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

Core-concrete interactions with overlying water pools

International Nuclear Information System (INIS)

Blose, R.E.; Powers, D.A.; Copus, E.R.; Brockmann, J.E.; Simpson, R.B.; Lucero, D.A.

1993-11-01

The WETCOR-1 test of simultaneous interactions of a high-temperature melt with water and a limestone/common-sand concrete is described. The test used a 34.1-kg melt of 76.8 w/o Al 2 O 3 , 16.9 w/o CaO, and 4.0 w/o SiO 2 heated by induction using tungsten susceptors. Once quasi-steady attack on concrete by the melt was established, an attempt was made to quench the melt at 1850 K with 295 K water flowing at 57 liters per minute. Net power into the melt at the time of water addition was 0.61 ± 0.19 W/cm 3 . The test configuration used in the WETCOR-1 test was designed to delay melt freezing to the walls of the test fixture. This was done to test hypotheses concerning the inherent stability of crust formation when high-temperature melts are exposed to water. No instability in crust formation was observed. The flux of heat through the crust to the water pool maintained over the melt in the test was found to be 0.52 ± 0.13 MW/m 2 . Solidified crusts were found to attenuate aerosol emissions during the melt concrete interactions by factors of 1.3 to 3.5. The combination of a solidified crust and a 30-cm deep subcooled water pool was found to attenuate aerosol emissions by factors of 3 to 15

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

Effects of surface shape on the geometry and surface topography of the melt pool in low-power density laser melting

KAUST Repository

Kim, Youngdeuk

2011-04-15

The quantitative correlations between workpiece volume and melt pool geometry, as well as the flow and thermal features of the melt pool are established. Thermocapillary convections in melt pool with a deformable free surface are investigated with respect to surface shape and laser intensity. When the contact angle between the tangent to the top surface and the vertical wall at the hot center is acute, the free surface flattens, compared with that of the initial free surface. Otherwise, the free surface forms a bowl-like shape with a deep crater and a low peripheral rim when the contact angle at the hot center is obtuse. Increasing the workpiece volume at a fixed laser intensity and a negative radial height gradient cause linear decreases in the geometric size and magnitude of flow and temperature of the melt pool. Conversely, linear increases are observed with a positive radial height gradient. © 2011 American Institute of Chemical Engineers (AIChE).

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

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

Melt-Pool Temperature and Size Measurement During Direct Laser Sintering

Energy Technology Data Exchange (ETDEWEB)

List, III, Frederick Alyious [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dinwiddie, Ralph Barton [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carver, Keith [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gockel, Joy E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-01

Additive manufacturing has demonstrated the ability to fabricate complex geometries and components not possible with conventional casting and machining. In many cases, industry has demonstrated the ability to fabricate complex geometries with improved efficiency and performance. However, qualification and certification of processes is challenging, leaving companies to focus on certification of material though design allowable based approaches. This significantly reduces the business case for additive manufacturing. Therefore, real time monitoring of the melt pool can be used to detect the development of flaws, such as porosity or un-sintered powder and aid in the certification process. Characteristics of the melt pool in the Direct Laser Sintering (DLS) process is also of great interest to modelers who are developing simulation models needed to improve and perfect the DLS process. Such models could provide a means to rapidly develop the optimum processing parameters for new alloy powders and optimize processing parameters for specific part geometries. Stratonics’ ThermaViz system will be integrated with the Renishaw DLS system in order to demonstrate its ability to measure melt pool size, shape and temperature. These results will be compared with data from an existing IR camera to determine the best approach for the determination of these critical parameters.

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

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

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.

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

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)

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

Effect of Laser Power and Scan Speed on Melt Pool Characteristics of Commercially Pure Titanium (CP-Ti)

Science.gov (United States)

Kusuma, Chandrakanth; Ahmed, Sazzad H.; Mian, Ahsan; Srinivasan, Raghavan

2017-07-01

Selective laser melting (SLM) is an additive manufacturing technique that creates complex parts by selectively melting metal powder layer-by-layer using a laser. In SLM, the process parameters decide the quality of the fabricated component. In this study, single beads of commercially pure titanium (CP-Ti) were melted on a substrate of the same material using an in-house built SLM machine. Multiple combinations of laser power and scan speed were used for single bead fabrication, while the laser beam diameter and powder layer thickness were kept constant. This experimental study investigated the influence of laser power, scan speed, and laser energy density on the melt pool formation, surface morphology, geometry (width and height), and hardness of solidified beads. In addition, the observed unfavorable effect such as inconsistency in melt pool width formation is discussed. The results show that the quality, geometry, and hardness of solidified melt pool are significantly affected by laser power, scanning speed, and laser energy density.

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)

Core-concrete interactions with overlying water pools. The WETCOR-1 test

Energy Technology Data Exchange (ETDEWEB)

Blose, R.E. [Ktech Corp., Albuquerque, NM (United States); Powers, D.A.; Copus, E.R.; Brockmann, J.E.; Simpson, R.B.; Lucero, D.A. [Sandia National Labs., Albuquerque, NM (United States)

1993-11-01

The WETCOR-1 test of simultaneous interactions of a high-temperature melt with water and a limestone/common-sand concrete is described. The test used a 34.1-kg melt of 76.8 w/o Al{sub 2}O{sub 3}, 16.9 w/o CaO, and 4.0 w/o SiO{sub 2} heated by induction using tungsten susceptors. Once quasi-steady attack on concrete by the melt was established, an attempt was made to quench the melt at 1850 K with 295 K water flowing at 57 liters per minute. Net power into the melt at the time of water addition was 0.61 {plus_minus} 0.19 W/cm{sup 3}. The test configuration used in the WETCOR-1 test was designed to delay melt freezing to the walls of the test fixture. This was done to test hypotheses concerning the inherent stability of crust formation when high-temperature melts are exposed to water. No instability in crust formation was observed. The flux of heat through the crust to the water pool maintained over the melt in the test was found to be 0.52 {plus_minus} 0.13 MW/m{sup 2}. Solidified crusts were found to attenuate aerosol emissions during the melt concrete interactions by factors of 1.3 to 3.5. The combination of a solidified crust and a 30-cm deep subcooled water pool was found to attenuate aerosol emissions by factors of 3 to 15.

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

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

Probabilistic analysis of some safety aspects of a swimming pool reactor

International Nuclear Information System (INIS)

Lieber, K.; Nicolescu, T.

1984-01-01

A probabilistic risk analysis of some safety aspects without the investigation of radioactivity release has been performed for the 10 MW (thermal) swimming-pool research reactor SAPHIR. Our presentation is focused on the 7 internal initiating events found to be relevant with respect to accident sequences that could result with core melt due to loss of coolant or overcriticality. The results are given by the core melt frequencies for the investigated accident sequences. It could be demonstrated by our investigation that the core melt hazard of the reactor is extremely low. (author)

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

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

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

Modelling the geometry of a moving laser melt pool and deposition track via energy and mass balances

Energy Technology Data Exchange (ETDEWEB)

Pinkerton, Andrew J; Li Lin [Laser Processing Research Centre, Department of Mechanical, Aerospace and Manufacturing Engineering, University of Manchester Institute of Science and Technology, PO Box 88, Sackville Street, Manchester M60 1QD (United Kingdom)

2004-07-21

The additive manufacturing technique of laser direct metal deposition allows multiple tracks of full density metallic material to be built to form complex parts for rapid tooling and manufacture. Practical results and theoretical models have shown that the geometries of the tracks are governed by multiple factors. Original work with single layer cladding identified three basic clad profiles but, so far, models of multiple layer, powder-feed deposition have been based on only two of them. At higher powder mass flow rates, experimental results have shown that a layer's width can become greater than the melt pool width at the substrate surface, but previous analytical models have not been able to accommodate this. In this paper, a model based on this third profile is established and experimentally verified. The model concentrates on mathematical analysis of the melt pool and establishes mass and energy balances based on one-dimensional heat conduction to the substrate. Deposition track limits are considered as arcs of circles rather than of ellipses, as used in most established models, reflecting the dominance of surface tension forces in the melt pool, and expressions for elongation of the melt pool with increasing traverse speed are incorporated. Trends in layer width and height with major process parameters are captured and predicted layer dimensions correspond well to the experimental values.

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

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

Core neutronics of a swimming pool research reactor

International Nuclear Information System (INIS)

Mannan, M.A.; Mondal, M.A.W.; Pervini, M.E.

1981-01-01

The initial cores of the 5 MW swimming pool research reactor of the Nuclear Research Centre, Tehran have been analyzed using the computer codes METHUSELAH and EQUIPOISE. The effective multiplication factor, critical mass, moderator temperature and void coefficients of the core have been calculated and compared with vendor's values. Calculated values agree reasonably well with the vendor's results. (author)

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

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 disruptive accident and recriticality analysis with FX2-POOL

International Nuclear Information System (INIS)

Abramson, P.B.

1976-01-01

The current state of development of FX2-POOL, a two-dimensional hydrodynamic, thermodynamic and neutronic scoping model for Hypothetical Core Disruptive Accident analysis is described. Checkout comparisons to VENUS for prompt burst conditions were good. Use of FX2-POOL to examine the importance of fuel to steel heat transfer during a prompt burst indicates that heat transfer plays no important role on that time scale. Scoping studies of material thermohydrodynamics for about 20 to 30 milliseconds following the prompt burst indicate that heat transfer is important on the time scale necessary for the CDA bubble to grow to the size of the original core. Preliminary results are presented for energetics of boiling fuel steel pools which are forced recritical by local surface pressurization

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.)

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

Ex-vessel melt-coolant interactions in deep water pool: Studies and accident management for Swedish BWRs

International Nuclear Information System (INIS)

Sienicki, J.J.; Chu, C.C.; Spencer, B.W.; Frid, W.; Loewenhielm, G.

1993-01-01

In Swedish BWRs having an annular suppression pool, the lower drywell beneath the reactor vessel is flooded with water to mitigate against the effects of melt release into the drywell during a severe accident. The THIRMAL code has been used to analyze the effectiveness of the water pool to protect lower drywell penetrations by fragmenting and quenching the melt as it relocates downward through the water. Experiments have also been performed to investigate the benefits of adding surfactants to the water to reduce the likelihood of fine-scale debris formation from steam explosions. This paper presents an overview of the accident management approach and surfactant investigations together with results from the THIRMAL analyses

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.

Seismic responses of a pool-type fast reactor with different core support designs

International Nuclear Information System (INIS)

Wu, Ting-shu; Seidensticker, R.W.

1989-01-01

In designing the core support system for a pool-type fast reactor, there are many issues which must be considered in order to achieve an optimum and balanced design. These issues include safety, reliability, as well as costs. Several design options are possible to support the reactor core. Different core support options yield different frequency ranges and responses. Seismic responses of a large pool-type fast reactor incorporated with different core support designs have been investigated. 4 refs., 3 figs

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

On-site releases of noble gases and iodine in the event of core meltdown in a swimming pool reactor

International Nuclear Information System (INIS)

Montaignac, E. de.

1976-10-01

Research aimed at defining a standard model accident for swimming pool type reactors, has led to the adoption to the so-called BORAX accident which involves complete meltdown of the reactor core. This type of accident-an accident related to dimensional problems- is useful for calculations concerning reactor components which have to withstand the mechanical forces resulting from the accident. A study of the radiobiological consequences of this type of accident, involving the entire reactor core, required research to determine as accurately as possible how the iodine, noble gases and solid fission products are distributed between the melted core and the site. The joint document in the annexure served as the basis for discussion at the meeting (BEVS/SESR) on 9th March 1973, at which the SESR set the standard parameter values to be used for estimating fission product distributions on the site. (author)

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

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

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

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

Validation of SPARC, a suppression pool aerosol capture model

International Nuclear Information System (INIS)

Owczarski, P.C.; Winegardner, W.K.

1985-09-01

A study of the potential for atmospheric release in hypothetical severe core melt accidents in BWRs with suppression pools was recently completed using a prototype of the SPARC code. The process of validating SPARC using an experimental data base is the concern of this paper

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.

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)

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

Theoretical description of laser melt pool dynamics, Task order number B239634, Quarter 3 report

Energy Technology Data Exchange (ETDEWEB)

Dykhne, A.

1995-05-10

Melting of solid matter under laser radiation is realized in almost every process of laser technology. The present paper addresses melted material flows in cases when melt zones are shallow, i.e., the zone width is appreciably greater than or of the same order as its depth. Such conditions are usually realized when hardening, doping or perforating thin plates or when using none-deep penetration. Melted material flowing under conditions of deep penetration, drilling of deep openings and cutting depends on a number of additional factors (as compared to the shallow-pool case), namely, formation of a vapor and gas cavern in the sample and propagation of the laser beam through the cavern. These extra circumstances complicate hydrodynamic consideration of the liquid bath and will be addressed is the paper to follow.

