be made. A brief description is given of a program addressing the nature of melt/concrete interactions which has been underway for three years at Sandia Laboratories. Emphasis in this program has been toward the behavior of prototypic melts of molten core materials with concrete representative of that found in existing or proposed reactors. The goals of the experimentation have been to identify phenomena particularly pertinent to questions of reactor safety, and phenomena particularly pertinent to questions of reactor safety, and provide quantitative data suitable for the purposes of risk assessment

Immobilization of carbon 14 contained in spent fuel hulls through melting-solidification treatment

International Nuclear Information System (INIS)

Mizuno, T.; Maeda, T.; Nakayama, S.; Banba, T.

2004-01-01

The melting-solidification treatment of spent nuclear fuel hulls is a potential technique to improve immobilization/stabilization of carbon-14 which is mobile in the environment due to its weakly absorbing properties. Carbon-14 can be immobilized in a solid during the treatment under an inert gas atmosphere, where carbon is not oxidized to gaseous form and remains in the solid. A series of laboratory scale experiments on retention of carbon into an alloy waste form was conducted. Metallic zirconium was melted with metallic copper (Zr/Cu=8/2 in weight) at 1200 deg C under an argon atmosphere. Almost all of the carbon remained in the resulting zirconium-copper alloy. (authors)
Retention Assessment of Core Operations Management Topics for Business Administration Students

Science.gov (United States)

Koppel, Nicole B.; Hollister, Kimberly Killmer

2009-01-01

To meet the new AACSB International standards regarding retention assessment and adequately determine "if and what students are learning," this research presents a framework within which expected learning outcomes and specific learning are assessed. This paper presents the framework and describes how the process can be implemented with…
GLASS MELTING PHENOMENA, THEIR ORDERING AND MELTING SPACE UTILISATION

Directory of Open Access Journals (Sweden)

Němec L.

2013-12-01

Full Text Available Four aspects of effective glass melting have been defined – namely the fast kinetics of partial melting phenomena, a consideration of the melting phenomena ordering, high utilisation of the melting space, and effective utilisation of the supplied energy. The relations were defined for the specific melting performance and specific energy consumption of the glass melting process which involve the four mentioned aspects of the process and indicate the potentials of effective melting. The quantity “space utilisation” has been treated in more detail as an aspect not considered in practice till this time. The space utilisation was quantitatively defined and its values have been determined for the industrial melting facility by mathematical modelling. The definitions of the specific melting performance and specific energy consumption have been used for assessment of the potential impact of a controlled melt flow and high space utilisation on the melting process efficiency on the industrial scale. The results have shown that even the partial control of the melt flow, leading to the partial increase of the space utilisation, may considerably increase the melting performance, whereas a decrease of the specific energy consumption was determined to be between 10 - 15 %.
Measuring technique of super high temperature thermal properties of reactor core materials

International Nuclear Information System (INIS)

Ono, Akira; Baba, Tetsuya; Watanabe, Hideo; Matsumoto, Tsuyoshi

1998-01-01

In this study, thermal properties of reactor core materials used for water cooled reactors and FBR were tried to develop a technique to measure their melt states at less than 3,000degC in order to contribute more correct evaluation of the reactor core behavior at severe accident. Then, a thermal property measuring method of high temperature melt by using floating method was investigated and its fundamental design was begun to investigate under a base of optimum judgement on the air flow floating throw-down method. And, in order to measure emissivity of melt specimen surface essential for correct temperature measurement using the throw down method, a spectroscopic emissivity measuring unit using an ellipsometer was prepared and induced. On the thermal properties measurement using the holding method, a specimen container to measure thermal diffusiveness of the high temperature melts by using laser flashing method was tried to prepare. (G.K.)
Structure of a mushy layer under hypergravity with implications for Earth's inner core

Science.gov (United States)

Huguet, Ludovic; Alboussière, Thierry; Bergman, Michael I.; Deguen, Renaud; Labrosse, Stéphane; Lesœur, Germain

2016-03-01

Crystallization experiments in the dendritic regime have been carried out in hypergravity conditions (from 1 to 1300 g) from an ammonium chloride solution (NH4Cl and H2O). A commercial centrifuge was equipped with a slip ring so that electric power (needed for a Peltier device and a heating element), temperature and ultrasonic signals could be transmitted between the experimental setup and the laboratory. Ultrasound measurements (2-6 MHz) were used to detect the position of the front of the mushy zone and to determine attenuation in the mush. Temperature measurements were used to control a Peltier element extracting heat from the bottom of the setup and to monitor the evolution of crystallization in the mush and in the liquid. A significant increase of solid fraction and attenuation in the mush is observed as gravity is increased. Kinetic undercooling is significant in our experiments and has been included in a macroscopic mush model. The other ingredients of the model are conservation of energy and chemical species, along with heat/species transfer between the mush and the liquid phase: boundary-layer exchanges at the top of the mush and bulk convection within the mush (formation of chimneys). The outputs of the model compare well with our experiments. We have then run the model in a range of parameters suitable for the Earth's inner core. This has shown the role of bulk mush convection for the inner core and the reason why a solid fraction very close to unity should be expected. We have also run melting experiments: after crystallization of a mush, the liquid has been heated from above until the mush started to melt, while the bottom cold temperature was maintained. These melting experiments were motivated by the possible local melting at the inner core boundary that has been invoked to explain the formation of the anomalously slow F-layer at the bottom of the outer core or inner core hemispherical asymmetry. Oddly, the consequences of melting are an increase in
Natural convection of the oxide pool in a three-layer configuration of core melts

Energy Technology Data Exchange (ETDEWEB)

Kim, Su-Hyeon; Park, Hae-Kyun; Chung, Bum-Jin, E-mail: bjchung@khu.ac.kr

2017-06-15

Highlights: • Natural convection of oxide pool in 3-layer configuration during IVR was investigated. • High Ra was achieved by using mass transfer experiments based on analogy concept. • Heat ratio to light metal layer was 14% higher for 3-layer configuration than 2-layer one. • Heat transfer to heavy metal layer was poor and hence heat load to side wall increased. • Angular heat loads to side wall showed strengthened heat focusing at uppermost location. - Abstract: We investigated the natural convection of the oxide layer in a three-layer configuration of core melts in a severe accident. In order to achieve high modified Rayleigh numbers of 10{sup 12}–10{sup 13}, mass transfer experiments were performed using a copper sulfate electroplating system based upon the analogy between heat and mass transfer. Four different cooling conditions of the top and the bottom plates were tested. The upward heat ratios were 14% higher for three-layer than for two-layer due to the reduced heights and the downward heat ratios were lower the same amount. The local Nusselt numbers for the top and the bottom plates were measured and compared with the two layer configuration. To explore the heat load to the reactor vessel, the angle-dependent heat fluxes at the side wall, were measured and compared with the two-layer configuration. Heat load to the side wall and peak heat at the uppermost location were intensified for the three-layer configuration.
Melting Penetration Simulation of Fe-U System at High Temperature Using MPS-LER

International Nuclear Information System (INIS)

Mustari, A P A; Irwanto, Dwi; Yamaji, A

2016-01-01

Melting penetration information of Fe-U system is necessary for simulating the molten core behavior during severe accident in nuclear power plants. For Fe-U system, the information is mainly obtained from experiment, i.e. TREAT experiment. However, there is no reported data on SS304 at temperature above 1350°C. The MPS-LER has been developed and validated to simulate melting penetration on Fe-U system. The MPS-LER modelled the eutectic phenomenon by solving the diffusion process and by applying the binary phase diagram criteria. This study simulates the melting penetration of the system at higher temperature using MPS-LER. Simulations were conducted on SS304 at 1400, 1450 and 1500°C. The simulation results show rapid increase of melting penetration rate. (paper)
The possibility and the effects of a steam explosion in the BWR lower head on recriticality of a BWR core

International Nuclear Information System (INIS)

Sehgal, B.R.; Dinh, T.N.

2002-12-01

The report describes an analysis considering a BWR postulated severe accident scenario during which the late vessel automatic depressurization brings the water below the level of the bottom core plate. The subsequent lack of ECCS leads to core heat up during which the control rods melt and the melt deposits on the core plate. At that point of time in the scenario, the core fuel bundles are still intact and the Zircaloy clad oxidation is about to start. The objective of the study is to provide the conditions of reflood into the hot core due to the level swell or a slug delivered from the lower head as the control rod melt drops into the water. These conditions are employed in the neutronic analysis with the RECRIT code to determine if the core recriticality may be achieved. (au)
OECD MCCI project Melt Eruption Test (MET) design report, Rev. 2. April 15, 2003

International Nuclear Information System (INIS)

Farmer, M.T.; Lomperski, S.; Kilsdonk, D.J.; Aeschlimann, R.W.; Basu, S.

2011-01-01

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
Simulation of steam explosion in stratified melt-coolant configuration

International Nuclear Information System (INIS)

Leskovar, Matjaž; Centrih, Vasilij; Uršič, Mitja

2016-01-01

Highlights: • Strong steam explosions may develop spontaneously in stratified configurations. • Considerable melt-coolant premixed layer formed in subcooled water with hot melts. • Analysis with MC3D code provided insight into stratified steam explosion phenomenon. • Up to 25% of poured melt was mixed with water and available for steam explosion. • Better instrumented experiments needed to determine dominant mixing process. - Abstract: A steam explosion is an energetic fuel coolant interaction process, which may occur during a severe reactor accident when the molten core comes into contact with the coolant water. In nuclear reactor safety analyses steam explosions are primarily considered in melt jet-coolant pool configurations where sufficiently deep coolant pool conditions provide complete jet breakup and efficient premixture formation. Stratified melt-coolant configurations, i.e. a molten melt layer below a coolant layer, were up to now believed as being unable to generate strong explosive interactions. Based on the hypothesis that there are no interfacial instabilities in a stratified configuration it was assumed that the amount of melt in the premixture is insufficient to produce strong explosions. However, the recently performed experiments in the PULiMS and SES (KTH, Sweden) facilities with oxidic corium simulants revealed that strong steam explosions may develop spontaneously also in stratified melt-coolant configurations, where with high temperature melts and subcooled water conditions a considerable melt-coolant premixed layer is formed. In the article, the performed study of steam explosions in a stratified melt-coolant configuration in PULiMS like conditions is presented. The goal of this analytical work is to supplement the experimental activities within the PULiMS research program by addressing the key questions, especially regarding the explosivity of the formed premixed layer and the mechanisms responsible for the melt-water mixing. To
Petrological Geodynamics of Mantle Melting I. AlphaMELTS + Multiphase Flow: Dynamic Equilibrium Melting, Method and Results

Directory of Open Access Journals (Sweden)

Massimiliano Tirone

2017-10-01

Full Text Available The complex process of melting in the Earth's interior is studied by combining a multiphase numerical flow model with the program AlphaMELTS which provides a petrological description based on thermodynamic principles. The objective is to address the fundamental question of the effect of the mantle and melt dynamics on the composition and abundance of the melt and the residual solid. The conceptual idea is based on a 1-D description of the melting process that develops along an ideal vertical column where local chemical equilibrium is assumed to apply at some level in space and time. By coupling together the transport model and the chemical thermodynamic model, the evolution of the melting process can be described in terms of melt distribution, temperature, pressure and solid and melt velocities but also variation of melt and residual solid composition and mineralogical abundance at any depth over time. In this first installment of a series of three contributions, a two-phase flow model (melt and solid assemblage is developed under the assumption of complete local equilibrium between melt and a peridotitic mantle (dynamic equilibrium melting, DEM. The solid mantle is also assumed to be completely dry. The present study addresses some but not all the potential factors affecting the melting process. The influence of permeability and viscosity of the solid matrix are considered in some detail. The essential features of the dynamic model and how it is interfaced with AlphaMELTS are clearly outlined. A detailed and explicit description of the numerical procedure should make this type of numerical models less obscure. The general observation that can be made from the outcome of several simulations carried out for this work is that the melt composition varies with depth, however the melt abundance not necessarily always increases moving upwards. When a quasi-steady state condition is achieved, that is when melt abundance does not varies significantly
Preparation of acetaminophen capsules containing beads prepared by hot-melt direct blend coating.

Science.gov (United States)

Pham, Loan; Christensen, John M

2014-02-01

Twelve hydrophobic coating agents were assessed for their effects on drug release after coating sugar cores by a flexible hot-melt coating method using direct blending. Drug-containing pellets were also produced and used as cores. The cores were coated with single or double wax layers containing acetaminophen (APAP). The harder the wax, the slower the resultant drug releases from single-coated beads. Wax coating can be deposited on cores up to 28% of the beads final weight and reaching 58% with wax and drug. Carnauba-coated beads dissolved in approximately 6 h releasing 80% of the loaded drug. Applying another wax layer extended drug release over 20 h, while still delivering 80% of the loaded drug. When drug-containing pellets (33-58% drug loading) were used as cores, double wax-coated pellets exhibited a near zero-order drug release for 16 h, releasing 80% of the loaded drug delivering 18 mg/h. The simple process of hot-melt coating by direct blending of pellet-containing drug-coated formulations provides excellent options for immediate and sustained release formulations when higher lipid coating or drug loading is warranted. Predicted plasma drug concentration time profiles using convolution and in vitro drug release properties of the beads were performed for optimal formulations.
Applications of nano-fluids to enhance LWR accidents management in in-vessel retention and emergency core cooling systems

International Nuclear Information System (INIS)

Chupin, A.; Hu, L. W.; Buongiorno, J.

2008-01-01

Water-based nano-fluid, colloidal dispersions of nano-particles in water; have been shown experimentally to increase the critical heat flux and surface wettability at very low concentrations. The use of nano-fluids to enhance accidents management would allow either to increase the safe margins in case of severe accidents or to upgrade the power of an existing power plant with constant margins. Building on the initial work, computational fluid dynamics simulations of the nano-fluid injection system have been performed to evaluate the feasibility of a nano-fluid injection system for in-vessel retention application. A preliminary assessment was also conducted on the emergency core cooling system of the European Pressurized Reactor (EPR) to implement a nano-fluid injection system for improving the management of loss of coolant accidents. Several design options were compared/or their respective merits and disadvantages based on criteria including time to injection, safety impact, and materials compatibility. (authors)
Superconducting tin core fiber

International Nuclear Information System (INIS)

Homa, Daniel; Liang, Yongxuan; Hill, Cary; Kaur, Gurbinder; Pickrell, Gary

2015-01-01

In this study, we demonstrated superconductivity in a fiber with a tin core and fused silica cladding. The fibers were fabricated via a modified melt-draw technique and maintained core diameters ranging from 50-300 microns and overall diameters of 125-800 microns. Superconductivity of this fiber design was validated via the traditional four-probe test method in a bath of liquid helium at temperatures on the order of 3.8 K. The synthesis route and fiber design are perquisites to ongoing research dedicated all-fiber optoelectronics and the relationships between superconductivity and the material structures, as well as corresponding fabrication techniques. (orig.)
Seasonal monitoring of melt and accumulation within the deep percolation zone of the Greenland Ice Sheet and comparison with simulations of regional climate modeling

Science.gov (United States)

Heilig, Achim; Eisen, Olaf; MacFerrin, Michael; Tedesco, Marco; Fettweis, Xavier

2018-06-01

Increasing melt over the Greenland Ice Sheet (GrIS) recorded over the past several years has resulted in significant changes of the percolation regime of the ice sheet. It remains unclear whether Greenland's percolation zone will act as a meltwater buffer in the near future through gradually filling all pore space or if near-surface refreezing causes the formation of impermeable layers, which provoke lateral runoff. Homogeneous ice layers within perennial firn, as well as near-surface ice layers of several meter thickness have been observed in firn cores. Because firn coring is a destructive method, deriving stratigraphic changes in firn and allocation of summer melt events is challenging. To overcome this deficit and provide continuous data for model evaluations on snow and firn density, temporal changes in liquid water content and depths of water infiltration, we installed an upward-looking radar system (upGPR) 3.4 m below the snow surface in May 2016 close to Camp Raven (66.4779° N, 46.2856° W) at 2120 m a.s.l. The radar is capable of quasi-continuously monitoring changes in snow and firn stratigraphy, which occur above the antennas. For summer 2016, we observed four major melt events, which routed liquid water into various depths beneath the surface. The last event in mid-August resulted in the deepest percolation down to about 2.3 m beneath the surface. Comparisons with simulations from the regional climate model MAR are in very good agreement in terms of seasonal changes in accumulation and timing of onset of melt. However, neither bulk density of near-surface layers nor the amounts of liquid water and percolation depths predicted by MAR correspond with upGPR data. Radar data and records of a nearby thermistor string, in contrast, matched very well for both timing and depth of temperature changes and observed water percolations. All four melt events transferred a cumulative mass of 56 kg m-2 into firn beneath the summer surface of 2015. We find that
Problem of corium melt coolability in passive protection systems against severe accidents in the containment

Directory of Open Access Journals (Sweden)

Ali Kalvand

2018-05-01

Full Text Available Paper is devoted to the development of the mathematical model and analysis of the problem of corium melt interaction with low-temperature melting blocks in the passive protection systems against severe accidents at the NPP, which is of high importance for substantiation of the nuclear power safety, for building and successful op-erating of passive protection systems. In the third-generation reactors passive protection systems against severe accidents at the NPP are mandatory, therefore this paper is of importance for the nuclear power safety. A few configurations for the cooling blocks’ distribution have been considered and an analysis of the blocks’ melting and corium’s cooling in the pool under reactor vessel have been done, which can serve more effective for further improvement of the safety current systems and for the development of new ones. The ways for solution of the problems and the methods for their successful elaboration were discussed. The developed mathematical models and the analysis performed in the paper might be helpful for the design of passive protection systems of the cori-um melt retention inside the containment after corium melt eruption from the broken reactor vessel.
Research activities at JAERI on core material behaviour under severe accident conditions

International Nuclear Information System (INIS)

Uetsuka, H.; Katanashi, S.; Ishijima, K.

1996-01-01

At the Japan Atomic Energy Research Institute (JAERI), experimental studies on physical phenomena under the condition of a severe accident have been conducted. This paper presents the progress of the experimental studies on fuel and core materials behaviour such as the thermal shock fracture of fuel cladding due to quenching, the chemical interaction of core materials at high temperatures and the examination of TMI-2 debris. The mechanical behaviour of fuel rod with heavily embrittled cladding tube due to the thermal shock during delayed reflooding have been investigated at the Nuclear Safety Research Reactor (NSSR) of JAERI. A test fuel rod was heated in steam atmosphere by both electric and nuclear heating using the NSSR, then the rod was quenched by reflooding at the test section. Melting of core component materials having relatively low melting points and their eutectic reaction with other materials significantly influence on the degradation and melt down of fuel bundles during severe accidents. Therefore basic information on the reaction of core materials is necessary to understand and analyze the progress of core melting and relocation. Chemical interactions have been widely investigated at high temperatures for various binary systems of core component materials including absorber materials such as Zircaloy/Inconel, Zircaloy/stainless steel, Zircaloy/(Ag-In-Cd), stainless steel B 4 C and Zircaloy/B 4 C. It was found that the reaction generally obeyed a parabolic rate law and the reaction rate was determined for each reaction system. Many debris samples obtained from the degraded core of TMI-2 were transported to JAERI for numerous examinations and analyses. The microstructural examination revealed that the most part of debris was ceramic and it was not homogeneous in a microscopic sense. The thermal diffusivity data was also obtained for the temperature range up to about 1800K. The data from the large scale integral experiments were also obtained through the
Validation of Code ASTEC with LIVE-L1 Experimental Results

International Nuclear Information System (INIS)

Bachrata, Andrea

2008-01-01

The severe accidents with core melting are considered at the design stage of project at Generation 3+ of Nuclear Power Plants (NPP). Moreover, there is an effort to apply the severe accident management to the operated NPP. The one of main goals of severe accidents mitigation is corium localization and stabilization. The two strategies that fulfil this requirement are: the in-vessel retention (e.g. AP-600, AP- 1000) and the ex-vessel retention (e.g. EPR). To study the scenario of in-vessel retention, a large experimental program and the integrated codes have been developed. The LIVE-L1 experimental facility studied the formation of melt pools and the melt accumulation in the lower head using different cooling conditions. Nowadays, a new European computer code ASTEC is being developed jointly in France and Germany. One of the important steps in ASTEC development in the area of in-vessel retention of corium is its validation with LIVE-L1 experimental results. Details of the experiment are reported. Results of the ASTEC (module DIVA) application to the analysis of the test are presented. (author)
SCDAP: a light water reactor computer code for severe core damage analysis

International Nuclear Information System (INIS)

Marino, G.P.; Allison, C.M.; Majumdar, D.

1982-01-01

Development of the first code version (MODO) of the Severe Core Damage Analysis Package (SCDAP) computer code is described, and calculations made with SCDAP/MODO are presented. The objective of this computer code development program is to develop a capability for analyzing severe disruption of a light water reactor core, including fuel and cladding liquefaction, flow, and freezing; fission product release; hydrogen generation; quenched-induced fragmentation; coolability of the resulting geometry; and ultimately vessel failure due to vessel-melt interaction. SCDAP will be used to identify the phenomena which control core behavior during a severe accident, to help quantify uncertainties in risk assessment analysis, and to support planning and evaluation of severe fuel damage experiments and data. SCDAP/MODO addresses the behavior of a single fuel bundle. Future versions will be developed with capabilities for core-wide and vessel-melt interaction analysis
Core catcher cooling for a gas-cooled fast breeder

International Nuclear Information System (INIS)

Dalle Donne, M.; Dorner, S.; Schretzmann, K.

1976-01-01

Water, molten salts, and liquid metals are under discussion as coolants for the core catcher of a gas-cooled fast breeder. The authors state that there is still no technically mature method of cooling a core melt. However, the investigations carried out so far suggest that there is a solution to this problem. (RW/AK) [de

Experiments and analyses on melt jet impingement during severe accidents

International Nuclear Information System (INIS)

Sehgal, B.R.; Green, J.A.; Dinh, T.N.; Dong, W.

1997-01-01

Relocation of melt from the core region, during a nuclear reactor severe accident, presents the potential for erosion of the reactor pressure vessel (RPV) wall as a result of melt jet impingement. The extent of vessel erosion will depend upon a variety of parameters, including jet diameter, velocity, composition, superheat, angle of inclination, and the presence of an overlying water or melt pool. Experiments have been conducted at the Royal Institute of Technology Division of Nuclear Power Safety (RIT/NPS) which employ a variety of melt and pressure vessel simulant materials, such as water, salt-ice, Cerrobend alloy and molten salt. These experiments have revealed that the erosion depth of the vessel simulant in the jet stagnation zone can be adequately predicted by the Saito correlation, which is based on turbulent heat transfer, while initial erosion rates are seen to be in line with the laminar-stagnation-zone model. A transition between the laminar and turbulent regimes was realized in most cases and is attributed to the roughness of the surface in the eroded cavity formed
Emerging melt quality control solution technologies for aluminium melt

Directory of Open Access Journals (Sweden)

Arturo Pascual, Jr

2009-11-01

Full Text Available The newly developed “MTS 1500” Melt Treatment System is performing the specifi cally required melt treatment operations like degassing, cleaning, modification and/or grain refinement by an automated process in one step and at the same location. This linked process is saving time, energy and metal losses allowing - by automated dosage of the melt treatment agents - the production of a consistent melt quality batch after batch. By linking the MTS Metal Treatment System with sensors operating on-line in the melt, i.e., with a hydrogen sensor “Alspek H”, a fully automated control of parts of the process chain like degassing is possible. This technology does guarantee a pre-specifi ed and documented melt quality in each melt treatment batch. Furthermore, to ensure that castings are consistent and predictable there is a growing realization that critical parameters such as metal cleanliness must be measured prior to casting. There exists accepted methods for measuring the cleanliness of an aluminum melt but these can be both slow and costly. A simple, rapid and meaningful method of measuring and bench marking the cleanliness of an aluminum melt has been developed to offer the foundry a practical method of measuring melt cleanliness. This paper shows the structure and performance of the integrated MTS melt treatment process and documents achieved melt quality standards after degassing, cleaning, modifi cation and grain refi nement operations under real foundry conditions. It also provides an insight on a melt cleanliness measuring device “Alspek MQ” to provide foundry men better tools in meeting the increasing quality and tighter specifi cation demand from the industry.
Constant electrical resistivity of Ni along the melting boundary up to 9 GPa

Science.gov (United States)

Silber, Reynold E.; Secco, Richard A.; Yong, Wenjun

2017-07-01

Characterization of transport properties of liquid Ni at high pressures has important geophysical implications for terrestrial planetary interiors, because Ni is a close electronic analogue of Fe and it is also integral to Earth's core. We report measurements of the electrical resistivity of solid and liquid Ni at pressures 3-9 GPa using a 3000 t multianvil large volume press. A four-wire method, in conjunction with a rapid acquisition meter and polarity switch, was used to overcome experimental challenges such as melt containment and maintaining sample geometry and to mitigate the extreme reactivity/solubility of liquid Ni with most thermocouple and electrode materials. Thermal conductivity is calculated using the Wiedemann-Franz law. Electrical resistivity of solid Ni exhibits the expected P dependence and is consistent with earlier experimental values. Within experimental uncertainties, our results indicate that resistivity of liquid Ni remains invariant along the P-dependent melting boundary, which is in disagreement with earlier prediction for liquid transition metals. The potential reasons for such behavior are examined qualitatively through the impact of P-independent local short-range ordering on electron mean free path and the possibility of constant Fermi surface at the onset of Ni melting. Correlation among metals obeying the Kadowaki-Woods ratio and the group of late transition metals with unfilled d-electron band displaying anomalously shallow melting curves suggests that on the melting boundary, Fe may exhibit the same resistivity behavior as Ni. This could have important implications for the heat flow in the Earth's core.
Retention of Halogens in Waste Glass

Energy Technology Data Exchange (ETDEWEB)

Hrma, Pavel R.

2010-05-01

In spite of their potential roles as melting rate accelerators and foam breakers, halogens are generally viewed as troublesome components for glass processing. Of five halogens, F, Cl, Br, I, and At, all but At may occur in nuclear waste. A nuclear waste feed may contain up to 10 g of F, 4 g of Cl, and ≤100 mg of Br and I per kg of glass. The main concern is halogen volatility, producing hazardous fumes and particulates, and the radioactive iodine 129 isotope of 1.7x10^7-year half life. Because F and Cl are soluble in oxide glasses and tend to precipitate on cooling, they can be retained in the waste glass in the form of dissolved constituents or as dispersed crystalline inclusions. This report compiles known halogen-retention data in both high-level waste (HLW) and low-activity waste (LAW) glasses. Because of its radioactivity, the main focus is on I. Available data on F and Cl were compiled for comparison. Though Br is present in nuclear wastes, it is usually ignored; no data on Br retention were found.
Investigation of activity release during light water reactor core meltdown

International Nuclear Information System (INIS)

Albrecht, H.; Matschoss, V.; Wild, H.

1978-01-01

A test facility was developed for the determination of activity release and of aerosol characteristics under realistic light water reactor core melting conditions. It is composed of a high-frequency induction furnace, a ThO 2 crucible system, and a collection apparatus consisting of membrane and particulate filters. Thirty-gram samples of a representative core material mixture (corium) were melted under air, argon, or steam at 0.8 to 2.2 bar. In air at 2700 0 C, for example, the relative release was 0.4 to 0.7% for iron, chromium, and cobalt and 4 to 11% for tin, antimony, and manganese. Higher release values of 20 to 40% at lower temperatures (2150 0 C, air) were found for selenium, cadmium, tellurium, and cesium. The size distribution of the aerosol particles was trimodal with maxima at diameters of 0.17, 0.30, and 0.73 μm. The result of a qualitative x-ray microanalysis was that the main elements of the melt were contained in each aerosol particle. Further investigations will include larger melt masses and the additional influence of concrete on the release and aerosol behavior
Apparatus for controlling molten core debris

International Nuclear Information System (INIS)

Golden, M.P.; Tilbrook, R.W.; Heylmun, N.F.

1972-01-01

Disclosed is an apparatus for containing, cooling, diluting, dispersing and maintaining subcritical the molten core debris assumed to melt through the bottom of a nuclear reactor pressure vessel in the unlikely event of a core meltdown. The apparatus is basically a sacrificial bed system which includes an inverted conical funnel, a core debris receptacle including a spherical dome, a spherically layered bed of primarily magnesia bricks, a cooling system of zig-zag piping in graphite blocks about and below the bed and a cylindrical liner surrounding the graphite blocks including a steel shell surrounded by firebrick. Tantalum absorber rods are used in the receptacle and bed. 9 claims, 22 figures
Reentrainment of aerosols during the filtered venting after a severe core melt accident

International Nuclear Information System (INIS)

Mueller, M.

1997-01-01

The major objective of this project is the experimental determination of the aerosol reentrainment from boiling pool during controlled filtered venting of the containment vessel after a severe core melt accident. For this reason a linear downscaled (1:20) model containment with an inner free volume of 5 m 3 is provided. Both, water soluble and unsoluble model substances are used as fission product simulants. The major advantage of the pilot plant is the ability to run it at steady state conditions of any period of time. Further, modelling of the aerosol reentrainment from boiling pool allows upscaling of results on nuclear power plants. The deterministic aerosol reentrainment model can also be used to calculate entrainment phenomena in the process industries such at distillation columns or at flash evaporators. Steady state experiments with water soluble model substances clearly reveal enhanced aerosol reentrainment from boiling pool due to increasing boiling pool concentration of fission product simulants and due to increasing gas velocities above the boiling pool surface. But there can be seen no influence of corium concrete interactions on the aerosol reentrainment. Compared to the steam production due to the decay heat the resulting gas volume flux is negligible. Next, there can be seen aerosol reentrainment from boiling pool only above boiling pool areas. Further, experiments under steady state conditions with unsoluble fission product simulants show on the one hand scrubbing effects in the boiling pool, on the other hand no aerosol reentrainment of solid particles 3 μm. The so called reentrainment factor - ratio between fission product simulant in the venting system and in the boiling pool - is for water soluble model substances in the range of 10 -5 , for unsoluble fission product simulants in the range of 10 -6 . (author) figs., tabs., 57 refs
Evaluation of downmotion time interval molten materials to core catcher during core disruptive accidents postulated in LMFR

International Nuclear Information System (INIS)

Voronov, S.A.; Kiryushin, A.I.; Kuzavkov, N.G.; Vlasichev, G.N.

1994-01-01

Hypothetical core disruptive accidents are postulated to clear potential of a reactor plant to withstand extreme conditions and to generate measures for management and mitigation of accidents consequence. In Russian advanced reactors there is a core catcher below the diagrid to prevent vessel bottom melting and to localize fuel debris. In this paper the calculation technique and estimation of relocation time of molten fuel and materials are presented in the case of core disruptive accidents postulated for LMFR reactor. To evaluate minimum interval of fuel relocation time the calculations for different initial data are provided. Large mass of materials between the core and the catcher in LMFR reactor hinders molten materials relocation toward the vessel bottom. That condition increases the time interval of reaching core catcher by molten fuel. Computations performed allowed to evaluate the minimum molten materials relocation time from the core to the core catcher. This time interval is in a range of 3.5-5.5 hours. (author)
Analysis of the loss of coolant accident for LEU cores of Pakistan research reactor-1

International Nuclear Information System (INIS)

Khan, L.A.; Bokhari, I.H.; Raza, S.H.

1993-12-01

Response of LEU cores for PARR-1 to a Loss of Coolant Accident (LOCA) has been studied. It has been assumed that pool water drains out to double ended rupture of primary coolant pipe or complete shearing of an experimental beam tube. Results show that for an operating power level of 10 MW, both the first high power and equilibrium cores would enter into melting conditions if the pool drain time is less than 22 h and 11 h respectively. However, an Emergency Core Cooling System (ECCS) capable of spraying the core at flow rate of 8.3 m/sup 3/h, for the above mentioned duration, would keep the peak core temperature much below the critical value. Maximum operating power levels below which melting would not occur have been assessed to 3.4 MW and 4.8 MW, respectively, for the first high power and equilibrium cores. (author) 5 figs
Some factors affecting radiative heat transport in PWR cores

International Nuclear Information System (INIS)

Hall, A.N.

1989-04-01

This report discusses radiative heat transport in Pressurized Water Reactor cores, using simple models to illustrate basic features of the transport process. Heat transport by conduction and convection is ignored in order to focus attention on the restrictions on radiative heat transport imposed by the geometry of the heat emitting and absorbing structures. The importance of the spacing of the emitting and absorbing structures is emphasised. Steady state temperature distributions are found for models of cores which are uniformly heated by fission product decay. In all of the models, a steady state temperature distribution can only be obtained if the central core temperature is in excess of the melting point of UO 2 . It has recently been reported that the MIMAS computer code, which takes into account radiative heat transport, has been used to model the heat-up of the Three Mile Island-2 reactor core, and the computations indicate that the core could not have reached the melting point of UO 2 at any time or any place. We discuss this result in the light of the calculations presented in this paper. It appears that the predicted stabilisation of the core temperatures at ∼ 2200 0 C may be a consequence of the artificially large spacing between the radial rings employed in the MIMAS code, rather than a result of physical significance. (author)
The Origin of the Compositional Diversity of Mercury's Surface Constrained From Experimental Melting of Enstatite Chondrites

Science.gov (United States)

Boujibar, A.; Righter, K.; Pando, K.; Danielson, L.

2015-01-01

Mercury is known as an endmember planet as it is the most reduced terrestrial planet with the highest core/mantle ratio. MESSENGER spacecraft has shown that its surface is FeO-poor (2-4 wt%) and Srich (up to 6-7 wt%), which confirms the reducing nature of its silicate mantle. Moreover, high resolution images revealed large volcanic plains and abundant pyroclastic deposits, suggesting important melting stages of the Mercurian mantle. This interpretation was confirmed by the high crustal thickness (up to 100 km) derived from Mercury's gravity field. This is also corroborated by a recent experimental result that showed that Mercurian partial melts are expected to be highly buoyant within the Mercurian mantle and could have risen from depths as high as the core-mantle boundary. In addition MESSENGER spacecraft provided relatively precise data on major elemental compositions of Mercury's surface. These results revealed important chemical and mineralogical heterogeneities that suggested several stages of differentiation and re-melting processes. However, the extent and nature of compositional variations produced by partial melting remains poorly constrained for the particular compositions of Mercury (very reducing conditions, low FeO-contents and high sulfur-contents). Therefore, in this study, we investigated the processes that lead to the various compositions of Mercury's surface. Melting experiments with bulk Mercury-analogue compositions were performed and compared to the compositions measured by MESSENGER.
Volatile diffusion in silicate melts and its effects on melt inclusions

Directory of Open Access Journals (Sweden)

P. Scarlato

2005-06-01

Full Text Available A compendium of diffusion measurements and their Arrhenius equations for water, carbon dioxide, sulfur, fluorine, and chlorine in silicate melts similar in composition to natural igneous rocks is presented. Water diffusion in silicic melts is well studied and understood, however little data exists for melts of intermediate to basic compositions. The data demonstrate that both the water concentration and the anhydrous melt composition affect the diffusion coefficient of water. Carbon dioxide diffusion appears only weakly dependent, at most, on the volatilefree melt composition and no effect of carbon dioxide concentration has been observed, although few experiments have been performed. Based upon one study, the addition of water to rhyolitic melts increases carbon dioxide diffusion by orders of magnitude to values similar to that of 6 wt% water. Sulfur diffusion in intermediate to silicic melts depends upon the anhydrous melt composition and the water concentration. In water-bearing silicic melts sulfur diffuses 2 to 3 orders of magnitude slower than water. Chlorine diffusion is affected by both water concentration and anhydrous melt composition; its values are typically between those of water and sulfur. Information on fluorine diffusion is rare, but the volatile-free melt composition exerts a strong control on its diffusion. At the present time the diffusion of water, carbon dioxide, sulfur and chlorine can be estimated in silicic melts at magmatic temperatures. The diffusion of water and carbon dioxide in basic to intermediate melts is only known at a limited set of temperatures and compositions. The diffusion data for rhyolitic melts at 800°C together with a standard model for the enrichment of incompatible elements in front of growing crystals demonstrate that rapid crystal growth, greater than 10-10 ms-1, can significantly increase the volatile concentrations at the crystal-melt interface and that any of that melt trapped
Structural failure analysis of reactor vessels due to molten core debris

International Nuclear Information System (INIS)

Pfeiffer, P.A.

1993-01-01

Maintaining structural integrity of the reactor vessel during a postulated core melt accident is an important safety consideration in the design of the vessel. This paper addresses the failure predictions of the vessel due to thermal and pressure loadings from the molten core debris depositing on the lower head of the vessel. Different loading combinations were considered based on a wet or dry cavity and pressurization of the vessel based on operating pressure or atmospheric (pipe break). The analyses considered both short term (minutes) and long term (days) failure modes. Short term failure modes include creep at elevated temperatures and plastic instabilities of the structure. Long term failure modes are caused by creep rupture that lead to plastic instability of the structure. The analyses predict the reactor vessel will remain intact after the core melt has deposited on the lower vessel head
In-core fuel disruption experiments simulating LOF accidents for homogeneous and heterogeneous core LMFBRs: FD2/4 series

International Nuclear Information System (INIS)

Wright, S.A.; Mast, P.K.; Schumacher, Gustav; Fischer, E.A.