Effect of pool rotation on three-dimensional flow in a shallow annular pool of silicon melt with bidirectional temperature gradients

Energy Technology Data Exchange (ETDEWEB)

Zhang, Quan-Zhuang; Peng, Lan; Liu, Jia [Key Laboratory of Low-grade Energy Utilization Technologies and Systems of Ministry of Education, College of Power Engineering, Chongqing University, Chongqing, 400044 (China); Wang, Fei, E-mail: penglan@cqu.edu.cn [Chongqing Special Equipment Inspection and Research Institute, Chongqing, 401121 (China)

2016-08-15

In order to understand the effect of pool rotation on silicon melt flow with the bidirectional temperature gradients, we conducted a series of unsteady three-dimensional (3D) numerical simulations in a shallow annular pool. The bidirectional temperature gradients are produced by the temperature difference between outer and inner walls as well as a constant heat flux at the bottom. Results show that when Marangoni number is small, a 3D steady flow is common without pool rotation. But it bifurcates to a 3D oscillatory flow at a low rotation Reynolds number. Subsequently, the flow becomes steady and axisymmetric at a high rotation Reynolds number. When the Marangoni number is large, pool rotation can effectively suppress the temperature fluctuation on the free surface, meanwhile, it improves the flow stability. The critical heat flux density diagrams are mapped, and the effects of radial and vertical temperature gradients on the flow are discussed. Additionally, the transition process from the flow dominated by the radial temperature gradient to the one dominated by the vertical temperature gradient is presented. (paper)

Ex-vessel melt-coolant interactions in deep water pool: studies and accident management for Swedish BWRs

International Nuclear Information System (INIS)

Chu, C.C.; Sienicki, J.J.; Spencer, B.W.; Frid, W.; Loewenhielm, G.

1995-01-01

In Swedish BWRs having an annular suppression pool, the lower drywell beneath the reactor vessel is flooded with water to mitigate against the effects of melt release into the drywell during a severe accident. The THIRMAL-1 code has been used to analyze the effectiveness of the water pool to protect lower drywell penetrations by fragmenting and quenching the melt as it relocates downward through the water. Experiments have also been performed to investigate the benefits of adding surfactants to the water to reduce the likelihood of fine-scale debris formation from steam explosions. This paper presents an overview of the accident management approach and surfactant investigations together with results from the THIRMAL-1 analyses. A description of the modeling incorporated in THIRMAL-1 is also provided. (orig.)

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

Debris and pool formation/heat transfer in FARO-LWR: experiments and analyses

International Nuclear Information System (INIS)

Magallon, D.; Annunziato, A.; Corradini, M.

1999-01-01

The FARO-LWR experiments examine the debris and pool formation from a pour of core melt materials (UO 2 /ZrO 2 and UO 2 /ZrO 2 /Zr) into a pool of water at prototypic accident conditions. The experiments give unique data on the debris bed initial conditions, morphology and heat transfer after the core melt has slump and (partly) quenched into the water of the lower head. Quantities of up to 170 kg of corium melt are poured by gravity into water of depth between 1 and 2 m through a nozzle of diameter 0.1 m at different system pressures. The debris is collected in a flat bottom catcher of diameter 0.66 m. It reaches heights up to 0.2 m depending on the melt quantity. In general, the melt reaches the bottom only partially fragmented. The debris which forms consists of a conglomerate ('cake') in contact with the collecting structure and overlaying fragments (loose debris). The mean particle size of the loose debris is in the range 3.5 - 4.8 mm. The upper surface of the debris is flat. A gap is present between the cake and the bottom plate. The paper reviews the debris formation and heat transfer to the bottom steel structure from these tests and describes the development of a model to predict the debris and pool formation process. Sensitivity analyses have been performed by the COMETA code to study the behaviour of the ratio between the cake mass and the total mass. (authors)

A computational model for viscous fluid flow, heat transfer, and melting in in situ vitrification melt pools

International Nuclear Information System (INIS)

McHugh, P.R.; Ramshaw, J.D.

1991-11-01

MAGMA is a FORTRAN computer code designed to viscous flow in in situ vitrification melt pools. It models three-dimensional, incompressible, viscous flow and heat transfer. The momentum equation is coupled to the temperature field through the buoyancy force terms arising from the Boussinesq approximation. All fluid properties, except density, are assumed variable. Density is assumed constant except in the buoyancy force terms in the momentum equation. A simple melting model based on the enthalpy method allows the study of the melt front progression and latent heat effects. An indirect addressing scheme used in the numerical solution of the momentum equation voids unnecessary calculations in cells devoid of liquid. Two-dimensional calculations can be performed using either rectangular or cylindrical coordinates, while three-dimensional calculations use rectangular coordinates. All derivatives are approximated by finite differences. The incompressible Navier-Stokes equations are solved using a new fully implicit iterative technique, while the energy equation is differenced explicitly in time. Spatial derivatives are written in conservative form using a uniform, rectangular, staggered mesh based on the marker and cell placement of variables. Convective terms are differenced using a weighted average of centered and donor cell differencing to ensure numerical stability. Complete descriptions of MAGMA governing equations, numerics, code structure, and code verification are provided. 14 refs

A computational model for viscous fluid flow, heat transfer, and melting in in situ vitrification melt pools

Energy Technology Data Exchange (ETDEWEB)

McHugh, P.R.; Ramshaw, J.D.

1991-11-01

MAGMA is a FORTRAN computer code designed to viscous flow in in situ vitrification melt pools. It models three-dimensional, incompressible, viscous flow and heat transfer. The momentum equation is coupled to the temperature field through the buoyancy force terms arising from the Boussinesq approximation. All fluid properties, except density, are assumed variable. Density is assumed constant except in the buoyancy force terms in the momentum equation. A simple melting model based on the enthalpy method allows the study of the melt front progression and latent heat effects. An indirect addressing scheme used in the numerical solution of the momentum equation voids unnecessary calculations in cells devoid of liquid. Two-dimensional calculations can be performed using either rectangular or cylindrical coordinates, while three-dimensional calculations use rectangular coordinates. All derivatives are approximated by finite differences. The incompressible Navier-Stokes equations are solved using a new fully implicit iterative technique, while the energy equation is differenced explicitly in time. Spatial derivatives are written in conservative form using a uniform, rectangular, staggered mesh based on the marker and cell placement of variables. Convective terms are differenced using a weighted average of centered and donor cell differencing to ensure numerical stability. Complete descriptions of MAGMA governing equations, numerics, code structure, and code verification are provided. 14 refs.

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

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

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)

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

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.)

Deep pooling of low degree melts and volatile fluxes at the 85°E segment of the Gakkel Ridge: Evidence from olivine-hosted melt inclusions and glasses

Science.gov (United States)

Shaw, Alison M.; Behn, Mark D.; Humphris, Susan E.; Sohn, Robert A.; Gregg, Patricia M.

2010-01-01

We present new analyses of volatile, major, and trace elements for a suite of glasses and melt inclusions from the 85°E segment of the ultra-slow spreading Gakkel Ridge. Samples from this segment include limu o pele and glass shards, proposed to result from CO 2-driven explosive activity. The major element and volatile compositions of the melt inclusions are more variable and consistently more primitive than the glass data. CO 2 contents in the melt inclusions extend to higher values (167-1596 ppm) than in the co-existing glasses (187-227 ppm), indicating that the melt inclusions were trapped at greater depths. These melt inclusions record the highest CO 2 melt concentrations observed for a ridge environment. Based on a vapor saturation model, we estimate that the melt inclusions were trapped between seafloor depths (˜ 4 km) and ˜ 9 km below the seafloor. However, the glasses are all in equilibrium with their eruption depths, which is inconsistent with the rapid magma ascent rates expected for explosive activity. Melting conditions inferred from thermobarometry suggest relatively deep (25-40 km) and cold (1240°-1325 °C) melting conditions, consistent with a thermal structure calculated for the Gakkel Ridge. The water contents and trace element compositions of the melt inclusions and glasses are remarkably homogeneous; this is an unexpected result for ultra-slow spreading ridges, where magma mixing is generally thought to be less efficient based on the assumption that steady-state crustal magma chambers are absent in these environments. All melts can be described by a single liquid line of descent originating from a pooled melt composition that is consistent with the aggregate melt calculated from a geodynamic model for the Gakkel Ridge. These data suggest a model in which deep, low degree melts are efficiently pooled in the upper mantle (9-20 km depth), after which crystallization commences and continues during ascent and eruption. Based on our melting model

Thermal-hydraulic analysis of the OSURR pool for power upgrade with natural convection core cooling

International Nuclear Information System (INIS)

Ha, J.J.; Aldemir, T.

1988-01-01

Natural convection mode core cooling will be maintained in the LEU conversion/power upgrade of The Ohio State University Research Reactor (OSURR) to 250-500 kW. The pool water will be cooled by a water-glycol-air and a water-water heat exchanger. A plume disperser will be installed in the pool to minimize evaporation from the pool top and to maintain the dose rate due to N-16 activity within allowable levels. The minimization of the pool heat removal system operation costs necessitates maximizing the inlet temperature to the water-glycol-air heat exchanger. For the maximization process, the change in the pool temperature and velocity fields have to be investigated as a function of: location and orientation of the heat removal system components and the plume disperser in the pool; mass flow rate through the plume disperser. The velocity and temperature fields in the pool are determined using COMMIX-1A. The computational system model accounts for the presence of all the pool components (i.e. core, thermal column, beam ports, ion chamber, guide tubes, rabbit, neutron source etc.). The results show that: (1) Both the heat removal system inlet point and the plume disperser have to be located close to the top of the core. (2) Using a disperser system consisting of several pipes may be more feasible than a single unit. (3) For high disperser flow, the disperser jet has to be almost parallel to the top of the core to prevent flow reversal in coolant channels. (4) More than one disperser system may be necessary to create an inversion layer in the pool

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

Heat and fission product transport in molten core material pool with crust

International Nuclear Information System (INIS)

Yun, J.I.; Suh, K.Y.; Kang, C.S.

2005-01-01

Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the reactor vessel during a severe accident. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products in the pool is estimated by product of the mass concentration and energy conversion factor of each fission product. Twenty-nine elements are chosen and classified by their chemical properties to calculate heat generation rate in the pool. The mass concentration of a fission product is obtained from released fraction and the tabular output of the ORIGEN 2 code. The initial core and pool inventories at each time can also be estimated using ORIGEN 2. The released fraction of each fission product is calculated based on the bubble dynamics and mass transport. Numerical analysis is performed for heat and fission product transport in a molten core material pool during the Three Mile Island Unit 2 (TMI-2) accident. The pool is assumed to be a partially filled hemisphere, whose change in geometry is neglected during the numerical calculation. Calculated results indicate that the peak temperature in the molten pool is significantly lowered, since a substantial amount of the volatile fission products is released from the molten pool during progression of the accident. The results may directly be applied to the existing severe accident analysis codes to more mechanistically determine the thermal load to the reactor vessel lower head during the in-vessel retention

Experimental study of simulant melt stream-water thermal interaction in pool and narrow geometries

International Nuclear Information System (INIS)

Narayanan, K.S.; Jasmin Sudha, A.; Murthy, S.S.; Rao, E.H.V.M.; Lydia, G.; Das, S.K.; Harvey, J.; Kannan, S.E.

2005-01-01

Full text of publication follows: Small scale experiments were carried out to investigate the thermal interaction characteristics of a few kilograms of Sn Pb, Bi and Zn as hot melt, in the film boiling region of water in an attempt to simulate a coherent fuel coolant interaction during a postulated severe accident in a nuclear reactor. Melt stream solidification and detached debris generation were studied with different melt superheat up to 200 deg. C, at different coolant temperatures of 30 deg. C, 50 deg. C, 70 deg. C, 90 deg. C, in pool geometry and in long narrow coolant column. The material was heated in an Alumina crucible and poured through a hot stainless steel funnel with a nozzle diameter of 7.7 mm, into the coolant. A stainless steel plate was used to collect the solidified mass after the interaction. Temperature monitoring was done in the coolant column close to the melt stream. The melt stream movement inside the coolant was imaged using a video camera at 25 fps. Measured melt stream entry velocity was around 1.5 m/sec. For low melt superheat and low coolant temperature, solidified porous tree like structure extended from the collector plate up to the melt release point. For water temperature of 70 deg. C, the solidified bed height at the center was found to decrease with increase in the melt superheat up to 150 deg. C. Fragmentation was found to occur when the melt superheat exceeded 200 deg. C. Particle size distribution was obtained for the fragmented debris. In 1D geometry, with 50 deg. C superheat, columnar solidification was observed with no fine debris. The paper gives the details of the results obtained in the experiments and highlights the role of Rayleigh-Taylor, Kelvin-Helmholtz instabilities and the melt physical properties on the fragmentation kinetics. (authors)

Welding pool measurement using thermal array sensor

Science.gov (United States)

Cho, Chia-Hung; Hsieh, Yi-Chen; Chen, Hsin-Yi

2015-08-01

Selective laser melting (SLM) is an additive manufacturing (AM) technology that uses a high-power laser beam to melt metal powder in chamber of inert gas. The process starts by slicing the 3D CAD data as a digital information source into layers to create a 2D image of each layer. Melting pool was formed by using laser irradiation on metal powders which then solidified to consolidated structure. In a selective laser melting process, the variation of melt pool affects the yield of a printed three-dimensional product. For three dimensional parts, the border conditions of the conductive heat transport have a very large influence on the melt pool dimensions. Therefore, melting pool is an important behavior that affects the final quality of the 3D object. To meet the temperature and geometry of the melting pool for monitoring in additive manufacturing technology. In this paper, we proposed the temperature sensing system which is composed of infrared photodiode, high speed camera, band-pass filter, dichroic beam splitter and focus lens. Since the infrared photodiode and high speed camera look at the process through the 2D galvanometer scanner and f-theta lens, the temperature sensing system can be used to observe the melting pool at any time, regardless of the movement of the laser spot. In order to obtain a wide temperature detecting range, 500 °C to 2500 °C, the radiation from the melting pool to be measured is filtered into a plurality of radiation portions, and since the intensity ratio distribution of the radiation portions is calculated by using black-body radiation. The experimental result shows that the system is suitable for melting pool to measure temperature.

Development of Lower Plenum Molten Pool Module of Severe Accident Analysis Code in Korea

International Nuclear Information System (INIS)

Son, Donggun; Kim, Dong-Ha; Park, Rae-Jun; Bae, Jun-Ho; Shim, Suk-Ku; Marigomen, Ralph

2014-01-01

To simulate a severe accident progression of nuclear power plant and forecast reactor pressure vessel failure, we develop computational software called COMPASS (COre Meltdown Progression Accident Simulation Software) for whole physical phenomena inside the reactor pressure vessel from a core heat-up to a vessel failure. As a part of COMPASS project, in the first phase of COMPASS development (2011 - 2014), we focused on the molten pool behavior in the lower plenum, heat-up and ablation of reactor vessel wall. Input from the core module of COMPASS is relocated melt composition and mass in time. Molten pool behavior is described based on the lumped parameter model. Heat transfers in between oxidic, metallic molten pools, overlying water, steam and debris bed are considered in the present study. The models and correlations used in this study are appropriately selected by the physical conditions of severe accident progression. Interaction between molten pools and reactor vessel wall is also simulated based on the lumped parameter model. Heat transfers between oxidic pool, thin crust of oxidic pool and reactor vessel wall are considered and we solve simple energy balance equations for the crust thickness of oxidic pool and reactor vessel wall. As a result, we simulate a benchmark calculation for APR1400 nuclear power plant, with assumption of relocated mass from the core is constant in time such that 0.2ton/sec. We discuss about the molten pool behavior and wall ablation, to validate our models and correlations used in the COMPASS. Stand-alone SIMPLE program is developed as the lower plenum molten pool module for the COMPASS in-vessel severe accident analysis code. SIMPLE program formulates the mass and energy balance for water, steam, particulate debris bed, molten corium pools and oxidic crust from the first principle and uses models and correlations as the constitutive relations for the governing equations. Limited steam table and the material properties are provided

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

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

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.)

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)

Influence of corium oxidation on fission product release from molten pool

International Nuclear Information System (INIS)

Bechta, S.V.; Krushinov, E.V.; Vitol, S.A.

2009-01-01

Release of low-volatile fission products and core materials from molten oxidic corium was investigated in the EVAN project under the auspices of ISTC. The experiments carried out in cold crucible with induction heating and RASPLAV test facility are described. The results are discussed in terms of reactor application; in particular, pool configuration, melt oxidation kinetics, critical influence of melt surface temperature and oxidation index on the fission product release rate and aerosol particle composition. The relevance of measured high release of Sr from the molten pool for the reactor application is highlighted. Comparisons of the experimental data with those from the COLIMA CA-U3 test and the VERCORS tests, as well as with predictions from IVTANTHERMO and GEMINI/NUCLEA are set. (author)

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

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

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.)

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.

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

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)

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

1999-08-01

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

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

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

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.

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

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

A simplified geometrical model for transient corium propagation in core for LWR with heavy reflector - 15271

International Nuclear Information System (INIS)

Saas, L.; Le Tellier, R.; Bajard, S.

2015-01-01

In this document, we present a simplified geometrical model (0D model) for both the in-core corium propagation transient and the characterization of the mode of corium transfer from the core to the vessel. A degraded core with a formed corium pool is used as an initial state. This initial state can be obtained from a simulation computed with an integral code. This model does not use a grid for the core as integral codes do. Geometrical shapes and 0D models are associated with the corium pool and the other components of the degraded core (debris, heavy reflector, core plate...). During the transient, these shapes evolve taking into account the thermal and stratification behavior of the corium pool and the melting of the core surrounding components. Some results corresponding to the corium pool propagation in core transients obtained with this model on a LWR with a heavy reflector are given and compared to grid approach of the integral codes MAAP4

Quenching behavior of molten pool with different strategies – A review

Energy Technology Data Exchange (ETDEWEB)

Shrikant,, E-mail: 2014rmt9018@mnit.ac.in; Pandel, U.; Duchaniya, R. K. [Department of Metallurgical & Materials Engineering, Malaviya National Institute of Technology Jaipur, Rajasthan 302017 (India); Nayak, A. K. [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400094 (India)

2016-05-06

After the major severe accident in nuclear reactor, there has been lot of concerns regarding long term core melt stabilization following a severe accident in nuclear reactors. Numerous strategies have been though 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 melt pool under bottom flooding as well as for indirect cooling. Besides these tests are very scattered because they involve different simulants material initial temperatures and masses of melt, which makes it very complex to judge the effectiveness of a particular technique and advantage over the other. In this review paper, a study has been carried on different cooling techniques of simulant materials with same mass. Three techniques have been compared here and the results are discussed. Under top flooding technique it took several hours to cool the melt under without decay heat condition. In bottom flooding technique was found to be the best technique among in indirect cooling technique, top flooded technique, and bottom flooded technique.

Quenching behavior of molten pool with different strategies – A review

International Nuclear Information System (INIS)

Shrikant,; Pandel, U.; Duchaniya, R. K.; Nayak, A. K.

2016-01-01

After the major severe accident in nuclear reactor, there has been lot of concerns regarding long term core melt stabilization following a severe accident in nuclear reactors. Numerous strategies have been though 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 melt pool under bottom flooding as well as for indirect cooling. Besides these tests are very scattered because they involve different simulants material initial temperatures and masses of melt, which makes it very complex to judge the effectiveness of a particular technique and advantage over the other. In this review paper, a study has been carried on different cooling techniques of simulant materials with same mass. Three techniques have been compared here and the results are discussed. Under top flooding technique it took several hours to cool the melt under without decay heat condition. In bottom flooding technique was found to be the best technique among in indirect cooling technique, top flooded technique, and bottom flooded technique.

PWR core and spent fuel pool analysis using scale and nestle

International Nuclear Information System (INIS)

Murphy, J. E.; Maldonado, G. I.; St Clair, R.; Orr, D.

2012-01-01

The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

PWR core and spent fuel pool analysis using scale and nestle

Energy Technology Data Exchange (ETDEWEB)

Murphy, J. E.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, Knoxville, TN 37996-2300 (United States); St Clair, R.; Orr, D. [Duke Energy, 526 S. Church St, Charlotte, NC 28202 (United States)

2012-07-01

The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

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.

Melt pool and keyhole behaviour analysis for deep penetration laser welding

International Nuclear Information System (INIS)

Fabbro, R

2010-01-01

One usually defines the main characteristic of the welding performances of a given laser system by its 'penetration curve' that corresponds to the welding depth as a function of the welding speed V w for a given set of operating parameters. Analysis of a penetration curve is interesting and gives very fruitful results. Coupled with high-speed video imaging of melt pool surface and ejected plume behaviour, the analysis of this penetration curve on a very large range of welding speeds, typically from 0 to 50 m min -1 , has allowed us to observe very different and characteristic regimes. These regimes are mainly characterized by the physical processes by which they impede the laser beam penetration inside the material. We show that it is only at rather high welding speeds that these limiting processes are reduced. Consequently, the scaling law of welding depth with welding speed is in agreement with adapted modelling of this process. On the other hand, as the welding speed is reduced, different effects depending on the weld pool dynamics and plume interaction strongly disturb the keyhole stability and are responsible for the deviation of the penetration curve from the previous modelling that agrees with a 1/V w scaling law. A corresponding criterion for the occurrence of this effect is defined.

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

Solid particle effects on heat transfer in a multi-layered molten pool with gas injection

International Nuclear Information System (INIS)

Bilbao y Leon, Rosa Marina; Corradini, Michael L.

2006-01-01

In the very unlikely event of a severe reactor accident involving core melt and pressure vessel failure, it is important to identify the circumstances that would allow the molten core material to cool down and resolidify, bringing core debris to a stable coolable state. To achieve this, it has been proposed to flood the cavity with water from above forming a layered structure where upward heat loss from the molten pool to the water will cause the core material to quench and solidify. In this situation the molten pool would become a three-phase mixture: e.g., a solid and liquid slurry formed by the molten pool as it cools to a temperature below the temperature of liquidus, agitated by the gases formed in the concrete ablation process. The present work quantifies the partition of the heat losses upward and downward in this multi-layered configuration, considering the influence of the viscosity and the solid fraction in the pool, from test data obtained from intermediate scale experiments at the University of Wisconsin-Madison. These experimental results show heat transfer behavior for multi-layered pools for a range of viscosities and solid fractions. These results are compared to previous experimental studies and well known correlations and models

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)

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)

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.

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.

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)

Influence of corium oxidation on fission product release from molten pool

Energy Technology Data Exchange (ETDEWEB)

Bechta, S.V., E-mail: bechta@sbor.spb.s [Alexandrov Scientific-Research Institute of Technology (NITI), Sosnovy Bor (Russian Federation); Krushinov, E.V.; Vitol, S.A.; Khabensky, V.B.; Kotova, S.Yu.; Sulatsky, A.A. [Alexandrov Scientific-Research Institute of Technology (NITI), Sosnovy Bor (Russian Federation); Gusarov, V.V.; Almyashev, V.I. [Grebenschikov Institute of Silicate Chemistry of the Russian Academy of Sciences (ISC RAS), St. Petersburg (Russian Federation); Ducros, G.; Journeau, C. [CEA, DEN, Cadarache, F-13108 St. Paul lez Durance (France); Bottomley, D. [Joint Research Centre Institut fuer Transurane (ITU), Karlsruhe (Germany); Clement, B. [Institut de Radioprotection et Surete Nucleaire (IRSN), St. Paul lez Durance (France); Herranz, L. [CIEMAT, Madrid (Spain); Guentay, S. [PSI, Wuerenlingen (Switzerland); Trambauer, K. [GRS, Muenchen (Germany); Auvinen, A. [VTT, Espoo (Finland); Bezlepkin, V.V. [SPbAEP, St. Petersburg (Russian Federation)

2010-05-15

Qualitative and quantitative determination of the release of low-volatile fission products and core materials from molten oxidic corium was investigated in the EVAN project under the auspices of ISTC. The experiments carried out in a cold crucible with induction heating and RASPLAV test facility are described. The results are discussed in terms of reactor application; in particular, pool configuration, melt oxidation kinetics, critical influence of melt surface temperature and oxidation index on the fission product release rate, aerosol particle composition and size distribution. The relevance of measured high release of Sr from the molten pool for the reactor application is highlighted. Comparisons of the experimental data with those from the COLIMA CA-U3 test and the VERCORS tests, as well as with predictions from IVTANTHERMO and GEMINI/NUCLEA codes are made. Recommendations for further investigations are proposed following the major observations and discussions.

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

Post shut-down decay heat removal from nuclear reactor core by natural convection loops in sodium pool

Energy Technology Data Exchange (ETDEWEB)

Rajamani, A. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Sundararajan, T., E-mail: tsundar@iitm.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Prasad, B.V.S.S.S. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Parthasarathy, U.; Velusamy, K. [Nuclear Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

2016-05-15

Highlights: • Transient simulations are performed for a worst case scenario of station black-out. • Inter-wrapper flow between various sub-assemblies reduces peak core temperature. • Various natural convection paths limits fuel clad temperatures below critical level. - Abstract: The 500 MWe Indian pool type Prototype Fast Breeder Reactor (PFBR) has a passive core cooling system, known as the Safety Grade Decay Heat Removal System (SGDHRS) which aids to remove decay heat after shut down phase. Immediately after reactor shut down the fission products in the core continue to generate heat due to beta decay which exponentially decreases with time. In the event of a complete station blackout, the coolant pump system may not be available and the safety grade decay heat removal system transports the decay heat from the core and dissipates it safely to the atmosphere. Apart from SGDHRS, various natural convection loops in the sodium pool carry the heat away from the core and deposit it temporarily in the sodium pool. The buoyancy driven flow through the small inter-wrapper gaps (known as inter-wrapper flow) between fuel subassemblies plays an important role in carrying the decay heat from the sub-assemblies to the hot sodium pool, immediately after reactor shut down. This paper presents the transient prediction of flow and temperature evolution in the reactor subassemblies and the sodium pool, coupled with the safety grade decay heat removal system. It is shown that with a properly sized decay heat exchanger based on liquid sodium and air chimney stacks, the post shutdown decay heat can be safely dissipated to atmospheric air passively.

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

Activity of corrosion products in pool type reactors with ascending flow in the core

International Nuclear Information System (INIS)

Andrade e Silva, Graciete S. de; Queiroz Bogado Leite, Sergio de

1995-01-01

A model for the activity of corrosion products in the water of a pool type reactor with ascending flow is presented. The problem is described by a set of coupled differential equations relating the radioisotope concentrations in the core and pool circuits and taking into account two types of radioactive sources: i) those from radioactive species formed in the fuel cladding, control elements, reflector, etc, and afterwards released to the primary stream by corrosion (named reactor sources) and ii) those formed from non radioactive isotopes entering the primary stream by corrosion of the circuit components and being activated when passing through the core (named circuit sources). (author). 6 refs, 3 figs, 4 tabs

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

Design report on SCDAP/RELAP5 model improvements - debris bed and molten pool behavior

International Nuclear Information System (INIS)

Allison, C.M.; Rempe, J.L.; Chavez, S.A.