1982-01-01

A series of Fuel Disruption (FD) experiments simulating LOF accidents transients for homogeneous- and heterogeneous-core LMFBRs is currently being performed in the Annular Core Research Reactor at SNL. The test fuel is observed with high-speed cinematography to determine the timing and the mode of the fuel disruption. The five experiments performed to date show that the timing and mode of fuel disruption depend on the power level, fuel temperature (after preheat and at disruption), and the fuel temperature gradient. Two basic modes of fuel disruption were observed; solid-state disruption and liquid-state swelling followed by slumping. Solid-state dispersive fuel behavior (several hundred degrees prior to fuel melting) is only observed at high power levels (6P 0 ), low preheat temperatures (2000 K), and high thermal gradients (2800 K/mm). The swelling/slumping behavior was observed in all cases near the time of fuel melting. Computational models have been developed that predict the fuel disruption modes and timing observed in the experiments
Gas Retention in a Heated Plastic Bonded Explosive (LX-14).

Energy Technology Data Exchange (ETDEWEB)

Hobbs, Michael L. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States). Engineering Sciences Center; Kaneshige, Michael J. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States). Energetics Components Center; Erikson, William W. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States). Engineering Sciences Center; Meirs, Kevin T. [U.S. Army Armament Research, Development and Engineering Center (ARDEC), Picatinny Arsenal, NJ (United States)

2017-09-01

In prior work, we found that the nitroplasticizer in the plastic bonded explosive PBX 9501 played a crucial role in cookoff, especially when predicting response in larger systems [1]. We have recently completed experiments with a similar explosive, LX-14, that has a relatively nonreactive binder. We expected the ignition times for LX-14 to be longer than PBX 9501 since PBX 9501 has a more reactive binder. However, our experiments show the opposite trend. This paradox can be explained by retention of reactive gases within the interior of LX-14 by the higher strength binder resulting in faster ignition times. In contrast, the binder in PBX 9501 melts at low temperatures and does not retain decomposition gases as well as the LX-14 binder. Retention of reactive gases in LX-14 may also explain the more violent response in oblique impact tests [2] when compared to PBX 9501.
Double melting in polytetrafluoroethylene γ-irradiated above its melting point

International Nuclear Information System (INIS)

Serov, S.A.; Khatipov, S.A.; Sadovskaya, N.V.; Tereshenkov, A.V.; Chukov, N.A.

2012-01-01

Highlights: ► PTFE irradiation leads to formation of double melting peaks in DSC curves. ► This is connected to dual crystalline morphology typical for PTFE. ► Two crystalline types exist in the PTFE irradiated in the melt. - Abstract: PTFE irradiation above its melting point leads to formation of double melting and crystallization peaks in DSC curves. Splitting of melting peaks is connected to dual crystalline morphology typical for PTFE irradiated in the melt. According to electron microscopy, two crystalline types with different size and packing density exist in the irradiated PTFE.
Molten LWR core material interactions with water and with concrete

International Nuclear Information System (INIS)

Dahlgren, D.A.; Buxton, L.D.; Muir, J.F.; Murfin, W.B.; Nelson, L.S.; Powers, D.A.

1977-01-01

Nuclear power reactors are designed and operated to minimize the possibility of fuel melting. Nevertheless, in order to assess the risks associated with reactor operation, a realistic assessment is required for postulated accident sequences in which melting occurs. To investigate the experimental basis of the fuel melt accident analyses, a comprehensive review was performed at Sandia Laboratories. The results of that study indicated several phenomenological areas where additional experimental data should be gathered to verify common assumptions made in risk studies. In particular, vapor explosions and molten core material/concrete interactions were identified for further study. Results of these studies are presented
Research and development on the melting test of low-level radioactive miscellaneous solid waste

International Nuclear Information System (INIS)

Nakashio, Nobuyuki; Hoshi, Akiko; Kameo, Yutaka; Nakashima, Mikio

2007-02-01

The Nuclear Science Research Institute of the Japan Atomic Energy Agency constructed the Advanced Volume Reduction Facilities (AVRF) in February 2003 for treatment of low-level radioactive miscellaneous solid waste (LLW). The waste volume reduction is carried out by a high-compaction process or melting processes in the AVRF. In advance of operating the melting process in the AVRF, melting tests of simulated LLW with RI tracers ( 60 Co, 137 Cs and 152 Eu) have been conducted by using the plasma melter in pilot scale. Viscosity of molten waste, chemical composition and physical properties of solidified products and distribution of the tracers in each product were investigated in various melting conditions. It was confirmed that the viscosity of molten waste was able to be controlled by adjusting chemical composition of molten waste. The RI tracer were almost uniformly distributed in the solidified products. The retention of 137 Cs depended on the basicity (CaO/SiO 2 ) of the solidified products. The solidified product possessed satisfactory compressive strength. In the case of basicity less than 0.8, the leachability of RI tracers from the solidified products was less than or equal to that of a high-level vitrified waste. In this review, experimental results of the melting tests were discussed in order to contribute to actual treatment of LLW in the AVRF. (author)
Scoping Analysis on Core Disruptive Accident in PGSFR (2015 Results)

Energy Technology Data Exchange (ETDEWEB)

Lee, Seung Won; Chang, Won-Pyo; Ha, Kwi-Seok; Ahn, Sang June; Kang, Seok Hun; Choi, Chi-Woong; Lee, Kwi Lim; Jeong, Jae-Ho; Kim, Jin Su; Jeong, Taekyeong [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In general, the severe accident is classified by three phases. The first phase is the initiation (pre-disassembly) phase that occurs the gradual core meltdown from accident initiation to the point of neutronic shutdown with an intact geometry. The second phase is the transition phase that happens the fuel transition from a solid to a liquid phase. Fuel and cladding can melt to form a molten pool and core can boil, then criticality conditions can recur. The third phase is the disassembly phase. In other words, this phase is Core Disruptive Accident (CDA). Power excursion is followed until the core is disassembled in this phase. In the early considerations of Liquid Metal Fast Breeder Reactor (LMFBR) energetics, the term Hypothetical Core Disruptive Accidents (HCDAs) was in common use. This was not only to connote the extremely low probability of initiation of such accidents, but also the tentative nature of our understanding of their behavior and resulting consequences. A numerical analysis is conducted to estimate the energy release, pressure behavior and core expansion behavior induced by CDA of PGSFR using CDA-ER and CDA-CEME codes. Conservatively, the calculated results of energy release and pressure behavior induced by CDA without Doppler effect in PGSFR when whole cores were melted (100 $/s) were 7.844 GJ and 4.845 GPa, respectively. With Doppler effect, the analyzed maximum energy release and pressure were 6.696 GJ and 3.449 GPa, respectively. The calculated results of the core expansion behavior during 0.015 seconds after the explosion without Doppler effect in PGSFR when whole cores were melted (100 $/s) were as follows: The total energy is calculated to be 1.87 GJ. At 0.01 s, the kinetic energy of the sodium is 1.85 GJ, while the expansion work and internal energy of the bubble are 19.7 MJ and 0.98 J, respectively. With Doppler effect, the total energy is calculated to be 1.33 GJ. At 0.01 s, the kinetic energy of the sodium is 1.31 GJ, while the expansion
Fundamental experiment on simulated molten core/concrete interaction

International Nuclear Information System (INIS)

Toda, S.; Katsumura, Y.

1994-01-01

If a complete and prolonged failure of coolant flow were to occur in a LWR or FBR, fission product decay heat would cause the fuel to overheat. If no available action to cool the fuel were taken, it would eventually melt. Ibis could lead to slumping of the molten core material and to the failure of the reactor pressure vessel and deposition of these materials into the concrete reactor cavity. Consequently, the molten core could melt and decompose the concrete. Vigorous agitation of the molten core pool by concrete decomposition gases is expected to enhance the convective heat transfer process. Besides the decomposition gases, melting concrete (slag) generated under the molten core pool will be buoyed up, and will also affect the downward heat transfer. Though, in this way, the heat transfer process across the interface is complicated by the slag and the gases evolved from the decomposed concrete, it is very important to make its process clear for the safety evaluation of nuclear reactors. Therefore, in this study, fundamental experiments were performed using simulated materials to observe the behaviors of the hot pool, slag and gases at the interface. Moreover, from the experimental observation, a correlation without empirical constants was proposed to calculate the interface heat transfer. The heat transfer across the interface would depend on thermo-physical interactions between the pool, slag and concrete which are changed by their thermal properties and interface temperature and so on. For example, the molten concrete is miscible in molten oxidic core debris, but is immiscible in metallic core debris. If a contact temperature between the molten core pool and the concrete falls below the solidus of the pool, solidification of the pool will occur. In this study, the case of immiscible slag in the pool is treated and solidification of the pool does not occur. Thus, water, paraffin and air were selected as the simulated molten core pool, concrete, and decomposition

Retention of anatomy knowledge by student radiographers

International Nuclear Information System (INIS)

Hall, A. Susanne; Durward, Brian R.

2009-01-01

Introduction: Anatomy has long been regarded as an integral part of the core curriculum. However, anecdotal evidence suggests that long-term retention of anatomy knowledge may be deficient. This study aims to evidence whether student radiographers demonstrate the same level of knowledge of anatomy after a period of time has elapsed and to correlate to approaches to learning and studying. Methodology: A repeated measures design was utilised to measure retention of anatomy knowledge for both MCQs and short-response answers to a Practical Radiographic Anatomy Examination; alpha value p < 0.05. Fifty-one students from levels 2 and 3 were retested after a time lapse of 10 and 22 months respectively. The students were not aware that their knowledge was being retested. Approaches to learning and studying were measured using the ASSIST inventory. Results: Statistical analysis found no difference in performance on MCQ assessment, in either the combined sample or levels 2 and 3 separately, from baseline to retention occasions; average retention rate being 99%. However, a statistical difference in performance on PRAE assessment was found, with level 2 experiencing a larger reduction in scores; retention rate of 67% compared to level 3 at 77%. The students perceived themselves to be principally strategic in their approach to learning and studying but no strong relationships were found when correlated to test scores. Conclusion: The student radiographers in this study demonstrated varied anatomy retention rates dependent on assessment method employed and time interval that had elapsed. It is recommended that diverse teaching and assessment strategies are adopted to encourage a deeper approach to learning and studying.
Melting method for miscellaneous radioactive solid waste and melting furnace

International Nuclear Information System (INIS)

Osaki, Toru; Furukawa, Hirofumi; Uda, Nobuyoshi; Katsurai, Kiyomichi

1998-01-01

A vessel containing miscellaneous solid wastes is inserted in a crucible having a releasable material on the inner surface, they are induction-heated from the outside of the crucible by way of low temperature heating coils to melt low melting point materials in the miscellaneous wastes within a temperature range at which the vessel does not melt. Then, they are induction-heated by way of high temperature heating coils to melt the vessel and not yet melted materials, those molten materials are cooled, solidified molten material and the releasable material are taken out, and then the crucible is used again. Then, the crucible can be used again, so that it can be applied to a large scaled melting furnace which treats wastes by a unit of drum. In addition, since the cleaning of the used crucible and the application of the releasable material can be conducted without interrupting the operation of the melting furnace, the operation cycle of the melting furnace can be shortened. (N.H.)
Patterns in new dimensionless quantities containing melting temperature, and their dependence on pressure

Directory of Open Access Journals (Sweden)

U. WALZER

1980-06-01

Full Text Available The relationships existing between melting temperature and other
macroscopic physical quantities are investigated. A new dimensionless
quantity Q(1 not containing the Grtineisen parameter proves to be suited for serving in future studies as a tool for the determination of the melting temperature in the outer core of the Earth. The pressure dependence of more general dimensionless quantities Q„ is determined analytically and, for the chemical elements, numerically, too. The patterns of various interesting dimensionless quantities are shown in the Periodic Table and compared.
Core Thermal-Hydraulic Conceptual Design for the Advanced SFR Design Concepts

International Nuclear Information System (INIS)

Cho, Chung Ho; Chang, Jin Wook; Yoo, Jae Woon; Song, Hoon; Choi, Sun Rock; Park, Won Seok; Kim, Sang Ji

2010-01-01

The Korea Atomic Energy Research Institute (KAERI) has developed the advanced SFR design concepts from 2007 to 2009 under the National longterm Nuclear R and D Program. Two types of core designs, 1,200 MWe breakeven and 600 MWe TRU burner core have been proposed and evaluated whether they meet the design requirements for the Gen IV technology goals of sustainability, safety and reliability, economics, proliferation resistance, and physical protection. In generally, the core thermal hydraulic design is performed during the conceptual design phase to efficiently extract the core thermal power by distributing the appropriate sodium coolant flow according to the power of each assembly because the conventional SFR core is composed of hundreds of ducted assemblies with hundreds of fuel rods. In carrying out the thermal and hydraulic design, special attention has to be paid to several performance parameters in order to assure proper performance and safety of fuel and core; the coolant boiling, fuel melting, structural integrity of the components, fuel-cladding eutectic melting, etc. The overall conceptual design procedure for core thermal and hydraulic conceptual design, i.e., flow grouping and peak pin temperature calculations, pressure drop calculations, steady-state and detailed sub-channel analysis is shown Figure 1. In the conceptual design phase, results of core thermal-hydraulic design for advanced design concepts, the core flow grouping, peak pin cladding mid-wall temperature, and pressure drop calculations, are summarized in this study
Probabilistic risk assessment (PRA) on the effectiveness of a core rescue system (SSN) for PWRs

International Nuclear Information System (INIS)

Petrangeli, G.; Valeri, A.

1983-01-01

Safety systems for the prevention of LWR core severe damage have recently been studied, which are based on automatic primary system depressurization and on borated water injection by low pressure accumulators. These systems have been named Core Rescue System (SSN). The present study evaluates the reduction in core melt probability brought about by the installation of a SSN system on the RSS (WASH 1400) PWR plant (Surry 1). The calculated result is a core melt probability reduction factor of about 250. Taking into account the possible effect of external or internal unknown events of negligible, yet undefined, probability it is concluded that a SSN system can make a plant ten times safer. The first part of a review report by Prof. N.C.Rasmussen, MIT, dealing with general comment, is attached
Rock Magnetic Study of IODP/ICDP Expedition 364 Site M0077A Drill Cores: Post-Impact Sediments, Impact Breccias, Melt, Granitic Basement and Dikes

Science.gov (United States)

Fucugauchi, J. U.; Perez-Cruz, L. L.; Rebolledo-Vieyra, M.; Tikoo, S.; Zylberman, W.; Lofi, J.

2017-12-01

Drilling at Site M0077 sampled post-impact sediments overlying a peak ring consisting of impact breccias, melt rock and granitoids. Here we focus on characterizing the peak ring using magnetic properties, which vary widely and depend on mineralogy, depositional and emplacement conditions and secondary alterations. Rock magnetic properties are integrated with Multi-Sensor Core Logger (MSCL) data, vertical seismic profile, physical properties, petrographic and chemical analyses and geophysical models. We measure low-field magnetic susceptibility at low- and high-frequencies, intensity and direction of natural remanent magnetization (NRM) and laboratory-induced isothermal (IRM) and anhysteretic (ARM) magnetizations, alternating-field demagnetization of NRM, IRM and NRM, susceptibility variation with temperature, anisotropy of magnetic susceptibility, hysteresis and IRM back-field demagnetization. Post-impact carbonates show low susceptibilities and NRM intensities, variable frequency-dependent susceptibilities and multivectorial remanences residing in low and high coercivity minerals. Hysteresis loops show low coercivity saturation magnetizations and variable paramagnetic mineral contents. Impact breccias (suevites) and melt rock show higher susceptibilities, low frequency-dependent susceptibilities, high NRM, ARM and IRM intensities and moderate ARM intensity/susceptibility ratios. Magnetic signal is dominated by fine-grained magnetite and titanomagnetites with PSD domain states. Melt rocks at the base of impactite section show the highest susceptibilities and remanence intensities. Basement section is characterized by low susceptibilities in the granites and higher values in the dikes, with NRM and ARM intensities increasing towards the base. The high susceptibilities and remanence intensities correlate with high seismic velocities, density and decreased porosity and electrical resistivity. Fracturing and alteration account for the reduced seismic velocities
Application of the core-concrete interaction code Wechsl to reactor case

International Nuclear Information System (INIS)

Cenerino, G.

1986-09-01

The WECHSL code, developed at Kernforschungszentrum Karlsruhe, West-Germany, is used for core melt accidents in nuclear power plants. The first calculations, considering silicate and limestone/common sand concretes of different compositions, analyze the influence of the initial mass of Zirconium in the corium and, in one case, the effect of sump water ingression on the top of the melt. Moreover, for a limestone concrete, a sensitivity study is made on the melting temperature of the concrete influencing the decomposition enthalpy. The main conclusion of that paper is that, in any case, the temperature of the melt drops rapidly from the initial temperature to a temperature level close to the solidification temperature of the metal phase in a relatively short period of time (approximately 15 minutes) and then a balance between the removed heat from the melt and heating sources inside the melt is established
An Iron-Rain Model for Core Formation on Asteroid 4 Vesta

Science.gov (United States)

Kiefer, Walter S.; Mittlefehldt, David W.

2016-01-01

Asteroid 4 Vesta is differentiated into a crust, mantle, and core, as demonstrated by studies of the eucrite and diogenite meteorites and by data from NASA's Dawn spacecraft. Most models for the differentiation and thermal evolution of Vesta assume that the metal phase completely melts within 20 degrees of the eutectic temperature, well before the onset of silicate melting. In such a model, core formation initially happens by Darcy flow, but this is an inefficient process for liquid metal and solid silicate. However, the likely chemical composition of Vesta, similar to H chondrites with perhaps some CM or CV chondrite, has 13-16 weight percent S. For such compositions, metal-sulfide melting will not be complete until a temperature of at least 1350 degrees Centigrade. The silicate solidus for Vesta's composition is between 1100 and 1150 degrees Centigrade, and thus metal and silicate melting must have substantially overlapped in time on Vesta. In this chemically and physically more likely view of Vesta's evolution, metal sulfide drops will sink by Stokes flow through the partially molten silicate magma ocean in a process that can be envisioned as "iron rain". Measurements of eucrites show that moderately siderophile elements such as Ni, Mo, and W reached chemical equilibrium between the metal and silicate phases, which is an important test for any Vesta differentiation model. The equilibration time is a function of the initial metal grain size, which we take to be 25-45 microns based on recent measurements of H6 chondrites. For these sizes and reasonable silicate magma viscosities, equilibration occurs after a fall distance of just a few meters through the magma ocean. Although metal drops may grow in size by merger with other drops, which increases their settling velocities and decreases the total core formation time, the short equilibration distance ensures that the moderately siderophile elements will reach chemical equilibrium between metal and silicate before
The parent magma of the nakhlite meteorites - Clues from melt inclusions

Science.gov (United States)

Harvey, Ralph P.; Mcsween, Harry Y., Jr.

1992-01-01

Several forms of trapped liquid found within nakhlite meteorites have been examined, including interstitial melt and magmatic inclusions within the cores of large olivine grains. Differences in the mineralogy and texture between two types of trapped melt inclusions, and between these inclusions and the mesostasis, indicate that vitrophyric inclusions are most appropriate for estimating the composition of a nakhlite parental magma in equilibrium with early-forming olivine and augite. Parent liquids were calculated from the mineralogy of large inclusions in Nakhla and Governador Valadares, using a system of mass-balance equations solved by linear regression methods. The chosen parental liquids were cosaturated in olivine and augite and had Mg/Fe values consistent with measured augite/liquid Kds. These parental magma compositions are similar to other published compositions for Nakhla, Chassigny, and Shergotty parental melts, and may correspond to a significant magma type on Mars.
Exploratory study of molten core material/concrete interactions, July 1975--March 1977

International Nuclear Information System (INIS)

Powers, D.A.; Dahlgren, D.A.; Muir, J.F.; Murfin, W.D.

1978-02-01

An experimental study of the interaction between high-temperature molten materials and structural concrete is described. The experimental efforts focused on the interaction of melts of reactor core materials weighing 12 to 200 kg at temperatures 1700 to 2800 0 C with calcareous and basaltic concrete representative of that found in existing light-water nuclear reactors. Observations concerning the rate and mode of melt penetration into concrete, the nature and generation rate of gases liberated during the interaction, and heat transfer from the melt to the concrete are described. Concrete erosion is shown to be primarily a melting process with little contribution from mechanical spallation. Water and carbon dioxide thermally released from the concrete are extensively reduced to hydrogen and carbon monoxide. Heat transfer from the melt to the concrete is shown to be dependent on gas generation rate and crucible geometry. Interpretation of results from the interaction experiments is supported by separate studies of the thermal decomposition of concretes, response of bulk concrete to intense heat fluxes (28 to 280 W/cm 2 ), and heat transfer from molten materials to decomposing solids. The experimental results are compared to assumptions made in previous analytic studies of core meltdown accidents in light-water nuclear reactors. A preliminary computer code, INTER, which models and extrapolates results of the experimental program is described. The code allows estimation of the effect of physical parameters on the nature of the melt/concrete interaction
Tin in granitic melts: The role of melting temperature and protolith composition

Science.gov (United States)

Wolf, Mathias; Romer, Rolf L.; Franz, Leander; López-Moro, Francisco Javier

2018-06-01

Granite bound tin mineralization typically is seen as the result of extreme magmatic fractionation and late exsolution of magmatic fluids. Mineralization, however, also could be obtained at considerably less fractionation if initial melts already had enhanced Sn contents. We present chemical data and results from phase diagram modeling that illustrate the dominant roles of protolith composition, melting conditions, and melt extraction/evolution for the distribution of Sn between melt and restite and, thus, the Sn content of melts. We compare the element partitioning between leucosome and restite of low-temperature and high-temperature migmatites. During low-temperature melting, trace elements partition preferentially into the restite with the possible exception of Sr, Cd, Bi, and Pb, that may be enriched in the melt. In high-temperature melts, Ga, Y, Cd, Sn, REE, Pb, Bi, and U partition preferentially into the melt whereas Sc, V, Cr, Co, Ni, Mo, and Ba stay in the restite. This contrasting behavior is attributed to the stability of trace element sequestering minerals during melt generation. In particular muscovite, biotite, titanite, and rutile act as host phases for Sn and, therefore prevent Sn enrichment in the melt as long as they are stable phases in the restite. As protolith composition controls both the mineral assemblage and modal contents of the various minerals, protolith composition eventually also controls the fertility of a rock during anatexis, restite mineralogy, and partitioning behavior of trace metals. If a particular trace element is sequestered in a phase that is stable during partial melting, the resulting melt is depleted in this element whereas the restite becomes enriched. Melt generation at high temperature may release Sn when Sn-hosts become unstable. If melt has not been lost before the breakdown of Sn-hosts, Sn contents in the melt will increase but never will be high. In contrast, if melt has been lost before the decomposition of Sn
Modelling of heat transfer between molten core and concrete with account of phase changes in the melt

International Nuclear Information System (INIS)

Petukhov, S.M.; Zemlianoukhin, V.V.

1992-01-01

The analysis of the process of heat transfer between molten corium and concrete in the case of severe accident in a PWR is performed. It is shown that Bradley's model may be improved for the case of an oxidic melt. A new model is developed and incorporated in the WECHSL-Mod2 Code. Post-test calculations of melt-concrete interaction experiments are carried out. The comparison and analysis of the experimental results and calculations are presented. (9 figures) (Author)
Comparative Study on Two Melting Simulation Methods: Melting Curve of Gold

International Nuclear Information System (INIS)

Liu Zhong-Li; Li Rui; Sun Jun-Sheng; Zhang Xiu-Lu; Cai Ling-Cang

2016-01-01

Melting simulation methods are of crucial importance to determining melting temperature of materials efficiently. A high-efficiency melting simulation method saves much simulation time and computational resources. To compare the efficiency of our newly developed shock melting (SM) method with that of the well-established two-phase (TP) method, we calculate the high-pressure melting curve of Au using the two methods based on the optimally selected interatomic potentials. Although we only use 640 atoms to determine the melting temperature of Au in the SM method, the resulting melting curve accords very well with the results from the TP method using much more atoms. Thus, this shows that a much smaller system size in SM method can still achieve a fully converged melting curve compared with the TP method, implying the robustness and efficiency of the SM method. (paper)
ELM-induced transient tungsten melting in the JET divertor

Science.gov (United States)

Coenen, J. W.; Arnoux, G.; Bazylev, B.; Matthews, G. F.; Autricque, A.; Balboa, I.; Clever, M.; Dejarnac, R.; Coffey, I.; Corre, Y.; Devaux, S.; Frassinetti, L.; Gauthier, E.; Horacek, J.; Jachmich, S.; Komm, M.; Knaup, M.; Krieger, K.; Marsen, S.; Meigs, A.; Mertens, Ph.; Pitts, R. A.; Puetterich, T.; Rack, M.; Stamp, M.; Sergienko, G.; Tamain, P.; Thompson, V.; Contributors, JET-EFDA

2015-02-01

The original goals of the JET ITER-like wall included the study of the impact of an all W divertor on plasma operation (Coenen et al 2013 Nucl. Fusion 53 073043) and fuel retention (Brezinsek et al 2013 Nucl. Fusion 53 083023). ITER has recently decided to install a full-tungsten (W) divertor from the start of operations. One of the key inputs required in support of this decision was the study of the possibility of W melting and melt splashing during transients. Damage of this type can lead to modifications of surface topology which could lead to higher disruption frequency or compromise subsequent plasma operation. Although every effort will be made to avoid leading edges, ITER plasma stored energies are sufficient that transients can drive shallow melting on the top surfaces of components. JET is able to produce ELMs large enough to allow access to transient melting in a regime of relevance to ITER. Transient W melt experiments were performed in JET using a dedicated divertor module and a sequence of IP = 3.0 MA/BT = 2.9 T H-mode pulses with an input power of PIN = 23 MW, a stored energy of ˜6 MJ and regular type I ELMs at ΔWELM = 0.3 MJ and fELM ˜ 30 Hz. By moving the outer strike point onto a dedicated leading edge in the W divertor the base temperature was raised within ˜1 s to a level allowing transient, ELM-driven melting during the subsequent 0.5 s. Such ELMs (δW ˜ 300 kJ per ELM) are comparable to mitigated ELMs expected in ITER (Pitts et al 2011 J. Nucl. Mater. 415 (Suppl.) S957-64). Although significant material losses in terms of ejections into the plasma were not observed, there is indirect evidence that some small droplets (˜80 µm) were released. Almost 1 mm (˜6 mm3) of W was moved by ˜150 ELMs within 7 subsequent discharges. The impact on the main plasma parameters was minor and no disruptions occurred. The W-melt gradually moved along the leading edge towards the high-field side, driven by j × B forces. The evaporation rate determined
Percolation blockage: A process that enables melt pond formation on first year Arctic sea ice

Science.gov (United States)

Polashenski, Chris; Golden, Kenneth M.; Perovich, Donald K.; Skyllingstad, Eric; Arnsten, Alexandra; Stwertka, Carolyn; Wright, Nicholas

2017-01-01

Melt pond formation atop Arctic sea ice is a primary control of shortwave energy balance in the Arctic Ocean. During late spring and summer, the ponds determine sea ice albedo and how much solar radiation is transmitted into the upper ocean through the sea ice. The initial formation of ponds requires that melt water be retained above sea level on the ice surface. Both theory and observations, however, show that first year sea ice is so highly porous prior to the formation of melt ponds that multiday retention of water above hydraulic equilibrium should not be possible. Here we present results of percolation experiments that identify and directly demonstrate a mechanism allowing melt pond formation. The infiltration of fresh water into the pore structure of sea ice is responsible for blocking percolation pathways with ice, sealing the ice against water percolation, and allowing water to pool above sea level. We demonstrate that this mechanism is dependent on fresh water availability, known to be predominantly from snowmelt, and ice temperature at melt onset. We argue that the blockage process has the potential to exert significant control over interannual variability in ice albedo. Finally, we suggest that incorporating the mechanism into models would enhance their physical realism. Full treatment would be complex. We provide a simple temperature threshold-based scheme that may be used to incorporate percolation blockage behavior into existing model frameworks.
Early planetesimal melting from an age of 4.5662 Gyr for differentiated meteorites

DEFF Research Database (Denmark)

Baker, J.; Bizzarro, Martin; Wittig, N.

2005-01-01

for these meteorites, however, are typically younger than age constraints for planetesimal differentiation. Such young ages indicate that the energy required to melt their parent bodies could not have come from the most likely heat source-radioactive decay of short-lived nuclides (Al and Fe) injected from a nearby...... decay could have triggered planetesimal melting. Small Mg excesses in bulk angrite samples confirm that Al decay contributed to the melting of their parent body. These results indicate that the accretion of differentiated planetesimals pre-dated that of undifferentiated planetesimals, and reveals......Long- and short-lived radioactive isotopes and their daughter products in meteorites are chronometers that can test models for Solar System formation. Differentiated meteorites come from parent bodies that were once molten and separated into metal cores and silicate mantles. Mineral ages...
Heat load imposed on reactor vessels during in-vessel retention of core melts

Energy Technology Data Exchange (ETDEWEB)

Kim, Su-Hyeon; Chung, Bum-Jin, E-mail: bjchung@khu.ac.kr

2016-11-15

Highlights: • Angular heat load on reactor vessel by natural convection of oxide pool was measured. • High Ra was achieved by using mass transfer experiments based on analogy concept. • Measured Nusselt numbers agreed reasonably with the other existing studies. • Three different types of volumetric heat sources were compared. • They didn’t affect the heat flux of the top plate but affected those of the reactor vessel. - Abstract: We measured the heat load imposed on reactor vessels by natural convection of the oxide pool in severe accidents. Based on the analogy between heat and mass transfer, mass transfer experiments were performed using a copper sulfate electroplating system. A modified Rayleigh number of the order 10{sup 14} was achieved in a small facility with a height of 0.1 m. Three different types of volumetric heat sources were compared and the average Nusselt number of the curved surface was 39% lower, whereas in the case of the top plate was 6% higher than in previous studies with a two-dimensional geometry due to the high Sc value of this study. Reliable experimental data on the angular heat flux ratios were reported compared to those of the BALI and SIGMA CP facilities in terms of fluctuations and consistency.
Further work on sodium borates as sacrificial materials for a core-catcher

International Nuclear Information System (INIS)

Dalle Donne, M.; Dorner, S.; Roth, A.; Werle, H.

1982-01-01

Sodium borates are suitable low melting point sacrificial materials for a core-catcher of a fast reactor. Concept, design and initial development work have been described previously. Here we report on the measurements of density, volumetric thermal expansion coefficients and viscosity of borax and sodium metaborate, pure and with various percentages of dissolved UO 2 . The density of these molten salts was measured with the buoyancy method in the temperature range 850 - 1300 0 C, while the viscosity was measured in the temperature range 700 - 1250 0 C with a Haake viscosity balance. Simulation experiments with low melting point materials were performed to investigate the ratio of the downward to sideward melt velocity. The results of these experiments show that this ratio is equal to 0.34 for a solid to liquid density ratio rho = 1.66. For the real borax core-catcher rho = 4 and this would correspond to a velocity ratio of about one
Olivine/melt transition metal partitioning, melt composition, and melt structure—Melt polymerization and Qn-speciation in alkaline earth silicate systems

Science.gov (United States)

Mysen, Bjorn O.

2008-10-01

The two most abundant network-modifying cations in magmatic liquids are Ca 2+ and Mg 2+. To evaluate the influence of melt structure on exchange of Ca 2+ and Mg 2+ with other geochemically important divalent cations ( m-cations) between coexisting minerals and melts, high-temperature (1470-1650 °C), ambient-pressure (0.1 MPa) forsterite/melt partitioning experiments were carried out in the system Mg 2SiO 4-CaMgSi 2O 6-SiO 2 with ⩽1 wt% m-cations (Mn 2+, Co 2+, and Ni 2+) substituting for Ca 2+ and Mg 2+. The bulk melt NBO/Si-range ( NBO/Si: nonbridging oxygen per silicon) of melt in equilibrium with forsterite was between 1.89 and 2.74. In this NBO/Si-range, the NBO/Si(Ca) (fraction of nonbridging oxygens, NBO, that form bonds with Ca 2+, Ca 2+- NBO) is linearly related to NBO/Si, whereas fraction of Mg 2+- NBO bonds is essentially independent of NBO/Si. For individual m-cations, rate of change of KD( m-Mg) with NBO/Si(Ca) for the exchange equilibrium, mmelt + Mg olivine ⇌ molivine + Mg melt, is linear. KD( m-Mg) decreases as an exponential function of increasing ionic potential, Z/ r2 ( Z: formal electrical charge, r: ionic radius—here calculated with oxygen in sixfold coordination around the divalent cations) of the m-cation. The enthalpy change of the exchange equilibrium, Δ H, decreases linearly with increasing Z/ r2 [Δ H = 261(9)-81(3)· Z/ r2 (Å -2)]. From existing information on (Ca,Mg)O-SiO 2 melt structure at ambient pressure, these relationships are understood by considering the exchange of divalent cations that form bonds with nonbridging oxygen in individual Qn-species in the melts. The negative ∂ KD( m-Mg) /∂( Z/ r2) and ∂(Δ H)/∂( Z/ r2) is because increasing Z/ r2 is because the cations forming bonds with nonbridging oxygen in increasingly depolymerized Qn-species where steric hindrance is decreasingly important. In other words, principles of ionic size/site mismatch commonly observed for trace and minor elements in crystals, also
Core size effects on safety performances of LMRs

Energy Technology Data Exchange (ETDEWEB)

Na, Byung Chan; Hahn, Do Hee [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

An oxide fuel small size core (1200 MWt) was analyzed in comparison with a large size core (3600 MWt) in order to evaluate the size effects on transient safety performances of liquid-metal reactors (LMRs). In the first part of the study, main static safety parameters (i.e., Doppler coefficient, sodium void effect, etc.) of the two cores were characterized, and the second part of the study was focused on the dynamic behavior of the cores in two representative transient events: the unprotected loss-of-flow (ULOF) and the unprotected transient overpower (UTOP). Margins to fuel melting and sodium boiling have been evaluated for these representative transients. Results show that the small core has a generally better or equivalent level of safety performances during these events. 6 refs., 4 figs., 2 tabs. (Author)

Core size effects on safety performances of LMRs

Energy Technology Data Exchange (ETDEWEB)

Na, Byung Chan; Hahn, Do Hee [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

An oxide fuel small size core (1200 MWt) was analyzed in comparison with a large size core (3600 MWt) in order to evaluate the size effects on transient safety performances of liquid-metal reactors (LMRs). In the first part of the study, main static safety parameters (i.e., Doppler coefficient, sodium void effect, etc.) of the two cores were characterized, and the second part of the study was focused on the dynamic behavior of the cores in two representative transient events: the unprotected loss-of-flow (ULOF) and the unprotected transient overpower (UTOP). Margins to fuel melting and sodium boiling have been evaluated for these representative transients. Results show that the small core has a generally better or equivalent level of safety performances during these events. 6 refs., 4 figs., 2 tabs. (Author)
Modelling of the Molten Core Concrete Interaction (MCCI)

International Nuclear Information System (INIS)

Guillaume, M.

2008-01-01

Severe accidents of nuclear power plants are very unlikely to occur, yet it is necessary to be able to predict the evolution of the accident. In some situations, heat generation due to the disintegration of fission products could lead to the melting of the core. If the molten core falls on the floor of the building, it would provoke the melting of the concrete floor. The objective of the studies is to calculate the melting rate of the concrete floor. The work presented in this report is in the continuity of the segregation phase model of Seiler and Froment. It is based on the results of the ARTEMIS experiments. Firstly, we have developed a new model to simulate the transfers within the interfacial area. The new model explains how heat is transmitted to concrete: by conduction, convection and latent heat generation. Secondly, we have modified the coupled modelling of the pool and the interfacial area. We have developed two new models: the first one is the 'liquidus model', whose main hypothesis is that there is no resistance to solute transfer between the pool and the interfacial area. The second one is 'the thermal resistance model', whose main hypothesis is that there is no solute transfer and no dissolution of the interfacial area. The second model is able to predict the evolution of the pool temperature and the melting rate in the tests 3 and 4, with the condition that the obstruction time of the interfacial area is about 10 5 s. The model is not able to explain precisely the origin of this value. The liquidus model is able to predict correctly the evolution of the pool temperature and the melting rate in the tests 2 and 6. (author) [fr
The melting mechanism in binary Pd0.25Ni0.75 nanoparticles: molecular dynamics simulations

Science.gov (United States)

Domekeli, U.; Sengul, S.; Celtek, M.; Canan, C.

2018-02-01

The melting mechanism for Pd0.25Ni0.75 alloy nanoparticles (NPs) was investigated using molecular dynamics (MD) simulations with quantum Sutton-Chen many-body potentials. NPs of six different sizes ranging from 682 to 22,242 atoms were studied to observe the effect of size on the melting point. The melting temperatures of the NPs were estimated by following the changes in both the thermodynamic and structural quantities such as the total energy, heat capacity and Lindemann index. We also used a thermodynamics model to better estimate the melting point and to check the accuracy of MD simulations. We observed that the melting points of the NPs decreased as their sizes decreased. Although the MD simulations for the bulk system yielded higher melting temperatures because of the lack of a seed for the liquid phase, the melting temperatures determined for both the bulk material and the NPs are in good agreement with those predicted from the thermodynamics model. The melting mechanism proceeds in two steps: firstly, a liquid-like shell is formed in the outer regions of the NP with increasing temperature. The thickness of the liquid-like shell increases with increasing temperature until the shell reaches a critical thickness. Then, the entire Pd-Ni NP including core-related solid-like regions melts at once.
Meltdown reactor core cooling facility

International Nuclear Information System (INIS)

Matsuoka, Tsuyoshi.