1994-11-01

The SCDAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and in combination with VICTORIA, fission product release and transport during severe accidents. Improvements for existing debris bed and molten pool models in the SCDAP/RELAP5/MOD3.1 code are described in this report. Model improvements to address (a) debris bed formation, heating, and melting; (b) molten pool formation and growth; and (c) molten pool crust failure are discussed. Relevant data, existing models, proposed modeling changes, and the anticipated impact of the changes are discussed. Recommendations for the assessment of improved models are provided

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

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

A simplified geometrical model for transient corium propagation in core for LWR with heavy reflector

Directory of Open Access Journals (Sweden)

Saas Laurent

2017-01-01

Full Text Available In the context of the simulation of the Severe Accidents (SA in Light Water Reactors (LWR, we are interested on the in-core corium pool propagation transient in order to evaluate the corium relocation in the vessel lower head. The goal is to characterize the corium and debris flows from the core to accurately evaluate the corium pool propagation transient in the lower head and so the associated risk of vessel failure. In the case of LWR with heavy reflector, to evaluate the corium relocation into the lower head, we have to study the risk associated with focusing effect and the possibility to stabilize laterally the corium in core with a flooded down-comer. It is necessary to characterize the core degradation and the stratification of the corium pool that is formed in core. We assume that the core degradation until the corium pool formation and the corium pool propagation could be modeled separately. In this document, we present a simplified geometrical model (0D model for the in-core corium propagation transient. A degraded core with a formed corium pool is used as an initial state. This state can be obtained from a simulation computed with an integral code. This model does not use a grid for the core as integral codes do. Geometrical shapes and 0D models are associated with the corium pool and the other components of the degraded core (debris, heavy reflector, core plate…. During the transient, these shapes evolve taking into account the thermal and stratification behavior of the corium pool and the melting of the core surrounding components. Some results corresponding to the corium pool propagation in core transients obtained with this model on a LWR with a heavy reflector are given and compared to grid approach of the integral codes MAAP4.

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)

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

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

Fission Product Release from Molten Pool: ceramic melt tests

Energy Technology Data Exchange (ETDEWEB)

Petrov, Yu.B.; Lopukh, D.B.; Petchenkov, A.Yu. [AO ' NP Sintez' , St. Petersburg (RU)] [and others

1999-07-01

Experimental results are presented on the volatilisation of UO{sub 2{+-}}{sub x}, SrO, BaO, CeO{sub 2} from corium melts. Corium melts were generated by high frequency induction melting in a cold crucible. The surface temperature of the melts was in the range from 1753 to 3023 K. Some results of the tests are discussed and a comparison with published data is made. (author)

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

Investigation on Selective Laser Melting AlSi10Mg Cellular Lattice Strut: Molten Pool Morphology, Surface Roughness and Dimensional Accuracy

Directory of Open Access Journals (Sweden)

Xuesong Han

2018-03-01

Full Text Available AlSi10Mg inclined struts with angle of 45° were fabricated by selective laser melting (SLM using different scanning speed and hatch spacing to gain insight into the evolution of the molten pool morphology, surface roughness, and dimensional accuracy. The results show that the average width and depth of the molten pool, the lower surface roughness and dimensional deviation decrease with the increase of scanning speed and hatch spacing. The upper surface roughness is found to be almost constant under different processing parameters. The width and depth of the molten pool on powder-supported zone are larger than that of the molten pool on the solid-supported zone, while the width changes more significantly than that of depth. However, if the scanning speed is high enough, the width and depth of the molten pool and the lower surface roughness almost keep constant as the density is still high. Therefore, high dimensional accuracy and density as well as good surface quality can be achieved simultaneously by using high scanning speed during SLMed cellular lattice strut.

Investigation on Selective Laser Melting AlSi10Mg Cellular Lattice Strut: Molten Pool Morphology, Surface Roughness and Dimensional Accuracy.

Science.gov (United States)

Han, Xuesong; Zhu, Haihong; Nie, Xiaojia; Wang, Guoqing; Zeng, Xiaoyan

2018-03-07

AlSi10Mg inclined struts with angle of 45° were fabricated by selective laser melting (SLM) using different scanning speed and hatch spacing to gain insight into the evolution of the molten pool morphology, surface roughness, and dimensional accuracy. The results show that the average width and depth of the molten pool, the lower surface roughness and dimensional deviation decrease with the increase of scanning speed and hatch spacing. The upper surface roughness is found to be almost constant under different processing parameters. The width and depth of the molten pool on powder-supported zone are larger than that of the molten pool on the solid-supported zone, while the width changes more significantly than that of depth. However, if the scanning speed is high enough, the width and depth of the molten pool and the lower surface roughness almost keep constant as the density is still high. Therefore, high dimensional accuracy and density as well as good surface quality can be achieved simultaneously by using high scanning speed during SLMed cellular lattice strut.

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.

Numerical Investigation of Turbulent Natural Convection Heat Transfer in an Internally-Heated Melt Pool and Metallic Layer

International Nuclear Information System (INIS)

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

1999-01-01

This paper presents results of numerical investigation of turbulent natural convection in an internally-heated oxidic pool, and in a metallic layer heated from below and cooled from top and sidewalls. Emphasis is placed upon applicability of the existing heat transfer correlations (obtained from simulant-material experiments) in assessments of a prototypic severe reactor accident. The objectives of this study are (i) to improve the current understanding of the physics of unstably stratified flows, and (ii) to reduce uncertainties associated with modeling and assessment of natural convection heat transfer in the above configuration. Prediction capabilities of different turbulence modeling approaches are first examined and discussed, based on extensive results of numerical investigations performed by present authors. Findings from numerical modeling of turbulent natural convection flow and heat transfer in melt pools and metallic layers are then described. (authors)

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

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.

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

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)

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

Experimental study of natural convection melting of ice in salt solutions

International Nuclear Information System (INIS)

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

1984-01-01

The solid-liquid interface morphology and the micro-physical process near the moving phase boundary during natural convection melting of a horizontal layer of ice by an overlying pool of salt solution were studied experimentally. A cathetometer which amplifies the interface region was used to measure the ice melting rate. Also measured were the temperature transients of the liquid pool. Within the temperature and the density ratio ranges explored, the ice melting rate was found to be very sensitive to the ratio of pool-to-ice melt density but independent of pool-to-ice temperature difference. By varying the density ratio, three different flow regimes and morphologies of the solid-liquid interface were observed, with melt streamers emanating from the crests of the wavy interface into the pool in all three cases. The measured wavelengths (spacing) between the streamers for four different pairs of materials were correlated with the density ratio and found to agree favorably with the predictions of Taylor instability theory

Numerical analysis of weld pool oscillation in laser welding

Energy Technology Data Exchange (ETDEWEB)

Cho, Jung Ho [Chungbuk National University, Cheongju (Korea, Republic of); Farson, Dave F [The Ohio State University, Columbus (United States); Hollis, Kendall; Milewski, John O. [Los Alamos National Laboratory, Los Alamos (United States)

2015-04-15

Volume of fluid (VOF) numerical simulation was used to investigate melt flow and volumetric oscillation of conduction-mode pulsed laser weld pools. The result is compared to high speed video stream of titanium laser spot welding experiment. The total simulation time is 10ms with the first 5 ms being heating and melting under constant laser irradiation and the remaining 5 ms corresponding to resolidification of the weld pool. During the melting process, the liquid pool did not exhibit periodic oscillation but was continually depressed by the evaporation recoil pressure. After the laser pulse, the weld pool was excited into volumetric oscillation by the release of pressure on its surface and oscillation of the weld pool surface was analyzed. The simulation model suggested adjusting thermal diffusivity to match cooling rate and puddle diameter during solidification which is distinguishable from previous weld pool simulation. The frequency continuously increased from several thousand cycles per second to tens of thousands of cycles per second as the weld pool solidified and its diameter decreased. The result is the first trial of investigation of small weld pool oscillation in laser welding although there have been several reports about arc welding.

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.

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

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)

Selective Laser Melting of Pure Copper

Science.gov (United States)

Ikeshoji, Toshi-Taka; Nakamura, Kazuya; Yonehara, Makiko; Imai, Ken; Kyogoku, Hideki

2018-03-01

Appropriate building parameters for selective laser melting of 99.9% pure copper powder were investigated at relatively high laser power of 800 W for hatch pitch in the range from 0.025 mm to 0.12 mm. The highest relative density of the built material was 99.6%, obtained at hatch pitch of 0.10 mm. Building conditions were also studied using transient heat analysis in finite element modeling of the liquidation and solidification of the powder layer. The estimated melt pool length and width were comparable to values obtained by observations using a thermoviewer. The trend for the melt pool width versus the hatch pitch agreed with experimental values.

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.)

Modeling of spreading of the melted corium jet inside the pool of emergency heat removal during severe accidents at NPP

Directory of Open Access Journals (Sweden)

I. V. Kazachkov

2012-03-01

Full Text Available Important nuclear power safety problem in touch with modeling of melted corium jet spreading inside the coolant pool is considered in the paper. It appears by development of the passive protection systems against se-vere accidents. The non-linear mathematical developed model is presented for the jet under reactor vessel pool for one of the perspective passive protection systems and the results of its analysis and studies are given. The performed analysis and the results of the numerical simulation done on the base of the model have allowed estab-lishing the interesting behaviors of the system, which may be useful for the scientists, as well as the engineers-constructors of the passive protection systems against severe accidents.

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.

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.

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)

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)

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

Study of hybrid laser / MAG welding process: characterization of the geometry and the hydrodynamics of the melt pool and development of a 3D thermal model

International Nuclear Information System (INIS)

Le Guen, E.

2010-11-01

Hybrid laser/MIG-MAG welding shows high advantages compared to laser welding or GMAW arc welding used separately. Thanks to this process, higher productivity can be gained through higher welding speed, higher squeeze tolerance moreover possible improvement of the metallurgical properties of the weld seam can be obtained. However, many operating parameters have to be set up in order to achieve optimal process. The complex physical phenomena, which govern welding process, have to be understood in order to use efficiently this technique in mass production. Understanding of these phenomena is also necessary to program numerical simulations which fit to this process. In the first step, experimental studies have been carried out with GMAW, laser and hybrid welding on samples of S355 steel. Influence of operating parameters has been analyzed through films performed with speed camera and macro-graphies of weld seam cross section. Surface deformations of the melt pool, induced by the arc pressure, weld pool length, droplet detachment and welding speed, have been analyzed precisely from images of the surface melt pool. In a second step, a numerical model was developed using the COMSOL Multiphysics software for MAG, laser and hybrid laser/MAG welding processes. A 3D quasi-stationary model has been calculated from the temperature field within the metal. The originality of the MAG and hybrid model lies in the prediction of the melt pool surface profile used to determine the 3D geometry, by taking into account the material input. The influence of different parameters such as arc power and speed welding on the efficiency as well as the distribution radius of the arc power and the arc pressure are analyzed through validations with different experimental results and different calculation configurations. (author)

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

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

Experimental investigation of particulate debris spreading in a pool

Energy Technology Data Exchange (ETDEWEB)

Konovalenko, A., E-mail: kono@kth.se [Division of Nuclear Power Safety, Royal Institute of Technology (KTH) , Roslagstullsbacken 21, Stockholm 106 91 (Sweden); Basso, S., E-mail: simoneb@kth.se [Division of Nuclear Power Safety, Royal Institute of Technology (KTH) , Roslagstullsbacken 21, Stockholm 106 91 (Sweden); Kudinov, P., E-mail: pkudinov@kth.se [Division of Nuclear Power Safety, Royal Institute of Technology (KTH) , Roslagstullsbacken 21, Stockholm 106 91 (Sweden); Yakush, S.E., E-mail: yakush@ipmnet.ru [Institute for Problems in Mechanics of the Russian Academy of Sciences, Ave. Vernadskogo 101 Bldg 1, Moscow 119526 (Russian Federation)

2016-02-15

Termination of severe accident progression by core debris cooling in a deep pool of water under reactor vessel is considered in several designs of light water reactors. However, success of this accident mitigation strategy is contingent upon the effectiveness of heat removal by natural circulation from the debris bed. It is assumed that a porous bed will be formed in the pool in the process of core melt fragmentation and quenching. Debris bed coolability depends on its properties and system conditions. The properties of the bed, including its geometry are the outcomes of the debris bed formation process. Spreading of the debris particles in the pool by two-phase turbulent flows induced by the heat generated in the bed can affect the shape of the bed and thus influence its coolability. The goal of this work is to provide experimental data on spreading of solid particles in the pool by large-scale two-phase flow. The aim is to provide data necessary for understanding of separate effects and for development and validation of models and codes. Validated codes can be then used for prediction of debris bed formation under prototypic severe accident conditions. In PDS-P (Particulate Debris Spreading in the Pool) experiments, air injection at the bottom of the test section is employed as a means to create large-scale flow in the pool in isothermal conditions. The test section is a rectangular tank with a 2D slice geometry, it has fixed width (72 mm), adjustable length (up to 1.5 m) and allows water filling to the depth of up to 1 m. Variable pool length and depth allows studying two-phase circulating flows of different characteristic sizes and patterns. The average void fraction in the pool is determined by video recording and subsequent image processing. Particles are supplied from the top of the facility above the water surface. Results of several series of PDS-P experiments are reported in this paper. The influence of the gas flow rate, pool dimensions, particle density

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)

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

Numerical Analysis on Heat Flux Distribution through the Steel Liner of the Ex-vessel Core Catcher

Energy Technology Data Exchange (ETDEWEB)

Oh, Se Hong; Choi, Choeng Ryul [ELSOLTEC, Yongin (Korea, Republic of); Kim, Byung Jo; Lee, Kyu Bok [KEPCO, Gimcheon (Korea, Republic of); Hwang, Do Hyun [KHNP-CRI, Daejeon (Korea, Republic of)

2016-05-15

In order to prevent material failure of steel container of the core catcher system due to high temperatures, heat flux through the steel liner wall must be kept below the critical heat flux (CHF), and vapor dry-out of the cooling channel must be avoided. In this study, CFD methodology has been developed to simulate the heat flux distribution in the core catcher system, involving following physical phenomena: natural convection in the corium pool, boiling heat transfer and solidification/melting of the corium. A CFD methodology has been developed to simulate the thermal/hydraulic phenomena in the core catcher system, and a numerical analysis has been carried out to estimate the heat flux through the steel liner of the core catcher. High heat flux values are formed at the free surface of the corium pool. However, the heat flux through the steel liner is maintained below the critical heat flux.