1992-01-01

The meltdown reactor core cooling facility comprises a meltdown reactor core cooling tank, a cooling water storage tank situates at a position higher than the meltdown reactor core cooling tank, an upper pipeline connecting the upper portions of the both of the tanks and a lower pipeline connecting the lower portions of them. Upon occurrence of reactor core meltdown, a high temperature meltdown reactor core is dropped on the cooling tank to partially melt the tank and form a hole, from which cooling water is flown out. Since the water source of the cooling water is the cooling water storage tank, a great amount of cooling water is further dropped and supplied and the reactor core is submerged and cooled by natural convection for a long period of time. Further, when the lump of the meltdown reactor core is small and the perforated hole of the meltdown reactor cooling tank is small, cooling water is boiled by the high temperature lump intruding into the meltdown reactor core cooling tank and blown out from the upper pipeline to the cooling water storage tank to supply cooling water from the lower pipeline to the meltdown reactor core cooling tank. Since it is constituted only with simple static facilities, the facility can be simplified to attain improvement of reliability. (N.H.)
Constitution and reaction behavior of LWR materials at core melting conditions

International Nuclear Information System (INIS)

Holleck, H.; Skokan, A.; Janzer, H.; Schlickeise, G.; Riemueller, K.; Stroemann, H.; Nold, E.; Schaefer, A.

1979-01-01

Crucible melting experiments were performed with mixtures of preoxidized corium and basaltic or limestone concrete in order to investigate the oxidation behavior of the fission products, esp. Mo and Ru, at elevated oxygen partial pressures by H 2 O and CO 2 released from concrete. - The solidification behavior of the metallic and oxide fractions of corium (A+R) and corium (E+R) in the course of the interaction with basaltic or limestone concrete was investigated by crucible experiments. -Thermoanalytical investigations were performed with concrete of different types ranging from pure basaltic to pure limestone aggregates in order to test the possibility of reactions between CaO and SiO 2 during the heating up period. (orig./RW) [de
Permeability and 3-D melt geometry in shear-induced high melt fraction conduits

Science.gov (United States)

Zhu, W.; Cordonnier, B.; Qi, C.; Kohlstedt, D. L.

2017-12-01

Observations of dunite channels in ophiolites and uranium-series disequilibria in mid-ocean ridge basalt suggest that melt transport in the upper mantle beneath mid-ocean ridges is strongly channelized. Formation of high melt fraction conduits could result from mechanical shear, pyroxene dissolution, and lithological partitioning. Deformation experiments (e.g. Holtzman et al., 2003) demonstrate that shear stress causes initially homogeneously distributed melt to segregate into an array of melt-rich bands, flanked by melt-depleted regions. At the same average melt fraction, the permeability of high melt fraction conduits could be orders of magnitude higher than that of their homogenous counterparts. However, it is difficult to determine the permeability of melt-rich bands. Using X-ray synchrotron microtomography, we obtained high-resolution images of 3-dimensional (3-D) melt distribution in a partially molten rock containing shear-induced high melt fraction conduits. Sample CQ0705, an olivine-alkali basalt aggregate with a nominal melt fraction of 4%, was deformed in torsion at a temperature of 1473 K and a confining pressure of 300 MPa to a shear strain of 13.3. A sub-volume of CQ0705 encompassing 3-4 melt-rich bands was imaged. Microtomography data were reduced to binary form so that solid olivine is distinguishable from basalt glass. At a spatial resolution of 160 nm, the 3-D images reveal the shape and connectedness of melt pockets in the melt-rich bands. Thin melt channels formed at grain edges are connected at large melt nodes at grain corners. Initial data analysis shows a clear preferred orientation of melt pockets alignment subparallel to the melt-rich band. We use the experimentally determined geometrical parameters of melt topology to create a digital rock with identical 3-D microstructures. Stokes flow simulations are conducted on the digital rock to obtain the permeability tensor. Using this digital rock physics approach, we determine how deformation
Chicxulub Impact Crater and Yucatan Carbonate Platform - Stratigraphy and Petrography of PEMEX Borehole Cores

Science.gov (United States)

Gutierrez-Cirlos, A. G.; Perez-Drago, G.; Perez-Cruz, L.; Urrutia-Fucugauchi, J.

2008-12-01

Chicxulub impact crater is the best preserved of the three large multi-ring structures documented in the terrestrial record. Chicxulub, formed 65 Ma ago, is associated with the Cretaceous/Tertiary (K/T) boundary layer and the impact related to the organism extinctions and events marking the boundary. The crater is buried under Tertiary sediments in the Yucatan carbonate platform in the southern Gulf of Mexico. The structure was initially recognized from gravity and magnetic anomalies in the PEMEX exploration surveys of the northwestern Yucatan peninsula. The exploration program included eight deep boreholes completed from 1952 through the 1970s. The investigations showing Chicxulub as a large complex impact crater formed at the K/T boundary have relayed on the PEMEX decades-long exploration program. However, despite frequent use of PEMEX information and core samples, significant parts of the database and cores remain to be evaluated, analyzed and incorporated with results from recent efforts. Access to PEMEX Core Repository has permitted to study the cores and collect new samples from some of the boreholes. We analyzed cores from Yucatan-6, Chicxulub-1, Sacapuc-1, Ticul-1, Yucatan-1 and Yucatan-4 boreholes to make new detailed stratigraphic correlations and petrographic characterization, using information from PEMEX database and the recent studies. In C-1 cores, breccias show 4-8 cm clasts of fine grained altered melt dispersed in a medium to coarse grained matrix composed of pyroxene and feldspar with little macroscopic alteration. Clasts contain 0.2 to 0.1 cm fragments of silicate material (basement) that show variable degrees of digestion. Melt samples from C-1 N10 comes from interval 1,393-1,394 m, and show a fine-to-medium grained coherent microcrystalline groundmass. Melt and breccias in Y-6 extend from about 1,100 m to more than 1,400 m. Sequence is well sorted, with an apparent gradation in both the lithic and melt clasts. In this presentation we report on
Differentiation of Asteroid 4 Vesta: Core Formation by Iron Rain in a Silicate Magma Ocean

Science.gov (United States)

Kiefer, Walter S.; Mittlefehldt, David W.

2017-01-01

Geochemical observations of the eucrite and diogenite meteorites, together with observations made by NASA's Dawn spacecraft while orbiting asteroid 4 Vesta, suggest that Vesta resembles H chondrites in bulk chemical composition, possible with about 25 percent of a CM-chondrite like composition added in. For this model, the core is 15 percent by mass (or 8 percent by volume) of the asteroid, with a composition of 73.7 percent by weight Fe, 16.0 percent by weight S, and 10.3 percent by weight Ni. The abundances of moderately siderophile elements (Ni, Co, Mo, W, and P) in eucrites require that essentially all of the metallic phase in Vesta segregated to form a core prior to eucrite solidification. The combination of the melting phase relationships for the silicate and metal phases, together with the moderately siderophile element concentrations together require that complete melting of the metal phase occurred (temperature is greater than1350 degrees Centigrade), along with substantial (greater than 40 percent) melting of the silicate material. Thus, core formation on Vesta occurs as iron rain sinking through a silicate magma ocean.
Designing Incentives for Marine Corps Cyber Workforce Retention

Science.gov (United States)

2014-12-01

the workplace. In 2005, Basset-Jones and Lloyd examined the relevance of Herzberg’s (1959) two factor theory in the current work environment...Herzberg’s (1959) two factor theory of intrinsic factors with several other empirical studies which identified extrinsic factors as key motivational...answering two core research questions:  What are the critical retention issues that face the Marine Corps’ cyber workforce and what are the factors
Considering clay rock heterogeneity in radionuclide retention

International Nuclear Information System (INIS)

Grambow, B.; Montavon, G.; Tournassat, C.; Giffaut, E.; Altmann, S.

2010-01-01

Document available in extended abstract form only. The Callovo-Oxfordian clay rock formation has a strong retention capacity for radionuclides, a favorable condition for the implementation of a nuclear waste repository. Principal retaining minerals are illite, and inter-stratified illite/smectite (I/S). Radionuclide retention has been studied on illite, illite/smectite and on clay rock obtained from different locations and data for retention on bentonite (80% smectite) are available. Sorption depends on the type of mineral, composition of mineralogical assemblages, individual mineral ion exchange capacities, ion distribution on exchange sites, specific surface areas, surface site types and densities for surface complexation as well as on water/rock ratios, temperature etc. As a consequence of mineralogical and textural variations, radionuclide retention properties are expected to vary with depth in the Callovo-Oxfordian formation. Using a simple additivity approach for the case of sorption of Cs and Ni it is shown that models and databases for illite and bentonite can be used to describe sorption in heterogeneous clay rock systems. A surface complexation/ion-exchange model as proposed by Bradbury and Baeyens without electrostatic contributions, was used directly as far as acid base properties are concerned but was modified with respect to sorption constants, in order to describe Na-, Ca, and Cs montmorillonite and bentonite MX-80 with a single set of surface complexation constants and also to account for carbonate and sulphate concentrations in groundwater. The model is integrated into the geochemical code PHREEQC considering dissolution/ precipitation/solubility constraints of accessory minerals (calcite, illite, celestite, quartz). Site densities for surface complexation and ion exchange are derived from the mass fractions of illite and of smectite in illite/smectite obtained from an overall fit of measured CEC data from all samples of the EST205 drill core
Bounding analysis of containment of high pressure melt ejection in advanced light water reactors

International Nuclear Information System (INIS)

Additon, S.L.; Fontana, M.H.; Carter, J.C.

1990-01-01

This paper reports on the loadings on containment due to direct containment heating (DCH) as a result of high pressure melt ejection (HPME) in advanced light water reactors (ALWR) which were estimated using conservative, bounding analyses. The purpose of the analyses was to scope the magnitude of the possible loadings and to indicate the performance needed from potential mitigation methods, such as a cavity configuration that limits energy transfer to the upper containment volume. Analyses were performed for three cases which examined the effect of availability of high pressure reactor coolant system water at the time of reactor vessel melt through and the effect of preflooding of the reactor cavity. The amount of core ejected from the vessel was varied from 100% to 0% for all cases. Results indicate that all amounts of core debris dispersal could be accommodated by the containment for the case where the reactor cavity was preflooded. For the worst case, all the energy from in-vessel hydrogen generation and combustion plus that from 45% of the entire molten core would be required to equilibrate with the containment upper volume in order to reach containment failure pressure
Extreme incompatibility of helium during mantle melting: Evidence from undegassed mid-ocean ridge basalts

Science.gov (United States)

Graham, David W.; Michael, Peter J.; Shea, Thomas

2016-11-01

We report total helium concentrations (vesicles + glass) for a suite of thirteen ultradepleted mid-ocean ridge basalts (UD-MORBs) that were previously studied for volatile contents (CO2, H2O) plus major and trace elements. The selected basalts are undersaturated in CO2 + H2O at their depths of eruption and represent rare cases of undegassed MORBs. Sample localities from the Atlantic (2), Indian (1) and Pacific (7) Oceans collectively show excellent linear correlations (r2 = 0.75- 0.92) between the concentrations of helium and the highly incompatible elements C, K, Rb, Ba, Nb, Th and U. Three basalts from Gakkel Ridge in the Arctic were also studied but show anomalous behavior marked by excess lithophile trace element abundances. In the Atlantic-Pacific-Indian suite, incompatible element concentrations vary by factors of 3-4.3, while helium concentration varies by a factor of 13. The strong correlations between the concentrations of helium and incompatible elements are explained by helium behavior as the most incompatible element during mantle melting. Partial melting of an ultradepleted mantle source, formed as a residue of earlier melt extraction, accounts for the observed concentrations. The earlier melting event involved removal of a small degree melt (∼1%) at low but non-zero porosity (0.01-0.5%), leading to a small amount of melt retention that strongly leveraged the incompatible element budget of the ultradepleted mantle source. Equilibrium melting models that produce the range of trace element and helium concentrations from this source require a bulk solid/melt distribution coefficient for helium that is lower than that for other incompatible elements by about a factor of ten. Alternatively, the bulk solid/melt distribution coefficient for helium could be similar to or even larger than that for other incompatible elements, but the much larger diffusivity of helium in peridotite leads to its more effective incompatibility and efficient extraction from a
FARO tests corium-melt cooling in water pool: Roles of melt superheat and sintering in sediment

Energy Technology Data Exchange (ETDEWEB)

Hwang, Gisuk [Department of Mechanical Engineering, Wichita State University, Wichita, KS 67260 (United States); Kaviany, Massoud [Department of Mechanical Engineering, University of Michigan, Ann Arbor, MI 48109 (United States); Division of Advance Nuclear Engineering, POSTECH, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Moriyama, Kiyofumi [Division of Advance Nuclear Engineering, POSTECH, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr [Division of Advance Nuclear Engineering, POSTECH, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Hwang, Byoungcheol; Lee, Mooneon; Kim, Eunho; Park, Jin Ho [Division of Advance Nuclear Engineering, POSTECH, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Nasersharifi, Yahya [Department of Mechanical Engineering, Wichita State University, Wichita, KS 67260 (United States)

2016-08-15

Highlights: • The numerical approach for FARO experimental data is suggested. • The cooling mechanism of ex-vessel corium is suggested. • The predicted minimum pool depth for no cake formation is suggested. - Abstract: The FARO tests have aimed at understanding an important severe accident mitigation action in a light water reactor when the accident progresses from the reactor pressure vessel boundary. These tests have aimed to measure the coolability of a molten core material (corium) gravity dispersed as jet into a water pool, quantifying the loose particle diameter distribution and fraction converted to cake under range of initial melt superheat and pool temperature and depth. Under complete hydrodynamic breakup of corium and consequent sedimentation in the pool, the initially superheated corium can result in debris bed consisting of discrete solid particles (loose debris) and/or a solid cake at the bottom of the pool. The success of the debris bed coolability requires cooling of the cake, and this is controlled by the large internal resistance. We postulate that the corium cake forms when there is a remelting part in the sediment. We show that even though a solid shell forms around the melt particles transiting in the water pool due to film-boiling heat transfer, the superheated melt allows remelting of the large particles in the sediment (depending on the water temperature and the transit time) using the COOLAP (Coolability Analysis with Parametric fuel-cooant interaction models) code. With this remelting and its liquid-phase sintering of the non-remelted particles, we predict the fraction of the melt particles converting to a cake through liquid sintering. Our predictions are in good agreement with the existing results of the FARO experiments. We address only those experiments with pool depths sufficient/exceeding the length required for complete breakup of the molten jet. Our analysis of the fate of molten corium aimed at devising the effective
Analysis of hypothetical nuclear excursions in the external core retention system

International Nuclear Information System (INIS)

Froehlich, R.; Kussmaul, G.; Schmuck, P.

1976-01-01

The core catcher system of the SNR 300 is outside the reactor tank. The probability of recriticality phenomena is reduced by its design, but the licensing procedures still call for the analysis of strong recriticality phenomena in the core catcher system outside the reactor tank in order to achieve a better understanding of the possible physical effects and to get to know the safety limits of the system. For their theoretical investigations, the authors used a two-partner model as presented in fig. 1. At the bottom of the core catcher - which consists of depleted UO 2 - there is a fuel cylinder. Another fuel cylinder (with the same axis) is dropped from a height of 250 cm. The two cylindrical masses are immersed in sodium, but a free fall is assumed since the possibility cannot be excluded that the reactor bottom may be empty or only partially filled with sodium. It was found that under these conditions the strongest excursions may be expected in those cases where prompt criticality does not occur until just before the two partners meet. (orig./AK) [de
An Experimental Investigation on APR1400 Penetration Weld Failure by Metallic Melt

International Nuclear Information System (INIS)

An, Sang Mo; Ha, Kwang Soon; Kim, Hwan Yeol

2014-01-01

The penetrations are considered as the most vulnerable parts with respect to the reactor vessel failure when a core melt severe accident occurs and the corium reaches the lower head. Penetration tube failure modes can be divided into two categories; tube ejection out of the vessel lower head and rupture of the penetration tube outside the vessel. Tube ejection begins with degrading the penetration tube weld strength to zero as the weld is exposed to temperatures as high as the weld melting temperature, which is called weld failure, and then overcoming any binding force in the hole in the vessel wall that results from differential thermal expansion of the tube and vessel wall. Tube rupture assumes that the debris bed has melted the instrument tube inside the reactor and melt migrates down into the tube to a location outside the vessel wall where a pressure rupture can occur, thus breaching the pressure boundary. In the present paper, we have a focus on the tube ejection failure mode, specifically on the APR1400 weld failure by direct contact with a metallic melt. The objective is to investigate experimentally the ablation kinetics of an APR1400 penetration weld during the interactions with a metallic melt and to suggest the modification of the existing weld failure model. This paper involves the interaction experiments of two different metallic melts (metallic corium and stainless steel melts) with a weld specimen, and rough estimation of weld failure time. The interaction experiments between the metallic melts and an APR1400 penetration weld were performed to investigate the ablation kinetics of the penetration weld. Metallic corium and stainless steel melts were generated using an induction heating technique and interacted with a penetration weld specimen. The ablation rate of the weld specimen showed a range from 0.109 to 0..244 mm/s and thus the APR1400 penetration weld was estimated to be failed at hundreds of times after the interaction with the melt
Observations of brine plumes below melting Arctic sea ice

Directory of Open Access Journals (Sweden)

A. K. Peterson

2018-02-01

Full Text Available In sea ice, interconnected pockets and channels of brine are surrounded by fresh ice. Over time, brine is lost by gravity drainage and flushing. The timing of salt release and its interaction with the underlying water can impact subsequent sea ice melt. Turbulence measurements 1 m below melting sea ice north of Svalbard reveal anticorrelated heat and salt fluxes. From the observations, 131 salty plumes descending from the warm sea ice are identified, confirming previous observations from a Svalbard fjord. The plumes are likely triggered by oceanic heat through bottom melt. Calculated over a composite plume, oceanic heat and salt fluxes during the plumes account for 6 and 9 % of the total fluxes, respectively, while only lasting in total 0.5 % of the time. The observed salt flux accumulates to 7.6 kg m−2, indicating nearly full desalination of the ice. Bulk salinity reduction between two nearby ice cores agrees with accumulated salt fluxes to within a factor of 2. The increasing fraction of younger, more saline ice in the Arctic suggests an increase in desalination processes with the transition to the new Arctic.
Observations of brine plumes below melting Arctic sea ice

Science.gov (United States)

Peterson, Algot K.

2018-02-01

In sea ice, interconnected pockets and channels of brine are surrounded by fresh ice. Over time, brine is lost by gravity drainage and flushing. The timing of salt release and its interaction with the underlying water can impact subsequent sea ice melt. Turbulence measurements 1 m below melting sea ice north of Svalbard reveal anticorrelated heat and salt fluxes. From the observations, 131 salty plumes descending from the warm sea ice are identified, confirming previous observations from a Svalbard fjord. The plumes are likely triggered by oceanic heat through bottom melt. Calculated over a composite plume, oceanic heat and salt fluxes during the plumes account for 6 and 9 % of the total fluxes, respectively, while only lasting in total 0.5 % of the time. The observed salt flux accumulates to 7.6 kg m-2, indicating nearly full desalination of the ice. Bulk salinity reduction between two nearby ice cores agrees with accumulated salt fluxes to within a factor of 2. The increasing fraction of younger, more saline ice in the Arctic suggests an increase in desalination processes with the transition to the new Arctic.
Amphibole and felsic veins from the gabbroic oceanic core complex of Atlantis Bank (Southwest Indian Ridge, IODP Hole U1473A): when the fluids meets the melts

Science.gov (United States)

Sanfilippo, A.; Tribuzio, R.; Antonicelli, M.; Zanetti, A.

2017-12-01

We present a petrological/geochemical investigation of brown amphibole and felsic veins drilled during IODP 360 expedition at Atlantis Bank, a gabbroic oceanic core complex from Southwest Indian Ridge. The main purpose of this study is to unravel the role of seawater and magmatic components in the origin of these veins. Brown amphibole veins were collected at 90-170 mbsf. These veins typically include minor modal amounts of plagioclase and are associated with alteration halos made up of brown amphibole and whitish milky plagioclase in host gabbros. Two sets of late magmatic felsic veins, which mostly consist of plagioclase and minor brown amphibole, were selected. Amphibole-plagioclase geothermometry (Holland and Blundy, 1994) documents that crystallization of brown amphibole and felsic veins occurred in the 850-700 °C interval. In the brown amphibole veins, amphibole and plagioclase have relatively low concentrations of incompatible trace elements and significant Cl (0.2-0.3 wt%). The development of these veins at near surface levels is therefore attributed to seawater-derived fluids migrating downward through cracks developing in the exhuming gabbro. To explain the high temperature estimates for the development of these shallow veins, however, the seawater-derived fluids must have interacted not only with the gabbros, but also with a high temperature magmatic component. This petrogenetic hypothesis is consistent with oxygen and hydrogen isotopic compositions of amphiboles from shallow veins in adjacent Hole 735B gabbros (Alt and Bach, 2006). Trace element compositions of amphibole and plagioclase from the felsic veins show formation by silicate melts rich in incompatible elements. In addition, Cl concentrations in amphibole from the felsic veins are low, thereby indicating that the melts feeding these veins had low or no seawater component. We cautiously propose that: (i) the felsic veins were generated by SiO2-rich melts residual after crystallization of Fe
The melting curve of MgSiO3 perovskite from molecular dynamics simulation

International Nuclear Information System (INIS)

Liu Zijiang; Zhang Cairong; Sun Xiaowei; Song Ting; Chu Yandong; Hu Jianbo

2011-01-01

The high-pressure melting curve of MgSiO 3 perovskite is simulated by using the constant temperature and pressure molecular dynamics method combined with effective pair potentials. The simulated structural properties of MgSiO 3 perovskite at ambient conditions reproduce the experiments and agree well with other theoretical works. The calculated equation of state is very successful in reproducing accurately the recent experimental data over wide pressure ranges. The predicted high-pressure melting curve is in good agreement with the recent experimental and the latest theoretical ones, and the melting curve up to the core-mantle boundary pressure, being very steep at lower pressures, rapidly flattens on increasing pressure. The present results also suggest the validity of the experimental data of Zerr and Boehler (1993 Science 262 553) and Shen and Lazor (1995 J. Geophys. Res. 100 17699).
Size-dependent melting modes and behaviors of Ag nanoparticles: a molecular dynamics study

Science.gov (United States)

Liang, Tianshou; Zhou, Dejian; Wu, Zhaohua; Shi, Pengpeng

2017-12-01

The size-dependent melting behaviors and mechanisms of Ag nanoparticles (NPs) with diameters of 3.5-16 nm were investigated by molecular dynamics (MD). Two distinct melting modes, non-premelting and premelting with transition ranges of about 7-8 nm, for Ag NPs were demonstrated via the evolution of distribution and transition of atomic physical states during annealing. The small Ag NPs (3.5-7 nm) melt abruptly without a stable liquid shell before the melting point, which is characterized as non-premelting. A solid-solid crystal transformation is conducted through the migration of adatoms on the surface of Ag NPs with diameters of 3.5-6 nm before the initial melting, which is mainly responsible for slightly increasing the melting point of Ag NPs. On the other hand, surface premelting of Ag NPs with diameters of 8-16 nm propagates from the outer shell to the inner core with initial anisotropy and late isotropy as the temperature increases, and the close-packed facets {111} melt by a side-consumed way which is responsible for facets {111} melting in advance relative to the crystallographic plane {111}. Once a stable liquid shell is formed, its size-independent minimum thickness is obtained, and a three-layer structure of atomic physical states is set up. Lastly, the theory of point defect-pair (vacancy-interstitial) severing as the mechanism of formation and movement of the solid-liquid interface was also confirmed. Our study provides a basic understanding and theoretical guidance for the research, production and application of Ag NPs.

Heat Transfer Analysis of the European Pressurized Water Reactor (EPR) Core Catcher Test Facility Volley

Energy Technology Data Exchange (ETDEWEB)

Pikkarainen, Mika; Laine, Jani; Purhonen, Heikki; Kyrki-Rajamaeki, Riitta [Lappeenranta University of Technology, P.O. 20 53851 Lappeenranta (Finland); Sairanen, Risto [Radiation and Nuclear Safety Authority, P.O. 14 00881 Helsinki (Finland)

2008-07-01

The EPR is designed to cope with severe accidents, involving core meltdown. A specific melt spreading area has been designed within the containment. This core catcher will be flooded by water, which transfers the decay heat to the containment heat removal system. To improve cooling, horizontal flow channels made of cast iron are located also below the core catcher. STUK, the radiation and nuclear safety authority in Finland, wanted an independent study of the functionality of the core catcher design. Effect of the presence of insulation material and boric acid in the cooling water was to be studied, as well as the general behavior of the system in different phases of the flooding of the core melt spreading area. To verify the function of the core catcher design, a scaled down test facility was built at Lappeenranta University of Technology. Since there are some physical restrictions of a test facility computational tools were applied especially for the tests where steady state conditions could not be reached without endangering the integrity of the test facility. This paper introduces the Volley test facility, computational simulations and compares them with the test results. Simulated temperatures of those Volley tests, which could be run until steady state conditions, are very close to the measured temperatures. It can be concluded also, that the temperatures are evidently below the cast iron melting point with heat fluxes used in the tests, if there is a small flow inside the cooling channels or even in case when only a few adjacent cooling channels are totally dry. (authors)
Heat Transfer Analysis of the European Pressurized Water Reactor (EPR) Core Catcher Test Facility Volley

International Nuclear Information System (INIS)

Pikkarainen, Mika; Laine, Jani; Purhonen, Heikki; Kyrki-Rajamaeki, Riitta; Sairanen, Risto

2008-01-01

The EPR is designed to cope with severe accidents, involving core meltdown. A specific melt spreading area has been designed within the containment. This core catcher will be flooded by water, which transfers the decay heat to the containment heat removal system. To improve cooling, horizontal flow channels made of cast iron are located also below the core catcher. STUK, the radiation and nuclear safety authority in Finland, wanted an independent study of the functionality of the core catcher design. Effect of the presence of insulation material and boric acid in the cooling water was to be studied, as well as the general behavior of the system in different phases of the flooding of the core melt spreading area. To verify the function of the core catcher design, a scaled down test facility was built at Lappeenranta University of Technology. Since there are some physical restrictions of a test facility computational tools were applied especially for the tests where steady state conditions could not be reached without endangering the integrity of the test facility. This paper introduces the Volley test facility, computational simulations and compares them with the test results. Simulated temperatures of those Volley tests, which could be run until steady state conditions, are very close to the measured temperatures. It can be concluded also, that the temperatures are evidently below the cast iron melting point with heat fluxes used in the tests, if there is a small flow inside the cooling channels or even in case when only a few adjacent cooling channels are totally dry. (authors)
ELM-induced transient tungsten melting in the JET divertor

International Nuclear Information System (INIS)

Coenen, J.W.; Clever, M.; Knaup, M.; Arnoux, G.; Matthews, G.F.; Balboa, I.; Meigs, A.; Bazylev, B.; Autricque, A.; Dejarnac, R.; Horacek, J.; Komm, M.; Coffey, I.; Corre, Y.; Gauthier, E.; Devaux, S.; Krieger, K.; Frassinetti, L.; Jachmich, S.; Marsen, S.

2015-01-01

The original goals of the JET ITER-like wall included the study of the impact of an all W divertor on plasma operation (Coenen et al 2013 Nucl. Fusion 53 073043) and fuel retention (Brezinsek et al 2013 Nucl. Fusion 53 083023). ITER has recently decided to install a full-tungsten (W) divertor from the start of operations. One of the key inputs required in support of this decision was the study of the possibility of W melting and melt splashing during transients. Damage of this type can lead to modifications of surface topology which could lead to higher disruption frequency or compromise subsequent plasma operation. Although every effort will be made to avoid leading edges, ITER plasma stored energies are sufficient that transients can drive shallow melting on the top surfaces of components. JET is able to produce ELMs large enough to allow access to transient melting in a regime of relevance to ITER. Transient W melt experiments were performed in JET using a dedicated divertor module and a sequence of I P = 3.0 MA/B T = 2.9 T H-mode pulses with an input power of P IN = 23 MW, a stored energy of ∼6 MJ and regular type I ELMs at ΔW ELM = 0.3 MJ and f ELM ∼ 30 Hz. By moving the outer strike point onto a dedicated leading edge in the W divertor the base temperature was raised within ∼1 s to a level allowing transient, ELM-driven melting during the subsequent 0.5 s. Such ELMs (δW ∼ 300 kJ per ELM) are comparable to mitigated ELMs expected in ITER (Pitts et al 2011 J. Nucl. Mater. 415 (Suppl.) S957–64). Although significant material losses in terms of ejections into the plasma were not observed, there is indirect evidence that some small droplets (∼80 µm) were released. Almost 1 mm (∼6 mm 3 ) of W was moved by ∼150 ELMs within 7 subsequent discharges. The impact on the main plasma parameters was minor and no disruptions occurred. The W-melt gradually moved along the leading edge towards the high-field side, driven by j
Enhancing retention of partial dentures using elastomeric retention rings

Directory of Open Access Journals (Sweden)

Kakkirala Revathi

2015-01-01

Full Text Available This report presents an alternative method for the retention of partial dentures that relies on the engagement of tooth undercuts by a lining material. The lab procedures are also presented. A new maxillary and mandibular acrylic partial dentures were fabricated using elastomeric retention technique for a partially dentate patient. A partially dentate man reported difficulty in retaining his upper removable partial denture (RPD. The maxillary RPD was designed utilizing elastomeric retention technique. During follow-up, it was necessary to replace the retention rings due to wear. The replacement of the retention rings, in this case, was done through a chairside reline technique. Elastomeric retention technique provides exceptionally good retention can be indicated to stabilize, cushion, splint periodontally involved teeth, no enough undercut for clasps, eliminate extractions, single or isolated teeth.
Assessment of core protection and monitoring systems for an advanced reactor SMART

International Nuclear Information System (INIS)

In, Wang Kee; Hwang, Dae Hyun; Yoo, Yeon Jong; Zee, Sung Qunn

2002-01-01

Analogue and digital core protection/monitoring systems were assessed for the implementation in an advanced reactor. The core thermal margins to nuclear fuel design limits (departure from nucleate boiling and fuel centerline melting) were estimated using the design data for a commercial pressurized water reactor and an advanced reactor. The digital protection system resulted in a greater power margin to the fuel centerline melting by at least 30% of rated power for both commercial and advanced reactors. The DNB margin with the digital system is also higher than that for the analogue system by 8 and 12.1% of rated power for commercial and advanced reactors, respectively. The margin gain with the digital system is largely due to the on-line calculations of DNB ratio and peak local power density from the live sensor signals. The digital core protection and monitoring systems are, therefore, believed to be more appropriate for the advanced reactor
Melting behaviour of gold-platinum nanoalloy clusters by molecular dynamics simulations

Energy Technology Data Exchange (ETDEWEB)

Ong, Yee Pin; Yoon, Tiem Leong [School of Physics, Universiti Sains Malaysia, 11800 USM, Penang (Malaysia); Lim, Thong Leng [Faculty of Engineering and Technology, Multimedia University, Melaka Campus, 75450 Melaka (Malaysia)

2015-04-24

The melting behavior of bimetallic gold-platinum nanoclusters is studied by applying Brownian-type isothermal molecular dynamics (MD) simulation, a program modified from the cubic coupling scheme (CCS). The process begins with the ground-state structures obtained from global minimum search algorithm and proceeds with the investigation of the effect of temperature on the thermal properties of gold-platinum nanoalloy clusters. N-body Gupta potential has been employed in order to account for the interactions between gold and platinum atoms. The ground states of the nanoalloy clusters, which are core-shell segregated, are heated until they become thermally segregated. The detailed melting mechanism of the nanoalloy clusters is studied via this approach to provide insight into the thermal stability of the nanoalloy clusters.
Application of metal oxide refractories for melting and casting reactive metals

International Nuclear Information System (INIS)

Jessen, N.C. Jr.; Holcombe, C.E. Jr.; Townsend, A.B.

1979-01-01

Extensive investigations have been conducted to develop metal oxide refractories for containment of molten uranium and uranium alloys. Since uranium and uranium alloys are readily susceptable to the formation of complex oxides, carbides, nitrides, intermetallic compounds, and suboxide reactions, severe problems exist for the production of quality castings. These contamination reactions are dependent on temperature, pressure, and molten metal interfacial reactions. The need for high purity metals to meet specification repeatedly has resulted in the development of improved metal oxide refractories and sophisticated furnace controls. Applications of Y 2 O 3 for use as a crucible and mold coating, precision molds and cores, and high temperature castable ceramics are discussed. Experimental results on melt impurity levels, thermal controls during melting, surface interactions and casting quality are presented
THE IMPACT OF BENEFITS AND SERVICES ON MANAGER ENGAGEMENT AND MANAGER RETENTION IN TOURISM INDUSTRY: ENHANCED BY THE EFFECT OF MANAGEMENT LEVEL

OpenAIRE

ALDATMAZ, Ipek; AYKAÇ, Cansu; DICLE, Ülkü

2016-01-01

Manager engagement and retention are vital to the success and organizational performance of manyservice sector organisations. Maintaining manager retention is a major challenge that many hotelenterprises face today. It is critical that organizations give greater importance to manager engagement,motivation and retention and therefore establish an efficient benefits and services strategy for retainingthese core managers for the persistence and achievement of the organization. Employee motivatio...
Investigating Planetesimal Evolution by Experiments with Fe-Ni Metallic Melts: Light Element Composition Effects on Trace Element Partitioning Behavior

Science.gov (United States)

Chabot, N. L.

2017-12-01

As planetesimals were heated up in the early Solar System, the formation of Fe-Ni metallic melts was a common occurrence. During planetesimal differentiation, the denser Fe-Ni metallic melts separated from the less dense silicate components, though some meteorites suggest that their parent bodies only experienced partial differentiation. If the Fe-Ni metallic melts did form a central metallic core, the core eventually crystallized to a solid, some of which we sample as iron meteorites. In all of these planetesimal evolution processes, the composition of the Fe-Ni metallic melt influenced the process and the resulting trace element chemical signatures. In particular, the metallic melt's "light element" composition, those elements present in the metallic melt in a significant concentration but with lower atomic masses than Fe, can strongly affect trace element partitioning. Experimental studies have provided critical data to determine the effects of light elements in Fe-Ni metallic melts on trace element partitioning behavior. Here I focus on combining numerous experimental results to identify trace elements that provide unique insight into constraining the light element composition of early Solar System Fe-Ni metallic melts. Experimental studies have been conducted at 1 atm in a variety of Fe-Ni systems to investigate the effects of light elements on trace element partitioning behavior. A frequent experimental examination of the effects of light elements in metallic systems involves producing run products with coexisting solid metal and liquid metal phases. Such solid-metal-liquid-metal experiments have been conducted in the Fe-Ni binary system as well as Fe-Ni systems with S, P, and C. Experiments with O-bearing or Si-bearing Fe-Ni metallic melts do not lend themselves to experiments with coexisting solid metal and liquid metal phases, due to the phase diagrams of these elements, but experiments with two immiscible Fe-Ni metallic melts have provided insight into
Rapakivi texture formation via disequilibrium melting in a contact partial melt zone, Antarctica

Science.gov (United States)

Currier, R. M.

2017-12-01

In the McMurdo Dry Valleys of Antarctica, a Jurassic aged dolerite sill induced partial melting of granite in the shallow crust. The melt zone can be traced in full, from high degrees of melting (>60%) along the dolerite contact, to no apparent signs of melting, 10s of meters above the contact. Within this melt zone, the well-known rapakivi texture is found, arrested in various stages of development. High above the contact, and at low degrees of melting, K-feldspar crystals are slightly rounded and unmantled. In the lower half of the melt zone, mantles of cellular textured plagioclase appear on K-feldspar, and thicken towards the contact heat source. At the highest degrees of melting, cellular-textured plagioclase completely replaces restitic K-feldspar. Because of the complete exposure and intact context, the leading models of rapakivi texture formation can be tested against this system. The previously proposed mechanisms of subisothermal decompression, magma-mixing, and hydrothermal exsolution all fail to adequately describe rapakivi generation in this melt zone. Preferred here is a closed system model that invokes the production of a heterogeneous, disequilibrium melt through rapid heating, followed by calcium and sodium rich melt reacting in a peritectic fashion with restitic K-feldspar crystals. This peritectic reaction results in the production of plagioclase of andesine-oligoclase composition—which is consistent with not just mantles in the melt zone, but globally as well. The thickness of the mantle is diffusion limited, and thus a measure of the diffusive length scale of sodium and calcium over the time scale of melting. Thermal modeling provides a time scale of melting that is consistent with the thickness of observed mantles. Lastly, the distribution of mantled feldspars is highly ordered in this melt zone, but if it were mobilized and homogenized—mixing together cellular plagioclase, mantled feldspars, and unmantled feldspars—the result would be
Density Determination of Metallic Melts from Diffuse X-Ray Scattering

Science.gov (United States)

Brauser, N.; Davis, A.; Greenberg, E.; Prakapenka, V. B.; Campbell, A.

2017-12-01

Liquids comprise several important structural components of the deep Earth, for example, the present outer core and a hypothesized magma ocean early in Earth history. However, the physical properties of the constituent materials of these structures at high pressures and temperatures are less well constrained than their crystalline counterparts. Determination of the physical properties of these liquids can inform geophysical models of the composition and structure of the Earth, but methods for studying the physical properties of liquids at high pressure and temperatures are underdeveloped. One proposed method for direct determination of density of a melt requires analysis of the diffuse scattered X-ray signal of the liquid. Among the challenges to applying this technique to high-pressure melts within a laser heated diamond anvil cell are the low signal-to-noise ratio and overlapping diffraction peaks from the crystalline components of the sample assembly interfering with the diffuse scattering from the liquid. Recent advances in instrumentation at synchrotron X-ray sources have made this method more accessible for determination of density of melted material. In this work we present the technique and report the densities of three high-pressure melts of the FCC metals iron, nickel, and gold derived from diffuse scattered X-ray spectra collected from in situ laser-heated diamond anvil cell synchrotron experiments. The results are compared to densities derived from shock wave experiments.
Experimental results for TiO2 melting and release using cold crucible melting

International Nuclear Information System (INIS)

Hong, S. W.; Min, B. T.; Park, I. G.; Kim, H. D.

2000-01-01

To simulate the severe accident phenomena using the real reactor material which melting point is about 2,800K, the melting and release method for materials with high melting point should be developed. This paper discusses the test results for TiO 2 materials using the cold crucible melting method to study the melting and release method of actual corium. To melt and release of few kg of TiO2, the experimental facility is manufactured through proper selection of design parameters such as frequency and capacity of R.F generator, crucible size and capacity of coolant. The melting and release of TiO 2 has been successfully performed in the cold crucible of 15cm in inner diameter and 30cm in height with 30kW RF power generator of 370 KHz. In the melt delivery experiment, about 2.6kg of molten TiO2, 60% of initial charged mass, is released. Rest of it is remained in the watercage in form of the rubble crust formed at the top of crucible and melt crust formed at the interface between the water-cage and melt. Especially, in the melt release test, the location of the working coil is important to make the thin crust at the bottom of the crucible
Models and correlations of the DEBRIS Late-Phase Melt Progression Model

International Nuclear Information System (INIS)

Schmidt, R.C.; Gasser, R.D.

1997-09-01

The DEBRIS Late Phase Melt Progression Model is an assembly of models, embodied in a computer code, which is designed to treat late-phase melt progression in dry rubble (or debris) regions that can form as a consequence of a severe core uncover accident in a commercial light water nuclear reactor. The approach is fully two-dimensional, and incorporates a porous medium modeling framework together with conservation and constitutive relationships to simulate the time-dependent evolution of such regions as various physical processes act upon the materials. The objective of the code is to accurately model these processes so that the late-phase melt progression that would occur in different hypothetical severe nuclear reactor accidents can be better understood and characterized. In this report the models and correlations incorporated and used within the current version of DEBRIS are described. These include the global conservation equations solved, heat transfer and fission heating models, melting and refreezing models (including material interactions), liquid and solid relocation models, gas flow and pressure field models, and the temperature and compositionally dependent material properties employed. The specific models described here have been used in the experiment design analysis of the Phebus FPT-4 debris-bed fission-product release experiment. An earlier DEBRIS code version was used to analyze the MP-1 and MP-2 late-phase melt progression experiments conducted at Sandia National Laboratories for the US Nuclear Regulatory Commission
Water jet intrusion into hot melt concomitant with direct-contact boiling of water

Energy Technology Data Exchange (ETDEWEB)

Sibamoto, Yasuteru [Japan Atomic Energy Research Inst., Tokai Research Establishment, Tokai, Ibaraki (Japan)

2005-08-01

Boiling of water poured on surface of high-temperature melt (molten metal or metal oxide) provides an efficient means for heat exchange or cooling of melt. The heat transfer surface area can be extended by forcing water into melt. Objectives of the present study are to elucidate key factors of the thermal and hydrodynamic interactions for the water jet injection into melt (Coolant Injection mode). Proposed applications include in in-vessel heat exchangers for liquid metal reactor and emergency measures for cooling of molten core debris in severe accidents of light water reactor. Water penetration into melt may occurs also as a result of fuel-coolant interaction (FCI) in modes other than CI, it is anticipated that the present study contributes to understand the fundamental mechanism of the FCI process. The previous works have been limited on understanding the melt-water interaction phenomena in the water-injection mode because of difficulty in experimental measurement where boiling occurs in opaque invisible hot melt unlike the melt-injection mode. We conducted visualization and measurement of melt-water-vapor multiphase flow phenomena by using a high-frame-rate neutron radiography technique and newly-developed probes. Although limited knowledge, however, has been gained even such an approach, the experimental data were analyzed deeply by comparing with the knowledge obtained from relevant matters. As a result, we succeeded in revealing several key phenomena and validity in the conditions under which stable heat transfer is established. Moreover, a non-intrusive technique for measurement of the velocity and pressure fields adjacent to a moving free surface is developed. The technique is based on the measurement of fluid surface profile, which is useful for elucidation of flow mechanism accompanied by a free surface like the present phenomena. (author)
Partial melting of lower oceanic crust gabbro: Constraints from poikilitic clinopyroxene primocrysts

Science.gov (United States)

Leuthold, Julien; Lissenberg, C. Johan; O'Driscoll, Brian; Karakas, Ozge; Falloon, Trevor; Klimentyeva, Dina N.; Ulmer, Peter

2018-03-01

Successive magma batches underplate, ascend, stall and erupt along spreading ridges, building the oceanic crust. It is therefore important to understand the processes and conditions under which magma differentiates at mid ocean ridges. Although fractional crystallization is considered to be the dominant mechanism for magma differentiation, open-system igneous complexes also experience Melting-Assimilation-Storage-Hybridization (MASH, Hildreth and Moorbath, 1988) processes. Here, we examine crystal-scale records of partial melting in lower crustal gabbroic cumulates from the slow-spreading Atlantic oceanic ridge (Kane Megamullion; collected with Jason ROV) and the fast-spreading East Pacific Rise (Hess Deep; IODP expedition 345). Clinopyroxene oikocrysts in these gabbros preserve marked intra-crystal geochemical variations that point to crystallization-dissolution episodes of the gabbro eutectic assemblage. Kane Megamullion and Hess Deep clinopyroxene core1 primocrysts and their plagioclase inclusions indicate crystallization from high temperature basalt (>1160 and >1200°C, respectively), close to clinopyroxene saturation temperature (fundamental mechanisms for generating the wide compositional variation observed in mid-ocean ridge basalts. We furthermore propose that such processes operate at both slow- and fast-spreading ocean ridges. Thermal numerical modelling shows that the degree of lower crustal partial melting at slow-spreading ridges can locally increase up to 50%, but the overall crustal melt volume is low (less than ca. 5% of total mantle-derived and crustal melts; ca. 20% in fast-spreading ridges).
Energy Saving Melting and Revert Reduction Technology (E-SMARRT): Melting Efficiency Improvement

Energy Technology Data Exchange (ETDEWEB)

Principal Investigator Kent Peaslee; Co-PIÃÂÃÂ¢ÃÂÃÂÃÂÃÂs: Von Richards, Jeffrey Smith

2012-07-31

Steel foundries melt recycled scrap in electric furnaces and typically consume 35-100% excess energy from the theoretical energy requirement required to pour metal castings. This excess melting energy is multiplied by yield losses during casting and finishing operations resulting in the embodied energy in a cast product typically being three to six times the theoretical energy requirement. The purpose of this research project was to study steel foundry melting operations to understand energy use and requirements for casting operations, define variations in energy consumption, determine technologies and practices that are successful in reducing melting energy and develop new melting techniques and tools to improve the energy efficiency of melting in steel foundry operations.
The Live program - Results of test L1 and joint analyses on transient molten pool thermal hydraulics

Energy Technology Data Exchange (ETDEWEB)

Buck, M.; Buerger, M. [Univ Stuttgart, Inst Kernenerget and Energiesyst, D-70569 Stuttgart (Germany); Miassoedov, A.; Gaus-Liu, X.; Palagin, A. [IRSN Forschungszentrum Karlsruhe GmbH, D-76021 Karlsruhe, (Germany); Godin-Jacqmin, L. [CEA Cadarache, DEN STRI LMA, F-13115 St Paul Les Durance (France); Tran, C. T.; Ma, W. M. [KTH, AlbaNova Univ Ctr, S-10691 Stockholm (Sweden); Chudanov, V. [Nucl Safety Inst, Moscow 113191 (Russian Federation)

2010-07-01

The development of a corium pool in the lower head and its behaviour is still a critical issue. This concerns, in general, the understanding of a severe accident with core melting, its course, major critical phases and timing, and the influence of these processes on the accident progression as well as, in particular, the evaluation of in-vessel melt retention by external vessel flooding as an accident mitigation strategy. Previous studies were especially related to the in-vessel retention question and often just concentrated on the quasi-steady state behaviour of a large molten pool in the lower head, considered as a bounding configuration. However, non-feasibility of the in-vessel retention concept for high power density reactors and uncertainties e. g. due to layering effects even for low or medium power reactors, turns this to be insufficient. Rather, it is essential to consider the whole evolution of the accident, including e. g. formation and growth of the in-core melt pool, characteristics of corium arrival in the lower head, and molten pool behaviour after the debris re-melting. These phenomena have a strong impact on a potential termination of a severe accident. The general objective of the LIVE program at FZK is to study these phenomena resulting from core melting experimentally in large-scale 3D geometry and in supporting separate-effects tests, with emphasis on the transient behaviour. Up to now, several tests on molten pool behaviour have been performed within the LIVE experimental program with water and with non-eutectic melts (KNO{sub 3}-NaNO{sub 3}) as simulant fluids. The results of these experiments, performed in nearly adiabatic and in isothermal conditions, allow a direct comparison with findings obtained earlier in other experimental programs (SIMECO, ACOPO, BALI, etc. ) and will be used for the assessment of the correlations derived for the molten pool behaviour. Complementary to other international programs with real corium melts, the results
MELT-IIIB: an updated version of the melt code

International Nuclear Information System (INIS)

Tabb, K.K.; Lewis, C.H.; O'Dell, L.D.; Padilla, A. Jr.; Smith, D.E.; Wilburn, N.P.

1979-04-01

The MELT series is a reactor modeling code designed to investigate a wide variety of hypothetical accident conditions, particularly the transient overpower sequence. MELT-IIIB is the latest in the series
Turbulence model for melt pool natural convection heat transfer

International Nuclear Information System (INIS)

Kelkar, K.M.; Patankar, S.V.

1994-01-01

Under severe reactor accident scenarios, pools of molten core material may form in the reactor core or in the hemispherically shaped lower plenum of the reactor vessel. Such molten pools are internally heated due to the radioactive decay heat that gives rise to buoyant flows in the molten pool. The flow in such pools is strongly influenced by the turbulent mixing because the expected Rayleigh numbers under accidents scenarios are very high. The variation of the local heat flux over the boundaries of the molten pools are important in determining the subsequent melt progression behavior. This study reports results of an ongoing effort towards providing a well validated mathematical model for the prediction of buoyant flow and heat transfer in internally heated pool under conditions expected in severe accident scenarios
Theoretical and experimental methods to determine the properties of molten core components and reaction products. Pt. 2

International Nuclear Information System (INIS)

Nazare, S.; Ondracek, G.; Schulz, B.

1975-10-01

In the course of a loss of coolant accident, a sequence of events would be initiated that ultimately could lead to core melting. The course of these events and the consequences of core meltdown would in part be determined by the properties of the core materials and the products of their interaction. On the basis of available theoretical and experimental results, the report attempts an estimation of properties such as: 1) work of adhesion between UO 2 - and (U,Zr) liquid phase, 2) heat of fusion of some melts, 3) heat capacity of liquid reaction products, 4) viscosity of liquid reaction products, 5) thermal conductivity of liquid reaction products. Experimental work is suggested for those cases, where the estimates need to be improved or verified. (orig.) [de

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

International Nuclear Information System (INIS)

Tran, Chi Thanh

2009-09-01

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, is
Containment loading during severe core damage accidents

International Nuclear Information System (INIS)

Fermandjian, J.; Evrard, J.M.; Cenerino, C.; Berthion, Y.; Carvallo, G.

1984-11-01

The objective of the article is to study the influence of the state of the reactor cavity (dry or flooded) and of the corium coolability on the thermal-hydraulics in the containment in the case of an accident sequence involving core melting and subsequent containment basemat erosion, in a 900 MWe PWR unit. Calculations are performed by using the JERICHO thermal hydraulics code
Cloud screening and melt water detection over melting sea ice using AATSR/SLSTR

Science.gov (United States)

Istomina, Larysa; Heygster, Georg

2014-05-01

With the onset of melt in the Arctic Ocean, the fraction of melt water on sea ice, the melt pond fraction, increases. The consequences are: the reduced albedo of sea ice, increased transmittance of sea ice and affected heat balance of the system with more heat passing through the ice into the ocean, which facilitates further melting. The onset of melt, duration of melt season and melt pond fraction are good indicators of the climate state of the Arctic and its change. In the absence of reliable sea ice thickness retrievals in summer, melt pond fraction retrieval from satellite is in demand as input for GCM as an indicator of melt state of the sea ice. The retrieval of melt pond fraction with a moderate resolution radiometer as AATSR is, however, a non-trivial task due to a variety of subpixel surface types with very different optical properties, which give non-unique combinations if mixed. In this work this has been solved by employing additional information on the surface and air temperature of the pixel. In the current work, a concept of melt pond detection on sea ice is presented. The basis of the retrieval is the sensitivity of AATSR reflectance channels 550nm and 860nm to the amount of melt water on sea ice. The retrieval features extensive usage of a database of in situ surface albedo spectra. A tree of decisions is employed to select the feasible family of in situ spectra for the retrieval, depending on the melt stage of the surface. Reanalysis air temperature at the surface and brightness temperature measured by the satellite sensor are analyzed in order to evaluate the melting status of the surface. Case studies for FYI and MYI show plausible retrieved melt pond fractions, characteristic for both of the ice types. The developed retrieval can be used to process the historical AATSR (2002-2012) dataset, as well as for the SLSTR sensor onboard the future Sentinel-3 mission (scheduled for launch in 2015), to keep the continuity and obtain longer time sequence
Flow Boiling on a Downward-Facing Inclined Plane Wall of Core Catcher

International Nuclear Information System (INIS)

Kim, Hyoung Tak; Bang, Kwang Hyun; Suh, Jung Soo

2013-01-01

In order to investigate boiling behavior on downward-facing inclined heated wall prior to the CHF condition, an experiment was carried out with 1.2 m long rectangular channel, inclined by 10 .deg. from the horizontal plane. High speed video images showed that the bubbles were sliding along the heated wall, continuing to grow and combining with the bubbles growing at their nucleation sites in the downstream. These large bubbles continued to slide along the heated wall and formed elongated slug bubbles. Under this slug bubble thin liquid film layer on the heated wall was observed and this liquid film prevents the wall from dryout. The length, velocity and frequency of slug bubbles sliding on the heated wall were measured as a function of wall heat flux and these parameters were used to develop wall boiling model for inclined, downward-facing heated wall. One approach to achieve coolable state of molten core in a PWR-like reactor cavity during a severe accident is to retain the core melt on a so-called core catcher residing on the reactor cavity floor after its relocation from the reactor pressure vessel. The core melt retained in the core catcher is cooled by water coolant flowing in an inclined cooling channel underneath as well as the water pool overlaid on the melt layer. Two-phase flow boiling with downward-facing heated wall such as this core catcher cooling channel has drawn a special attention because this orientation of heated wall may reach boiling crisis at lower heat flux than that of a vertical or upward-facing heated wall. Nishikawa and Fujita, Howard and Mudawar, Qiu and Dhir have conducted experiments to study the effect of heater orientation on boiling heat transfer and CHF. SULTAN experiment was conducted to study inclined large-scale structure coolability by water in boiling natural convection. In this paper, high-speed visualization of boiling behavior on downward-facing heated wall inclined by 10 .deg. is presented and wall boiling model for the
Computer Simulation To Assess The Feasibility Of Coring Magma

Science.gov (United States)

Su, J.; Eichelberger, J. C.

2017-12-01

Lava lakes on Kilauea Volcano, Hawaii have been successfully cored many times, often with nearly complete recovery and at temperatures exceeding 1100oC. Water exiting nozzles on the diamond core bit face quenches melt to glass just ahead of the advancing bit. The bit readily cuts a clean annulus and the core, fully quenched lava, passes smoothly into the core barrel. The core remains intact after recovery, even when there are comparable amounts of glass and crystals with different coefficients of thermal expansion. The unique resulting data reveal the rate and sequence of crystal growth in cooling basaltic lava and the continuous liquid line of descent as a function of temperature from basalt to rhyolite. Now that magma bodies, rather than lava pooled at the surface, have been penetrated by geothermal drilling, the question arises as to whether similar coring could be conducted at depth, providing fundamentally new insights into behavior of magma. This situation is considerably more complex because the coring would be conducted at depths exceeding 2 km and drilling fluid pressures of 20 MPa or more. Criteria that must be satisfied include: 1) melt is quenched ahead of the bit and the core itself must be quenched before it enters the barrel; 2) circulating drilling fluid must keep the temperature of the coring assembling cooled to within operational limits; 3) the drilling fluid column must nowhere exceed the local boiling point. A fluid flow simulation was conducted to estimate the process parameters necessary to maintain workable temperatures during the coring operation. SolidWorks Flow Simulation was used to estimate the effect of process parameters on the temperature distribution of the magma immediately surrounding the borehole and of drilling fluid within the bottom-hole assembly (BHA). A solid model of the BHA was created in SolidWorks to capture the flow behavior around the BHA components. Process parameters used in the model include the fluid properties and
Post retention and post/core shear bond strength of four post systems.

Science.gov (United States)

Stockton, L W; Williams, P T; Clarke, C T

2000-01-01

As clinicians we continue to search for a post system which will give us maximum retention while maximizing resistance to root fracture. The introduction of several new post systems, with claims of high retentive and resistance to root fracture values, require that independent studies be performed to evaluate these claims. This study tested the tensile and shear dislodgment forces of four post designs that were luted into roots 10 mm apical of the CEJ. The Para Post Plus (P1) is a parallel-sided, passive design; the Para Post XT (P2) is a combination active/passive design; the Flexi-Post (F1) and the Flexi-Flange (F2) are active post designs. All systems tested were stainless steel. This study compared the test results of the four post designs for tensile and shear dislodgment. All mounted samples were loaded in tension until failure occurred. The tensile load was applied parallel to the long axis of the root, while the shear load was applied at 450 to the long axis of the root. The Flexi-Post (F1) was significantly different from the other three in the tensile test, however, the Para Post XT (P2) was significantly different to the other three in the shear test and had a better probability for survival in the Kaplan-Meier survival function test. Based on the results of this study, our recommendation is for the Para Post XT (P2).
The MELTSPREAD Code for Modeling of Ex-Vessel Core Debris Spreading Behavior, Code Manual – Version3-beta

Energy Technology Data Exchange (ETDEWEB)

Farmer, M. T. [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-09-01

MELTSPREAD3 is a transient one-dimensional computer code that has been developed to predict the gravity-driven flow and freezing behavior of molten reactor core materials (corium) in containment geometries. Predictions can be made for corium flowing across surfaces under either dry or wet cavity conditions. The spreading surfaces that can be selected are steel, concrete, a user-specified material (e.g., a ceramic), or an arbitrary combination thereof. The corium can have a wide range of compositions of reactor core materials that includes distinct oxide phases (predominantly Zr, and steel oxides) plus metallic phases (predominantly Zr and steel). The code requires input that describes the containment geometry, melt “pour” conditions, and cavity atmospheric conditions (i.e., pressure, temperature, and cavity flooding information). For cases in which the cavity contains a preexisting water layer at the time of RPV failure, melt jet breakup and particle bed formation can be calculated mechanistically given the time-dependent melt pour conditions (input data) as well as the heatup and boiloff of water in the melt impingement zone (calculated). For core debris impacting either the containment floor or previously spread material, the code calculates the transient hydrodynamics and heat transfer which determine the spreading and freezing behavior of the melt. The code predicts conditions at the end of the spreading stage, including melt relocation distance, depth and material composition profiles, substrate ablation profile, and wall heatup. Code output can be used as input to other models such as CORQUENCH that evaluate long term core-concrete interaction behavior following the transient spreading stage. MELTSPREAD3 was originally developed to investigate BWR Mark I liner vulnerability, but has been substantially upgraded and applied to other reactor designs (e.g., the EPR), and more recently to the plant accidents at Fukushima Daiichi. The most recent round of
Influence of New Sol-gel Refractory Coating on the Casting Properties of Cold Box and Furan Cores for Grey Cast iron

DEFF Research Database (Denmark)

Nwaogu, Ugochukwu Chibuzoh; Poulsen, T.; Bischoff, C

2010-01-01

New Sol-Gel coated sand cores made from coldbox and furan binder systems were investigated. The idea of the coating was to improve the surface quality of castings. Grey iron was cast on the cores in a sand casting process. The effect of the high temperature of the melt on the cores was assessed...... by measuring the heating curves. The viscosity of the coating, moisture content and the permeability of the cores were evaluated. The surface quality of the castings was investigated using SEM and OM. The results show that the moisture content of the cores affected the permeability. In furan cores the vapour...... transport zone (VTZ) when in contact with the melt is larger than it is in a coldbox which means the furan cores have higher moisture content. The new sol-gel coating has the potential for improving the surface quality of castings without negative effects on the graphite distribution. The surface...
Fukushima Daiichi Unit 1 Ex-Vessel Prediction: Core Concrete Interaction

International Nuclear Information System (INIS)

Robb, Kevin R; Farmer, Mitchell; Francis, Matthew W

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.
The high-pressure phase diagram of Fe(0.94)O - A possible constituent of the earth's core

Science.gov (United States)

Knittle, Elise; Jeanloz, Raymond

1991-01-01

Electrical resistivity measurements to pressures of 83 GPa and temperatures ranging from 300 K to 4300 K confirm the presence of both crystalline and liquid metallic phases of FeO at pressures above 60-70 GPa and temperatures above 1000 K. By experimentally determinig the melting temperature of FeO to 100 GPa and of a model-core composition at 83 GPa, it is found that the solid-melt equilibria can be described by complete solid solution across the Fe-FeO system at pressures above 70 GPa. The results indicate that oxygen is a viable and likely candidate for the major light alloying element of the earth's liquid outer core. The data suggest that the temperature at the core-mantle boundary is close to 4800 K and that heat lost out of the core accounts for more than 20 percent of the heat flux observed at the surface.
Modeling of molten core-concrete interactions and fission-product release

International Nuclear Information System (INIS)

Norkus, J.K.; Corradini, M.L.

1991-09-01

The study of molten core-concrete interaction is important in estimating the possible consequences of a severe nuclear reactor accident. CORCON-Mod2 is a computer program which models the thermal, chemical, and physical phenomena associated with molten core-concrete interactions. Models have been added to extend and improve the modeling of these phenomena. An ideal solution chemical equilibrium methodology is presented to predict the fission-product vaporization release. Additional chemical species have been added, and the calculation of chemical equilibrium has been expanded to the oxidic layer and to the mixed layer configuration. Recent experiments performed at Argonne National Laboratory are compared to CORCON predictions of melt temperature, erosion depth, and release fraction of fission products. The results consistently underpredicted the melt temperatures and erosion rates. However, the predictions of release of Te, Ba, Sr, and U were good. A sensitivity study of the effects of initial temperature, concrete type, use of the mixing option, degree of zirconium oxidation, cavity size, and amount of control material on erosion, gas production, and release of radioactive materials was performed for a PWR and a BWR. The initial melt temperature had the greatest effect on the results of interest. Concrete type and cavity size also had important effects. 78 refs., 35 figs., 40 tabs
Behavior of concrete in contact with molten corium in the case of a hypothetical core melt accident

International Nuclear Information System (INIS)

Peehs, M.; Skokan, A.; Reimann, M.

1979-01-01

The temperature-dependent properties of basaltic and limestone concrete as needed for predicting Corium melt propagation in concrete (elongation behavior, specific heat and degradation enthalpy, thermal diffusivity, and conductivity) are determined experimentally together with the chemical and physical reactions occurring in heated concrete. The determined oxidation potential of -335 kJ/mole for molten Corium interacting with the concrete is in accordance with the observed H 2 generation due to the melt internal oxidation of zirconium, chromium, and iron. The liquefaction temperatures of the different concretes investigated are approx. 1300 to 1400 0 C. The relatively high degradation enthalpy of basaltic and limestone concrete is the reason for the barrier effect of concrete against propagating molten Corium
Behavior of a corium jet in high pressure melt ejection from a reactor pressure vessel

International Nuclear Information System (INIS)

Frid, W.

1988-04-01

Discharge of the molten core debris from a pressurized reactor vessel has been recognized as an important accident scenario for pressurized water reactors. Recent high-pressure melt streaming experiments conducted at Sandia National Laboratories, designed to study cavity and containment events related to melt ejection, have resulted in two important observations: (1) Expansion and breakup of the ejected molten jet. (2) Significant aerosol generation during the ejection process. The expansion and breakup of the jet in the experiments are attributed to rapid evolution of the pressurizing gas (nitrogen or hydrogen) dissolved in the melt. It has been concluded that aerosol particles may be formed by condensation of melt vapor and mechanical breakup of the melt and generation. It was also shown that the above stated phenomena are likely to occur in reactor accidents. This report provides results from analytical and experimental investigations on the behavior of a gas supersaturated molten jet expelled from a pressurized vessel. Aero-hydrodynamic stability of liquid jets in gas, stream degassing of molten metals, and gas bubble nucleation in molten metals are relevant problems that are addressed in this work
Plastic deformation of FeSi at high pressures: implications for planetary cores

Science.gov (United States)

Kupenko, Ilya; Merkel, Sébastien; Achorner, Melissa; Plückthun, Christian; Liermann, Hanns-Peter; Sanchez-Valle, Carmen

2017-04-01

The cores of terrestrial planets is mostly comprised of a Fe-Ni alloy, but it should additionally contain some light element(s) in order to explain the observed core density. Silicon has long been considered as a likely candidate because of geochemical and cosmochemical arguments: the Mg/Si and Fe/Si ratios of the Earth does not match those of the chondrites. Since silicon preferentially partition into iron-nickel metal, having 'missing' silicon in the core would solve this problem. Moreover, the evidence of present (e.g. Mercury) or ancient (e.g. Mars) magnetic fields on the terrestrial planets is a good indicator of (at least partially) liquid cores. The estimated temperature profiles of these planets, however, lay below iron melting curve. The addition of light elements in their metal cores could allow reducing their core-alloy melting temperature and, hence, the generation of a magnetic field. Although the effect of light elements on the stability and elasticity of Fe-Ni alloys has been widely investigated, their effect on the plasticity of core materials remains largely unknown. Yet, this information is crucial for understanding how planetary cores deform. Here we investigate the plastic deformation of ɛ-FeSi up to 50 GPa at room temperature employing a technique of radial x-ray diffraction in diamond anvil cells. Stoichiometric FeSi endmember is a good first-order approximation of the Fe-FeSi system and a good starting material to develop new experimental perspectives. In this work, we focused on the low-pressure polymorph of FeSi that would be the stable phase in the cores of small terrestrial planets. We will present the analysis of measured data and discuss their potential application to constrain plastic deformation in planetary cores.
Plasma wall interaction and tritium retention in TFTR

International Nuclear Information System (INIS)

Skinner, C.H.; Amarescu, E.; Ascione, G.

1996-01-01

The Tokamak Fusion Test Reactor (TFTR) has been operating safely and routinely with deuterium-tritium fuel for more than two years. In this time, TFTR has produced an impressive number of record breaking results including core fusion power, ∼ 2 MW/m 3 , comparable to that expected for ITER. Advances in wall conditioning via lithium pellet injection have played an essential role in achieving these results. Deuterium-tritium operation has also provided a special opportunity to address the issues of tritium recycling and retention. Tritium retention over two years of operation was approximately 40%. Recently, the in-torus tritium inventory was reduced by half through a combination of glow discharge cleaning, moist-air soaks, and plasma discharge cleaning. The tritium inventory is not a constraint in continued operations. The authors present recent results from TFTR in the context of plasma wall interactions and deuterium-tritium issues
Plasma wall interaction and tritium retention in TFTR

International Nuclear Information System (INIS)

Skinner, C.H.; Amarescu, E.; Ascione, G.

1997-01-01

The Tokamak Fusion Test Reactor (TFTR) has been operating safely and routinely with deuterium-tritium fuel for more than two years. In this time, TFTR has produced a number of record breaking results including core fusion power, ∝2 MW/m 3 , comparable to that expected for ITER. Advances in wall conditioning via lithium pellet injection have played an essential role in achieving these results. Deuterium-tritium operation has also provided a special opportunity to address the issues of tritium recycling and retention. Tritium retention over two years of operation was approximately 40%. Recently the in-torus tritium inventory was reduced by half through a combination of glow discharge cleaning, moist-air soaks, and plasma discharge cleaning. The tritium inventory is not a constraint in continued operations. Recent results from TFTR in the context of plasma wall interactions and deuterium-tritium issues are presented. (orig.)
Oxidation during reflood of reactor core with melting cladding

Energy Technology Data Exchange (ETDEWEB)

Siefken, L.J.; Allison, C.M.; Davis, K.L. [and others

1995-09-01

Models were recently developed and incorporated into the SCDAP/RELAP5 code for calculating the oxidation of fuel rods during cladding meltdown and reflood. Experiments have shown that a period of intense oxidation may occur when a hot partially oxidized reactor core is reflooded. This paper offers an explanation of the cladding meltdown and oxidation processes that cause this intense period of oxidation. Models for the cladding meltdown and oxidation processes are developed. The models are assessed by simulating a severe fuel damage experiment that involved reflood. The models for cladding meltdown and oxidation were found to improve calculation of the temperature and oxidation of fuel rods during the period in which hot fuel rods are reflooded.
Melting of Dense Sodium

International Nuclear Information System (INIS)

Gregoryanz, Eugene; Degtyareva, Olga; Hemley, Russell J.; Mao, Ho-kwang; Somayazulu, Maddury

2005-01-01

High-pressure high-temperature synchrotron diffraction measurements reveal a maximum on the melting curve of Na in the bcc phase at ∼31 GPa and 1000 K and a steep decrease in melting temperature in its fcc phase. The results extend the melting curve by an order of magnitude up to 130 GPa. Above 103 GPa, Na crystallizes in a sequence of phases with complex structures with unusually low melting temperatures, reaching 300 K at 118 GPa, and an increased melting temperature is observed with further increases in pressure
A 2D double-porosity model for melting and melt migration beneath mid-oceanic ridges

Science.gov (United States)

Liu, B.; Liang, Y.; Parmentier, E.

2017-12-01

Several lines of evidence suggest that the melting and melt extraction region of the MORB mantle is heterogeneous consisting of an interconnected network of high permeability dunite channels in a low porosity harzburgite or lherzolite matrix. In principle, one can include channel formation into the tectonic-scale geodynamic models by solving conservation equations for a chemically reactive and viscously deformable porous medium. Such an approach eventually runs into computational limitations such as resolving fractal-like channels that have a spectrum of width. To better understand first order features of melting and melt-rock interaction beneath MOR, we have formulated a 2D double porosity model in which we treat the triangular melting region as two overlapping continua occupied by the low-porosity matrix and interconnected high-porosity channels. We use melt productivity derived from a thermodynamic model and melt suction rate to close our problem. We use a high-order accurate numerical method to solve the conservation equations in 2D for porosity, solid and melt velocities and concentrations of chemical tracers in the melting region. We carry out numerical simulations to systematically study effects of matrix-to-channel melt suction and spatially distributed channels on the distributions of porosity and trace element and isotopic ratios in the melting region. For near fractional melting with 10 vol% channel in the melting region, the flow field of the matrix melt follows closely to that of the solid because the small porosity (exchange between the melt and the solid. The smearing effect can be approximated by dispersion coefficient. For slowly diffusing trace elements (e.g., LREE and HFSE), the melt migration induced dispersion can be as effective as thermal diffusion. Therefore, sub-kilometer scale heterogeneities of Nd and Hf isotopes are significantly damped or homogenized in the melting region.
Energy Saving Melting and Revert Reduction Technology: Melting Efficiency in Die Casting Operations

Energy Technology Data Exchange (ETDEWEB)

David Schwam

2012-12-15

This project addressed multiple aspects of the aluminum melting and handling in die casting operations, with the objective of increasing the energy efficiency while improving the quality of the molten metal. The efficiency of melting has always played an important role in the profitability of aluminum die casting operations. Consequently, die casters need to make careful choices in selecting and operating melting equipment and procedures. The capital cost of new melting equipment with higher efficiency can sometimes be recovered relatively fast when it replaces old melting equipment with lower efficiency. Upgrades designed to improve energy efficiency of existing equipment may be well justified. Energy efficiency is however not the only factor in optimizing melting operations. Melt losses and metal quality are also very important. Selection of melting equipment has to take into consideration the specific conditions at the die casting shop such as availability of floor space, average quantity of metal used as well as the ability to supply more metal during peaks in demand. In all these cases, it is essential to make informed decisions based on the best available data.

Constraints on The Coupled Thermal Evolution of the Earth's Core and Mantle, The Age of The Inner Core, And The Origin of the 186Os/188Os Core(?) Signal in Plume-Derived Lavas

Science.gov (United States)

Lassiter, J. C.

2005-12-01

-14 TW. In the absence of heat-producing elements in the core, such high heat flow rates require an inner core younger than ~1 Ga and preclude the development of significant 186Os enrichment in the outer core. Experimental studies suggest that potassium may partition into Fe-S-O liquids during core formation. Radioactive decay of potassium in the core could provide an additional heat source and reconcile geophysical evidence for high core/mantle heat flow with apparent geochemical evidence for an ancient inner core. However, high concentrations of chalcophile elements such as Cu in the mantle are inconsistent with significant segregation of a S-rich liquid during core formation, precluding K partitioning into the core by this mechanism. Furthermore, core formation scenarios that would lead to high K content in the core (e.g., core formation prior to terrestrial volatile depletion) also result in high core Pb concentrations. Core/mantle interaction would then produce strong negative correlations between 186Os/188Os and 207Pb/204Pb ratios, but such correlations are not observed. In summary, elevated 186Os/188Os ratios in some plume-derived lavas are unlikely to reflect core/mantle interaction because the inner core is too young for this isotopic signature to have developed in the outer core. Melt generation from pyroxenite or fractionation of PGEs between sulfide melts and monosulfide solid solutions provide alternative mechanisms for generating ancient mantle reservoirs with elevated Pt/Os and 186Os/188Os.
Comparison of structure, morphology, and leach characteristics of multi-phase ceramics produced via melt processing and hot isostatic pressing

Science.gov (United States)

Dandeneau, Christopher S.; Hong, Tao; Brinkman, Kyle S.; Vance, Eric R.; Amoroso, Jake W.

2018-04-01

Melt processing of multi-phase ceramic waste forms offers potential advantages over traditional solid-state synthesis methods given both the prevalence of melters currently in use and the ability to reduce the possibility of airborne radionuclide contamination. In this work, multi-phase ceramics with a targeted hollandite composition of Ba1.0Cs0.3Cr1.0Al0.3Fe1.0Ti5.7O16 were fabricated by melt processing at 1675 °C and hot isostatic pressing (HIP) at 1250 and 1300 °C. X-ray diffraction analysis (XRD) confirmed hollandite as the major phase in all specimens. Zirconolite/pyrochlore peaks and weaker perovskite reflections were observed after melt processing, while HIP samples displayed prominent perovskite peaks and low-intensity zirconolite reflections. Melt processing produced specimens with large (>50 μm) well-defined hollandite grains, while HIP yielded samples with a more fine-grained morphology. Elemental analysis showed "islands" rich in Cs and Ti across the surface of the 1300 °C HIP sample, suggesting partial melting and partitioning of Cs into multiple phases. Photoemission data revealed multiple Cs 3d spin-orbit pairs for the HIP samples, with the lower binding energy doublets likely corresponding to Cs located in more leachable phases. Among all specimens examined, the melt-processed sample exhibited the lowest fractional release rates for Rb and Cs. However, the retention of Sr and Mo was greater in the HIP specimens.
The WECHSL-Mod2 code: A computer program for the interaction of a core melt with concrete including the long term behavior

International Nuclear Information System (INIS)

Reimann, M.; Stiefel, S.

1989-06-01

The WECHSL-Mod2 code is a mechanistic computer code developed for the analysis of the thermal and chemical interaction of initially molten LWR reactor materials with concrete in a two-dimensional, axisymmetrical concrete cavity. The code performs calculations from the time of initial contact of a hot molten pool over start of solidification processes until long term basemat erosion over several days with the possibility of basemat penetration. The code assumes that the metallic phases of the melt pool form a layer at the bottom overlayed by the oxide melt atop. Heat generation in the melt is by decay heat and chemical reactions from metal oxidation. Energy is lost to the melting concrete and to the upper containment by radiation or evaporation of sumpwater possibly flooding the surface of the melt. Thermodynamic and transport properties as well as criteria for heat transfer and solidification processes are internally calculated for each time step. Heat transfer is modelled taking into account the high gas flux from the decomposing concrete and the heat conduction in the crusts possibly forming in the long term at the melt/concrete interface. The WECHSL code in its present version was validated by the BETA experiments. The test samples include a typical BETA post test calculation and a WECHSL application to a reactor accident. (orig.) [de
Mass transfer experiments for the heat load during in-vessel retention of core melt

Energy Technology Data Exchange (ETDEWEB)

Park, Hae Kyun; Chung, Bum Jin [Dept. of Nuclear Engineering, Kyung Hee University, Seoul (Korea, Republic of)

2016-08-15

We investigated the heat load imposed on the lower head of a reactor vessel by the natural convection of the oxide pool in a severe accident. Mass transfer experiments using a CuSO{sub 4}–H{sub 2}SO{sub 4} electroplating system were performed based on the analogy between heat and mass transfer. The Ra′{sub H} of 10{sup 14} order was achieved with a facility height of only 0.1 m. Three different volumetric heat sources were compared; two had identical configurations to those previously reported, and the other was designed by the authors. The measured Nu's of the lower head were about 30% lower than those previously reported. The measured angular heat flux ratios were similar to those reported in existing studies except for the peaks appearing near the top. The volumetric heat sources did not affect the Nu of the lower head but affected the Nu of the top plate by obstructing the rising flow from the bottom.
Package of programs for calculating accidents involving melting of the materials in a fast-reactor vessel

International Nuclear Information System (INIS)

Vlasichev, G.N.