Transformation and fragmentation behavior of molten aluminum in sodium pool

International Nuclear Information System (INIS)

Nishimura, S.; Kinoshita, I.; Ueda, N.; Sugiyama, K. I.

2003-01-01

In order to investigate the possibility of fragmentation of the metallic alloy fuel on liquid phase formed by metallurgical reactions, which is important in evaluating the sequence of core disruptive accidents for metallic fuel fast reactors, a series of experiments was carried out using molten aluminum and sodium under the condition that the boiling of sodium on the surface of the melt does not occur. The melting point of aluminum (933K) is roughly equivalent to the liquefaction temperature between the U-Pu-Zr alloy fuel and the SUS cladding (about 923K). The thermal fragmentation of a molten aluminum with a solid crust in the sodium pool is caused by the transient pressurization within the melt confined by the solid crust even under the condition that the instantaneous contact interface temperature between the melt and the sodium is below the boiling point of sodium. This indicates the possibility that the metallic alloy fuel on liquid phase formed by metallurgical reactions can be fragmented without occurring the boiling of sodium on the surface of the melt. The transient pressurization within the melt is considered to be caused by following two mechanisms. i) the overheating of the coolant entrapped hydrodynamically inside the aluminum melt confined by solid crust ii) the progression of solid crust inward and the squeeze of inner liquid part of the aluminum melt confined by solid crust It is found that the degree of fragmentation defined by mass median diameter has the same tendency for different dropping modes (drop or jet) with different mass and ambient Weber number of the melt in the present experimental conditions

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.

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)

Thermal behavior and densification mechanism during selective laser melting of copper matrix composites: Simulation and experiments

International Nuclear Information System (INIS)

Dai, Donghua; Gu, Dongdong

2014-01-01

Highlights: • Thermal behavior and densification activity during SLM of composites are simulated. • Temperature distributions and melt pool dimensions during SLM are disclosed. • Motion behaviors of gaseous bubbles in laser induced melt pool are elucidated. • Simulation results show good agreement with the obtained experimental results. - Abstract: Simulation of temperature distribution and densification process of selective laser melting (SLM) WC/Cu composite powder system has been performed, using a finite volume method (FVM). The transition from powder to solid, the surface tension induced by temperature gradient, and the movement of laser beam power with a Gaussian energy distribution are taken into account in the physical model. The effect of the applied linear energy density (LED) on the temperature distribution, melt pool dimensions, behaviors of gaseous bubbles and resultant densification activity has been investigated. It shows that the temperature distribution is asymmetric with respect to the laser beam scanning area. The center of the melt pool does not locate at the center of the laser beam but slightly shifts towards the side of the decreasing X-axis. The dimensions of the melt pool are in sizes of hundreds of micrometers and increase with the applied LED. For an optimized LED of 17.5 kJ/m, an enhanced efficiency of gas removal from the melt pool is realized, and the maximum relative density of laser processed powder reaches 96%. As the applied LED surpasses 20 kJ/m, Marangoni flow tends to retain the entrapped gas bubbles. The flow pattern has a tendency to deposit the gas bubbles at the melt pool bottom or to agglomerate gas bubbles by the rotating flow in the melt pool, resulting in a higher porosity in laser processed powder. The relative density and corresponding pore size and morphology are experimentally acquired, which are in a good agreement with the results predicted by simulation

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

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

Loss of cooling accident simulation of nuclear power station spent-fuel pool

Energy Technology Data Exchange (ETDEWEB)

Lee, M.; Liang, K-S., E-mail: mlee@ess.nthu.edu.tw, E-mail: ksliang_1@hotmail.com [National Tsing Hua Univ., Hsinchu, Taiwan (China); Lin, K-Y., E-mail: syrup760914@gmail.com [Taiwan Power Company, Taiwan (China)

2014-07-01

The core melt down accident of Fukushima Nuclear Power Station on March 11th, 2011 alerted nuclear industry that the long term loss of cooling of spent fuel pool may need some attention. The target plant analyzed is the Chinshan Nuclear Power Station of Taiwan Power Company. The 3-Dimensional RELAP5 input deck of the spent fuel pool of the station is built. The results indicate that spent fuel of Chinshan Nuclear Power Station is uncovered at 6.75 days after an accident of loss cooling takes place and cladding temperature rises above 2,200{sup o}F around 8 days. The time is about 13 hours earlier than the results predicted using simple energy balance method. The results also show that the impact of Counter Current Flow Limitation (CCFL) and radiation heat transfer model is marginal. (author)

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)

Phenomenological Studies on Melt-Structure-Water Interactions (MSWI) during Postulated Severe Accidents

International Nuclear Information System (INIS)

Sehgal, B.R.; Park, H.S.; Giri, A.; Karbojian, A.; Jasiulevicius, A.; Hansson, R.C.; Chikkanagoudar, U.; Shiferaw, D.; Stepanyan, A.

2004-01-01

This is the annual report for the work performed in year 2003 in the research project 'Melt-Structure-Water Interactions (MSWI) During Severe Accidents in LWRs', under the auspices of the APRI Project, jointly funded by SKI, HSK, and the Swedish and Finnish power companies. The emphasis of the work was placed on phenomena and parameters, which govern the droplet fragmentation in steam explosions, in-vessel and ex-vessel melt/debris coolability, melt pool convection, and the thermal and mechanical loadings of a pressure vessel during melt-vessel interaction. Most research projects in 2002, such as the COMECO, POMECO and MISTEE programs, were continued. An analysis of the FOREVER experiments using the RELAP code to investigate the melt coolability, bubble dynamics and bubble stability to investigate the dynamic behavior of vapor bubble during steam explosions and associated melt fragmentation, quenching boiling experiment to investigate the thermal behavior of single melt droplet were newly initiated. The SIMECO experiment to investigate the three-layer melt pool convection was restarted. The experimental facilities for these projects were fully functional during year 2003. Many of the investigations performed during the course of the MSWI project have produced papers, which have been published in the proceedings of technical meetings and Journals. Significant technical advances were achieved during the course of these studies. These were: A series of experiments on single drop steam explosions was performed to investigate the fine fragmentation process of a metallic melt drop in various thermal conditions. For the first time, transient fine fragmentation process of a melt drop during explosion phase of a steam explosion was visualized continuously and quantified. Different triggering behavior with respect to the coolant subcooling was observed. The analyses on bubble dynamics during a single drop steam explosion and vapor bubble stability estimated the dynamic

Development of LILAC-meltpool for the thermo-hydraulic analysis of core melt relocated in a reactor vessel

International Nuclear Information System (INIS)

Kim, Jong Tae; Kim, Sang Baik; Kim, Hee Dong

2002-03-01

LILAC-meltpool has been developed to study thermo-hydraulic behavior of molten pool and thermal behavior of vessel wall during severe accident. To validate LILAC-meltpool code several two and three dimensional thermo-hydraulic problems were selected and solved. The benchmark problems have experimental results or verified numerical results. Through the validation it was found that LILAC-meltpool reproduces very accurate numerical results. Two-layered semicircular pool was solved to study thermal and hydraulic characteristics of pool stratification. The LAVA experiment using alumina/ferrite molten pool was calculated and compared with computed results. Cooling of alumina/ferrite two-layered pool was affected by stratification. In the numerical results temperature of vessel inner was highest at a location below the interface. Crust was developed from upper surface and lower outer surface, but in the area near the interface corium simulant existed as molten state for long time. LAVA-4 experiment was studied using gap-cooling model in LILAC-meltpool code. Temperature increase of LAVA vessel after alumina melt relocation was strongly dependent on gap formation mechanism. Calculated cooling rates of the vessel were very similar to experimental results. For LAVA experiments which do not have heat generation coolant penetrates easily into a gap and it is found that gap-cooling is very effective for cooling of vessel, but it is thought that coolant penetration could be limited near upper part of gap because of decay heat and high temperature of corium crust

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

Development of severe accident analysis code - A study on the molten core-concrete interaction under severe accidents

Energy Technology Data Exchange (ETDEWEB)

Jung, Chang Hyun; Lee, Byung Chul; Huh, Chang Wook; Kim, Doh Young; Kim, Ju Yeul [Seoul National University, Seoul (Korea, Republic of)

1996-07-01

The purpose of this study is to understand the phenomena of the molten core/concrete interaction during the hypothetical severe accident, and to develop the model for heat transfer and physical phenomena in MCCIs. The contents of this study are analysis of mechanism in MCCIs and assessment of heat transfer models, evaluation of model in CORCON code and verification in CORCON using SWISS and SURC Experiments, and 1000 MWe PWR reactor cavity coolability, and establishment a model for prediction of the crust formation and temperature of melt-pool. The properties and flow condition of melt pool covering with the conditions of severe accident are used to evaluate the heat transfer coefficients in each reviewed model. Also, the scope and limitation of each model for application is assessed. A phenomenological analysis is performed with MELCOR 1.8.2 and MELCOR 1.8.3 And its results is compared with corresponding experimental reports of SWISS and SURC experiments. And the calculation is performed to assess the 1000 MWe PWR reactor cavity coolability. To improve the heat transfer model between melt-pool and overlying coolant and analyze the phase change of melt-pool, 2 dimensional governing equations are established using the enthalpy method and computational program is accomplished in this study. The benchmarking calculation is performed and its results are compared to the experiment which has not considered effects of the coolant boiling and the gas injection. Ultimately, the model shall be developed for considering the gas injection effect and coolant boiling effect. 66 refs., 10 tabs., 29 refs. (author)

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.

A multi-component evaporation model for beam melting processes

Science.gov (United States)

Klassen, Alexander; Forster, Vera E.; Körner, Carolin

2017-02-01

In additive manufacturing using laser or electron beam melting technologies, evaporation losses and changes in chemical composition are known issues when processing alloys with volatile elements. In this paper, a recently described numerical model based on a two-dimensional free surface lattice Boltzmann method is further developed to incorporate the effects of multi-component evaporation. The model takes into account the local melt pool composition during heating and fusion of metal powder. For validation, the titanium alloy Ti-6Al-4V is melted by selective electron beam melting and analysed using mass loss measurements and high-resolution microprobe imaging. Numerically determined evaporation losses and spatial distributions of aluminium compare well with experimental data. Predictions of the melt pool formation in bulk samples provide insight into the competition between the loss of volatile alloying elements from the irradiated surface and their advective redistribution within the molten region.

Analysis of the Processes in Spent Fuel Pools in Case of Loss of Heat Removal due to Water Leakage

Directory of Open Access Journals (Sweden)

Algirdas Kaliatka

2013-01-01

Full Text Available The safe storage of spent fuel assemblies in the spent fuel pools is very important. These facilities are not covered by leaktight containment; thus, the consequences of overheating and melting of fuel in the spent fuel pools can be very severe. On the other hand, due to low decay heat of fuel assemblies, the processes in pools are slow in comparison with processes in reactor core during LOCA accident. Thus, the accident management measures play a very important role in case of some accidents in spent fuel pools. This paper presents the analysis of possible consequences of fuel overheating due to leakage of water from spent fuel pool. Also, the accident mitigation measure, the late injection of water was evaluated. The analysis was performed for the Ignalina NPP Unit 2 spent fuel pool, using system thermal hydraulic code for severe accident analysis ATHLET-CD. The phenomena, taking place during such accident, are discussed. Also, benchmarking of results of the same accident calculation using ASTEC and RELAP/SCDAPSIM codes is presented here.

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

Evaluation of LOCA in a swimming-pool type reactor using the 3D-AIRLOCA code

International Nuclear Information System (INIS)

Nagler, A.; Gilat, J.; Hirshfeld, H.

1991-01-01

The 3D-AIRLOCA code was used to calculate core temperature evolution curves in the wake of a full LOCA in a swimming pool type reactor, resulting in complete core exposure and dryout within about 1000 sec of the initiating event. The results show that fuel integrity loss thresholds (450 C for softening and 650 C for melting) are reached and exceeded over large fractions of the core at powr levels as low as 2 MW. At 4.5 MW, the softening threshold is reached even when the accident occurs up to 12 hours after reactor shutdown for continuous operation, and up to 2 hrs after shutdown for intermittent (6 hrs/day, 4 days a week) operation. The situation is even more severe in blockage cases, when the air flow through the core is blocked by residual water at the grid plate level. It is concluded that substantial fission product releases are quite likely in this class of accidents. (orig.)

Swimming-pool piles

International Nuclear Information System (INIS)

Trioulaire, M.

1959-01-01

In France two swimming-pool piles, Melusine and Triton, have just been set in operation. The swimming-pool pile is the ideal research tool for neutron fluxes of the order of 10 13 . This type of pile can be of immediate interest to many research centres, but its cost must be reduced and a break with tradition should be observed in its design. It would be an advantage: - to bury the swimming-pool; - to reject the experimental channel; - to concentrate the cooling circuit in the swimming-pool; - to carry out all manipulations in the water; - to double the core. (author) [fr

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

Vaporization of elemental mercury from pools of molten lead at low concentrations

International Nuclear Information System (INIS)

Greene, G.A.; Finfrock, C.C.