1994-01-01

Methods for calculating one-dimensional nonstationary temperature distribution in a system of physically coupled materials are described. Six computer programs developed for calculating accident processes for fast reactor core melt are described in the article. The methods and computer programs take into account melting, solidification, and, in some cases, vaporization of materials. The programs perform calculations for heterogeneous systems consisting of materials with arbitrary but constant composition and heat transfer conditions at material boundaries. Additional modules provide calculations of specific conditions of heat transfer between materials, the change in these conditions and configuration of the materials as a result of coolant boiling, melting and movement of the fuel and structural materials, temperature dependences of thermophysical properties of the materials, and heat release in the fuel. 11 refs., 3 figs
Molten core material holding device in a nuclear reactor

International Nuclear Information System (INIS)

Nakamura, Hisashi; Tanaka, Nobuo; Takahashi, Katsuro.

1985-01-01

Purpose: To improve the function of cooling to hold molten core materials in a molten core material holding device. Constitution: Plenum structures are formed into a pan-like configuration, in which liners made of metal having high melting point and relatively high heat conductivity such as tantalum, tungsten, rhenium or alloys thereof are integrally appended to hold and directly cool the molten reactor core materials. Further, a plurality of heat pipes, passing through the plenum structures, facing the cooling portion thereof to the coolants at the outer side and immersing the heating portion into the molten core materials fallen to deposit in the inner liners are disposed radially. Furthermore, heat pipes embodded in the plenum structure are disposed in the same manner below the liners. Thus, the plenum structures and the molten reactor core materials can be cooled at a high efficiency. (Seki, T.)
Simulation experiments on the radial pool growth in gas-releasing melting system

International Nuclear Information System (INIS)

Farhadieh, R.; Purviance, R.; Carlson, N.

1983-01-01

Following an HCDA, molten core-debris can contact the concrete foundation of the reactor building resulting in a molten UO 2 /concrete interaction and considerable gas release. The released gas can pressurize the containment building potentially leading to radiological releases. Furthermore, directional growth of the molten core-debris pool can reduce the reactor building structural integrity. To implement design changes that insure structural integrity, an understanding of the thermal-hydraulic and mass-transfer process associated with such a growth is most desirable. Owing to the complex nature of the combined heat, mass, and hydrodynamic processes associated with the two-dimensional problem of gas release and melting, the downward and radial penetration problems have been investigated separately. The present experimental study addresses the question of sideward penetration of the molten core debris into a gas-releasing, meltable, miscible solid
Diversified emergency core cooling in CANDU with a passive moderator heat rejection system

Energy Technology Data Exchange (ETDEWEB)

Spinks, N [AECL Research, Chalk River Labs., Chalk River, ON (Canada)

1996-12-01

A passive moderator heat rejection system is being developed for CANDU reactors which, combined with a conventional emergency-coolant injection system, provides the diversity to reduce core-melt frequency to order 10{sup -7} per unit-year. This is similar to the approach used in the design of contemporary CANDU shutdown systems which leads to a frequency of order 10{sup -8} per unit-year for events leading to loss of shutdown. Testing of a full height 1/60 power-and-volume-scaled loop has demonstrated the feasibility of the passive system for removal of moderator heat during normal operation and during accidents. With the frequency of core-melt reduced, by these measures, to order 10{sup -7} per unit year, no need should exist for further mitigation. (author). 3 refs, 2 figs.
Fission product retention in TRISO coated UO2 particle fuels subjected to HTR simulated core heating tests

International Nuclear Information System (INIS)

Baldwin, C.A.; Kania, M.J.

1991-01-01

Results of the examination and analysis of 25,730 individual microspheres from spherical fuel elements HFR-K3/1 and HFR-K3/3 are reported. The parent spheres were irradiated in excess of end-of-life exposure and subsequently subjected to simulated core heating tests in a special high-temperature furnace at Forschungszentrum, Juelich, GmbH (KFA). Following the heating tests, the spheres were electrolytically deconsolidated to obtain unbounded fuel particles for Irradiated Microsphere Gamma Analyzer (IMGA) analysis. For sphere HFR-K3/1, which was heated for 500 h at 1600 deg. C, only four particles were identified as having released fission products. The remaining particles from the sphere showed no statistical evidence of fission product release. Scanning Electron Microscopy (SEM) examination showed that three of the defect particles had large sections of the TRISO coating missing, while the fourth appeared normal. For sphere HFR-K3/3, which was heated for 100 h at 1800 deg. C, the IMGA data revealed that fission product release (cesium) from individual particles was significant and that there was large particle-to-particle variation in retention capabilities. Individual particle release (cesium) averaged ten times the KFA-measured integral spherical fuel element release value. In addition, the bimodal distribution of the individual particle data indicated that two distinct modes of failure at fuel temperatures of 1800 deg. C and above may exist. (author). 6 refs, 6 figs, 4 tabs
Fission product retention in TRISCO coated UO2 particle fuels subjected to HTR simulated core heating tests

International Nuclear Information System (INIS)

Baldwin, C.A.; Kania, M.J.

1990-11-01

Results of the examination and analysis of 25,730 individual microspheres from spherical fuel elements HFR-K3/1 and HFR-K3/3 are reported. The parent spheres were irradiated in excess of end-of-life exposure and subsequently subjected to simulated core heating tests in a special high-temperature furnace at Forschungszentrum, Juelich, GmbH (KFA). Following the heating tests, the spheres were electrolytically deconsolidated to obtain unbonded fuel particles for Irradiated Microsphere Gamma Analyzer (IMGA) analysis. For sphere HFR-K3/1, which was heated for 500 h at 1600 degree C, only four particles were identified as having released fission products. The remaining particles from the sphere showed no statistical evidence of fission product release. Scanning Electron Microscopy (SEM) examination showed that three of the defect particles had large sections of the TRISO coating missing, while the fourth appeared normal. For sphere HFR-K3/3, which was heated for 100 h at 1800 degree C, the IMGA data revealed that fission product release (cesium) from individual particles was significant and that there was large particle-to-particle variation in retention capabilities. Individual particle release (cesium) averaged ten times the KFA-measured integral spherical fuel element release value. In addition, the bimodal distribution of the individual particle data indicated that two distinct modes of failure at fuel temperatures of 1800 degree C and above may exist. 6 refs., 6 figs., 4 tabs
Fragmentation and quench behavior of corium melt streams in water

International Nuclear Information System (INIS)

Spencer, B.W.; Wang, K.; Blomquist, C.A.; McUmber, L.M.; Schneider, J.P.

1994-02-01

The interaction of molten core materials with water has been investigated for the pour stream mixing mode. This interaction plays a crucial role during the later stages of in-vessel core melt progression inside a light water reactor such as during the TMI-2 accident. The key issues which arise during the molten core relocation include: (i) the thermal attack and possible damage to the RPV lower head from the impinging molten fuel stream and/or the debris bed, (ii) the molten fuel relocation pathways including the effects of redistribution due to core support structure and the reactor lower internals, (iii) the quench rate of the molten fuel through the water in the lower plenum, (iv) the steam generation and hydrogen generation during the interaction, (v) the transient pressurization of the primary system, and (vi) the possibility of a steam explosion. In order to understand these issues, a series of six experiments (designated CCM-1 through -6) was performed in which molten corium passed through a deep pool of water in a long, slender pour stream mode. Results discussed include the transient temperatures and pressures, the rate and magnitude of steam/hydrogen generation, and the posttest debris characteristics
Shock melting method to determine melting curve by molecular dynamics: Cu, Pd, and Al.

Science.gov (United States)

Liu, Zhong-Li; Zhang, Xiu-Lu; Cai, Ling-Cang

2015-09-21

A melting simulation method, the shock melting (SM) method, is proposed and proved to be able to determine the melting curves of materials accurately and efficiently. The SM method, which is based on the multi-scale shock technique, determines melting curves by preheating and/or prepressurizing materials before shock. This strategy was extensively verified using both classical and ab initio molecular dynamics (MD). First, the SM method yielded the same satisfactory melting curve of Cu with only 360 atoms using classical MD, compared to the results from the Z-method and the two-phase coexistence method. Then, it also produced a satisfactory melting curve of Pd with only 756 atoms. Finally, the SM method combined with ab initio MD cheaply achieved a good melting curve of Al with only 180 atoms, which agrees well with the experimental data and the calculated results from other methods. It turned out that the SM method is an alternative efficient method for calculating the melting curves of materials.
Shock melting method to determine melting curve by molecular dynamics: Cu, Pd, and Al

International Nuclear Information System (INIS)

Liu, Zhong-Li; Zhang, Xiu-Lu; Cai, Ling-Cang

2015-01-01

A melting simulation method, the shock melting (SM) method, is proposed and proved to be able to determine the melting curves of materials accurately and efficiently. The SM method, which is based on the multi-scale shock technique, determines melting curves by preheating and/or prepressurizing materials before shock. This strategy was extensively verified using both classical and ab initio molecular dynamics (MD). First, the SM method yielded the same satisfactory melting curve of Cu with only 360 atoms using classical MD, compared to the results from the Z-method and the two-phase coexistence method. Then, it also produced a satisfactory melting curve of Pd with only 756 atoms. Finally, the SM method combined with ab initio MD cheaply achieved a good melting curve of Al with only 180 atoms, which agrees well with the experimental data and the calculated results from other methods. It turned out that the SM method is an alternative efficient method for calculating the melting curves of materials
Analytical and Experimental Study for Validation of the Device to Confine BN Reactor Melted Fuel

International Nuclear Information System (INIS)

Rogozhkin, S.; Osipov, S.; Sobolev, V.; Shepelev, S.; Kozhaev, A.; Mavrin, M.; Ryabov, A.

2013-01-01

To validate the design and confirm the design characteristics of the special retaining device (core catcher) used for protection of BN reactor vessel in the case of a severe beyond-design basis accident with core melting, computational and experimental studies were carried out. The Tray test facility that uses water as coolant was developed and fabricated by OKBM; experimental studies were performed. To verify the methodical approach used for the computational study, experimental results obtained in the Tray test facility were compared with numerical simulation results obtained by the STAR-CCM+ CFD code
Insights into Mercury's Core Evolution from the Thermodynamic Properties of Fe-S-Si

Science.gov (United States)

Edgington, A.; Vocadlo, L.; Stixrude, L. P.; Wood, I. G.; Lord, O. T.

2015-12-01

The structure, composition and evolution of Mercury, the innermost planet, are puzzling, as its high uncompressed density implies a body highly enriched in metallic iron, whilst the existence of Mercury's magnetic field and observations of its longitude librations [1] suggest at least a partially molten core. This study uses a combination of experimental and ab-initio computer simulation techniques to determine the properties of Fe-S-Si (relative atomic percentages, 80:10:10) throughout the conditions of the interior of the planet Mercury, and evaluates the implications of this material for the structure and evolution of the planet's core. Previous studies have considered the addition of sulphur to the pure iron system, as this can significantly depress the melting curve of iron, and so may possibly allow Mercury's core to remain molten to the present day [2]. However, important constraints placed by the MESSENGER spacecraft on Mercury's surface abundance of iron [3] suggest that the planet formed in highly reduced conditions, in which significant amounts of silicon could have also dissolved into the core [4]. First-principles molecular dynamics simulations of the thermodynamic properties of liquid Fe-S-Si, alongside laser-heated diamond-anvil-cell experiments to determine the melting behaviour of the same composition, reveal the slopes of the adiabatic gradient and melting curve respectively, which together may allow insight into the evolution of our solar system's smallest planet. [1] Margot, J. L. et al. (2007) Science, 316: 710-714[2] Schubert, G. et al. (1988) in 'Mercury' 429-460[3] Nittler, L. R. et al. (2011) Science, 333, 1847-1850[4] Malavergne, V. et al. (2010) Icarus, 206:199-209
The Microwave Properties of Simulated Melting Precipitation Particles: Sensitivity to Initial Melting

Science.gov (United States)

Johnson, B. T.; Olson, W. S.; Skofronick-Jackson, G.

2016-01-01

A simplified approach is presented for assessing the microwave response to the initial melting of realistically shaped ice particles. This paper is divided into two parts: (1) a description of the Single Particle Melting Model (SPMM), a heuristic melting simulation for ice-phase precipitation particles of any shape or size (SPMM is applied to two simulated aggregate snow particles, simulating melting up to 0.15 melt fraction by mass), and (2) the computation of the single-particle microwave scattering and extinction properties of these hydrometeors, using the discrete dipole approximation (via DDSCAT), at the following selected frequencies: 13.4, 35.6, and 94.0GHz for radar applications and 89, 165.0, and 183.31GHz for radiometer applications. These selected frequencies are consistent with current microwave remote-sensing platforms, such as CloudSat and the Global Precipitation Measurement (GPM) mission. Comparisons with calculations using variable-density spheres indicate significant deviations in scattering and extinction properties throughout the initial range of melting (liquid volume fractions less than 0.15). Integration of the single-particle properties over an exponential particle size distribution provides additional insight into idealized radar reflectivity and passive microwave brightness temperature sensitivity to variations in size/mass, shape, melt fraction, and particle orientation.
Influence of spatial discretization, underground water storage and glacier melt on a physically-based hydrological model of the Upper Durance River basin

Science.gov (United States)

Lafaysse, M.; Hingray, B.; Etchevers, P.; Martin, E.; Obled, C.

2011-06-01

SummaryThe SAFRAN-ISBA-MODCOU hydrological model ( Habets et al., 2008) presents severe limitations for alpine catchments. Here we propose possible model adaptations. For the catchment discretization, Relatively Homogeneous Hydrological Units (RHHUs) are used instead of the classical 8 km square grid. They are defined from the dilineation of hydrological subbasins, elevation bands, and aspect classes. Glacierized and non-glacierized areas are also treated separately. In addition, new modules are included in the model for the simulation of glacier melt, and retention of underground water. The improvement resulting from each model modification is analysed for the Upper Durance basin. RHHUs allow the model to better account for the high spatial variability of the hydrological processes (e.g. snow cover). The timing and the intensity of the spring snowmelt floods are significantly improved owing to the representation of water retention by aquifers. Despite the relatively small area covered by glaciers, accounting for glacier melt is necessary for simulating the late summer low flows. The modified model is robust over a long simulation period and it produces a good reproduction of the intra and interannual variability of discharge, which is a necessary condition for its application in a modified climate context.
Survey of the state of the German safety study

International Nuclear Information System (INIS)

Heuser, F.W.; Kotthoff, K.

1977-01-01

In spring 1976 the Federal Ministry of Research and Technology has ordered a safety study to assess the risk for a nuclear power plant with a PWR. Giving first a survey on the main subtasks of the study the present state of work and some first results are discussed. Assuming a failure of safety systems a core melt event and a subsequent failure of the containment could occur. Corresponding accident sequences are discussed in some detail. Related hereto the results of some calculations for fission product release with respect to different containment failure modes are given. According to the results obtained so far the consequences of a core melt event can essentially be restricted by the retention function of the containment. (orig.) [de
Metamorphism and partial melting of ordinary chondrites: Calculated phase equilibria

Science.gov (United States)

Johnson, T. E.; Benedix, G. K.; Bland, P. A.

2016-01-01

Constraining the metamorphic pressures (P) and temperatures (T) recorded by meteorites is key to understanding the size and thermal history of their asteroid parent bodies. New thermodynamic models calibrated to very low P for minerals and melt in terrestrial mantle peridotite permit quantitative investigation of high-T metamorphism in ordinary chondrites using phase equilibria modelling. Isochemical P-T phase diagrams based on the average composition of H, L and LL chondrite falls and contoured for the composition and abundance of olivine, ortho- and clinopyroxene, plagioclase and chromite provide a good match with values measured in so-called equilibrated (petrologic type 4-6) samples. Some compositional variables, in particular Al in orthopyroxene and Na in clinopyroxene, exhibit a strong pressure dependence when considered over a range of several kilobars, providing a means of recognising meteorites derived from the cores of asteroids with radii of several hundred kilometres, if such bodies existed at that time. At the low pressures (recorders of peak conditions. The intersection of isopleths of these variables may allow pressures to be quantified, even at low P, permitting constraints on the minimum size of parent asteroid bodies. The phase diagrams predict the onset of partial melting at 1050-1100 °C by incongruent reactions consuming plagioclase, clinopyroxene and orthopyroxene, whose compositions change abruptly as melting proceeds. These predictions match natural observations well and support the view that type 7 chondrites represent a suprasolidus continuation of the established petrologic types at the extremes of thermal metamorphism. The results suggest phase equilibria modelling has potential as a powerful quantitative tool in investigating, for example, progressive oxidation during metamorphism, the degree of melting and melt loss or accumulation required to produce the spectrum of differentiated meteorites, and whether the onion shell or rubble pile
DEPENDENCY OF SULFATE SOLUBILITY ON MELT COMPOSITION AND MELT POLYMERIZATION

International Nuclear Information System (INIS)

JANTZEN, CAROL M.

2004-01-01

Sulfate and sulfate salts are not very soluble in borosilicate waste glass. When sulfate is present in excess it can form water soluble secondary phases and/or a molten salt layer (gall) on the melt pool surface which is purported to cause steam explosions in slurry fed melters. Therefore, sulfate can impact glass durability while formation of a molten salt layer on the melt pool can impact processing. Sulfate solubility has been shown to be compositionally dependent in various studies, (e.g. , B2O3, Li2O, CaO, MgO, Na2O, and Fe2O3 were shown to increase sulfate solubility while Al2O3 and SiO2 decreased sulfate solubility). This compositional dependency is shown to be related to the calculated melt viscosity at various temperatures and hence the melt polymerization

Manufacturing and characterization of encapsulated microfibers with different molecular weight poly(ε-caprolactone) (PCL) resins using a melt electrospinning technique

International Nuclear Information System (INIS)

Lee, Jason K; Ko, Junghyuk; Jun, Martin B G; Lee, Patrick C

2016-01-01

Encapsulated structures of poly(ε-caprolactone) microfibers were successfully fabricated through two distinct melt electrospinning methods: melt coaxial and melt-blending electrospinning methods. Both methods resulted in encapsulated microfibers, but the resultant microfibers had different morphologies. Melt coaxial electrospinning formed a dual, semi-concentric structure, whereas melt-blending electrospinning resulted in an islands-in-a-sea fiber structure (i.e. a multiple-core structure). The encapsulated microfibers were produced using a custom-designed melt coaxial electrospinning device and the microfibers were characterized using a scanning electron microscope. To analyze the properties of the melt blended encapsulated fibers and coaxial fibers, the microfiber mesh specimens were collected. The mechanical properties of each microfiber mesh were analyzed through a tensile test. The coaxial microfiber meshes were post processed with a femtosecond laser machine to create dog-bone shaped tensile test specimens, while the melt blended microfiber meshes were kept as-fabricated. The tensile experiments undertaken with coaxial microfiber specimens resulted in an increase in tensile strength compared to 10 k and 45 k monolayer specimens. However, melt blended microfiber meshes did not result in an increase in tensile strength. The melt blended microfiber mesh results indicate that by using greater amounts of 45 k PCL resin within the microstructure, the resulting fibers obtain a higher tensile strength. (paper)
2D model for melt progression through rods and debris

International Nuclear Information System (INIS)

Fichot, F.

2001-01-01

During the degradation of a nuclear core in a severe accident scenario, the high temperatures reached lead to the melting of materials. The formation of liquid mixtures at various elevations is followed by the flow of molten materials through the core. Liquid mixture may flow under several configurations: axial relocation along the rods, horizontal motion over a plane surface such as the core support plate or a blockage of material, 2D relocation through a debris bed, etc.. The two-dimensional relocation of molten material through a porous debris bed, implemented for the simulation of late degradation phases, has opened a new way to the elaboration of the relocation model for the flow of liquid mixture along the rods. It is based on a volume averaging method, where wall friction and capillary effects are taken into account by introducing effective coefficients to characterize the solid matrix (rods, grids, debris, etc.). A local description of the liquid flow is necessary to derive the effective coefficients. Heat transfers are modelled in a similar way. The derivation of the conservation equations for the liquid mixture falling flow (momentum) in two directions (axial and radial-horizontal) and for the heat exchanges (energy) are the main points of this new model for simulating melt progression. In this presentation, the full model for the relocation and solidification of liquid materials through a rod bundle or a debris bed is described. It is implemented in the ICARE/CATHARE code, developed by IPSN in Cadarache. The main improvements and advantages of the new model are: A single formulation for liquid mixture relocation, in 2D, either through a rod bundle or a porous debris bed, Extensions to complex structures (grids, by-pass, etc..), The modeling of relocation of a liquid mixture over plane surfaces. (author)
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
In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)

Energy Technology Data Exchange (ETDEWEB)

F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

2005-01-01

In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.
Enhanced⁹⁹Tc retention in glass waste form using Tc(IV)-incorporated Fe minerals

OpenAIRE

Um, W; Luksic, SA; Wang, G; Saslow, S; Kim, DS; Schweiger, MJ; Soderquist, CZ; Bowden, ME; Lukens, WW; Kruger, AA

2017-01-01

© 2017 Elsevier B.V. Technetium ( 99 Tc) immobilization by doping into iron oxide mineral phases may alleviate the problems with Tc volatility during vitrification of nuclear waste. Because reduced Tc, Tc(IV), substitutes for Fe(III) in the crystal structure by a process of Tc reduction from Tc(VII) to Tc(IV) followed by co-precipitation of Fe oxide minerals, two Tc-incorporated Fe minerals (Tc-goethite and Tc-magnetite/maghemite) were prepared and tested for Tc retention in glass melt sample...
Shock Melting of Iron Silicide as Determined by In Situ X-ray Diffraction.

Science.gov (United States)

Newman, M.; Kraus, R. G.; Wicks, J. K.; Smith, R.; Duffy, T. S.

2016-12-01

The equation of state of core alloys at pressures and temperatures near the solid-liquid coexistence curve is important for understanding the dynamics at the inner core boundary of the Earth and super-Earths. Here, we present a series of laser driven shock experiments on textured polycrystalline Fe-15Si. These experiments were conducted at the Omega and Omega EP laser facilities. Particle velocities in the Fe-15Si samples were measured using a line VISAR and were used to infer the thermodynamic state of the shocked samples. In situ x-ray diffraction measurements were used to probe the melting transition and investigate the potential decomposition of Fe-15Si in to hcp and B2 structures. This work examines the kinetic effects of decomposition due to the short time scale of dynamic compression experiments. In addition, the thermodynamic data collected in these experiments adds to a limited body of information regarding the equation of state of Fe-15Si, which is a candidate for the composition in Earth's outer core. Our experimental results show a highly textured solid phase upon shock compression to pressures ranging from 170 to 300 GPa. Below 320 GPa, we observe diffraction peaks consistent with decomposition of the D03 starting material in to an hcp and a cubic (potentially B2) structure. Upon shock compression above 320 GPa, the intense and textured solid diffraction peaks give way to diffuse scattering and loss of texture, consistent with melting along the Hugoniot. When comparing these results to that of pure iron, we can ascertain that addition of 15 wt% silicon increases the equilibrium melting temperature significantly, or that the addition of silicon significantly increases the metastability of the solid phase, relative to the liquid. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.
Station blackout core damage frequency in an advanced nuclear reactor

International Nuclear Information System (INIS)

Carvalho, Luiz Sergio de

2004-01-01

Even though nuclear reactors are provided with protection systems so that they can be automatically shut down in the event of a station blackout, the consequences of this event can be severe. This is because many safety systems that are needed for removing residual heat from the core and for maintaining containment integrity, in the majority of the nuclear power plants, are AC dependent. In order to minimize core damage frequency, advanced reactor concepts are being developed with safety systems that use natural forces. This work shows an improvement in the safety of a small nuclear power reactor provided by a passive core residual heat removal system. Station blackout core melt frequencies, with and without this system, are both calculated. The results are also compared with available data in the literature. (author)
Extraction of trapped gases in ice cores for isotope analysis

International Nuclear Information System (INIS)

Leuenberger, M.; Bourg, C.; Francey, R.; Wahlen, M.

2002-01-01

The use of ice cores for paleoclimatic investigations is discussed in terms of their application for dating, temperature indication, spatial time marker synchronization, trace gas fluxes, solar variability indication and changes in the Dole effect. The different existing techniques for the extraction of gases from ice cores are discussed. These techniques, all to be carried out under vacuum, are melt-extraction, dry-extraction methods and the sublimation technique. Advantages and disadvantages of the individual methods are listed. An extensive list of references is provided for further detailed information. (author)
Heatup of the TMI-2 lower head during core relocation

International Nuclear Information System (INIS)

Wang, S.K.; Sienicki, J.J.; Spencer, B.W.

1989-01-01

An analysis has been carried out to assess the potential of a melting attack upon the reactor vessel lower head and incore instrument nozzle penetration weldments during the TMI core relocation event at 224 minutes. Calculations were performed to determine the potential for molten corium to undergo breakup into droplets which freeze and form a debris bed versus impinging upon the lower head as one or more coherent streams. The effects of thermal-hydraulic interactions between corium streams and water inside the lower plenum, the effects of the core support assembly structure upon the corium, and the consequences of corium relocation by way of the core former region were examined. 19 refs., 24 figs
Radionuclide release and aerosol generation during core debris interactions with concrete

International Nuclear Information System (INIS)

Powers, D.A.

1986-01-01

During severe accidents at nuclear power plants, it is possible for the reactor fuel to melt and penetrate the reactor vessel. This can lead to vigorous interaction of core materials (UO 2 , ZrO 2 , Zr, and stainless steel) with structural concrete. Sparging of the molten core debris by gases (H 2 O and CO 2 ) liberated from the concrete can lead to rapid release of radionuclides from the core debris. A theoretical description of this release process has been developed and is called the VANESA model. The treatments in the VANESA model of the thermodynamics of radionuclide vaporization and the kinetic barriers to vaporization will be described. Predictions obtained from the model will be compared to the results of tests of core debris/concrete interactions
Melting of contaminated metallic waste

International Nuclear Information System (INIS)

Lee, Y.-S.; Cheng, S.-Y.; Kung, H.-T.; Lin, L.-F.

2004-01-01

Approximately 100 tons of contaminated metallic wastes were produced each year due to maintenance for each TPC's nuclear power reactor and it was roughly estimated that there will be 10,000 tons of metallic scraps resulted from decommissioning of each reactor in the future. One means of handling the contaminated metal is to melt it. Melting process owns not only volume reduction which saves the high cost of final disposal but also resource conservation and recycling benefits. Melting contaminated copper and aluminum scraps in the laboratory scale have been conducted at INER. A total of 546 kg copper condenser tubes with a specific activity of about 2.7 Bq/g was melted in a vacuum induction melting facility. Three types of products, ingot, slag and dust were derived from the melting process, with average activities of 0.10 Bq/g, 2.33 Bq/g and 84.3 Bq/g respectively. After the laboratory melting stage, a pilot plant with a 500 kg induction furnace is being designed to melt the increasingly produced contaminated metallic scraps from nuclear facilities and to investigate the behavior of different radionuclides during melting. (author)
Dynamics of Melting and Melt Migration as Inferred from Incompatible Trace Element Abundance in Abyssal Peridotites

Science.gov (United States)

Peng, Q.; Liang, Y.

2008-12-01

To better understand the melting processes beneath the mid-ocean ridge, we developed a simple model for trace element fractionation during concurrent melting and melt migration in an upwelling steady-state mantle column. Based on petrologic considerations, we divided the upwelling mantle into two regions: a double- lithology upper region where high permeability dunite channels are embedded in a lherzolite/harzburgite matrix, and a single-lithology lower region that consists of partially molten lherzolite. Melt generated in the single lithology region migrates upward through grain-scale diffuse porous flow, whereas melt in the lherzolite/harzburgite matrix in the double-lithology region is allowed to flow both vertically through the overlying matrix and horizontally into its neighboring dunite channels. There are three key dynamic parameters in our model: degree of melting experienced by the single lithology column (Fd), degree of melting experienced by the double lithology column (F), and a dimensionless melt suction rate (R) that measures the accumulated rate of melt extraction from the matrix to the channel relative to the accumulated rate of matrix melting. In terms of trace element fractionation, upwelling and melting in the single lithology column is equivalent to non-modal batch melting (R = 0), whereas melting and melt migration in the double lithology region is equivalent to a nonlinear combination of non-modal batch and fractional melting (0 abyssal peridotite, we showed, with the help of Monte Carlo simulations, that it is difficult to invert for all three dynamic parameters from a set of incompatible trace element data with confidence. However, given Fd, it is quite possible to constrain F and R from incompatible trace element abundances in residual peridotite. As an illustrative example, we used the simple melting model developed in this study and selected REE and Y abundance in diopside from abyssal peridotites to infer their melting and melt migration
The thermal evolution of Mercury's Fe-Si core

Science.gov (United States)

Knibbe, Jurriën Sebastiaan; van Westrenen, Wim

2018-01-01

We have studied the thermal and magnetic field evolution of planet Mercury with a core of Fe-Si alloy to assess whether an Fe-Si core matches its present-day partially molten state, Mercury's magnetic field strength, and the observed ancient crustal magnetization. The main advantages of an Fe-Si core, opposed to a previously assumed Fe-S core, are that a Si-bearing core is consistent with the highly reduced nature of Mercury and that no compositional convection is generated upon core solidification, in agreement with magnetic field indications of a stable layer at the top of Mercury's core. This study also present the first implementation of a conductive temperature profile in the core where heat fluxes are sub-adiabatic in a global thermal evolution model. We show that heat migrates from the deep core to the outer part of the core as soon as heat fluxes at the outer core become sub-adiabatic. As a result, the deep core cools throughout Mercury's evolution independent of the temperature evolution at the core-mantle boundary, causing an early start of inner core solidification and magnetic field generation. The conductive layer at the outer core suppresses the rate of core growth after temperature differences between the deep and shallow core are relaxed, such that a magnetic field can be generated until the present. Also, the outer core and mantle operate at higher temperatures than previously thought, which prolongs mantle melting and mantle convection. The results indicate that S is not a necessary ingredient of Mercury's core, bringing bulk compositional models of Mercury more in line with reduced meteorite analogues.
Environmentally friendly and highly productive bi-component melt spinning of thermoregulated smart polymer fibres with high latent heat capacity

Directory of Open Access Journals (Sweden)

Ch. Cherif

2018-03-01

Full Text Available A stable and reproducible bi-component melt spinning process on an industrial scale incorporating Phase Change Material (PCM into textile fibres has been successfully developed and carried out using a melt spinning machine. The key factor for a successful bi-component melt spinning process is that a deep insight into the thermal and rheological behaviour of PCM using Difference Scanning Calorimetry (DSC, Thermogravimetric Analysis (TGA, and an oscillatory rheological investigation. PCM is very sensitive to the temperature and residence time of the melt spinning process. It is found that the optimal process temperature of PCM is 210 °C. The textile-physical properties and the morphology of the melt spun and further drawn bi-component core and sheath fibres (bico fibres were investigated and interpreted. The heat capacities of PCM incorporated in bico fibres were also determined by means of DSC. The melt spun bico fibres integrating PCM provide a high latent heat of up to 22 J/g, which is three times higher than that of state-of-the-art fibres, which were also obtained using the melt spinning process. Therefore, they have the potential to be used as smart polymer fibres for textile and other technical applications.
Core-shell polymer nanorods by a two-step template wetting process

International Nuclear Information System (INIS)

Dougherty, S; Liang, J

2009-01-01

One-dimensional core-shell polymer nanowires offer many advantages and great potential for many different applications. In this paper we introduce a highly versatile two-step template wetting process to fabricate two-component core-shell polymer nanowires with controllable shell thickness. PLLA and PMMA were chosen as model polymers to demonstrate the feasibility of this process. Solution wetting with different concentrations of polymer solutions was used to fabricate the shell layer and melt wetting was used to fill the shell with the core polymer. The shell thickness was analyzed as a function of the polymer solution concentration and viscosity, and the core-shell morphology was observed with TEM. This paper demonstrates the feasibility of fabricating polymer core-shell nanostructures using our two-step template wetting process and opens the arena for optimization and future experiments with polymers that are desirable for specific applications.
Siderophile Volatile Element Partitioning during Core Formation.

Science.gov (United States)

Loroch, D. C.; Hackler, S.; Rohrbach, A.; Klemme, S.

2017-12-01

Since the nineteen sixties it is known, that the Earth's mantle is depleted relative to CI chondrite in numerous elements as a result of accretion and core-mantle differentiation. Additionally, if we take the chondritic composition as the initial solar nebular element abundances, the Earth lacks 85 % of K and up to 98 % of other volatiles. However one potentially very important group of elements has received considerably less attention in this context and these elements are the siderophile but volatile elements (SVEs). SVEs perhaps provide important information regarding the timing of volatile delivery to Earth. Especially for the SVEs the partitioning between metal melt and silicate melt (Dmetal/silicate) at core formation conditions is poorly constrained, never the less they are very important for most of the core formation models. This study is producing new metal-silicate partitioning data for a wide range of SVEs (S, Se, Te, Tl, Ag, As, Au, Cd, Bi, Pb, Sn, Cu, Ge, Zn, In and Ga) with a focus on the P, T and fO2dependencies. The initial hypothesis that we are aiming to test uses the accretion of major portions of volatile elements while the core formation was still active. The key points of this study are: - What are the effects of P, T and fO2 on SVE metal-silicate partioning? - What is the effect of compositional complexity on SVE metal-silicate partioning? - How can SVE's D-values fit into current models of core formation? The partitioning experiments will be performed using a Walker type multi anvil apparatus in a pressure range between 10 and 20 GPa and temperatures of 1700 up to 2100 °C. To determine the Dmetal/silicate values we are using a field emission high-resolution JEOL JXA-8530F EPMA for major elements and a Photon Machines Analyte G2 Excimer laser (193 nm) ablation system coupled to a Thermo Fisher Element 2 single-collector ICP-MS (LA-ICP-MS) for the trace elements. We recently finished the first sets of experiments and can provide the
Analysis of ex-vessel melt jet breakup and coolability. Part 1: Sensitivity on model parameters and accident conditions

Energy Technology Data Exchange (ETDEWEB)

Moriyama, Kiyofumi; Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr; Hwang, Byoungcheol; Jung, Woo Hyun

2016-06-15

Highlights: • Application of JASMINE code to melt jet breakup and coolability in APR1400 condition. • Coolability indexes for quasi steady state breakup and cooling process. • Typical case in complete breakup/solidification, film boiling quench not reached. • Significant impact of water depth and melt jet size; weak impact of model parameters. - Abstract: The breakup of a melt jet falling in a water pool and the coolability of the melt particles produced by such jet breakup are important phenomena in terms of the mitigation of severe accident consequences in light water reactors, because the molten and relocated core material is the primary heat source that governs the accident progression. We applied a modified version of the fuel–coolant interaction simulation code, JASMINE, developed at Japan Atomic Energy Agency (JAEA) to a plant scale simulation of melt jet breakup and cooling assuming an ex-vessel condition in the APR1400, a Korean advanced pressurized water reactor. Also, we examined the sensitivity on seven model parameters and five initial/boundary condition variables. The results showed that the melt cooling performance of a 6 m deep water pool in the reactor cavity is enough for removing the initial melt enthalpy for solidification, for a melt jet of 0.2 m initial diameter. The impacts of the model parameters were relatively weak and that of some of the initial/boundary condition variables, namely the water depth and melt jet diameter, were very strong. The present model indicated that a significant fraction of the melt jet is not broken up and forms a continuous melt pool on the containment floor in cases with a large melt jet diameter, 0.5 m, or a shallow water pool depth, ≤3 m.
Modelling of the controlled melt flow in a glass melting space – Its melting performance and heat losses

Czech Academy of Sciences Publication Activity Database

Jebavá, Marcela; Dyrčíková, Petra; Němec, Lubomír

2015-01-01

Roč. 430, DEC 15 (2015), s. 52-63 ISSN 0022-3093 Institutional support: RVO:67985891 Keywords : glass melt flow * mathematical modelling * energy distribution * space utilizatios * melting performance Subject RIV: JH - Ceramics, Fire-Resistant Materials and Glass Impact factor: 1.825, year: 2015
A study of the decontamination procedures used for chemical analysis of polar deep ice cores

Directory of Open Access Journals (Sweden)

Takayuki Miyake

2009-11-01

Full Text Available We investigated the decontamination procedures used on polar deep ice cores before chemical analyses such as measurements of the concentrations of iron species and dust (microparticles. We optimized cutting and melting protocols for decontamination using chemically ultraclean polyethylene bags and simulated ice samples made from ultrapure water. For dust and ion species including acetate, which represented a high level of contamination, we were able to decrease contamination to below several μg l^ for ion concentrations and below 10000 particles ml^ for the dust concentration. These concentration levels of ion species and dust are assumed to be present in the Dome Fuji ice core during interglacial periods. Decontamination of the ice core was achieved by cutting away approximately 3 mm of the outside of a sample and by melting away approximately 30% of a sample's weight. Furthermore, we also report the preparation protocols for chemical analyses of the 2nd Dome Fuji ice core, including measurements of ion and dust concentrations, pH, electric conductivity (EC, and stable isotope ratios of water (δD and δO, based on the results of the investigation of the decontamination procedures.
Basic experimental study with visual observation on elimination of the re-criticality issue using the MELT-II facility. Simulated fuel-escape behavior through a coolant channel

International Nuclear Information System (INIS)

Matsuba, Ken-ichi; Imahori, Shinji; Isozaki, Mikio

2004-11-01

In a core disruptive accident of fast reactors, fuel escape from the reactor core is a key phenomenon for prevention of re-criticality with significant mechanical-energy release subsequent to formation of a large-scale fuel pool with high mobility. Therefore, it is effective to study possibility of early fuel escape through probable escape paths such as a control-rod-guide-tube space well before high-mobility-pool formation. The purpose of the present basic experimental study is to clarify the mechanism of fuel-escape under a condition expected in the reactor situation, in which some amount of coolant may be entrapped into the molten-fuel pool. The following results have been obtained through basic experiments in which molten Wood's metal (components: 60wt%Bi-20wt%Sn-20wt%In, density at the room temperature: 8700 kg/m 3 , melting point: 78.8degC) is ejected into an coolant channel filled with water. (1) In the course of melt ejection, a small quantity of coolant is forced to be entrapped into the melt pool as a result of thermal interactions leading to high-pressure rise within the coolant channel. (2) Melt ejection is accelerated by pressure build-up which results from vapor pressure of entrapped coolant within the melt pool. (3) Average melt-ejection rate tends to increase in lower coolant-subcooling conditions, in which pressure build-up within the melt pool is enhanced. These results indicate a probability of a phenomenon in which melt ejection is accelerated by entrapment of coolant within a melt pool. Through application of the mechanism of confirmed phenomenon into the reactor condition, it is suggested that fuel escape is enhanced by entrapment of coolant within a fuel pool. (author)

On the failure modes of alternative containment designs following postulated core meltdown

International Nuclear Information System (INIS)

Chan, C.K.; Knee, H.E.; Okrent, D.