2000-01-01

Should coolant accidentally be lost to the APT (Accelerator Production of Tritium) blanket and target, and the decay heat in the target be deposited in the surrounding blanket by thermal radiation, temperatures in the blanket modules could exceed structural limits and cause a physical collapse of the blanket modules into a non-coolable geometry. Such a sequence of unmitigated events could result in some melting of the APT blanket and create the potential for the release of mercury into the target-blanket cavity air space. Experiments were conducted which simulate such hypothetical accident conditions in order to measure the rate of vaporization of elemental mercury from pools of molten lead to quantify the possible severe accident source term for the APT blanket region. Molten pools of from 0.01% to 0.10% mercury in lead were prepared under inert conditions. Experiments were conducted, which varied in duration from several hours to as long as a month, to measure the mercury vaporization from the lead pools. The melt pools and gas atmospheres were held fixed at 340 C during the tests. Parameters which were varied in the tests included the mercury concentration, gas flow rate over the melt and agitation of the melt, gas atmosphere composition and the addition of aluminum to the melt. The vaporization of mercury was found to scale roughly linearly with the concentration of mercury in the pool. Variations in the gas flow rates were not found to have any effect on the mass transfer, however agitation of the melt by a submerged stirrer did enhance the mercury vaporization rate. The rate of mercury vaporization with an argon (inert) atmosphere was found to exceed that for an air (oxidizing) atmosphere by as much as a factor of from ten to 20; the causal factor in this variation was the formation of an oxide layer over the melt pool with the air atmosphere which served to retard mass transfer across the melt-atmosphere interface. Aluminum was introduced into the melt to

An experimental study of the behaviour of fission products following an accident on a swimming pool reactor

International Nuclear Information System (INIS)

Dadillon, J.

1976-11-01

In the estimation of nuclear risks connected with the running of a reactor an essential factor, sometimes neglected because insufficiently known, is the knowledge of the type, amount and behaviour of the contamination actually released inside the containment in the case of an accident. In the special case of swimming pool reactors the cooling fluid proves to be a very efficient barrier against contamination. Three experiments were carried out in the reactor CABRI, during which several fuel element plates were melted inside the core itself. (Author)

A simplified model of aerosol scrubbing by a water pool overlying core debris interacting with concrete

International Nuclear Information System (INIS)

Powers, D.A.; Sprung, J.L.

1993-11-01

A classic model of aerosol scrubbing from bubbles rising through water is applied to the decontamination of gases produced during core debris interactions with concrete. The model, originally developed by Fuchs, describes aerosol capture by diffusion, sedimentation, and inertial impaction. This original model for spherical bubbles is modified to account for ellipsoidal distortion of the bubbles. Eighteen uncertain variables are identified in the application of the model to the decontamination of aerosols produced during core debris interactions with concrete by a water pool of specified depth and subcooling. These uncertain variables include properties of the aerosols, the bubbles, the water and the ambient pressure. Results are analyzed using a nonparametric, order statistical analysis that allows quantitative differentiation of stochastic and phenomenological uncertainty. The sampled values of the decontamination factors are used to construct estimated probability density functions for the decontamination factor at confidence levels of 50%, 90% and 95%. The decontamination factors for pools 30, 50, 100, 200, 300, and 500 cm deep and subcooling levels of 0, 2, 5, 10, 20, 30, 50, and 70 degrees C are correlated by simple polynomial regression. These polynomial equations can be used to estimate decontamination factors at prescribed confidence levels

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

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

Void fraction for gas bubbling in shallow viscous pools-application to molten core concrete interaction

International Nuclear Information System (INIS)

Journeau, C.; Haquet, J.F.

2005-01-01

During Molten Core-Concrete Interaction, the concrete will release gases (mainly steam and carbon oxides) that will flow through the corium pool. To obtain reliable heat transfer prediction, it is necessary to model the void fraction in the pool as a function of the gas mass flow (or superficial velocity at the interface). A series of simulant-materials have been performed with water-air and sugar syrup-air in order to study how the drift model could be applied to a shallow pool (where the bubbly flow is not fully developed) and to liquids which are more viscous (with higher Morton numbers) than water. The bubble average diameter was estimated around 3 mm with spherical to ellipsoidal shapes. For all the configurations, even with the shallowest pools (6 cm height for 38 cm diameter) the experimental void fractions follow the drift-model relationship. In water, the distribution coefficient C 0 tends to the classical value of 1.2 while the drift velocity V jg tends to the 23 cm/s predicted by Ishii (1975) model for churn flows. For the more viscous syrup, the drift velocity tends to 13 cm/s which is significantly lower than the value obtained from the Ishii correlation for bubbly or churn flows (established for water). These results are then applied to MCCI experimental configurations. (authors)

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)

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.

Kinetic approach in numerical modeling of melting and crystallization at laser cladding with powder injection

Energy Technology Data Exchange (ETDEWEB)

Mirzade, F. Kh., E-mail: fmirzade@rambler.ru [Institute on Laser and Information Technology, Russian Academy of Sciences, 1 Svyatoozerskaya Street, Shatura, Moscow Region 140700 (Russian Federation); Niziev, V.G.; Panchenko, V. Ya.; Khomenko, M.D.; Grishaev, R.V. [Institute on Laser and Information Technology, Russian Academy of Sciences, 1 Svyatoozerskaya Street, Shatura, Moscow Region 140700 (Russian Federation); Pityana, S.; Rooyen, Corney van [CSIR-National Laser Centre, Building 46A, Meiring Nauder Road, Brummeria, Pretoria (South Africa)

2013-08-15

The numerical model of laser cladding with coaxial powder injection includes the equations for heat transfer, melting and crystallization kinetics. It has been shown that the main parameters influencing the melt pool dynamics and medium maximum temperature are mass feed rate, laser power and scanning velocity. It has been observed that, due to the phase change occurring with superheating/undercooling, the melt zone has the boundary distinguished from melting isotherm. The calculated melt pool dimensions and dilution are in a good agreement with the experimental results for cladding of 431 martensitic stainless steel onto carbon steel substrate.

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

Heat and mass transfer in a liquid pool with wall ablation and composition effects

International Nuclear Information System (INIS)

Pham, Q.T.

2013-01-01

This work deals with the thermal-hydraulics of a melt pool coupled with the physical chemistry for the purpose of describing the behaviour of mixtures of materials (non-eutectic). Evolution of transient temperature in a liquid melt pool heated by volumetric power dissipation has been described with solidification on the cooled wall. The model has been developed and is validated for the experimental results given by LIVE experiment, performed at Karlsruhe Institute of Technology (KIT) in Germany. Under the conditions of these tests, it is shown that the interface temperature follows the liquidus temperature (corresponding to the composition of the liquid bath) during the whole transient. Assumption of interface temperature as liquidus temperature allows recalculating the evolution of the maximum melt temperature as well as the local crust thickness. Furthermore, we propose a model for describing the interaction between a non-eutectic liquid melt pool (subjected to volumetric power dissipation) and an ablated wall whose melting point is below the liquidus temperature of the melt. The model predictions are compared with results of ARTEMIS 2D tests. A new formulation of the interface temperature between the liquid melt and the solid wall (below liquidus temperature) has been proposed. (author) [fr

Numerical simulation of passive heat removal under severe core meltdown scenario in a sodium cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

David, Dijo K.; Mangarjuna Rao, P., E-mail: pmr@igcar.gov.in; Nashine, B.K.; Selvaraj, P.; Chellapandi, P.

2015-09-15

Highlights: • PAHR in SFR under large core relocation to in-vessel core catcher is numerically analyzed. • A 1-D thermal conduction model and a 2-D axisymmetric CFD model are developed for turbulent natural convection phenomenon. • The side pool (cold pool) was found out to be instrumental in storing heat and dissipating it to the heat sink. • Single tray type in-vessel core catcher is found to be thermally effective under one-fourth core relocation. - Abstract: A sequence of highly unlikely events leading to significant meltdown of the Sodium cooled Fast Reactor (SFR) core can cause the failure of reactor vessel if the molten fuel debris settles at the bottom of the reactor main vessel. To prevent this, pool type SFRs are usually provided with an in-vessel core catcher above the bottom wall of the main vessel. The core catcher should collect, retain and passively cool these debris by facilitating decay heat removal by natural convection. In the present work, the heat removal capability of the existing single tray core catcher design has been evaluated numerically by analyzing the transient development of natural convection loops inside SFR pool. A 1-D heat diffusion model and a simplified 2-D axi-symmetric CFD model are developed for the same. Maximum temperature of the core catcher plate evaluated for different core meltdown scenarios using these models showed that there is much higher heat removal potential for single tray in-vessel SFR core catcher compared to the design basis case of melting of 7 subassemblies under total instantaneous blockage of a subassembly. The study also revealed that the side pool of cold sodium plays a significant role in decay heat removal. The maximum debris bed temperature attained during the initial hours of PAHR does not depend much on when the Decay Heat Exchanger (DHX) gets operational, and it substantiates the inherent safety of the system. The present study paves the way for better understanding of the thermal

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

Influence of scan strategy and molten pool configuration on microstructures and tensile properties of selective laser melting additive manufactured aluminum based parts

Science.gov (United States)

Dai, Donghua; Gu, Dongdong; Zhang, Han; Xiong, Jiapeng; Ma, Chenglong; Hong, Chen; Poprawe, Reinhart

2018-02-01

Selective laser melting additive manufacturing of the AlSi12 material parts through the re-melting of the previously solidified layer using the continuous two layers 90° rotate scan strategy was conducted. The influence of the re-melting behavior and scan strategy on the formation of the ;track-track; and ;layer-layer; molten pool boundaries (MPBs), dimensional accuracy, microstructure feature, tensile properties, microscopic sliding behavior and the fracture mechanism as loaded a tensile force has been studied. It showed that the defects, such as the part distortion, delamination and cracks, were significantly eliminated with the deformation rate less than 1%. The microstructure of a homogeneous distribution of the Si phase, no apparent grain orientation on both sides of the MPBs, was produced in the as-fabricated part, promoting the efficient transition of the load stress. Cracks preferentially initiate at the ;track-track; MPBs when the tensile stress increases to a certain value, resulting in the formation of the cleavage steps along the tensile loading direction. The cracks propagate along the ;layer-layer; MPBs, generating the fine dimples. The mechanical behavior of the SLM-processed AlSi12 parts can be significantly enhanced with the ultimate tensile strength, yield strength and elongation of 476.3 MPa, 315.5 MPa and 6.7%, respectively.

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

Thermal and hydraulic behaviour of CANDU cores under severe accident conditions - final report. Vol. 1

International Nuclear Information System (INIS)

Rogers, J.T.

1984-06-01

This report gives the results of a study of the thermo-hydraulic aspects of severe accident sequences in CANDU reactors. The accident sequences considered are the loss of the moderator cooling system and the loss of the moderator heat sink, each following a large loss-of-coolant accident accompanied by loss of emergency coolant injection. Factors considered include expulsion and boil-off of the moderator, uncovery, overheating and disintegration of the fuel channels, quenching of channel debris, re-heating of channel debris following complete moderator expulsion, formation and possible boiling of a molten pool of core debris and the effectiveness of the cooling of the calandria wall by the shield tank water during the accident sequences. The effects of these accident sequences on the reactor containment are also considered. Results show that there would be no gross melting of fuel during moderator expulsion from the calandria, and for a considerable time thereafter, as quenched core debris re-heats. Core melting would not begin until about 135 minutes after accident initiation in a loss of the moderator cooling system and until about 30 minutes in a loss of the moderator heat sink. Eventually, a pool of molten material would form in the bottom of the calandria, which may or may not boil, depending on property values. In all cases, the molten core would be contained within the calandria, as long as the shield tank water cooling system remains operational. Finally, in the period from 8 to 50 hours after the initiation of the accident, the molten core would re-solidify within the calandria. There would be no consequent damage to containment resulting from these accident sequences, nor would there be a significant increase in fission product releases from containment above those that would otherwise occur in a dual failure LOCA plus LOECI

Melt pool modelling, simulation and experimental validation for SLM

NARCIS (Netherlands)

Wits, Wessel

2017-01-01

SLM parts are built by successively melting layers of powder in a powder bed. Process parameters are often optimized experimentally by laser scanning a number of single tracks and subsequently determining which settings lead to a good compromise between quality and build speed. However,

Coatings with laser melt injection of ceramic particles

NARCIS (Netherlands)

de Hosson, J.T.M.; Ocelik, V.; de Oliveira, U.; Seal, S; Dahotre, NB; Moore, JJ; Suryanarayana, C; Agarwal, A

2003-01-01

The conditions for a successful Laser Melt Injection (LMI) of SiC and WC particles into the melt pool of Al8Si and Ti6Al4V alloys were studied experimentally and theoretically by FEM calculations. The laser employed is a high power Nd:YAG The formation of a relatively thick aluminium oxide layer on

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.

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

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.

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

International Nuclear Information System (INIS)

Seiichi, Koshizuka; Shoji, Matsuura; Mizue, Sekine; Yoshiaki, Oka

2001-01-01

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

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

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

Melt pool vorticity in deep penetration laser material welding

Indian Academy of Sciences (India)

weld pool has been evaluated in case of high power CO2 laser beam welding. The ... The experiments based on twin or triple spot interaction geometry have also ... while the other one is between the liquid and the solid states of the metal.

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

Loss of spent fuel pool cooling PRA: Model and results

International Nuclear Information System (INIS)

Siu, N.; Khericha, S.; Conroy, S.; Beck, S.; Blackman, H.

1996-09-01

This letter report documents models for quantifying the likelihood of loss of spent fuel pool cooling; models for identifying post-boiling scenarios that lead to core damage; qualitative and quantitative results generated for a selected plant that account for plant design and operational practices; a comparison of these results and those generated from earlier studies; and a review of available data on spent fuel pool accidents. The results of this study show that for a representative two-unit boiling water reactor, the annual probability of spent fuel pool boiling is 5 x 10 -5 and the annual probability of flooding associated with loss of spent fuel pool cooling scenarios is 1 x 10 -3 . Qualitative arguments are provided to show that the likelihood of core damage due to spent fuel pool boiling accidents is low for most US commercial nuclear power plants. It is also shown that, depending on the design characteristics of a given plant, the likelihood of either: (a) core damage due to spent fuel pool-associated flooding, or (b) spent fuel damage due to pool dryout, may not be negligible

Post-accident heat removal ''information exchange''

International Nuclear Information System (INIS)

Plein, H.G.; Carlson, G.A.

1975-01-01

The in-core molten pool experiments are designed to produce a pool of fission heated temperature and flow patterns of such pools, and evaluate the barrier melt-through potential of the molten UO 2 . The first experiments, to be conducted this fiscal year in the Annular Core Pulse Reactor, will be uncomplicated and multiply-contained to prove containment design and to provide initial information on fission heated molten pool characteristics. Concurrent with the in-core experiments, high temperature ultrasonic techniques are being developed to measure UO 2 temperatures up to and above the melting point for use in later more definitive experiments scheduled for FY77

XCT analysis of the influence of melt strategies on defect population in Ti–6Al–4V components manufactured by Selective Electron Beam Melting

International Nuclear Information System (INIS)

Tammas-Williams, S.; Zhao, H.; Léonard, F.; Derguti, F.; Todd, I.; Prangnell, P.B.

2015-01-01

Selective Electron Beam Melting (SEBM) is a promising powder bed Additive Manufacturing technique for near-net-shape manufacture of high-value titanium components. However without post-manufacture HIPing the fatigue life of SEBM parts is currently dominated by the presence of porosity. In this study, the size, volume fraction, and spatial distribution of the pores in model samples have been characterised in 3D, using X-ray Computed Tomography, and correlated to the process variables. The average volume fraction of the pores (< 0.2%) was measured to be lower than that usually observed in competing processes, such as selective laser melting, but a strong relationship was found with the different beam strategies used to contour, and infill by hatching, a part section. The majority of pores were found to be small spherical gas pores, concentrated in the infill hatched region; this was attributed to the lower energy density and less focused beam used in the infill strategy allowing less opportunity for gas bubbles to escape the melt pool. Overall, increasing the energy density or focus of the beam was found to correlate strongly to a reduction in the level of gas porosity. Rarer irregular shaped pores were mostly located in the contour region and have been attributed to a lack of fusion between powder particles. - Graphical abstract: Display Omitted - Highlights: • Vast majority of defects detected were small spherical gas pores. • Gas bubbles trapped in the powder granules expand and coalesce in the melt pool. • Pores have been shown not to be randomly distributed. • Larger and deeper melt pools give more opportunity for gas to escape. • Minor changes to melt strategy result in significant reductions in pore population

XCT analysis of the influence of melt strategies on defect population in Ti–6Al–4V components manufactured by Selective Electron Beam Melting

Energy Technology Data Exchange (ETDEWEB)

Tammas-Williams, S., E-mail: Samuel.tammas-wiliams@manchester.ac.uk [School of Materials, University of Manchester, Manchester M13 9PL (United Kingdom); Department of Materials Science and Engineering, University of Sheffield, Sheffield S1 3JD (United Kingdom); Zhao, H. [School of Materials, University of Manchester, Manchester M13 9PL (United Kingdom); Department of Materials Science and Engineering, University of Sheffield, Sheffield S1 3JD (United Kingdom); Léonard, F. [School of Materials, University of Manchester, Manchester M13 9PL (United Kingdom); Derguti, F.; Todd, I. [Department of Materials Science and Engineering, University of Sheffield, Sheffield S1 3JD (United Kingdom); Prangnell, P.B. [School of Materials, University of Manchester, Manchester M13 9PL (United Kingdom)

2015-04-15

Selective Electron Beam Melting (SEBM) is a promising powder bed Additive Manufacturing technique for near-net-shape manufacture of high-value titanium components. However without post-manufacture HIPing the fatigue life of SEBM parts is currently dominated by the presence of porosity. In this study, the size, volume fraction, and spatial distribution of the pores in model samples have been characterised in 3D, using X-ray Computed Tomography, and correlated to the process variables. The average volume fraction of the pores (< 0.2%) was measured to be lower than that usually observed in competing processes, such as selective laser melting, but a strong relationship was found with the different beam strategies used to contour, and infill by hatching, a part section. The majority of pores were found to be small spherical gas pores, concentrated in the infill hatched region; this was attributed to the lower energy density and less focused beam used in the infill strategy allowing less opportunity for gas bubbles to escape the melt pool. Overall, increasing the energy density or focus of the beam was found to correlate strongly to a reduction in the level of gas porosity. Rarer irregular shaped pores were mostly located in the contour region and have been attributed to a lack of fusion between powder particles. - Graphical abstract: Display Omitted - Highlights: • Vast majority of defects detected were small spherical gas pores. • Gas bubbles trapped in the powder granules expand and coalesce in the melt pool. • Pores have been shown not to be randomly distributed. • Larger and deeper melt pools give more opportunity for gas to escape. • Minor changes to melt strategy result in significant reductions in pore population.

Numerical analysis of the effects of non-conventional laser beam geometries during laser melting of metallic materials

International Nuclear Information System (INIS)

Safdar, Shakeel; Li, Lin; Sheikh, M A

2007-01-01

Laser melting is an important industrial activity encountered in a variety of laser manufacturing processes, e.g. selective laser melting, welding, brazing, soldering, glazing, surface alloying, cladding etc. The majority of these processes are carried out by using either circular or rectangular beams. At present, the melt pool characteristics such as melt pool geometry, thermal gradients and cooling rate are controlled by the variation of laser power, spot size or scanning speed. However, the variations in these parameters are often limited by other processing conditions. Although different laser beam modes and intensity distributions have been studied to improve the process, no other laser beam geometries have been investigated. The effect of laser beam geometry on the laser melting process has received very little attention. This paper presents an investigation of the effects of different beam geometries including circular, rectangular and diamond shapes on laser melting of metallic materials. The finite volume method has been used to simulate the transient effects of a moving beam for laser melting of mild steel (EN-43A) taking into account Marangoni and buoyancy convection. The temperature distribution, melt pool geometry, fluid flow velocities and heating/cooling rates have been calculated. Some of the results have been compared with the experimental data

Evaporation-induced gas-phase flows at selective laser melting

Science.gov (United States)

Zhirnov, I.; Kotoban, D. V.; Gusarov, A. V.

2018-02-01

Selective laser melting is the method for 3D printing from metals. A solid part is built from powder layer-by-layer. A continuum-wave laser beam scans every powder layer to fuse powder. The process is studied with a high-speed CCD camera at the frame rate of 104 fps and the resolution up to 5 µm per pixel. Heat transfer and evaporation in the laser-interaction zone are numerically modeled. Droplets are ejected from the melt pool in the direction around the normal to the melt surface and the powder particles move in the horizontal plane toward the melt pool. A vapor jet is observed in the direction of the normal to the melt surface. The velocities of the droplets, the powder particles, and the jet flow and the mass loss due to evaporation are measured. The gas flow around the vapor jet is calculated by Landau's model of submerged jet. The measured velocities of vapor, droplets, and powder particles correlate with the calculated flow field. The obtained results show the importance of evaporation and the flow of the vapor and the ambient gas. These gas-dynamic phenomena can explain the formation of the denudated zones and the instability at high-energy input.

Oxidation kinetics of corium pool

International Nuclear Information System (INIS)

Sulatsky, A.A.; Smirnov, S.A.; Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Kotova, S.Yu.; Fischer, M.; Hellmann, S.; Tromm, W.; Miassoedov, A.; Bottomley, D.; Piluso, P.; Barrachin, M.

2013-01-01

Highlights: • The analysis of experimental data on molten corium oxidation was been carried out. • The analysis has revealed the main factors influencing the oxidation kinetics. • The analysis was used for developing a qualitative analytical model. • The numerical modeling has confirmed the results of experimental data analysis. -- Abstract: Experimental, theoretical and numerical studies of oxidation kinetics of an open surface corium pool have been reported. The experiments have been carried out within OECD MASCA program and ISTC METCOR, METCOR-P and EVAN projects. It has been shown that the melt oxidation is controlled by an oxidant supply to the melt free surface from the atmosphere, not by the reducer supply from the melt. The project experiments have not detected any input of the zirconium oxidation kinetics into the process chemistry. The completed analysis puts forward a simple analytical model, which gives an explanation of the main features of melt oxidation process. The numerical modeling results are in good agreement with experimental data and theoretical considerations

Oxidation kinetics of corium pool

Energy Technology Data Exchange (ETDEWEB)

Sulatsky, A.A., E-mail: andrei314@mail.ru [Alexandrov Research Institute of Technologies (NITI), Sosnovy Bor (Russian Federation); Smirnov, S.A. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (NIIEFA), St. Petersburg (Russian Federation); Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Kotova, S.Yu. [Alexandrov Research Institute of Technologies (NITI), Sosnovy Bor (Russian Federation); Fischer, M.; Hellmann, S. [AREVA NP GmbH, Erlangen (Germany); Tromm, W.; Miassoedov, A. [Forschungzentrum Karlsruhe (FZK), Karlsruhe (Germany); Bottomley, D. [EUROPÄISCHE KOMMISSION, Joint Research Centre Institut für Transurane (ITU), Karlsruhe (Germany); Piluso, P. [CEA Cadarache-DEN/DTN/STRI, St.Paul-lez-Durance (France); Barrachin, M. [Institut de Radioprotection et Sûreté Nucléaire, St.Paul-lez-Durance (France)

2013-09-15

Highlights: • The analysis of experimental data on molten corium oxidation was been carried out. • The analysis has revealed the main factors influencing the oxidation kinetics. • The analysis was used for developing a qualitative analytical model. • The numerical modeling has confirmed the results of experimental data analysis. -- Abstract: Experimental, theoretical and numerical studies of oxidation kinetics of an open surface corium pool have been reported. The experiments have been carried out within OECD MASCA program and ISTC METCOR, METCOR-P and EVAN projects. It has been shown that the melt oxidation is controlled by an oxidant supply to the melt free surface from the atmosphere, not by the reducer supply from the melt. The project experiments have not detected any input of the zirconium oxidation kinetics into the process chemistry. The completed analysis puts forward a simple analytical model, which gives an explanation of the main features of melt oxidation process. The numerical modeling results are in good agreement with experimental data and theoretical considerations.

Convective cooling in a pool-type research reactor

Science.gov (United States)

Sipaun, Susan; Usman, Shoaib

2016-01-01

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

Controlling liquid pool depth in VAR of a 21.6 cm diameter ingot of Alloy 718

Science.gov (United States)

Lopez, Felipe; Beaman, Joseph; Williamson, Rodney; Taleff, Eric; Watt, Trevor

It is believed that the final microstructure in vacuum arc remelted (VAR) ingots is strongly influenced by the molten metal pool profile. Thus, if the pool profile was properly controlled during the melt then defect-free microstructures would be obtained. The recent development of a reduced-order model of VAR solidification allowed the design of a pool depth controller to accomplish this task. The controller used a linear quadratic regulator and a Kalman filter to stabilize the melt pool solidification front under the effect of uncertain process dynamics and noisy measurements. Basic Axisymmetric Remelting (BAR), a high-fidelity VAR ingot model, was used in real time to provide pool depth measurements that were incorporated in the control loop. The controller was tested at Los Alamos National Laboratory in a 21.6 diameter Alloy 718 ingot. Details of the controller design will be presented, along with comparisons to experimentally-measured pool depths.

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

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.

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

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.

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

Fundamental study on flow characteristics of disrupted core pool at a low energy level (Joint research)

International Nuclear Information System (INIS)

Morita, Koji; Liu, Ping; Matsumoto, Tatsuya; Fukuda, Kenji; Tobita, Yoshiharu; Sato, Ikken

2007-03-01

Dynamic behaviors of solid particle beds in a liquid pool against pressure transients were investigated to model the mobility of core materials in a low-energy disrupted core of a liquid metal fast reactor. A series of experiments was performed with a particle bed of different heights, comprising different monotype solid particles, where variable initial pressures of the originally pressurized nitrogen gas were adopted as the pressure source. Computational simulations of the experiments were performed using SIMMER-III, a fast reactor safety analysis code. Experimental analyses using the SIMMER-III code show that physical models and method used in the code can reasonably represent the transient behaviors of multiphase flows with rich solid phase as observed in the experiments. The validation of several key models of SIMMER-III was also discussed for treating transient behaviors of the solid-particle phase in multiphase flows. (author)

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

The postirradiation examination of the DC [Dry Capsule] melt dynamics experiments

International Nuclear Information System (INIS)

Fryer, C.P.; Hitchcock, J.T.