1977-01-01

The containment response to a postulated core meltdown accident in a PWR ice condenser containment, a BWR Mark III containment and a BWR non-inerted Mark I containment has been examined to see if the WASH-1400 containment failure mode judgement for the Surry large, dry containment and the Peach Bottom Mark I inerted-containment are likely to be appropriate for these alternative containment plant designs. For the PWR, the representative accident chosen for the analysis is a large cold leg break accompanied by a loss of all electric power while the BWR respresentative event chosen is a recirculation line break without adequate core cooling function. Two containment event paths are studied for each of these two cases, depending on whether or not containment vapor suppression function is assumed to be available. Both the core and the containment pressure and temperature response to the accident events are computed for the four time intervals which characterize (a) blowdown of the pipe break, (b) core melt, (c) vessel melt-through, and (d) containment foundation penetration. The calculations are based on a best esimate of the most probable sequence, but certain phenomena and events were followed down multiple tracks. It appears that the non-inerted Mark I containment is not so vulnerable to overpressurization from hydrogen burning as the Mark III; however, acceptable temperatures may be exceeded. (Auth.)
Assessment of Core Failure Limits for Light Water Reactor Fuel under Reactivity Initiated Accidents

International Nuclear Information System (INIS)

Jernkvist, Lars Olof; Massih, Ali R.

2004-12-01

Core failure limits for high-burnup light water reactor UO 2 fuel rods, subjected to postulated reactivity initiated accidents (RIAs), are here assessed by use of best-estimate computational methods. The considered RIAs are the hot zero power rod ejection accident (HZP REA) in pressurized water reactors and the cold zero power control rod drop accident (CZP CRDA) in boiling water reactors. Burnup dependent core failure limits for these events are established by calculating the fuel radial average enthalpy connected with incipient fuel pellet melting for fuel burnups in the range of 30 to 70 MWd/kgU. The postulated HZP REA and CZP CRDA result in lower enthalpies for pellet melting than RIAs that take place at rated power. Consequently, the enthalpy thresholds presented here are lower bounds to RIAs at rated power. The calculations are performed with best-estimate models, which are applied in the FRAPCON-3.2 and SCANAIR-3.2 computer codes. Based on the results of three-dimensional core kinetics analyses, the considered power transients are simulated by a Gaussian pulse shape, with a fixed width of either 25 ms (REA) or 45 ms (CRDA). Notwithstanding the differences in postulated accident scenarios between the REA and the CRDA, the calculated core failure limits for these two events are similar. The calculated enthalpy thresholds for fuel pellet melting decrease gradually with fuel burnup, from approximately 960 J/gUO 2 at 30 MWd/kgU to 810 J/gUO 2 at 70 MWd/kgU. The decline is due to depression of the UO 2 melting temperature with increasing burnup, in combination with burnup related changes to the radial power distribution within the fuel pellets. The presented fuel enthalpy thresholds for incipient UO 2 melting provide best-estimate core failure limits for low- and intermediate-burnup fuel. However, pulse reactor tests on high-burnup fuel rods indicate that the accumulation of gaseous fission products within the pellets may lead to fuel dispersal into the coolant at
Coolability in the frame of core melt accidents in light water reactors. Model development and validation for ATHLET-CD and ASTEC. Final report; Kuehlbarkeit im Rahmen von Kernschmelzunfaellen bei Leichtwasserreaktoren. Modellentwicklung und Validierung fuer ATHLET-CD und ASTEC. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Buck, Michael; Pohlner, Georg; Rahman, Saidur; Berkhan, Ana

2015-07-15

The code system ATHLET/ATHLET-CD is being developed in the frame of the reactor safety research of the German Federal Ministry for Economic Affairs and Energy (BMWi) within the topic analysis of transients and accident sequences. It serves for simulation of transients and accidents to be used in safety analyses for light water reactors. In the present project the development and validation of models for ATHLET-CD for description of the processes during severe accidents are continued. These works should enable broad safety analyses by a mechanistic description of the processes even during late phases of a degrading core and by this a profound estimation on coolability and accident management options during every phase. With the actual status of modelling in ATHLET-CD analyses on coolability are made to give a solid base for estimates about stabilization by cooling or accident progression, dependent on the scenario. The modeling in the MEWA module, describing the processes in a severely degraded core in ATHLET-CD, is extended on the processes in the lower plenum. For this, the model on melt pool behavior is extended and linked to the RPV wall. The coupling between MEWA and the thermal-hydraulics of ATHLET-CD is improved. The validation of the models is continued by calculations on new experiments and comparing analyses done in the frame of the European Network SARNET-2. For the European integral code ASTEC contributions from the modeling for ATHLET-CD will be done, especially by providing a model for the melt behavior in the lower plenum of a LWR. This report illustrates the work carried out in the frame of this project, and shows results of calculations and the status of validation by recalculations on experiments for debris bed coolability, melt pool behavior as well as jet fragmentation and debris bed formation.
Recent Changes in Arctic Sea Ice Melt Onset, Freeze-Up, and Melt Season Length

Science.gov (United States)

Markus, Thorsten; Stroeve, Julienne C.; Miller, Jeffrey

2010-01-01

In order to explore changes and trends in the timing of Arctic sea ice melt onset and freeze-up and therefore melt season length, we developed a method that obtains this information directly from satellite passive microwave data, creating a consistent data set from 1979 through present. We furthermore distinguish between early melt (the first day of the year when melt is detected) and the first day of continuous melt. A similar distinction is made for the freeze-up. Using this method we analyze trends in melt onset and freeze-up for 10 different Arctic regions. In all regions except for the Sea of Okhotsk, which shows a very slight and statistically insignificant positive trend (O.4 days/decade), trends in melt onset are negative, i.e. towards earlier melt. The trends range from -1.0day/decade for the Bering Sea to -7.3 days/decade for the East Greenland Sea. Except for the Sea of Okhotsk all areas also show a trend towards later autumn freeze onset. The Chukchi/Beaufort Seas and Laptev/East Siberian Seas observe the strongest trends with 7 days/decade. For the entire Arctic, the melt season length has increased by about 20 days over the last 30 years. Largest trends of over 1O days/decade are seen for Hudson Bay, the East Greenland Sea the Laptev/East Siberian Seas, and the Chukchi/Beaufort Seas. Those trends are statistically significant a1 the 99% level.
Reaction of soda-lime-silica glass melt with water vapour at melting temperatures

Czech Academy of Sciences Publication Activity Database

Vernerová, Miroslava; Kloužek, Jaroslav; Němec, Lubomír

2015-01-01

Roč. 416, MAY 15 (2015), s. 21-30 ISSN 0022-3093 R&D Projects: GA TA ČR TA01010844 Institutional support: RVO:67985891 Keywords : glass melt * sulfate * water vapour * bubble nucleation * melt foaming * glass melting Subject RIV: JH - Ceramics, Fire-Resistant Materials and Glass Impact factor: 1.825, year: 2015
Distribution of siderophile and other trace elements in melt rock at the Chicxulub impact structure

Science.gov (United States)

Schuraytz, B. C.; Lindstrom, D. J.; Martinez, R. R.; Sharpton, V. L.; Marin, L. E.

1994-01-01

Recent isotopic and mineralogical studies have demonstrated a temporal and chemical link between the Chicxulub multiring impact basin and ejecta at the Cretaceous-Tertiary boundary. A fundamental problem yet to be resolved, however, is identification of the projectile responsible for this cataclysmic event. Drill core samples of impact melt rock from the Chichxulub structure contain Ir and Os abundances and Re-Os isotopic ratios indicating the presence of up to approx. 3 percent meteoritic material. We have used a technique involving microdrilling and high sensitivity instrumental neutron activation analysis (INAA) in conjunction with electron microprobe analysis to characterize further the distribution of siderophile and other trace elements among phases within the C1-N10 melt rock.
Transition metal ions in silicate melts. I. Manganese in sodium silicate melts

Energy Technology Data Exchange (ETDEWEB)

Nelson, C; White, W B

1980-01-01

Optical absorption spectra obtained on glasses quenched from sodium silicate melts show Mn/sup 3 +/ to be the dominant species for melts heated in air and Mn/sup 2 +/ to be the dominant species for melts heated at P/sub O/sub 2// = 10/sup -17/ bar. The absorption spectrum of Mn/sup 3 +/ consists of an intense band at 20,000 cm/sup -1/ with a 15,000 cm/sup -1/ satellite possibly arising from the Jahn-Teller effect. The independence of the spectrum from melt composition and the high band intensity is offered as evidence for a distinct Mn/sup 3 +/ complex in the melt. The spectrum of Mn/sup 2 +/ is weak and many expected bands are not observed. A two-band luminescence spectrum from Mn/sup 2 +/ has been tentatively interpreted as due to Mn/sup 2 +/ in interstitial sites in the network and Mn/sup 2 +/ coordiated by non-bridging oxygens.
On high-pressure melting of tantalum

Science.gov (United States)

Luo, Sheng-Nian; Swift, Damian C.

2007-01-01

The issues related to high-pressure melting of Ta are discussed within the context of diamond-anvil cell (DAC) and shock wave experiments, theoretical calculations and common melting models. The discrepancies between the extrapolations of the DAC melting curve and the melting point inferred from shock wave experiments, cannot be reconciled either by superheating or solid-solid phase transition. The failure to reproduce low-pressure DAC melting curve by melting models such as dislocation-mediated melting and the Lindemann law, and molecular dynamics and quantum mechanics-based calculations, undermines their predictions at moderate and high pressures. Despite claims to the contrary, the melting curve of Ta (as well as Mo and W) remains inconclusive at high pressures.
Transient debris freezing and potential wall melting during a severe reactivity initiated accident experiment

International Nuclear Information System (INIS)

El-Genk, M.S.; Moore, R.L.

1981-01-01

It is important to light water reactor (LWR) safety analysis to understand the transient freezing of molten core debris on cold structures following a hypothetical core meltdown accident. The purpose of this paper is to (a) present the results of a severe reactivity initiated accident (RIA) in-pile experiment with regard to molten debris distribution and freezing following test fuel rod failure, (b) analyze the transient freezing of molten debris (primarily a mixture of UO/sub 2/ fuel and Zircaloy cladding) deposited on the inner surface of the test shroud wall upon rod failure, and (c) assess the potential for wall melting upon being contacted by the molten debris. 26 refs
Retention of uranium(VI) by laumontite, a fracture-filling material of granite

International Nuclear Information System (INIS)

Baik, M.H.; Lee, S.Y.; Shon, W.J.

2009-01-01

Retention of U(VI) by laumontite, a fracture-filling material of granite as investigated by conducting dynamic and batch sorption experiments in a love-box using a granite core with a natural fracture. The hydrodynamic properties of the granite core were obtained from the elution curve of a on-sorbing tracer, Br - . The elution curve of U(VI) showed a similar behavior to Br - . This reveals that the retention of U(VI) by the fracture-filling material was not significant when migrating through the fracture at a given condition. From the dynamic sorption experiment, the retardation factor R a and the distribution coefficient K a of U(VI) were obtained as about 2.9 and 0.16 cm, respectively. The distribution coefficient K d ) of U(VI) onto laumontite obtained by conducting a batch sorption experiment resulted in a small value of 2.3±0.5 mL/g. This low K d value greed with the result of the dynamic sorption experiment. For the distribution of uranium on the granite surface investigated by an X-ray image mapping, the fracture region filled with laumontite showed a relatively lower content of uranium compared to the surrounding granite surface. Thus, the low retention of U(VI) by the fracture-filling material can be explained by following two mechanisms. One is that U(VI) exists as anionic uranyl hydroxides or uranyl carbonates at a given groundwater condition and the other is the remarkably low sorption capacity of the laumontite for U(VI). author)
TMI-2 core damage: a summary of present knowledge

International Nuclear Information System (INIS)

Owen, D.E.; Mason, R.E.; Meininger, R.D.; Franz, W.A.

1983-01-01

Extensive fuel damage (oxidation and fragmentation) has occurred and the top approx. 1.5 m of the center portion of the TMI-2 core has relocated. The fuel fragmentation extends outward to slightly beyond one-half the core radius in the direction examined by the CCTV camera. While the radial extent of core fragmentation in other directions was not directly observed, control and spider drop data and in-core instrument data suggest that the core void is roughly symmetrical, although there are a few indications of severe fuel damage extending to the core periphery. The core material fragmented into a broad range of particle sizes, extending down to a few microns. APSR movement data, the observation of damaged fuel assemblies hanging unsupported from the bottom of the reactor upper plenum structure, and the observation of once-molten stainless steel immediately above the active core indicate high temperatures (up to at least 1720 K) extended to the very top of the core. The relative lack of damage to the underside of the plenum structure implies a sharp temperature demarcation at the core/plenum interface. Filter debris and leadscrew deposit analyses indicate extensive high temperature core materials interaction, melting of the Ag-In-Cd control material, and transport of particulate control material to the plenum and out of the vessel
Hot-melt sub- and outercoating combined with enteric aqueous coating to improve the stability of aspirin tablets

Directory of Open Access Journals (Sweden)

Xiuzhi Wang

2017-05-01

Full Text Available Aspirin is apt to hydrolyze. In order to improve its stability, a new method has been developed involving the application of hot-melt sub- and outercoating combined with enteric aqueous coating. The main aim was to investigate the influence of these factors on the stability of ASA and understand how they work. Satisfactory storage stability were obtained when the aspirin tablet core coated with Eudragit L30D55 film was combined with glycerin monostearate (GMS as an outercoat. Hygroscopicity testing indicated that the moisture penetrating into the tablet may result in a significant change in the physical properties of the coating film observed by scanning electron microscopy. Investigation of the compatibility between the drug and film excipients shows that the talc and methacrylic acid had a significant catalytic effect on ASA. A hypothesis was proposed that the hydrolysis of ASA enteric coated tablets (ASA-ECT was mostly concentrated in the internal film and the interfaces between the film and tablet core. In conclusion, hot-melt coating technology is an alternative to subcoating or outercoating. Also, GMS sub-coating was a better choice for forming a stable barrier between the tablet core and the polymer coating layer, and increases the structure and chemical stability.
Optical properties of melting first-year Arctic sea ice

Science.gov (United States)

Light, Bonnie; Perovich, Donald K.; Webster, Melinda A.; Polashenski, Christopher; Dadic, Ruzica

2015-11-01

The albedo and transmittance of melting, first-year Arctic sea ice were measured during two cruises of the Impacts of Climate on the Eco-Systems and Chemistry of the Arctic Pacific Environment (ICESCAPE) project during the summers of 2010 and 2011. Spectral measurements were made for both bare and ponded ice types at a total of 19 ice stations in the Chukchi and Beaufort Seas. These data, along with irradiance profiles taken within boreholes, laboratory measurements of the optical properties of core samples, ice physical property observations, and radiative transfer model simulations are employed to describe representative optical properties for melting first-year Arctic sea ice. Ponded ice was found to transmit roughly 4.4 times more total energy into the ocean, relative to nearby bare ice. The ubiquitous surface-scattering layer and drained layer present on bare, melting sea ice are responsible for its relatively high albedo and relatively low transmittance. Light transmittance through ponded ice depends on the physical thickness of the ice and the magnitude of the scattering coefficient in the ice interior. Bare ice reflects nearly three-quarters of the incident sunlight, enhancing its resiliency to absorption by solar insolation. In contrast, ponded ice absorbs or transmits to the ocean more than three-quarters of the incident sunlight. Characterization of the heat balance of a summertime ice cover is largely dictated by its pond coverage, and light transmittance through ponded ice shows strong contrast between first-year and multiyear Arctic ice covers.
CFD-calculations to a core catcher benchmark

International Nuclear Information System (INIS)

Willschuetz, H.G.

1999-04-01

There are numerous experiments for the exploration of the corium spreading behaviour, but comparable data have not been available up to now in the field of the long term behaviour of a corium expanded in a core catcher. The difficulty consists in the experimental simulation of the decay heat that can be neglected for the short-run course of events like relocation and spreading, which must, however, be considered during investigation of the long time behaviour. Therefore the German GRS, defined together with Battelle Ingenieurtechnik a benchmark problem in order to determine particular problems and differences of CFD codes simulating an expanded corium and from this, requirements for a reasonable measurement of experiments, that will be performed later. First the finite-volume-codes Comet 1.023, CFX 4.2 and CFX-TASCflow were used. To be able to make comparisons to a finite-element-code, now calculations are performed at the Institute of Safety Research at the Forschungszentrum Rossendorf with the code ANSYS/FLOTRAN. For the benchmark calculations of stage 1 a pure and liquid melt with internal heat sources was assumed uniformly distributed over the area of the planned core catcher of a EPR plant. Using the Standard-k-ε-turbulence model and assuming an initial state of a motionless superheated melt several large convection rolls will establish within the melt pool. The temperatures at the surface do not sink to a solidification level due to the enhanced convection heat transfer. The temperature gradients at the surface are relatively flat while there are steep gradients at the ground where the no slip condition is applied. But even at the ground no solidification temperatures are observed. Although the problem in the ANSYS-calculations is handled two-dimensional and not three-dimensional like in the finite-volume-codes, there are no fundamental deviations to the results of the other codes. (orig.)
Retention of key employees in the oil field service sector

International Nuclear Information System (INIS)

Lee, S.

1999-01-01

Before 1994, Core Laboratories Canada Ltd. adopted local country benefit plans as stipulated by the government of the day. This approach meant that the company had many different benefit plans in place or in some situations no benefit plans at all, if the law of the land allowed such an approach. The company at this time viewed the lack of or minimal benefit plans as a cost saving venture. The parent company did not take onto account the effect on morale, employee retention and loyalty that these limited plans provided. A change in ownership in 1994 presented the opportunity for Core to re-assess its benefits package and introduce an incentive plan for its worldwide employees. The introduction of a pension with profits plan proved to be satisfying to employees, and the manager's incentive plan enabled the company to retain, with the exception of people who retired from the business, its entire management staff over a four year period. The stock option plan led to the retention of essential employees and reduced the turnover in this area. Discretionary bonuses succeeded in promoting recognition amongst employees as well as providing monetary reward, and the combination of benefits, incentive and stock option plans enabled the company to retain the vast majority of key employees and to entice selected individuals to the company from other organizations. 3 refs
Melt inclusions: Chapter 6

Science.gov (United States)

,; Lowenstern, J. B.

2014-01-01

Melt inclusions are small droplets of silicate melt that are trapped in minerals during their growth in a magma. Once formed, they commonly retain much of their initial composition (with some exceptions) unless they are re-opened at some later stage. Melt inclusions thus offer several key advantages over whole rock samples: (i) they record pristine concentrations of volatiles and metals that are usually lost during magma solidification and degassing, (ii) they are snapshots in time whereas whole rocks are the time-integrated end products, thus allowing a more detailed, time-resolved view into magmatic processes (iii) they are largely unaffected by subsolidus alteration. Due to these characteristics, melt inclusions are an ideal tool to study the evolution of mineralized magma systems. This chapter first discusses general aspects of melt inclusions formation and methods for their investigation, before reviewing studies performed on mineralized magma systems.
Isostructural solid-solid phase transition in monolayers of soft core-shell particles at fluid interfaces: structure and mechanics.

Science.gov (United States)

Rey, Marcel; Fernández-Rodríguez, Miguel Ángel; Steinacher, Mathias; Scheidegger, Laura; Geisel, Karen; Richtering, Walter; Squires, Todd M; Isa, Lucio

2016-04-21

We have studied the complete two-dimensional phase diagram of a core-shell microgel-laden fluid interface by synchronizing its compression with the deposition of the interfacial monolayer. Applying a new protocol, different positions on the substrate correspond to different values of the monolayer surface pressure and specific area. Analyzing the microstructure of the deposited monolayers, we discovered an isostructural solid-solid phase transition between two crystalline phases with the same hexagonal symmetry, but with two different lattice constants. The two phases corresponded to shell-shell and core-core inter-particle contacts, respectively; with increasing surface pressure the former mechanically failed enabling the particle cores to come into contact. In the phase-transition region, clusters of particles in core-core contacts nucleate, melting the surrounding shell-shell crystal, until the whole monolayer moves into the second phase. We furthermore measured the interfacial rheology of the monolayers as a function of the surface pressure using an interfacial microdisk rheometer. The interfaces always showed a strong elastic response, with a dip in the shear elastic modulus in correspondence with the melting of the shell-shell phase, followed by a steep increase upon the formation of a percolating network of the core-core contacts. These results demonstrate that the core-shell nature of the particles leads to a rich mechanical and structural behavior that can be externally tuned by compressing the interface, indicating new routes for applications, e.g. in surface patterning or emulsion stabilization.
High-pressure melting curve of KCl: Evidence against lattice-instability theories of melting

International Nuclear Information System (INIS)

Ross, M.; Wolf, G.

1986-01-01

We show that the large curvature in the T-P melting curve of KCl is the result of a reordering of the liquid to a more densely packed arrangement. As a result theories of melting, such as the instability model, which do not take into account the structure of the liquid fail to predict the correct pressure dependence of the melting curve
Dynamics of upper mantle rocks decompression melting above hot spots under continental plates

Science.gov (United States)

Perepechko, Yury; Sorokin, Konstantin; Sharapov, Victor

2014-05-01

temperature (THS) > 1900oC asthenolens size ~700 km. When THS = of 2000oC the maximum melting degree of the primitive mantle is near 40%. An increase in the TB > 1900oC the maximum degree of melting could rich 100% with the same size of decompression melting zone (700 km). We examined decompression melting above the HS having LHS = 100 km - 780 km at a TB 1850- 2100oC with the thickness of lithosphere = 100 km.It is shown that asthenolens size (Lln) does not change substantially: Lln=700 km at LHS = of 100 km; Lln= 800 km at LHS = of 780 km. In presence of asymmetry of large HS the region of advection is developed above the HS maximum with the formation of asymmetrical cell. Influence of lithospheric plate thicknesses on appearance and evolution of asthenolens above the HS were investigated for the model stepped profile for the TB ≤ of 1750oS with Lhs = 100km and maximum of THS =2350oC. With an increase of TB the Lln difference beneath lithospheric steps is leveled with retention of a certain difference to melting degrees and time of the melting appearance a top of the HS. RFBR grant 12-05-00625.
Changes in Black Carbon Deposition to Antarctica from Two Ice Core Records, A.D. 1850-2000

Science.gov (United States)

Bisiaux, Marion M.; Edward, Ross; McConnell, Joseph R.; Curran, Mark A. J.; VanOmmen, Tas D.; Smith, Andrew M.; Neumann, Thomas A.; Pasteris, Daniel R.; Penner, Joyce E.; Taylor, Kendrick

2012-01-01

Continuous flow analysis was based on a steady sample flow and in-line detection of BC and other chemical substances as described in McConnell et al. (2007). In the cold room, previously cut one meter ice core sticks of 3x3cm, are melted continuously on a heated melter head specifically designed to eliminate contamination from the atmosphere or by the external parts of the ice. The melted ice from the most inner part of the ice stick is continuously pumped by a peristaltic pump and carried to a clean lab by Teflon lines. The recorded signal is continuous, integrating a sample volume of about 0.05 mL, for which the temporal resolution depends on the speed of melting, ice density and snow accumulation rate at the ice core drilling site. For annual accumulation derived from the WAIS and Law Dome ice cores, we assumed 3.1 cm water equivalent uncertainty in each year's accumulation from short scale spatial variability (glaciological noise) which was determined from several measurements of annual accumulation in multiple parallel ice cores notably from the WAIS Divide ice core site (Banta et al., 2008) and from South Pole site (McConnell et al., 1997; McConnell et al., 2000). Refractory black carbon (rBC) concentrations were determined using the same method as in (Bisiaux et al., 2011) and adapted to continuous flow measurements as described by (McConnell et al., 2007). The technique uses a single particle intracavity laser induced incandescence photometer (SP2, Droplet Measurement Technologies, Boulder, Colorado) coupled to an ultrasonic nebulizer/desolvation (CETAC UT5000) Flow Injection Analysis (FIA). All analyses, sample preparation etc, were performed in a class 100 cleanroom using anti contamination "clean techniques". The samples were not acidified.

The results of the CCI-3 reactor material experiment investigating 2-D core-concrete interaction and debris coolability with a siliceous concrete crucible

International Nuclear Information System (INIS)

Farmer, M.T.; Basu, S.

2006-01-01

The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) program is conducting reactor material experiments and associated analysis with the objectives of resolving the ex-vessel debris coolability issue, and to 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. Despite years of international research, there are remaining uncertainties in the models that evaluate the lateral vs. axial power split during core-concrete interaction because of a lack of truly two-dimensional experiment data. As a result, there are differences in the 2-D cavity erosion predicted by codes such as MELCOR, WECHSL, and COSACO. In the continuing effort to bridge this data gap, the third in a series of large scale Core-Concrete Interaction experiments (CCI-3) has been conducted as part of the MCCI program. This test involved the interaction of a 375 kg core-oxide melt within a two-dimensional siliceous concrete crucible. The initial phase of the test was conducted under dry conditions. After a predetermined ablation depth was reached, the cavity was flooded to obtain data on the coolability of a core melt after core-concrete interaction has progressed for some time. This paper provides a summary description of the test facility and an overview of test results
Influence of core-finishing intervals on tensile strength of cast posts-and-cores luted with zinc phosphate cement

Directory of Open Access Journals (Sweden)

Michele Andrea Lopes Iglesias

2012-08-01

Full Text Available The core finishing of cast posts-and-cores after luting is routine in dental practice. However, the effects of the vibrations produced by the rotary cutting instruments over the luting cements are not well-documented. This study evaluated the influence of the time intervals that elapsed between the cementation and the core-finishing procedures on the tensile strength of cast posts-and-cores luted with zinc phosphate cement. Forty-eight bovine incisor roots were selected, endodontically treated, and divided into four groups (n = 12: GA, control (without finishing; GB, GC, and GD, subjected to finishing at 20 minutes, 60 minutes, and 24 hours after cementation, respectively. Root canals were molded, and the resin patterns were cast in copper-aluminum alloy. Cast posts-and-cores were luted with zinc phosphate cement, and the core-finishing procedures were applied according to the groups. The tensile tests were performed at a crosshead speed of 0.5 mm/min for all groups, 24 hours after the core-finishing procedures. The data were subjected to one-way analysis of variance (ANOVA and Tukey's test (α = 0.05. No significant differences were observed in the tensile strengths between the control and experimental groups, regardless of the time interval that elapsed between the luting and finishing steps. Within the limitations of the present study, it was demonstrated that the core-finishing procedures and time intervals that elapsed after luting did not appear to affect the retention of cast posts-and-cores when zinc phosphate cement was used.
Annual layering in the NGRIP ice core during the Eemian

Directory of Open Access Journals (Sweden)

A. Svensson

2011-12-01

Full Text Available The Greenland NGRIP ice core continuously covers the period from present day back to 123 ka before present, which includes several thousand years of ice from the previous interglacial period, MIS 5e or the Eemian. In the glacial part of the core, annual layers can be identified from impurity records and visual stratigraphy, and stratigraphic layer counting has been performed back to 60 ka. In the deepest part of the core, however, the ice is close to the pressure melting point, the visual stratigraphy is dominated by crystal boundaries, and annual layering is not visible to the naked eye. In this study, we apply a newly developed setup for high-resolution ice core impurity analysis to produce continuous records of dust, sodium and ammonium concentrations as well as conductivity of melt water. We analyzed three 2.2 m sections of ice from the Eemian and the glacial inception. In all of the analyzed ice, annual layers can clearly be recognized, most prominently in the dust and conductivity profiles. Part of the samples is, however, contaminated in dust, most likely from drill liquid. It is interesting that the annual layering is preserved despite a very active crystal growth and grain boundary migration in the deep and warm NGRIP ice. Based on annual layer counting of the new records, we determine a mean annual layer thickness close to 11 mm for all three sections, which, to first order, confirms the modeled NGRIP time scale (ss09sea. The counting does, however, suggest a longer duration of the climatically warmest part of the NGRIP record (MIS5e of up to 1 ka as compared to the model estimate. Our results suggest that stratigraphic layer counting is possible basically throughout the entire NGRIP ice core, provided sufficiently highly-resolved profiles become available.
The Analysis of Surrounding Structure Effect on the Core Degradation Progress with COMPASS Code

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

In line with the importance of severe accident analysis after Fukushima accident, the development of integrated severe accident code has been launched by the collaboration of three institutes in Korea. KAERI is responsible to develop modules related to the in-vessel phenomena, while other institutes are to the containment and severe accident mitigation facility, respectively. In the first phase, the individual severe accident module has been developed and the construction of integrated analysis code is planned to perform in the second phase. The basic strategy is to extend the design basis analysis codes of SPACE and CAP, which are being validated in Korea for the severe accident analysis. In the first phase, KAERI has targeted to develop the framework of severe accident code, COMPASS (COre Meltdown Progression Accident Simulation Software), covering the severe accident progression in a vessel from a core heat-up to a vessel failure as a stand-alone fashion. In order to analyze the effect of surrounding structure, the melt progression has been compared between the central zone and the most outer zone under the condition of constant radial power peaking factor. Figure 2 and 3 shows the fuel element temperature and the clad mass at the central zone, respectively. Due to the axial power peaking factor, the axial node No.3 has the highest temperature, while the top and bottom nodes have the lowest temperature. When the clad temperature reaches to the Zr melting temperature (2129.15K), the Zr starts to melt. The axial node No.2 reaches to the fuel melting temperature about 5000 sec and the molten fuel relocates to the node No.1, which results to the blockage of flow area in node No.1. The blocked flow area becomes to open about 6100 sec due to the molten ZrO{sub 2} mass relocation to core support plate. Figure 4 and 5 shows the fuel element temperature and the clad mass at the most outer zone, respectively. It is shown that the fuel temperature increase more slowly
Eruption Depths, Magma Storage and Magma Degassing at Sumisu Caldera, Izu-Bonin Arc: Evidence from Glasses and Melt Inclusions

Science.gov (United States)

Johnson, E. R.

2015-12-01

Island arc volcanoes can become submarine during cataclysmal caldera collapse. The passage of a volcanic vent from atmospheric to under water environment involves complex modifications of the eruption style and subsequent transport of the pyroclasts. Here, we use FTIR measurements of the volatile contents of glass and melt inclusions in the juvenile pumice clasts in the Sumisu basin and its surroundings (Izu-Bonin arc) to investigate changes in eruption depths, magma storage and degassing over time. This study is based on legacy cores from ODP 126, where numerous unconsolidated (250 m), massive to normally graded pumice lapilli-tuffs were recovered over four cores (788C, 790A, 790B and 791A). Glass and clast geochemistry indicate the submarine Sumisu caldera as the source of several of these pumice lapilli-tuffs. Glass chips and melt inclusions from these samples were analyzed using FTIR for H2O and CO2 contents. Glass chips record variable H2O contents; most chips contain 0.6-1.6 wt% H2O, corresponding to eruption depths of 320-2100 mbsl. Variations in glass H2O and pressure estimates suggest that edifice collapse occurred prior-to or during eruption of the oldest of these samples, and that the edifice may have subsequently grown over time. Sanidine-hosted melt inclusions from two units record variably degassed but H2O-rich melts (1.1-5.6 wt% H2O). The lowest H2O contents overlap with glass chips, consistent with degassing and crystallization of melts until eruption, and the highest H2O contents suggest that large amounts of degassing accompanied likely explosive eruptions. Most inclusions, from both units, contain 2-4 wt% H2O, which further indicates that the magmas crystallized at pressures of ~50-100 MPa, or depths ~400-2800 m below the seafloor. Further glass and melt inclusion analyses, including major element compositions, will elucidate changes in magma storage, degassing and evolution over time.
Electron beam melting of sponge titanium

International Nuclear Information System (INIS)

Kanayama, Hiroshi; Kusamichi, Tatsuhiko; Muraoka, Tetsuhiro; Onouye, Toshio; Nishimura, Takashi

1991-01-01

Fundamental investigations were done on electron beam (EB) melting of sponge titanium by using 80 kW EB melting furnace. Results obtained are as follows: (1) To increase the melting yield of titanium in EB melting of sponge titanium, it is important to recover splashed metal by installation of water-cooled copper wall around the hearth and to decrease evaporation loss of titanium by keeping the surface temperature of molten metal just above the melting temperature of titanium without local heating. (2) Specific power consumption of drip melting of pressed sponge titanium bar and hearth melting of sponge titanium are approximately 0.9 kWh/kg-Ti and 0.5-0.7 kWh/kg-Ti, respectively. (3) Ratios of the heat conducted to water-cooled mould in the drip melting and to water-cooled hearth in the hearth melting to the electron beam input power are 50-65% and 60-65%, respectively. (4) Surface defects of EB-melted ingots include rap which occurs when the EB output is excessively great, and transverse cracks when the EB output is excessively small. To prevent surface defects, the up-down withdrawal method is effective. (author)
Design and preliminary analysis of in-vessel core catcher made of high-temperature ceramics material in PWR

International Nuclear Information System (INIS)

Xu Hong; Ma Li; Wang Junrong; Zhou Zhiwei

2011-01-01

In order to protect the interior wall of pressure vessel from melting, as an additional way to external reactor vessel cooling (ERVC), a kind of in-vessel core catcher (IVCC) made of high-temperature ceramics material was designed. Through the high-temperature and thermal-resistance characteristic of IVCC, the distributing of heat flux was optimized. The results show that the downward average heat flux from melt in ceramic layer reduces obviously and the interior wall of pressure vessel doesn't melt, keeping its integrity perfectly. Increasing of upward heat flux from metallic layer makes the upper plenum structure's temperature ascend, but the temperature doesn't exceed its melting point. In conclusion, the results indicate the potential feasibility of IVCC made of high-temperature ceramics material. (authors)
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
Melting relations in the Fe-rich portion of the system FeFeS at 30 kb pressure

Science.gov (United States)

Brett, R.; Bell, P.M.

1969-01-01

The melting relations of FeFeS mixtures covering the composition range from Fe to Fe67S33 have been determined at 30 kb pressure. The phase relations are similar to those at low pressure. The eutectic has a composition of Fe72.9S27.1 and a temperature of 990??C. Solubility of S in Fe at elevated temperatures at 30 kb is of the same order of magnitude as at low pressure. Sulfur may have significantly lowered the melting point of iron in the upper mantle during the period of coalescence of metal prior to core formation in the primitive earth. ?? 1969.
Second OECD (NEA) CSNI specialist meeting on molten core debris-concrete interactions

International Nuclear Information System (INIS)

Alsmeyer, H.

1992-11-01

The 37 contributions concentrated on two main topics. The first topic is the 'classical' core debris-concrete interaction, both experimental and theoretical. Integral effects and separate effects were addressed in thermal hydraulics and heat transfer, material interaction, and aerosol release during concrete erosion, with some applications to prototypical nuclear power plants. The second topic is the possibility of controlling and ending the erosion of the concrete by spreading of the core melt, and/or achieving coolability by the addition of water. (orig./HP) [de
Analysis and optimisation of vertical surface roughness in micro selective laser melting

International Nuclear Information System (INIS)

Abele, Eberhard; Kniepkamp, Michael

2015-01-01

Surface roughness is a major disadvantage of many additive manufacturing technologies like selective laser melting (SLM) compared to established processes like milling or drilling. With recent advancements the resolution of the SLM process could be increased to layer heights of less than 10 μm leading to a new process called micro selective laser melting (μSLM). The purpose of this paper is to analyze the influence of the μSLM process parameters and exposure strategies on the morphology of vertical surfaces. Contour scanning using varying process parameters was used to increase the surface quality. It is shown that it is possible to achieve average surface roughness of less than 1.7 μm using low scan speeds compared to 8–10 μm without contour scanning. Furthermore it is shown that a contour exposure prior to the core exposure leads to surface defects and thus increased roughness. (paper)
Melting point of yttria

International Nuclear Information System (INIS)

Skaggs, S.R.

1977-06-01

Fourteen samples of 99.999 percent Y 2 O 3 were melted near the focus of a 250-W CO 2 laser. The average value of the observed melting point along the solid-liquid interface was 2462 +- 19 0 C. Several of these same samples were then melted in ultrahigh-purity oxygen, nitrogen, helium, or argon and in water vapor. No change in the observed temperature was detected, with the exception of a 20 0 C increase in temperature from air to helium gas. Post test examination of the sample characteristics, clarity, sphericity, and density is presented, along with composition. It is suggested that yttria is superior to alumina as a secondary melting-point standard
TMI-2 core debris grab samples: Examination and analysis: Part 1

International Nuclear Information System (INIS)

Akers, D.W.; Carlson, E.R.; Cook, B.A.; Ploger, S.A.; Carlson, J.O.

1986-09-01

Six samples of particulate debris were removed from the TMI-2 core rubble bed during September and October 1983, and five more samples were obtained in March 1984. The samples (up to 174 g each) were obtained at two locations in the core: H8 (center) and E9 (mid-radius). Ten of the eleven samples were examined at the Idaho National Engineering Laboratory to obtain data on the physical and chemical nature of the debris and the postaccident condition of the core. Portions of the samples also were subjected to differential thermal analysis at Rockwell Hanford Operations and metallurgical and chemical examinations at Argonne National Laboratories. This report presents results of the examination of the core debris grab samples, including physical, metallurgical, chemical, and radiochemical analyses. The results indicate that temperatures in the core reached at least 3100 K during the TMI-2 accident, fuel melting and significant mixing of core structural material occurred, and large fractions of some radionuclides (e.g., 90 Sr and 144 Ce) were retained in the core
Chemical and X-ray diffraction analysis on selected samples from the TMI-2 reactor core

International Nuclear Information System (INIS)

Kleykamp, H.; Pejsa, R.

1991-05-01

Selected samples from different positions of the damaged TMI-2 reactor core were investigated by X-ray microanalysis and X-ray diffraction. The measurements yield the following resolidified phases after cooling: Cd and In depleted Ag absorber material, intermetallic Zr-steel compounds, fully oxidized Zircaloy, UO 2 -ZrO 2 solid solutions and their decomposed phases, and Fe-Al-Cr-Zr spinels. The composition of the phases and their lattice parameters as well as the eutectic and monotectic character can serve as indicators of local temperatures of the core. The reaction sequences are estimated from the heterogeneous equilibria of these phases. The main conclusions are: (1) Liquefaction onset is locally possible by Inconel-Zircaloy and steel-Zircaloy reactions of spacers and absorber guide tubes at 930deg C. However, increased rates of dissolution occur above 1200deg C. (2) UO 2 dissolution in the Inconel-steel-Zircaloy melt starts at 1300deg C with increased rates above 1900deg C. (3) Fuel temperatures in the core centre are increased above 2550deg C, liquid (U,Zr)O 2 is generated. (4) Square UO 2 particles are reprecipitated from the Incoloy-steel-Zircaloy-UO 2 melt during cooling, the remaining metallic melt is oxygen poor; two types of intermetallic phases are formed. (5) Oxidized Fe and Zr and Al 2 O 3 from burnable absorber react to spinels which form a low melting eutectic with the fuel at 1500deg C. The spinel acts as lubricant for fuel transport to the lower reactor plenum above 1500deg C. (6) Ruthenium (Ru-106) is dissolved in the steel phase, antimony (Sb-125) in the α-Ag absorber during liquefaction. (7) Oxidation of the Zircaloy-steel phases takes place mainly in the reflood stage 3 of the accident scenario. (orig.) [de
Continuous Improvement: A Way of Integrating Student Enrollment, Advising, and Retention Systems in a Metropolitan University.

Science.gov (United States)

Beeler, Karl J.; Moehl, Pamela J.

1996-01-01

The University of Missouri-St. Louis has discovered the value of continuous quality improvement methods in upgrading its core student-related administrative processes. As a result, it is increasing efficiency and personalizing a traditionally bureaucratic system of student service. Concurrent goals are to increase retention and decrease time to…
Method of melting solid waste

International Nuclear Information System (INIS)

Ootsuka, Katsuyuki; Mizuno, Ryokichi; Kuwana, Katsumi; Sawada, Yoshihisa; Komatsu, Fumiaki.

1982-01-01

Purpose: To enable the volume reduction treatment of a HEPA filter containing various solid wastes, particularly acid digestion residue, or an asbestos separator at a relatively low temperature range. Method: Solid waste to be heated and molten is high melting point material treated by ''acid digestion treatment'' for treating solid waste, e.g. a HEPA filter or polyvinyl chloride, etc. of an atomic power facility treated with nitric acid or the like. When this material is heated and molten by an electric furnace, microwave melting furnace, etc., boron oxide, sodium boride, sodium carbonate, etc. is added as a melting point lowering agent. When it is molten in this state, its melting point is lowered, and it becomes remarkably fluid, and the melting treatment is facilitated. Solidified material thus obtained through the melting step has excellent denseness and further large volume reduction rate of the solidified material. (Yoshihara, H.)
Tests on the release of fission and activation products during core meltdown

Energy Technology Data Exchange (ETDEWEB)

Albrecht, H; Krause, W; Wild, H

1976-01-01

The first available results are related to tests in which the release of the main components of the core melt, namely the steel, zircaloy and uranium components, was determined using ThO/sub 2/ crucibles. The release products are dispersed onto the pipe walls of the transport system and the measuring filters which were installed at about 1 m distance from the melt crucibles. Of these, only the precipitates on the filters have been analyzed so far. In the tests under air, the release was clearly dependent on the maximum temperature reached. The release values for Mo and Mn were the highest with 5-10%; uranium with 0.1% on the other hand, was the lowest. In a steam atmosphere over the melt, the analysis of the filter precipitates for all elements gave considerably lower values than with the tests in air.
Rapidly disintegrating vagina retentive cream suppositories of progesterone: development, patient satisfaction and in vitro/in vivo studies.

Science.gov (United States)

Bendas, Ehab Rasmy; Basalious, Emad B

2016-01-01

Our objective was to develop novel vagina retentive cream suppositories (VRCS) of progesterone having rapid disintegration and good vaginal retention. VRCS of progesterone were prepared using oil in water (o/w) emulsion of mineral oil or theobroma oil in hard fat and compared with conventional vaginal suppositories (CVS) prepared by hard fat. VRCS formulations were tested for content uniformity, disintegration, melting range, in vitro release and stability studies. The most stable formulation (VRCS I) was subjected to scaling-up manufacturing and patients' satisfaction test. The rapid disintegration, good retentive properties are applicable through the inclusion of emulsified theobroma oil rather than hydrophilic surfactant into the hard fat bases. The release profile of progesterone from VRCS I showed a biphasic pattern due to the formation of progesterone reservoir in the emulsified theobroma oil. All volunteers involved in patients' satisfaction test showed high satisfactory response to the tested formulation (VRCS). The in vivo pharmacokinetic study suggests that VRCS of progesterone provided higher rate and extent of absorption compared to hard fat based suppositories. Our results proposed that emulsified theobroma oil could be promising to solve the problems of poor patients' satisfaction and variability of drug absorption associated with hard fat suppositories.
Melting of KCl and pressure calibration from in situ ionic conductivity measurements in a multi-anvil apparatus

Science.gov (United States)

Li, J.; Dong, J.; Zhu, F.

2017-12-01

Melting plays an unparalleled role in planetary differentiation processes including the formation of metallic cores, basaltic crusts, and atmospheres. Knowledge of the melting behavior of Earth materials provides critical constraints for establishing the Earth's thermal structure, interpreting regional seismic anomalies, and understanding the nature of chemical heterogeneity. Measuring the melting points of compressed materials, however, have remained challenging mainly because melts are often mobile and reactive, and temperature and pressure gradients across millimeter or micron-sized samples introduce large uncertainties in melting detection. Here the melting curve of KCl was determined through in situ ionic conductivity measurements, using the multi-anvil apparatus at the University of Michigan. The method improves upon the symmetric configuration that was used recently for studying the melting behaviors of NaCl, Na2CO3, and CaCO3 (Li and Li 2015 American Mineralogist, Li et al. 2017 Earth and Planetary Science Letters). In the new configuration, the thermocouple and electrodes are placed together with the sample at the center of a cylindrical heater where the temperature is the highest along the axis, in order to minimize uncertainties in temperature measurements and increase the stability of the sample and electrodes. With 1% reproducibility in melting point determination at pressures up to 20 GPa, this method allows us to determine the sample pressure to oil load relationship at high temperatures during multiple heating and cooling cycles, on the basis of the well-known melting curves of ionic compounds. This approach enables more reliable pressure measurements than relying on a small number of fixed-point phase transitions. The new data on KCl bridge the gap between the piston-cylinder results up to 4 GPa (Pistorius 1965 J. of Physics and Chemistry of Solids) and several diamond-anvil cell data points above 20 GPa (Boehler et al. 1996 Physical Review). We
Thermodynamics of Oligonucleotide Duplex Melting

Science.gov (United States)

Schreiber-Gosche, Sherrie; Edwards, Robert A.

2009-01-01

Melting temperatures of oligonucleotides are useful for a number of molecular biology applications, such as the polymerase chain reaction (PCR). Although melting temperatures are often calculated with simplistic empirical equations, application of thermodynamics provides more accurate melting temperatures and an opportunity for students to apply…

Automatic Control of Silicon Melt Level

Science.gov (United States)

Duncan, C. S.; Stickel, W. B.

1982-01-01

A new circuit, when combined with melt-replenishment system and melt level sensor, offers continuous closed-loop automatic control of melt-level during web growth. Installed on silicon-web furnace, circuit controls melt-level to within 0.1 mm for as long as 8 hours. Circuit affords greater area growth rate and higher web quality, automatic melt-level control also allows semiautomatic growth of web over long periods which can greatly reduce costs.
Fluid Core Size of Mars from Detection of the Solar Tide

Science.gov (United States)

Yoder, C. F.; Konopliv, A. S.; Yuan, D. N.; Standish, E. M.; Folkner, W. M.

2003-04-01

The solar tidal deformation of Mars, measured by its k2 potential Love number, has been obtained from an analysis of Mars Global Surveyor radio tracking. The observed k2 of 0.153 +/- 0.017 is large enough to rule out a solid iron core and so indicates that at least the outer part of the core is liquid. The inferred core radius is between 1520 and 1840 kilometers and is independent of many interior properties, although partial melt of the mantle is one factor that could reduce core size. Ice-cap mass changes can be deduced from the seasonal variations in air pressure and the odd gravity harmonic J3, given knowledge of cap mass distribution with latitude. The south cap seasonal mass change is about 30 to 40% larger than that of the north cap.
Using high resolution tritium profiles to quantify the effects of melt on two Spitsbergen ice cores

NARCIS (Netherlands)

van der Wel, L.G.; Streurman, H.J.; Isaksson, E.; Helsen, M.M.; van de Wal, R.S.W.; Martma, T.; Pohjola, V.A.; Moore, J.C.; Meijer, H.A.J.

2011-01-01

Ice cores from small ice caps provide valuable climatic information, additional to that of Greenland and Antarctica. However, their integrity is usually compromised by summer meltwater percolation. To determine to what extent this can affect such ice cores, we performed high-resolution tritium
Using high-resolution tritium profiles to quantify the effects of melt on two Spitsbergen ice cores

NARCIS (Netherlands)

Wel, L.G. van der; Streurman, H.J.; Isaksson, E.; Helsen, M.M.; Wal, R.S.W. van de; Martma, T.; Pohjola, V.A.; Moore, J.C.; Meijer, H.A.J.

2011-01-01

Ice cores from small ice caps provide valuable climatic information, additional to that of Greenland and Antarctica. However, their integrity is usually compromised by summer meltwater percolation. To determine to what extent this can affect such ice cores, we performed high-resolution tritium
Melting of polydisperse hard disks

NARCIS (Netherlands)

Pronk, S.; Frenkel, D.

2004-01-01

The melting of a polydisperse hard-disk system is investigated by Monte Carlo simulations in the semigrand canonical ensemble. This is done in the context of possible continuous melting by a dislocation-unbinding mechanism, as an extension of the two-dimensional hard-disk melting problem. We find
HIV Stigma, Retention in Care, and Adherence Among Older Black Women Living With HIV.

Science.gov (United States)

Sangaramoorthy, Thurka; Jamison, Amelia M; Dyer, Typhanye V

Stigma is recognized as a barrier to the prevention, care, and treatment of HIV, including engagement in the HIV care continuum. HIV stigma in older Black women may be compounded by preexisting social inequities based on gender, age, and race. Using semi-structured interviews and survey questionnaires, we explore experiences of HIV stigma, retention in care, and antiretroviral therapy (ART) adherence in 35 older Black women with HIV from Prince George's County, Maryland. Study findings indicated that older Black women experienced high levels of HIV stigma, retention in care, and ART adherence. Findings suggest that experiences of HIV stigma were intensified for older Black women due to multiple stigmatized social positions. Participants also reported experiences of marginalization in health care that hindered retention in care and ART adherence. Interventions aimed at improving HIV prevention, care, and treatment outcomes should incorporate HIV stigma reduction strategies as core elements. Copyright © 2017 Association of Nurses in AIDS Care. Published by Elsevier Inc. All rights reserved.
A comparison of core degradation phenomena in the CORA, QUENCH, Phébus SFD and Phébus FP experiments

Energy Technology Data Exchange (ETDEWEB)

Haste, T., E-mail: tim.haste@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, IRSN, BP 3, F-13115 St. Paul-lez-Durance Cedex (France); Steinbrück, M., E-mail: martin.steinbrueck@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Barrachin, M., E-mail: marc.barrachin@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, IRSN, BP 3, F-13115 St. Paul-lez-Durance Cedex (France); Luze, O. de, E-mail: olivier.de-luze@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, IRSN, BP 3, F-13115 St. Paul-lez-Durance Cedex (France); Grosse, M., E-mail: mirco.grosse@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Stuckert, J., E-mail: juri.stuckert@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)

2015-03-15

Highlights: • The results of the experiments CORA, QUENCH and Phébus SFD/FP are summarised. • All phenomena expected up to melt movement to the lower head are shown consistently. • Separate-effect tests performed at KIT and IRSN aid improve their modelling. • Data from the integral tests help independent validation of new and improved models. • The improved codes will help reduce uncertainties in safety-critical areas for core degradation. - Abstract: Over the past 20 years, integral fuel bundle experiments performed at IRSN Cadarache, France (Phébus-SFD and Phébus FP – fission heated) and at Karlsruhe Institute of Technology, Germany (CORA and QUENCH – electrically heated), accompanied by separate-effect tests, have provided a wealth of detailed information on core degradation phenomena that occur under severe accident conditions, relevant to such safety issues as in-vessel retention of the core, recovery of the core by water reflood, hydrogen generation and fission product release. These data form an important basis for development and validation of severe accident analysis codes such as ASTEC (IRSN/GRS, EC) and MELCOR (USNRC/SNL, USA) that are used to assess the safety of current and future reactor designs, so helping to reduce the uncertainty associated with such code predictions. Following the recent end of the Phébus FP project, it is appropriate now to compare the core degradation phenomena observed in these four major experimental series, indicating the main conclusions that have been drawn. This covers subjects such as early phase degradation up to loss of rod-like geometry (all the series), late phase degradation and the link between fission product release and core degradation (Phébus FP), oxidation phenomena (all the series), reflood behaviour (CORA and QUENCH), as well as particular topics such as the effects of control rod material and fuel burn-up on core degradation. It also outlines the separate-effects experiments performed to
Slab and Sediment Melting during Subduction Initiation: Mantle Plagiogranites from the Oman Ophiolite

Science.gov (United States)

Rollinson, H. R.

2014-12-01

Granitoid dykes up to several hundred metres wide and 2 km long are found in depleted harzburgites in the mantle section of the Oman ophiolite. They vary in composition from tonalite to potassic granite and are generally more potassic than the crustal plagiogranites found within the sheeted dyke complex higher up within the ophiolite stratigraphy. Some granites are strongly peraluminous and contain garnet and andalusite. They are geochemically variable, some with REE that are relatively unfractionated ((La/Yb)n= 3.5-6.0, flat middle to heavy REE, steep light REE) to those which are highly fractionated ((La/Yb)n= 28-220). On primitive-mantle normalised plots some have very high concentrations of fluid-mobile elements - Cs, Rb, Th, U and Pb. Few have significant Ta-Nb anomalies. On the Ca-Fe-Mg-Ti discrimination diagram of Patino Douce (J. Petrol., 1999) whole-rock compositions define a spectrum between felsic-pelite derived melts and amphibolite-derived melts. There is a chemical similarity between the least REE fractionated plagiogranites (generally tonalites and granodiorites) and melts of an amphibolitic parent. This is supported by the occurrence of mafic xenoliths in some dykes, the presence of hornblende and highly calcic cores (up to An85) in some plagioclase grains. Trace element modelling using Oman Geotimes lavas as the starting composition indicates that melting took place in the garnet stability field, although enrichment in the melt in Cs, Rb, Ba and Pb suggests that there was another component present in addition to the mafic parent. Other plagiogranites (trondhjemites and granites) have a strongly peraluminous chemistry and mineralogy and geochemical similarities with the Himalayan leucogranites implying that they were derived from a sedimentary protolith. These mantle plagiogranites are more prevalent in the northern outcrops of the ophiolite. The volume of granitoid melt and the depth of melting preclude their derivation from the sole of the
Clinic teaching made easy: a prospective study of the American Academy of Dermatology core curriculum in primary care learners.

Science.gov (United States)

McCleskey, Patrick E

2013-08-01

Dermatology instruction for primary care learners is limited, and the American Academy of Dermatology (AAD) has developed a new core curriculum for dermatology. This study sought to prospectively evaluate short-term knowledge acquisition and long-term knowledge retention after using the AAD core curriculum during a clinical dermatology clerkship. Resident physicians and physician assistant students performing clerkships at military dermatology clinics were given access to the AAD core curriculum teaching modules before their public availability. Knowledge acquisition was measured with pretests and posttests, and a follow-up quiz was given up to a year after the dermatology rotation to assess knowledge retention. In all, 82 primary care learners met inclusion criteria. Knowledge improved significantly from pretest to posttest (60.1 vs 77.4, P dermatology clerkship. Copyright © 2012 American Academy of Dermatology, Inc. Published by Mosby, Inc. All rights reserved.
Studies on core melt behaviour in a BWR pressure vessel lower head

International Nuclear Information System (INIS)

Lindholm, I.; Ikonen, K.; Hedberg, K.

1999-01-01

Core debris behaviour in the Nordic BWR lower head was investigated numerically using MELCOR and MAAP4 codes. Lower head failure due to penetration failure was studied with more detailed PASULA code taking thermal boundary conditions from MELCOR calculations. Creep rupture failure mode was examined with the two integral codes. Also, the possibility to prevent vessel failure by late reflooding was assessed in this study. (authors)
Model of interfacial melting

DEFF Research Database (Denmark)

Mouritsen, Ole G.; Zuckermann, Martin J.

1987-01-01

A two-dimensional model is proposed to describe systems with phase transitions which take place in terms of crystalline as well as internal degrees of freedom. Computer simulation of the model shows that the interplay between the two sets of degrees of freedom permits observation of grain-boundar......-boundary formation and interfacial melting, a nonequilibrium process by which the system melts at the boundaries of a polycrystalline domain structure. Lipid membranes are candidates for systems with pronounced interfacial melting behavior....
OECD MCCI project long-term 2-D molten core concrete interaction test design report, Rev. 0. September 30, 2002

International Nuclear Information System (INIS)

Farmer, M.T.; Kilsdonk, D.J.; Lomperski, S.; Aeschliman, R.W.; Basu, S.

2011-01-01

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
The Results of the CCI-3 Reactor Material Experiment Investigating 2-D Core-Concrete Interaction and Debris Coolability with a Siliceous Concrete Crucible

International Nuclear Information System (INIS)

Farmer, M.T.; Lomperski, S.; Basu, S.

2006-01-01

The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) program conducted reactor materials experiments and associated analysis to achieve the following two objectives: 1) resolve the ex-vessel debris coolability issue, 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 of future plants. With respect to the second objective, there are remaining uncertainties in the models that evaluate the lateral vs. axial power split during core-concrete interaction because of a lack of truly two-dimensional experiment data. As a result, there are differences in the 2-D cavity erosion profiles predicted by codes such as WECHSL, COSACO, TOLBIAC, MEDICIS, and MELCOR. In the continuing effort to bridge this data gap, the third in a series of large scale Core-Concrete Interaction experiments (CCI-3) has been conducted as part of the MCCI program. This test investigated the long-term interaction of a 375 kg core-oxide melt within a two-dimensional siliceous concrete crucible. The initial phase of the test was conducted under dry conditions. After a predetermined time interval, the cavity was flooded with water to obtain data on the coolability of a core melt after core-concrete interaction has progressed for some time. This paper provides a description of the facility and an overview of results from this test. (authors)
Non-chondritic iron isotope ratios in planetary mantles as a result of core formation

Science.gov (United States)

Elardo, Stephen M.; Shahar, Anat

2017-02-01

Information about the materials and conditions involved in planetary formation and differentiation in the early Solar System is recorded in iron isotope ratios. Samples from Earth, the Moon, Mars and the asteroid Vesta reveal significant variations in iron isotope ratios, but the sources of these variations remain uncertain. Here we present experiments that demonstrate that under the conditions of planetary core formation expected for the Moon, Mars and Vesta, iron isotopes fractionate between metal and silicate due to the presence of nickel, and enrich the bodies' mantles in isotopically light iron. However, the effect of nickel diminishes at higher temperatures: under conditions expected for Earth's core formation, we infer little fractionation of iron isotopes. From our experimental results and existing conceptual models of magma ocean crystallization and mantle partial melting, we find that nickel-induced fractionation can explain iron isotope variability found in planetary samples without invoking nebular or accretionary processes. We suggest that near-chondritic iron isotope ratios of basalts from Mars and Vesta, as well as the most primitive lunar basalts, were achieved by melting of isotopically light mantles, whereas the heavy iron isotope ratios of terrestrial ocean floor basalts are the result of melting of near-chondritic Earth mantle.
Assessment of Mass Fraction and Melting Temperature for the Application of Limestone Concrete and Siliceous Concrete to Nuclear Reactor Basemat Considering Molten Core–Concrete Interaction

Directory of Open Access Journals (Sweden)

Hojae Lee

2016-04-01

Full Text Available Severe accident scenarios in nuclear reactors, such as nuclear meltdown, reveal that an extremely hot molten core may fall into the nuclear reactor cavity and seriously affect the safety of the nuclear containment vessel due to the chain reaction caused by the reaction between the molten core and concrete. This paper reports on research focused on the type and amount of vapor produced during the reaction between a high-temperature molten core and concrete, as well as on the erosion rate of concrete and the heat transfer characteristics at its vicinity. This study identifies the mass fraction and melting temperature as the most influential properties of concrete necessary for a safety analysis conducted in relation to the thermal interaction between the molten core and the basemat concrete. The types of concrete that are actually used in nuclear reactor cavities were investigated. The H2O content in concrete required for the computation of the relative amount of gases generated by the chemical reaction of the vapor, the quantity of CO2 necessary for computing the cooling speed of the molten core, and the melting temperature of concrete are evaluated experimentally for the molten core–concrete interaction analysis.
Can Nano-Particle Melt below the Melting Temperature of Its Free Surface Partner?

International Nuclear Information System (INIS)

Sui Xiao-Hong; Qin Shao-Jing; Wang Zong-Guo; Kang Kai; Wang Chui-Lin

2015-01-01

The phonon thermal contribution to the melting temperature of nano-particles is inspected. The discrete summation of phonon states and its corresponding integration form as an approximation for a nano-particle or for a bulk system have been analyzed. The discrete phonon energy levels of pure size effect and the wave-vector shifts of boundary conditions are investigated in detail. Unlike in macroscopic thermodynamics, the integration volume of zero-mode of phonon for a nano-particle is not zero, and it plays an important role in pure size effect and boundary condition effect. We find that a nano-particle will have a rising melting temperature due to purely finite size effect; a lower melting temperature bound exists for a nano-particle in various environments, and the melting temperature of a nano-particle with free boundary condition reaches this lower bound. We suggest an easy procedure to estimation the melting temperature, in which the zero-mode contribution will be excluded, and only several bulk quantities will be used as input. We would like to emphasize that the quantum effect of discrete energy levels in nano-particles, which is not present in early thermodynamic studies on finite size corrections to melting temperature in small systems, should be included in future researches. (condensed matter: structural, mechanical, and thermal properties)
Enhanced ice sheet melting driven by volcanic eruptions during the last deglaciation.

Science.gov (United States)

Muschitiello, Francesco; Pausata, Francesco S R; Lea, James M; Mair, Douglas W F; Wohlfarth, Barbara

2017-10-24

Volcanic eruptions can impact the mass balance of ice sheets through changes in climate and the radiative properties of the ice. Yet, empirical evidence highlighting the sensitivity of ancient ice sheets to volcanism is scarce. Here we present an exceptionally well-dated annual glacial varve chronology recording the melting history of the Fennoscandian Ice Sheet at the end of the last deglaciation (∼13,200-12,000 years ago). Our data indicate that abrupt ice melting events coincide with volcanogenic aerosol emissions recorded in Greenland ice cores. We suggest that enhanced ice sheet runoff is primarily associated with albedo effects due to deposition of ash sourced from high-latitude volcanic eruptions. Climate and snowpack mass-balance simulations show evidence for enhanced ice sheet runoff under volcanically forced conditions despite atmospheric cooling. The sensitivity of past ice sheets to volcanic ashfall highlights the need for an accurate coupling between atmosphere and ice sheet components in climate models.
Chemical interactions of reactor core materials up to very high temperatures

International Nuclear Information System (INIS)

Hofmann, P.; Hagen, S.; Schanz, G.; Skokan, A.

1989-01-01

The paper describes which chemical interactions may occur in a LWR fuel rod bundle containing (Ag, In, Cd) absorber rods or (Al 2 O 3 /B 4 C) burnable poison rods with increasing temperature up to the complete melting of the components and the formed reaction products. The kinetics of the most important chemical interactions has been investigated and the results are described. In most cases the reaction products have lower melting points or ranges than the original components. This results in a relocation of liquefied components often far below their melting points. There exist three distinct temperature regimes in which liquid phases can form in the core in differently large quantities. These temperature regimes are described in detail. The phase relations in the important ternary (U, Zr, O) system have been extensively studied. The effect of steel constituents on the phase relations is given in addition. All the considerations are focused on PWR conditions only. (orig.) [de
ESFR core optimization and uncertainty studies

International Nuclear Information System (INIS)

Rineiski, A.; Vezzoni, B.; Zhang, D.; Marchetti, M.; Gabrielli, F.; Maschek, W.; Chen, X.-N.; Buiron, L.; Krepel, J.; Sun, K.; Mikityuk, K.; Polidoro, F.; Rochman, D.; Koning, A.J.; DaCruz, D.F.; Tsige-Tamirat, H.; Sunderland, R.

2015-01-01

In the European Sodium Fast Reactor (ESFR) project supported by EURATOM in 2008-2012, a concept for a large 3600 MWth sodium-cooled fast reactor design was investigated. In particular, reference core designs with oxide and carbide fuel were optimized to improve their safety parameters. Uncertainties in these parameters were evaluated for the oxide option. Core modifications were performed first to reduce the sodium void reactivity effect. Introduction of a large sodium plenum with an absorber layer above the core and a lower axial fertile blanket improve the total sodium void effect appreciably, bringing it close to zero for a core with fresh fuel, in line with results obtained worldwide, while not influencing substantially other core physics parameters. Therefore an optimized configuration, CONF2, with a sodium plenum and a lower blanket was established first and used as a basis for further studies in view of deterioration of safety parameters during reactor operation. Further options to study were an inner fertile blanket, introduction of moderator pins, a smaller core height, special designs for pins, such as 'empty' pins, and subassemblies. These special designs were proposed to facilitate melted fuel relocation in order to avoid core re-criticality under severe accident conditions. In the paper further CONF2 modifications are compared in terms of safety and fuel balance. They may bring further improvements in safety, but their accurate assessment requires additional studies, including transient analyses. Uncertainty studies were performed by employing a so-called Total Monte-Carlo method, for which a large number of nuclear data files is produced for single isotopes and then used in Monte-Carlo calculations. The uncertainties for the criticality, sodium void and Doppler effects, effective delayed neutron fraction due to uncertainties in basic nuclear data were assessed for an ESFR core. They prove applicability of the available nuclear data for ESFR
Corium spreading: hydrodynamics, rheology and solidification of a high-temperature oxide melt

International Nuclear Information System (INIS)

Journeau, Ch.

2006-06-01

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)

Influence of Support Configurations on the Characteristics of Selective Laser-Melted Inconel 718

Science.gov (United States)

Nadammal, Naresh; Kromm, Arne; Saliwan-Neumann, Romeo; Farahbod, Lena; Haberland, Christoph; Portella, Pedro Dolabella

2018-03-01

Samples fabricated using two different support configurations by following identical scan strategies during selective laser melting of superalloy Inconel 718 were characterized in this study. Characterization methods included optical microscopy, electron back-scattered diffraction and x-ray diffraction residual stress measurement. For the scan strategy considered, microstructure and residual stress development in the samples were influenced by the support structures. However, crystallographic texture intensity and the texture components formed within the core part of the samples were almost independent of the support. The formation of finer grains closer to the support as well as within the columnar grain boundaries resulted in randomization and texture intensity reduction by nearly half for the sample built on a lattice support. Heat transfer rates dictated by the support configurations in addition to the scan strategy influenced the microstructure and residual stress development in selective laser-melted Inconel 718 samples.
Force induced DNA melting

International Nuclear Information System (INIS)

Santosh, Mogurampelly; Maiti, Prabal K

2009-01-01

When pulled along the axis, double-strand DNA undergoes a large conformational change and elongates by roughly twice its initial contour length at a pulling force of about 70 pN. The transition to this highly overstretched form of DNA is very cooperative. Applying a force perpendicular to the DNA axis (unzipping), double-strand DNA can also be separated into two single-stranded DNA, this being a fundamental process in DNA replication. We study the DNA overstretching and unzipping transition using fully atomistic molecular dynamics (MD) simulations and argue that the conformational changes of double-strand DNA associated with either of the above mentioned processes can be viewed as force induced DNA melting. As the force at one end of the DNA is increased the DNA starts melting abruptly/smoothly above a critical force depending on the pulling direction. The critical force f m , at which DNA melts completely decreases as the temperature of the system is increased. The melting force in the case of unzipping is smaller compared to the melting force when the DNA is pulled along the helical axis. In the case of melting through unzipping, the double-strand separation has jumps which correspond to the different energy minima arising due to sequence of different base pairs. The fraction of Watson-Crick base pair hydrogen bond breaking as a function of force does not show smooth and continuous behavior and consists of plateaus followed by sharp jumps.
Study of corium radial spreading between fuel rods in a PWR core

International Nuclear Information System (INIS)

Roche, S.; Gatt, J.M.

1996-01-01

In the framework of severe accident studies for PWR like Three Mile Island Unit 2 (TMI-2), the reactor core essentially constituted of fuel rods begins to heat and then to melt. During the early degradation phase, a melt (essentially UO2 and ZrO2) that constitutes the corium flows first along the rods, and after a blockage formation, may radially propagate towards the core periphery. A simplified model has been elaborated to study the corium freezing phenomena during its crossflow between the fuel rods. The corium spreads on an horizontal support made, of either a corium crust, or a grid assembly. The model solves numerically the interface energy balance equation at the solid-liquid corium interface and the monodimensional heat balance equation in transient process with convective terms and heat source (residual power). ''Zukauskas'' correlations are used to calculate heat transfer coefficients. The model can be integrated in severe accident codes like ICARE II (IPSN) describing the in-vessel degradation scenarios. (author). 5 refs, 10 figs
The COMET-L3 experiment on long-term melt. Concrete interaction and cooling by surface flooding

International Nuclear Information System (INIS)

Alsmeyer, H.; Cron, T.; Fluhrer, B.; Messemer, G.; Miassoedov, A.; Schmidt-Stiefel, S.; Wenz, T.

2007-02-01

The COMET-L3 experiment considers the long-term situation of corium/concrete interaction in an anticipated core melt accident of a light-water-reactor, after the metal melt is layered beneath the oxide melt. The experimental focus is on cavity formation in the basemat and the risk of long term basemat penetration. The experiment investigates the two-dimensional concrete erosion in a cylindrical crucible fabricated from siliceous concrete in the first phase of the test, and the influence of surface flooding in the second phase. Decay heating in the two-component metal and oxide melt is simulated by sustained induction heating of the metal phase that is overlaid by the oxide melt. The inner diameter of the concrete crucible was 60 cm, the initial mass of the melt was 425 kg steel and 211 kg oxide at 1665 C, resulting in a melt height of 450 mm. The net power to the metal melt was about 220 kW from 0 s to 1880 s, when the maximum erosion limit of the crucible was reached and heating was terminated. In the initial phase of the test (less than 100 s), the overheated, highly agitated metal melt causes intense interaction with the concrete, which leads to fast decrease of the initial melt overheat and reduction of the initially high concrete erosion rate. Thereafter, under quasistationary conditions until about 800 s, the erosion by the metal melt slows down to some 0.07 mm/s into the axial direction. Lateral erosion is a factor 3 smaller. Video observation of the melt surface shows an agitated melt with ongoing gas release from the decomposing concrete. Several periods of more intense gas release, gas driven splashing, and release of crusts from the concrete interface indicate the existence and iterative break-up of crusts that probably form at the steel/concrete interface. Surface flooding of the melt is initiated at 800 s by a shower from the crucible head with 0.375 litre water/s. Flooding does not lead to strong melt/water interactions, and no entrapment reactions or
Efforts toward rapid construction of the cortistatin A carbocyclic core via enyne-ene metathesis

KAUST Repository

Baumgartner, Corinne; Ma, Sandy; Liu, Qi; Stoltz, Brian M.

2010-01-01

Our efforts toward the construction of the carbocylic core of cortistatin A via an enyne-ene metathesis are disclosed. Interestingly, an attempted S N2 inversion of a secondary mesylate in our five-membered D-ring piece gave a product with retention
How to arrest a core meltdown accident (doing nothing)

International Nuclear Information System (INIS)

Baron, Jorge H.

2000-01-01

In the eventual situation of a severe accident in a nuclear reactor, the molten core is able to relocate inside the pressure vessel. This may lead to the vessel failure, due to the thermal attack of the molten core (at approximation of 3000K) on the vessel steel wall. The vessel failure implies the failure of a very important barrier that contains the radioactive materials generated during the reactor operation, with a significant risk of producing high radiation doses both on operators and on the public. It is expected, for the new generation of nuclear reactors, that these will be required to withstand (by design) a core melt down accident, without the need for an immediate evacuation of the surrounding population. In this line, the use of a totally passive system is postulated, which fulfills the objective of containing the molten core inside the pressure vessel, at low temperature (approximation 1200K) precluding its failure. The conceptual design of a passive in-vessel core catcher is presented in this paper, built up of zinc, and designed for the CAREM-25 nuclear power plant. (author)
Degraded core studies at INEL

International Nuclear Information System (INIS)

Buescher, B.J.; Howe, T.M.; Miller, R.W.

1982-01-01

During 1980, planning of prototypical severe fuel damage tests to be conducted in the Power Burst Facility (PBF) to investigate fuel behavior in severe accidents up to temperatures of 2400 0 K was initiated. This first series of tests is designated Phase I. Also, a code development effort was initiated to provide a reliable predictive tool for core behavior during severe accidents. During 1981, an assessment of capabilities and preliminary planning were begun for an in-pile experimental program to investigate the behavior of larger arrays of previously irradiated fuel rods at temperatures through UO 2 melting. This latter series of tests is designated Phase II
Melting under shock compression

International Nuclear Information System (INIS)

Bennett, B.I.

1980-10-01

A simple model, using experimentally measured shock and particle velocities, is applied to the Lindemann melting formula to predict the density, temperature, and pressure at which a material will melt when shocked from room temperature and zero pressure initial conditions
Electric melting furnace of solidifying radioactive waste by utilizing magnetic field and melting method

International Nuclear Information System (INIS)

Igarashi, Hiroshi.

1990-01-01

An electric melting furnace for solidification of radioactive wastes utilizing magnetic fields in accordance with the present invention comprises a plurality of electrodes supplying AC current to molten glass in a glass melting furnace and a plurality of magnetic poles for generating AC magnetic fields. Interactions between the current and the magnetic field, generated forces in the identical direction in view of time in the molten glass. That is, forces for promoting the flow of molten glass in the melting furnace are resulted due to the Fleming's left-hand rule. As a result, the following effects can be obtained. (1) The amount of heat ransferred from the molten glass to the starting material layer on the molten surface is increased to improve the melting performance. (2) For an identical melting performance, the size and the weight of the melting furnace can be reduced to decrease the amount of secondary wastes when the apparatus-life is exhausted. (3) Bottom deposits can be suppressed and prevented from settling and depositing to the reactor bottom by the promoted flow in the layer. (4) Further, the size of auxiliary electrodes for directly supplying electric current to heat the molten glass near the reactor bottom can be decreased. (I.S.)
An Equation Governing Ultralow-Velocity Zones: Implications for Holes in the ULVZ, Lateral Chemical Reactions at the Core-Mantle Boundary, and Damping of Heat Flux Variations in the Core

Science.gov (United States)

Hernlund, J. W.; Matsui, H.

2017-12-01

Ultralow-velocity zones (ULVZ) are increasingly illuminated by seismology, revealing surprising diversity in size, shape, and physical characteristics. The only viable hypotheses are that ULVZs are a compositionally distinct FeO-enriched dense material, which could have formed by fractional crystallization of a basal magma ocean, segregation of subducted banded iron formations, precipitation of solids from the outer core, partial melting and segregation of iron-rich melts from subducted basalts, or most likely a combination of many different processes. But many questions remain: Are ULVZ partially molten in some places, and not in others? Are ULVZ simply the thicker portions of an otherwise global thin layer, covering the entire CMB and thus blocking or moderating chemical interactions between the core and overlying mantle? Is such a layer inter-connected and able to conduct electrical currents that allow electro-magnetic coupling of core and mantle angular momentum? Are they being eroded and shrinking in size due to viscous entrainment, or is more material being added to ULVZ over time? Here we derive an advection-diffusion-like equation that governs the dynamical evolution of a chemically distinct ULVZ. Analysis of this equation shows that ULVZ should become readily swept aside by viscous mantle flows at the CMB, exposing "ordinary mantle" to the top of the core, thus inducing chemical heterogeneity that drives lateral CMB chemical reactions. These reactions are correlated with heat flux, thus maintaining large-scale pressure variations atop the core that induce cyclone-like flows centered around ULVZ and ponded subducted slabs. We suggest that turbulent diffusion across adjacent cyclone streams inside a stratified region atop the core readily accommodates lateral transport and re-distribution of components such as O and Si, in addition to heat. Our model implies that the deeper core is at least partly shielded from the influence of strong heat flux variations at
Holocene evolution of a drowned melt-water valley in the Danish Wadden Sea

DEFF Research Database (Denmark)

Pedersen, Jørn Bjarke Torp; Svinth, Steffen; Bartholdy, Jesper

2009-01-01

Cores from the salt marshes along the drowned melt-water valley of river Varde Å in the Danish Wadden Sea have been dated and analysed (litho- and biostratigraphically) to reconstruct the Holocene geomorphologic evolution and relative sea level history of the area. The analysed cores cover...... the total post-glacial transgression, and the reconstructed sea level curve represents the first unbroken curve of this kind from the Danish Wadden Sea, including all phases from the time where sea level first reached the Pleistocene substrate of the area. The sea level has been rising from - 12 m below...... the present level at c. 8400 cal yr BP, interrupted by two minor drops of sea level rise, and the Holocene sequence consists in most places of clay atop...
SCDAP/RELAP5 Modeling of Movement of Melted Material Through Porous Debris in Lower Head

International Nuclear Information System (INIS)

Siefken, L. J.

1998-01-01

Designs are described for implementing models for calculating the movement of melted material through the interstices in a matrix of porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and nonporous debris that results from core material slumping into the lower head during a severe accident in a Light Water Reactor. Currently, the COUPLE model has no capability to model the movement of material that melts within a matrix of porous material. The COUPLE model also does not have the capability to model the movement of liquefied core plate material that slumps onto a porous debris bed in the lower head. In order to advance beyond the assumption the liquefied material always remains stationary, designs are developed for calculations of the movement of liquefied material through the interstices in a matrix of porous material. Correlations are identified for calculating the permeability of the porous debris and for calculating the rate of flow of liquefied material through the interstices in the debris bed. Correlations are also identified for calculating the relocation of solid debris that has a large amount of cavities due to the flowing away of melted material. Equations are defined for calculating the effect on the temperature distribution in the debris bed of heat transported by moving material and for changes in effective thermal conductivity and heat capacity due to the movement of material. The implementation of these models is expected to improve the calculation of the material distribution and temperature distribution of debris in the lower head for cases in which the debris is porous and liquefied material is present within the porous debris
[Three-dimensional computer aided design for individualized post-and-core restoration].

Science.gov (United States)

Gu, Xiao-yu; Wang, Ya-ping; Wang, Yong; Lü, Pei-jun

2009-10-01

To develop a method of three-dimensional computer aided design (CAD) of post-and-core restoration. Two plaster casts with extracted natural teeth were used in this study. The extracted teeth were prepared and scanned using tomography method to obtain three-dimensional digitalized models. According to the basic rules of post-and-core design, posts, cores and cavity surfaces of the teeth were designed using the tools for processing point clouds, curves and surfaces on the forward engineering software of Tanglong prosthodontic system. Then three-dimensional figures of the final restorations were corrected according to the configurations of anterior teeth, premolars and molars respectively. Computer aided design of 14 post-and-core restorations were finished, and good fitness between the restoration and the three-dimensional digital models were obtained. Appropriate retention forms and enough spaces for the full crown restorations can be obtained through this method. The CAD of three-dimensional figures of the post-and-core restorations can fulfill clinical requirements. Therefore they can be used in computer-aided manufacture (CAM) of post-and-core restorations.
Hypoeutectic melting in the UO{sub 2-x}-Gd{sub 2}O{sub 3} system

Energy Technology Data Exchange (ETDEWEB)

Journeau, Christophe, E-mail: christophe.journeau@cea.fr [CEA, DEN, SMTA, LPMA, Cadarache, F13108 St Paul lez Durance (France); Fouquart, Pascal [CEA, DEN, SMTA, LPMA, Cadarache, F13108 St Paul lez Durance (France); Domenger, Renaud; Allegri, Patrick [CEA, DEN, SGCS, LMAC, Marcoule, F30207 Bagnols sur Cèze (France)

2017-05-15

Gadolinium is one of the best neutron absorber materials and its use can be considered as a sacrificial material in a Sodium Fast Reactor core catcher in view of preventing recriticallity. A series of experiments have been conducted in the VITI induction-heated facility to study the melting in the UO{sub 2-x}-Gd{sub 2}O{sub 3} system with 60–87 mol% gadolinia. These experiments have indicated that the eutectic composition is around 92 mol% Gd{sub 2}O{sub 3} – 8 mol% UO{sub 2-x} and that the liquidus line is close to that of Popov et al. [Atom. Energ. 110 (2011) pp. 221–229] phase diagram. - Highlights: •Melting/Solidification experiments with UO{sub 2-x} and Gd{sub 2}O{sub 3} in reducing environment. •Eutectic composition around 92 mol% Gd{sub 2}O{sub 3}-8 mol% UO{sub 2-x}. •UO{sub 2-x} - Gd{sub 2}O{sub 3} liquidus line seems close to that of the pseudobinary phase diagram proposed by Popov et al. •Results will support the assessment of Gd{sub 2}O{sub 3} as a sacrificial material to mitigate criticality risk in SFR core catchers.
Melting in trivalent metal chlorides

International Nuclear Information System (INIS)

Saboungi, M.L.; Price, D.L.; Scamehorn, C.; Tosi, M.P.

1990-11-01

We report a neutron diffraction study of the liquid structure of YCl 3 and combine the structural data with macroscopic melting and transport data to contrast the behaviour of this molten salt with those of SrCl 2 , ZnCl 2 and AlCl 3 as prototypes of different melting mechanisms for ionic materials. A novel melting mechanism for trivalent metal chlorides, leading to a loose disordered network of edge-sharing octahedral units in the liquid phase, is thereby established. The various melting behaviours are related to bonding character with the help of Pettifor's phenomenological chemical scale. (author). 25 refs, 4 figs, 3 tabs
Spinnability and Characteristics of Polyvinylidene Fluoride (PVDF)-based Bicomponent Fibers with a Carbon Nanotube (CNT) Modified Polypropylene Core for Piezoelectric Applications.

Science.gov (United States)

Glauß, Benjamin; Steinmann, Wilhelm; Walter, Stephan; Beckers, Markus; Seide, Gunnar; Gries, Thomas; Roth, Georg

2013-07-03

This research explains the melt spinning of bicomponent fibers, consisting of a conductive polypropylene (PP) core and a piezoelectric sheath (polyvinylidene fluoride). Previously analyzed piezoelectric capabilities of polyvinylidene fluoride (PVDF) are to be exploited in sensor filaments. The PP compound contains a 10 wt % carbon nanotubes (CNTs) and 2 wt % sodium stearate (NaSt). The sodium stearate is added to lower the viscosity of the melt. The compound constitutes the fiber core that is conductive due to a percolation CNT network. The PVDF sheath's piezoelectric effect is based on the formation of an all-trans conformation β phase, caused by draw-winding of the fibers. The core and sheath materials, as well as the bicomponent fibers, are characterized through different analytical methods. These include wide-angle X-ray diffraction (WAXD) to analyze crucial parameters for the development of a crystalline β phase. The distribution of CNTs in the polymer matrix, which affects the conductivity of the core, was investigated by transmission electron microscopy (TEM). Thermal characterization is carried out by conventional differential scanning calorimetry (DSC). Optical microscopy is used to determine the fibers' diameter regularity (core and sheath). The materials' viscosity is determined by rheometry. Eventually, an LCR tester is used to determine the core's specific resistance.
The ımpact of benefıts and servıces on manager engagement and manager retentıon ın tourısm ındustry: Enhanced by the effect of management level

OpenAIRE

ALDATMAZ, Ipek; AYKAÇ, Cansu; DICLE, Ülkü

2017-01-01

Manager engagement and retention are vital to the success and organizational performance of many service sector organisations. Maintaining manager retention is a major challenge that many hotel enterprises face today. It is critical that organizations give greater importance to manager engagement, motivation and retention and therefore establish an efficient benefits and services strategy for retaining these core managers for the persistence and achievement of the organization...
Molten Core - Concrete interactions in nuclear accidents. Theory and design of an experimental facility

International Nuclear Information System (INIS)

Sevon, T.

2005-11-01

In a hypothetical severe accident in a nuclear power plant, the molten core of the reactor may flow onto the concrete floor of containment building. This would cause a molten core . concrete interaction (MCCI), in which the heat transfer from the hot melt to the concrete would cause melting of the concrete. In assessing the safety of nuclear reactors, it is important to know the consequences of such an interaction. As background to the subject, this publication includes a description of the core melt stabilization concept of the European Pressurized water Reactor (EPR), which is being built in Olkiluoto in Finland. The publication includes a description of the basic theory of the interaction and the process of spalling or cracking of concrete when it is heated rapidly. A literature survey and some calculations of the physical properties of concrete and corium. concrete mixtures at high temperatures have been conducted. In addition, an equation is derived for conservative calculation of the maximum possible concrete ablation depth. The publication also includes a literature survey of experimental research on the subject of the MCCI and discussion of the results and deficiencies of the experiments. The main result of this work is the general design of an experimental facility to examine the interaction of molten metals and concrete. The main objective of the experiments is to assess the probability of spalling, or cracking, of concrete under pouring of molten material. A program of five experiments has been designed, and pre-test calculations of the experiments have been conducted with MELCOR 1.8.5 accident analysis program and conservative analytic calculations. (orig.)
Urinary retention in women.

Science.gov (United States)

Juma, Saad

2014-07-01

This review is a summary of the most pertinent published studies in the literature in the last 18 months that address cause, diagnosis, and management of urinary retention in women. Symptoms, uroflow, and pressure-flow studies have a low predictive value for and do not correlate with elevated postvoid residual urine (PVR). Anterior and posterior colporrhaphy do not cause de-novo bladder outlet obstruction in the majority of patients with elevated PVR, and the cause of elevated PVR may be other factors such as pain or anxiety causing abnormal relaxation of the pelvic floor and contributing to voiding difficulty. The risk of urinary retention in a future pregnancy after mid-urethral sling (MUS) is small. The risk of urinary tract infection and urinary retention after chemodenervation of the bladder with onabotulinumtoxin-A (100 IU) in patients with non-neurogenic urge incontinence is 33 and 5%, respectively. There is a lack of consensus among experts on the timing of sling takedown in the management of acute urinary retention following MUS procedures. There has been a significant progress in the understanding of the causation of urinary retention. Important areas that need further research (basic and clinical) are post-MUS and pelvic organ prolapse repair urinary retention and obstruction, and urinary retention owing to detrusor underactivity.
Constraints on the Parental Melts of Enriched Shergottites from Image Analysis and High Pressure Experiments

Science.gov (United States)

Collinet, M.; Medard, E.; Devouard, B.; Peslier, A.

2012-01-01

Martian basalts can be classified in at least two geochemically different families: enriched and depleted shergottites. Enriched shergottites are characterized by higher incompatible element concentrations and initial Sr-87/Sr-86 and lower initial Nd-143/Nd-144 and Hf-176/Hf-177 than depleted shergottites [e.g. 1, 2]. It is now generally admitted that shergottites result from the melting of at least two distinct mantle reservoirs [e.g. 2, 3]. Some of the olivine-phyric shergottites (either depleted or enriched), the most magnesian Martian basalts, could represent primitive melts, which are of considerable interest to constrain mantle sources. Two depleted olivine-phyric shergottites, Yamato (Y) 980459 and Northwest Africa (NWA) 5789, are in equilibrium with their most magnesian olivine (Fig. 1) and their bulk rock compositions are inferred to represent primitive melts [4, 5]. Larkman Nunatak (LAR) 06319 [3, 6, 7] and NWA 1068 [8], the most magnesian enriched basalts, have bulk Mg# that are too high to be in equilibrium with their olivine megacryst cores. Parental melt compositions have been estimated by subtracting the most magnesian olivine from the bulk rock composition, assuming that olivine megacrysts have partially accumulated [3, 9]. However, because this technique does not account for the actual petrography of these meteorites, we used image analysis to study these rocks history, reconstruct their parent magma and understand the nature of olivine megacrysts.

Investigation of melt agglomeration process with a hydrophobic binder in combination with sucrose stearate.

Science.gov (United States)

Heng, Paul Wan Sia; Wong, Tin Wui; Cheong, Wai See

2003-08-01

The melt agglomeration process of lactose powder with hydrogenated cottonseed oil (HCO) as the hydrophobic meltable binder was investigated by studying the physicochemical properties of molten HCO modified by sucrose stearates S170, S770 and S1570. The size, size distribution, micromeritic and adhesion properties of agglomerates as well as surface tension, contact angle, viscosity and specific volume of molten HCO, with and without sucrose stearates, were examined. The viscosity, specific volume and surface tension of molten HCO were found to be modified to varying extents by sucrose stearates which are available in different HLB values and melt properties. The growth of melt agglomerates was promoted predominantly by an increase in viscosity, an increase in specific volume or a decrease in surface tension of the molten binding liquid. The agglomerate growth propensity was higher with an increase in inter-particulate binding strength, agglomerate surface wetness and extent of agglomerate consolidation which enhanced the liquid migration from agglomerate core to periphery leading to an increased surface plasticity for coalescence. The inclusion of high concentrations of completely meltable sucrose stearate S170 greatly induced the growth of agglomerates through increased specific volume and viscosity of the molten binding liquid. On the other hand, the inclusion of incompletely meltable sucrose stearates S770 and S1570 promoted the agglomeration mainly via the reduction in surface tension of the molten binding liquid with declining agglomerate growth propensity at high sucrose stearate concentrations. In addition to being an agglomeration modifier, sucrose stearate demonstrated anti-adherent property in melt agglomeration process. The properties of molten HCO and melt agglomerates were dependent on the type and concentration of sucrose stearate added.
Efficient approach to compute melting properties fully from ab initio with application to Cu

Science.gov (United States)

Zhu, Li-Fang; Grabowski, Blazej; Neugebauer, Jörg

2017-12-01

Applying thermodynamic integration within an ab initio-based free-energy approach is a state-of-the-art method to calculate melting points of materials. However, the high computational cost and the reliance on a good reference system for calculating the liquid free energy have so far hindered a general application. To overcome these challenges, we propose the two-optimized references thermodynamic integration using Langevin dynamics (TOR-TILD) method in this work by extending the two-stage upsampled thermodynamic integration using Langevin dynamics (TU-TILD) method, which has been originally developed to obtain anharmonic free energies of solids, to the calculation of liquid free energies. The core idea of TOR-TILD is to fit two empirical potentials to the energies from density functional theory based molecular dynamics runs for the solid and the liquid phase and to use these potentials as reference systems for thermodynamic integration. Because the empirical potentials closely reproduce the ab initio system in the relevant part of the phase space the convergence of the thermodynamic integration is very rapid. Therefore, the proposed approach improves significantly the computational efficiency while preserving the required accuracy. As a test case, we apply TOR-TILD to fcc Cu computing not only the melting point but various other melting properties, such as the entropy and enthalpy of fusion and the volume change upon melting. The generalized gradient approximation (GGA) with the Perdew-Burke-Ernzerhof (PBE) exchange-correlation functional and the local-density approximation (LDA) are used. Using both functionals gives a reliable ab initio confidence interval for the melting point, the enthalpy of fusion, and entropy of fusion.
A Brief Review of Past INL Work Assessing Radionuclide Content in TMI-2 Melted Fuel Debris: The Use of 144Ce as a Surrogate for Pu Accountancy

Energy Technology Data Exchange (ETDEWEB)

D. L. Chichester; S. J. Thompson

2013-09-01

This report serves as a literature review of prior work performed at Idaho National Laboratory, and its predecessor organizations Idaho National Engineering Laboratory (INEL) and Idaho National Engineering and Environmental Laboratory (INEEL), studying radionuclide partitioning within the melted fuel debris of the reactor of the Three Mile Island 2 (TMI-2) nuclear power plant. The purpose of this review is to document prior published work that provides supporting evidence of the utility of using 144Ce as a surrogate for plutonium within melted fuel debris. When the TMI-2 accident occurred no quantitative nondestructive analysis (NDA) techniques existed that could assay plutonium in the unconventional wastes from the reactor. However, unpublished work performed at INL by D. W. Akers in the late 1980s through the 1990s demonstrated that passive gamma-ray spectrometry of 144Ce could potentially be used to develop a semi-quantitative correlation for estimating plutonium content in these materials. The fate and transport of radioisotopes in fuel from different regions of the core, including uranium, fission products, and actinides, appear to be well characterized based on the maximum temperature reached by fuel in different parts of the core and the melting point, boiling point, and volatility of those radioisotopes. Also, the chemical interactions between fuel, fuel cladding, control elements, and core structural components appears to have played a large role in determining when and how fuel relocation occurred in the core; perhaps the most important of these reaction appears to be related to the formation of mixed-material alloys, eutectics, in the fuel cladding. Because of its high melting point, low volatility, and similar chemical behavior to plutonium, the element cerium appears to have behaved similarly to plutonium during the evolution of the TMI-2 accident. Anecdotal evidence extrapolated from open-source literature strengthens this logical feasibility for
Premixing and steam explosion phenomena in the tests with stratified melt-coolant configuration and binary oxidic melt simulant materials

Energy Technology Data Exchange (ETDEWEB)

Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se; Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Konovalenko, Alexander, E-mail: kono@kth.se; Karbojian, Aram, E-mail: karbojan@kth.se

2017-04-01

Highlights: • Steam explosion in stratified melt-coolant configuration is studied experimentally. • Different binary oxidic melt simulant materials were used. • Five spontaneous steam explosions were observed. • Instability of melt-coolant interface and formation of premixing layer was observed. • Explosion strength is influenced by melt superheat and water subcooling. - Abstract: Steam explosion phenomena in stratified melt-coolant configuration are considered in this paper. Liquid corium layer covered by water on top can be formed in severe accident scenarios with (i) vessel failure and release of corium melt into a relatively shallow water pool; (ii) with top flooding of corium melt layer. In previous assessments of potential energetics in stratified melt-coolant configuration, it was assumed that melt and coolant are separated by a stable vapor film and there is no premixing prior to the shock wave propagation. This assumption was instrumental for concluding that the amount of energy that can be released in such configuration is not of safety importance. However, several recent experiments carried out in Pouring and Under-water Liquid Melt Spreading (PULiMS) facility with up to 78 kg of binary oxidic corium simulants mixtures have resulted in spontaneous explosions with relatively high conversion ratios (order of one percent). The instability of the melt-coolant interface, melt splashes and formation of premixing layer were observed in the tests. In this work, we present results of experiments carried out more recently in steam explosion in stratified melt-coolant configuration (SES) facility in order to shed some light on the premixing phenomena and assess the influence of the test conditions on the steam explosion energetics.
Melting of superheated molecular crystals

Science.gov (United States)

Cubeta, Ulyana; Bhattacharya, Deepanjan; Sadtchenko, Vlad

2017-07-01

Melting dynamics of micrometer scale, polycrystalline samples of isobutane, dimethyl ether, methyl benzene, and 2-propanol were investigated by fast scanning calorimetry. When films are superheated with rates in excess of 105 K s-1, the melting process follows zero-order, Arrhenius-like kinetics until approximately half of the sample has transformed. Such kinetics strongly imply that melting progresses into the bulk via a rapidly moving solid-liquid interface that is likely to originate at the sample's surface. Remarkably, the apparent activation energies for the phase transformation are large; all exceed the enthalpy of vaporization of each compound and some exceed it by an order of magnitude. In fact, we find that the crystalline melting kinetics are comparable to the kinetics of dielectric α-relaxation in deeply supercooled liquids. Based on these observations, we conclude that the rate of non-isothermal melting for superheated, low-molecular-weight crystals is limited by constituent diffusion into an abnormally dense, glass-like, non-crystalline phase.
Comparison of the behaviour of two core designs for ASTRID in case of severe accidents

Energy Technology Data Exchange (ETDEWEB)

Bertrand, F., E-mail: frederic.bertrand@cea.fr [CEA, DEN, DER, F-13108 Saint Paul-lez-Durance (France); Marie, N.; Prulhière, G.; Lecerf, J. [CEA, DEN, DER, F-13108 Saint Paul-lez-Durance (France); Seiler, J.M. [CEA, DEN, DTN, F-38054 Grenoble (France)

2016-02-15

Highlights: • Low void worth CFV and SFRv2 cores are compared for ASTRID pre-conceptual design. • Severe accident behaviour is assessed with a simplified calculation approach and tools. • Mitigation to limit reactivity inserted by core compaction is easier for CFV than for SFRv2 core. • When facing arbitrary reactivity ramps, CFV core would lead to lower energy release than SFRv2 core. • Time scale for core degradation is one order of magnitude larger for CFV than for SFRv2. - Abstract: The present paper is dedicated to the studies carried out during the first stage of the pre-conceptual design of the French demonstrator of fourth generation SFR reactors (ASTRID) in order to compare the behaviour of two envisaged core concepts under severe accident transients. Among the two studied core concepts, whose powers are 1500 MWth, the first one is a classical homogeneous core (called SFRv2) with large pin diameter whose the sodium overall voiding reactivity effect is 5 $. The second concept is an axially heterogeneous core (called CFV) whose global void reactivity effect is negative (−1.2 $ at the end of cycle at the equilibrium). The comparison of the cores relies on two typical accident families: a reactivity insertion (unprotected transient overpower, UTOP) and an overall loss of core cooling (unprotected loss of flow, ULOF). In the first part of the comparison, the primary phase of an UTOP is studied in order to assess typical features of the transient behaviour: power and reactivity evolutions, material heating and melting/vaporization and mechanical energy release due to fuel vapor expansion. The second part of the comparison deals with the calculation of the reactivity potential for degraded states (molten pools) representative of the secondary phase of a mild UTOP and of a strong UTOP (strong or mild qualifies the reactivity ramp inserted). According to the reactivity potential, the amount of fuel to extract from the core and the amount of absorber
Waterlike anomalies in a two-dimensional core-softened potential

Science.gov (United States)

Bordin, José Rafael; Barbosa, Marcia C.

2018-02-01

We investigate the structural, thermodynamic, and dynamic behavior of a two-dimensional (2D) core-corona system using Langevin dynamics simulations. The particles are modeled by employing a core-softened potential which exhibits waterlike anomalies in three dimensions. In previous studies in a quasi-2D system a new region in the pressure versus temperature phase diagram of structural anomalies was observed. Here we show that for the two-dimensional case two regions in the pressure versus temperature phase diagram with structural, density, and diffusion anomalies are observed. Our findings indicate that, while the anomalous region at lower densities is due the competition between the two length scales in the potential at higher densities, the anomalous region is related to the reentrance of the melting line.
Estimation of hydraulic conductivities of Yucca Mountain tuffs from sorptivity and water retention measurements

International Nuclear Information System (INIS)

Zimmerman, R.W.; Bodvarsson, G.S.

1995-06-01

The hydraulic conductivity functions of the matrix rocks at Yucca Mountain, Nevada, are among the most important data needed as input for the site-scale hydrological model of the unsaturated zone. The difficult and time-consuming nature of hydraulic conductivity measurements renders it infeasible to directly measure this property on large numbers of cores. Water retention and sorptivity measurements, however, can be made relatively rapidly. The sorptivity is, in principle, a unique functional of the conductivity and water retention functions. It therefore should be possible to invert sorptivity and water retention measurements in order to estimate the conductivity; the porosity is the only other parameter that is required for this inversion. In this report two methods of carrying out this inversion are presented, and are tested against a limited data set that has been collected by Flint et al. at the USGS on a set of Yucca Mountain tuffs. The absolute permeability is usually predicted by both methods to within an average error of about 0.5 - 1.0 orders of magnitude. The discrepancy appears to be due to the fact that the water retention curves have only been measured during drainage, whereas the imbibition water retention curve is the one that is relevant to sorptivity measurements. Although the inversion methods also yield predictions of the relative permeability function, there are yet no unsaturated hydraulic conductivity data against which to test these predictions
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
Pressure melting and ice skating

Science.gov (United States)

Colbeck, S. C.

1995-10-01

Pressure melting cannot be responsible for the low friction of ice. The pressure needed to reach the melting temperature is above the compressive failure stress and, if it did occur, high squeeze losses would result in very thin films. Pure liquid water cannot coexist with ice much below -20 °C at any pressure and friction does not increase suddenly in that range. If frictional heating and pressure melting contribute equally, the length of the wetted contact could not exceed 15 μm at a speed of 5 m/s, which seems much too short. If pressure melting is the dominant process, the water films are less than 0.08 μm thick because of the high pressures.
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
New finite element-based modeling of reactor core support plate failure

Energy Technology Data Exchange (ETDEWEB)

Pandazis, Peter; Lovasz, Liviusz [Gesellschaft fuer Anlagen- und Reaktorsicherheit gGmbH, Garching (Germany). Forschungszentrum; Babcsany, Boglarka [Budapest Univ. of Technology and Economics, Budapest (Hungary). Inst. of Nuclear Techniques; Hajas, Tamas

2017-12-15

ATHLET-CD is the severe accident module of the code system AC{sup 2} that is designed to simulate the core degradation phenomena including fission product release and transport in the reactor circuit, as well as the late phase processes in the lower plenum. In case of a severe accident degradation of the reactor core occurs, the fuel assemblies start to melt. The evolution of such processes is usually accompanied with the failure of the core support plate and relocation of the molten core to the lower plenum. Currently, the criterion for the failure of the support plate applied by ATHLET-CD is a user-defined signal which can be a specific time or process variable like mass, temperature, etc. A new method, based on FEM approach, was developed that could lead in the future to a more realistic criterion for the failure of the core support plate. This paper presents the basic idea and theory of this new method as well as preliminary verification calculations and an outlook on the planned future development.
Improvements in EBR-2 core depletion calculations

International Nuclear Information System (INIS)

Finck, P.J.; Hill, R.N.; Sakamoto, S.

1991-01-01

The need for accurate core depletion calculations in Experimental Breeder Reactor No. 2 (EBR-2) is discussed. Because of the unique physics characteristics of EBR-2, it is difficult to obtain accurate and computationally efficient multigroup flux predictions. This paper describes the effect of various conventional and higher order schemes for group constant generation and for flux computations; results indicate that higher-order methods are required, particularly in the outer regions (i.e. the radial blanket). A methodology based on Nodal Equivalence Theory (N.E.T.) is developed which allows retention of the accuracy of a higher order solution with the computational efficiency of a few group nodal diffusion solution. The application of this methodology to three-dimensional EBR-2 flux predictions is demonstrated; this improved methodology allows accurate core depletion calculations at reasonable cost. 13 refs., 4 figs., 3 tabs
Melting temperatures of MgO under high pressure determined by micro-texture observation

Science.gov (United States)

Kimura, T.; Ohfuji, H.; Nishi, M.; Irifune, T.

2016-12-01

Periclase (MgO) is the second abundant mineral after bridgmanite in the Earth's lower mantle, and its melting temperature (Tm) under pressure is important to constrain the chemical composition of ultra-deep magma formed near the mantle-core boundary. However, the melting behavior is highly controversial among previous studies: a laser-heated diamond anvil cell (LHDAC) study reported a melting curve with a dTm/dP of 30 K/GPa at zero pressure [1], while several theoretical computations gave substantially higher dTm/dP of 90 100 K/GPa [2,3]. We performed a series of LHDAC experiments for measurements of Tm of MgO under high pressure, using single crystal MgO as the starting material. The melting was detected by using micro-texture observations of the quenched samples. We found that the laser-heated area of the sample quenched from the Tm in previous LHDAC experiments [1] showed randomly aggregated granular crystals, which was not caused by melting, but by plastic deformation of the sample. This suggests that the Tms of their study were substantially underestimated. On the other hand, the sample recovered from the temperature higher by 1500-1700 K than the Tms in previous LHDAC experiments showed a characteristic internal texture comparable to the solidification texture typically shown in metal casting. We determined the Tms based on the observation of this texture up to 32 GPa. Fitting our Tms to the Simon equation yields dTm/dP of 82 K/GPa at zero pressure, which is consistent with those of the theoretical predictions (90 100 K/GPa) [2,3]. Extrapolation of the present melting curve of MgO to the pressure of the CMB (135 GPa) gives a melting temperature of 8900 K. The present steep melting slope offers the eutectic composition close to peridotite (in terms of Mg/Si ratio) throughout the lower mantle conditions. According to the model for sink/float relationship between the solid mantle and the magma [4], a considerable amount of iron (Fe/(Mg+Fe) > 0.24) is expected
Inner Core Structure of Hurricane Alicia from Airborne Doppler Radar Observations.

Science.gov (United States)

Marks, Frank D., Jr.; Houze, Robert A., Jr.

1987-05-01

Airborne Doppler radar measurements are used to determine the horizontal winds, vertical air motions, radar reflectivity and hydrometer fallspeeds over much of the inner-core region (within 40 km of the eye) of Hurricane Alicia (1983). The reconstructed flow field is more complete and detailed than any obtained previously. The data show both the primary (azimuthal) and secondary (radial-height) circulations. The primary circulation was characterized by an outward sloping maximum of tangential wind. The secondary circulation was characterized by a deep layer of radial inflow in the lower troposphere and a layer of intense outflow above 10 km altitude. The rising branch of the secondary circulation was located in the eyewall and sloped radially outward. Discrete convective-scale bubbles of more intense upward motion were superimposed on this mean rising current, and convective-scale downdrafts were located throughout and below the core of maximum precipitation in the eyewall.Precipitation particles in the eyewall rainshaft circulated 18-20 km downwind as they fell, consistent with the typical upwind slope with increasing altitude of eyewall precipitation cores Outside the eyewall, the precipitation was predominantly stratiform. A radar bright band was evident at the melting level. Above the melting level, ice particles were advected into the stratiform region from the upper levels of the eyewall and drifted downward through a mesoscale region of ascent. Hypothetical precipitation particle trajectories showed that as these particles fell slowly through the mesoscale updraft toward the melting level, they were carried azimuthally as many as 1 1/2 times around the storm. During this spiraling descent, the particles evidently grew vigorously. The amount of water condensed by the ambient mesoscale ascent exceeded that transported into the stratiform region by the eyewall outflow by a factor of 3. As the particles fell into the lower troposphere, they entered a mesoscale
Synthesis of CuO-NiO core-shell nanoparticles by homogeneous precipitation method

International Nuclear Information System (INIS)

Bayal, Nisha; Jeevanandam, P.

2012-01-01

Highlights: ► CuO-NiO core-shell nanoparticles have been synthesized using a simple homogeneous precipitation method for the first time. ► Mechanism of the formation of core-shell nanoparticles has been investigated. ► The synthesis route may be extended for the synthesis of other mixed metal oxide core-shell nanoparticles. - Abstract: Core-shell CuO–NiO mixed metal oxide nanoparticles in which CuO is the core and NiO is the shell have been successfully synthesized using homogeneous precipitation method. This is a simple synthetic method which produces first a layered double hydroxide precursor with core-shell morphology which on calcination at 350 °C yields the mixed metal oxide nanoparticles with the retention of core-shell morphology. The CuO–NiO mixed metal oxide precursor and the core-shell nanoparticles were characterized by powder X-ray diffraction, FT-IR spectroscopy, thermal gravimetric analysis, elemental analysis, scanning electron microscopy, transmission electron microscopy, and diffuse reflectance spectroscopy. The chemical reactivity of the core-shell nanoparticles was tested using catalytic reduction of 4-nitrophenol with NaBH 4 . The possible growth mechanism of the particles with core-shell morphology has also been investigated.
Effectiveness of In-Vessel Retention Strategies and Minimum Safety Injection Flow over Postulated Severe Accidents of OPR1000

International Nuclear Information System (INIS)

Kim, Sung Joong; Seo, Seungwon; Lee, Seongnyeon; KIm, Hwan Yeol; Ha, Kwang Soon; Park, Jonghwa; Park, Raejoon

2013-01-01

The objective of this study is first to evaluate various serious severe accident scenarios of OPR1000 with and without in-vessel retention strategies using MELCOR code. Second is to develop a mechanistic model of minimum safety injection flow using the thermal-hydraulic parameters of CET and collapsed water level obtained from the MELCOR simulation results. Effectiveness of RCS depressurization of OPR1000 is investigated for postulated severe accidents of SBLOCA, SBO, and TLOF. It is seen that timely operator action is important to achieve the best mitigation. Also The MELCOR simulation results of SBLOCA, SBO, and TLOFW are utilized to develop a model for minimum safety injection flow. The model suggests that if HPSI is available with RCS pressure lower than 120 bars, the core coolability can be guaranteed. In this study, several MELCOR simulations are conducted in search for effective in-vessel retention strategies over postulated severe accidents of SBLOCA, SBO, and TLOFW of OPR1000. Detailed accident sequences are presented and indicative parameters diagnosing the reactor thermal-hydraulic state are interrogated to provide useful information to the operator actions. To properly assist operator's action during the severe accident, the thermal-hydraulic parameters should be virtual, intuitive, and reliable. In addition, the parameters should be collected through the instrumentations close to the reactor core. In this regard, Core Exit Temperature (CET) and collapsed core water level are deemed as the commensurate parameters
Effectiveness of In-Vessel Retention Strategies and Minimum Safety Injection Flow over Postulated Severe Accidents of OPR1000

Energy Technology Data Exchange (ETDEWEB)

Kim, Sung Joong; Seo, Seungwon; Lee, Seongnyeon [Hanyang Univ., Seoul (Korea, Republic of); KIm, Hwan Yeol; Ha, Kwang Soon; Park, Jonghwa; Park, Raejoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

The objective of this study is first to evaluate various serious severe accident scenarios of OPR1000 with and without in-vessel retention strategies using MELCOR code. Second is to develop a mechanistic model of minimum safety injection flow using the thermal-hydraulic parameters of CET and collapsed water level obtained from the MELCOR simulation results. Effectiveness of RCS depressurization of OPR1000 is investigated for postulated severe accidents of SBLOCA, SBO, and TLOF. It is seen that timely operator action is important to achieve the best mitigation. Also The MELCOR simulation results of SBLOCA, SBO, and TLOFW are utilized to develop a model for minimum safety injection flow. The model suggests that if HPSI is available with RCS pressure lower than 120 bars, the core coolability can be guaranteed. In this study, several MELCOR simulations are conducted in search for effective in-vessel retention strategies over postulated severe accidents of SBLOCA, SBO, and TLOFW of OPR1000. Detailed accident sequences are presented and indicative parameters diagnosing the reactor thermal-hydraulic state are interrogated to provide useful information to the operator actions. To properly assist operator's action during the severe accident, the thermal-hydraulic parameters should be virtual, intuitive, and reliable. In addition, the parameters should be collected through the instrumentations close to the reactor core. In this regard, Core Exit Temperature (CET) and collapsed core water level are deemed as the commensurate parameters.
Analysis of molten fuel behavior in coolant channel during severe accidents in KALIMER

International Nuclear Information System (INIS)

Suk, Soo Dong; Lee, Yong Bum; Hahn, Do Hee

2004-11-01

Preliminary safety analyses of the KALIMER-600 design have shown that the design has inherent safety characteristics and is capable of accommodating double fault initiators such as ATWS events without boiling coolant or melting fuel. For the future design of liquid metal reactor, however, the evaluation of the safety performance and the determination of containment requirements may require consideration of tripe-fault accident sequences of extremely low probability of occurrence that leads to fuel melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will required as a design requirement for the future design of LMR. For sodium-cooled core designs with metallic fuel, one of the major phenomenological modeling uncertainties to be resolved is the potential for freezing and plugging of molten metallic fuel in above- and below-core structures and possibly in inter-subassembly spaces. In this study, scoping analyses were carried out to evaluate the penetration depths in the coolant channels by molten fuel mixture during the unprotected loss-of-flow accidents in the core of the KALIMER-600. It is assumed in the analyses that a solid fuel crust would start to form upon contact with the coolant channel structure temperature of which is below the fuel solidus. The analysis results predict that the coolant channels would be plugged by the freezing molten fuel in the inlet lower shield as well as in the outlet, fission-gas-plenum region for the KALIMER-600 design
Overturn of magma ocean ilmenite cumulate layer: Implications for lunar magmatic evolution and formation of a lunar core

Science.gov (United States)

Hess, P. C.; Parmentier, E. M.

1993-01-01

We explore a model for the chemical evolution of the lunar interior that explains the origin and evolution of lunar magmatism and possibly the existence of a lunar core. A magma ocean formed during accretion differentiates into the anorthositic crust and chemically stratified cumulate mantle. The cumulative mantle is gravitationally unstable with dense ilmenite cumulate layers overlying olivine-orthopyroxene cumulates with Fe/Mg that decreases with depth. The dense ilmenite layer sinks to the center of the moon forming the core. The remainder of the gravitationally unstable cumulate pile also overturns. Any remaining primitive lunar mantle rises to its level of neutral buoyancy in the cumulate pile. Perhaps melting of primitive lunar mantle due to this decompression results in early lunar Mg-rich magmatism. Because of its high concentration of incompatible heat producing elements, the ilmenite core heats the overlying orthopyroxene-bearing cumulates. As a conductively thickening thermal boundary layer becomes unstable, the resulting mantle plumes rise, decompress, and partially melt to generate the mare basalts. This model explains both the timing and chemical characteristics of lunar magmatism.

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