1988-07-01

The Dry Capsule experiment series investigates the coolability of dry fast reactor core debris through nuclear heating of actual reactor materials in order to obtain the thermal properties of dry debris, the nature of the transition from a debris bed to a molten pool, and the thermal and kinetic behavior of molten pools. The purpose is to develop a data base in support of model development. The work is jointly sponsored by the US Nuclear Regulatory Commission (USNRC), the Power Reactor and Nuclear Fuel Development Corporation (PNC, Japan), and Joint Research Centre, Ispra (EURATOM). This report provides a brief description of the two experiments in the Dry Capsule series and presents the results of the postirradiation examination. These tests investigated dry debris beds (/approximately/2kg) composed of pure UO 2 and mixed UO 2 and stainless steel. The beds were taken into melt to observe the growth of a molten pool in the UO 2 bed and the agglomeration and migration of steel in a composite bed. The peak measured temperature in the UO 2 bed was above 3100/degree/C. Approximately 50 percent of the urania formed a molten pool. Surrounding the molten pool and an overlying void was a high density urania crust. Lenticular pore formation and migration caused by urania vapor transport in the strong thermal gradient were seen in the urania crust. In the mixed UO 2 and steel bed, the peak measured temperature was 2600/degree/ C. All of the steel had been molten. Steel and urania migration were observed; both were thought to be caused by vapor transport mechanisms

Technical specification for fabrication of HANARO pool cover

Energy Technology Data Exchange (ETDEWEB)

Ryu, Jeong Soo; Woo, Sang Ik

2001-06-01

This technical specification details the requirements and the acceptance criteria for design, seismic analysis, function test, installation and quality assurance for HANARO pool cover which will be installed at the top of reactor pool. The pool cover is classified as non-nuclear safety, seismic category II and quality class T. The basic design of the pool cover for increasing HANARO applications has been carried out for supporting the driving devices which can load, unload and rotate the irradiation targets in the in-core and out-core vertical irradiation holes under on-power operation. The comments of HANARO user group related with irradiation tests have optimally considered in the process of design. The interference between fuel handling and control absorber units in the reactor pool and activities to load, unload and rotate the irradiation targets at the top of the reactor pool have been minimized. The pool cover can be moved for maintenance and can protect the reactor pool from unexpected drop of foreign materials. It provides the space to vertical access of driving devices for NTD, CT/IR and OR4/OR5 under on-power operation. And the pool cover assembly must maintain its structural integrity under seismic load. Based on the above design concept, the HANARO pool cover has been proposed as supporting structure of driving devices for NTD, fission moly and RI production under on-power operation.

Convective cooling in a pool-type research reactor

Energy Technology Data Exchange (ETDEWEB)

Sipaun, Susan, E-mail: susan@nm.gov.my [Malaysian Nuclear Agency, Industrial Technology Division, Blok 29T, Bangi 43200, Selangor (Malaysia); Usman, Shoaib, E-mail: usmans@mst.edu [Missouri University of Science and Technology, Nuclear Engineering, 222 Fulton Hall 301 W.14th St., Rolla 64509 MO (United States)

2016-01-22

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.

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.

Controlling pool depth during VAR of Alloy 718

Science.gov (United States)

Lopez, F.; Beaman, J.; Williamson, R.; Evans, D.

2016-07-01

A longtime goal of superalloy producers has been to control the geometry of the liquid pool in solidifying ingots. Accurate pool depth control at appropriate values is expected to result in ingots free of segregation defects. This article describes an industrial VAR experiment in which a 430mm (17 in) diameter Alloy 718 electrode was melted into a 510mm (20 in) ingot. In the experiment, the depth of the liquid pool at the mid-radius was controlled to three different set-points: 137 mm (nominal), 193 mm (deep) and 118 mm (shallow). At each level, the pool depth was marked by a power cutback of several minutes. The ingot was sectioned and longitudinal slices were cut out. Analysis of the photographed ingot revealed that accurate control was obtained for both the nominal and deep pool cases, while the third one was not conclusive.

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)

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

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.

Heat transfer modelling and stability analysis of selective laser melting

International Nuclear Information System (INIS)

Gusarov, A.V.; Yadroitsev, I.; Bertrand, Ph.; Smurov, I.

2007-01-01

The process of direct manufacturing by selective laser melting basically consists of laser beam scanning over a thin powder layer deposited on a dense substrate. Complete remelting of the powder in the scanned zone and its good adhesion to the substrate ensure obtaining functional parts with improved mechanical properties. Experiments with single-line scanning indicate, that an interval of scanning velocities exists where the remelted tracks are uniform. The tracks become broken if the scanning velocity is outside this interval. This is extremely undesirable and referred to as the 'balling' effect. A numerical model of coupled radiation and heat transfer is proposed to analyse the observed instability. The 'balling' effect at high scanning velocities (above ∼20 cm/s for the present conditions) can be explained by the Plateau-Rayleigh capillary instability of the melt pool. Two factors stabilize the process with decreasing the scanning velocity: reducing the length-to-width ratio of the melt pool and increasing the width of its contact with the substrate

Convection-driven melting in an n-octane pool fire bounded by an ice wall

DEFF Research Database (Denmark)

Farahani, Hamed Farmahini; Alva, Wilson Ulises Rojas; Rangwala, Ali S.

2017-01-01

×3cm) placed on one side of the tray. The melting front velocity, as an indicator of the melting rate of the ice, increased from 0.04cm/min to 1cm/min. The measurement of the burning rates and flame heights showed two distinctive behaviors; an induction period from the initial self-sustained flame...... to the multi-roll location. The multi-roll structure could be the main reason for the transport of the heat received from the flame toward the ice wall which causes the melting....

Experimental study of the fragmentation and quench behavior of corium melts in water

International Nuclear Information System (INIS)

Wang, S.K.; Blomquist, C.A.; Spencer, B.W.; McUmber, L.M.; Schneider, J.P.; Illinois Univ., Urbana, IL

1989-01-01

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: (1) the thermal attack and possible damage to the RPV lower head from the impinging molten fuel stream and/or the debris bed, (2) the molten fuel relocation pathways including the effects of redistribution due to core support structure and the reactor lower internals, (3) the quench rate of the molten fuel through the water in the lower plasma, (4) the steam generation and hydrogen generation during the interaction, (5) the transient pressurization of the primary system, and (6) 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. 9 refs., 29 figs

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.)

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.)

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

Location specific solidification microstructure control in electron beam melting of Ti-6Al-4V

Energy Technology Data Exchange (ETDEWEB)

Narra, Sneha P.; Cunningham, Ross; Beuth, Jack; Rollett, Anthony D.

2018-01-01

Relationships between prior beta grain size in solidified Ti-6Al-4V and melting process parameters in the Electron Beam Melting (EBM) process are investigated. Samples are built by varying a machine-dependent proprietary speed function to cover the process space. Optical microscopy is used to measure prior beta grain widths and assess the number of prior beta grains present in a melt pool in the raster region of the build. Despite the complicated evolution of beta grain sizes, the beta grain width scales with melt pool width. The resulting understanding of the relationship between primary machine variables and prior beta grain widths is a key step toward enabling the location specific control of as-built microstructure in the EBM process. Control of grain width in separate specimens and within a single specimen is demonstrated.

"Ripples" in an Aluminum Pool?

Science.gov (United States)

Rohr, James; Wang, Si-Yin; Nesterenko, Vitali F.

2018-05-01

Our motivation for this article is for students to realize that opportunities for discovery are all around them. Discoveries that can still puzzle present day researchers. Here we explore an observation by a middle school student concerning the production of what appears to be water-like "ripples" produced in aluminum foil when placed between two colliding spheres. We both applaud and explore the student's reasoning that the ripples were formed in a melted aluminum pool.

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

Analysis of a molten pool natural convection in the APR1400 RPV at a severe accident

International Nuclear Information System (INIS)

Kim, Jong Tae; Park, Rae Joon; Kim, Sang Baik

2005-01-01

During a hypothetical severe accident, reactor fuel rods and structures supporting them are melted and relocated in the lower head of the reactor vessel. These relocated molten materials could be separated by their density difference and construct metal/oxide stratified pools in the lower head. A decay heat generated from the fuel material is transferred to the vessel wall and upper structures remaining in the reactor vessel by natural convection. As shown in Fig. 1 two-layered stratified molten pool is developed in the reactor lower vessel. The oxidic pool usually constructed by the mixture of uranium oxide and zirconium oxide. The melting temperature of the oxidic material is very high compared to the steel vessel and metallic layer. And highly turbulent natural convection generated by the decay heat enhances heat transfer to the boundary of the oxidic pool. By this thermal mechanism, oxide curst is developed around the oxidic layer as shown in Fig. 1. The oxidic pool is bounded thermally and fluid-dynamically by the developed crust. By this boundedness, the heat transfer structure in the stratified oxidic/metallic pool can be solved separately. The thermal boundary condition of the oxidic pool is isothermal with constant melting temperature of the oxidic material. The decay heat is transfer to side wall and upper interface between oxidic and metallic layer. Turbulent natural convection is dominant heat transfer mechanism in the oxidic pool. The heat transferred from the bottom oxidic layer is imposed to the upper metallic layer. This transferred heat in the metallic pool is removed through side and upper surface, which is augmented also by natural convection developed in the pool. In this study, a molten pool natural convection in the APR1400 RPV during a severe accident is simulated using the Lilac code and the calculated heat flux distribution on the reactor vessel wall is compared with a lumped-parameter (LP) prediction

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)

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.

Quench cooling of superheated debris beds in containment during LWR core meltdown accidents

International Nuclear Information System (INIS)

Ginsberg, T.; Chen, J.C.

1984-01-01

Light water reactor core meltdown accident sequence studies suggest that superheated debris beds may settle on the concrete floor beneath the reactor vessel. A model for the heat transfer processes during quench of superheated debris beds cooled by an overlying pool of water has been presented in a prior paper. This paper discusses the coolability of decay-heated debris beds from the standpoint of their transient quench characteristics. It is shown that even though a debris bed configuration may be coolable from the point of view of steady-state decay heat removal, the quench behavior from an initially elevated temperature may lead to bed melting prior to quench of the debris

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

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.

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

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

Influence of the oxidation of a molten pool on fission product release by the computer code RELOS

International Nuclear Information System (INIS)

Kleinhietpass, I.D.

2006-01-01

The objective of reactor safety research is to prevent the release of radionuclides into the environment of a nuclear power plant. Although Light Water Reactors are well configured by a multitude of safety systems and a severe accident seems very unlikely to happen, risk analyses are made implying a severe accident. In that the reactor core might melt in consequence of insuffient cooling, relocate and accumulate forming a molten pool and thus providing a potential contribution to the aerosol source term. For the investigation of the release of radionuclides from a molten pool the code RELOS is under development at the Chair of Energy Systems and Energy Economics of the Ruhr- Universitaet Bochum. RELOS calculates for a multi-component/multi-phase system that amount of a component which evaporates from a hot liquid phase into a cooler gas atmosphere. The release behaviour of a radionuclide is determined by its volatility. Chemical reactions of less volatile elements could produce components being more volatile. Particularly the formation of (higher) metallic oxides could influence the volatility in evidence. In RELOS chemical compositions are determined by thermochemical equilibrium analyses in terms of free enthalpy minimization. In order to investigate the impact of pool oxidation on the fission product release behaviour RELOS is calculating, mechanistic models were implemented calculating the amount of oxygen being available at the pool surface and its transfer into the liquid corium phase. For this a diffusion and a convection model were provided. The diffusion model is based on Fick's second law and was realised using Crank-Nicolson, an implicit differential method. The convection model considers a mass transfer coefficient and the difference of the concentrations at the phase boundary and the bulk of the liquid corium. Moreover, the convection model includes both the temperature induced convection resulting from a cooler pool surface and a heated bottom - as it

Vaporization of chemical species and the production of aerosols during a core debris/concrete interaction

International Nuclear Information System (INIS)

Butland, A.T.D.; Mignanelli, M.A.; Potter, P.E.; Smith, P.N.

1987-01-01

The equilibrium chemical composition within gas bubbles sparging through isothermal molten corium-concrete mixtures has been evaluated theoretically. A series of sensitivity calculations gives some insight into a number of factors which are of importance in determining the radionuclide and non-radioactive releases during core-concrete interaction. The degree of mixing or layering of the pool has turned out to be of paramount importance in determining the magnitudes of the releases. The presence of unoxidized zirconium in the melt tends to enhance the release of a number of species and the type of concrete used for the base mat can have a significant effect. The predictions can be sensitive to the thermodynamic data used in the calculations. The vaporization of various species into the gas bubbles can require large amounts of heat; the loss of this heat from the melt can have an effect on the extent of the vaporization

Simulation on the Effects of Surfactants and Observed Thermocapillary Motion for Laser Melting Physics

Science.gov (United States)

Nourgaliev, Robert; Barney, Rebecca; Weston, Brian; Delplanque, Jean-Pierre; McCallen, Rose

2017-11-01

A newly developed, robust, high-order in space and time, Newton-Krylov based reconstructed discontinuous Galerkin (rDG) method is used to model and analyze thermocapillary convection in melt pools. The application of interest is selective laser melting (SLM) which is an Additive Manufacturing (AM, 3D metal laser printing) process. These surface tension driven flows are influenced by temperature gradients and surfactants (impurities), and are known as the Marangoni flow. They have been experimentally observed in melt pools for welding applications, and are thought to influence the microstructure of the re-solidified material. We study the effects of the laser source configuration (power, beam size and scanning speed), as well as surfactant concentrations. Results indicate that the surfactant concentration influences the critical temperature, which governs the direction of the surface thermocapillary traction. When the surface tension traction changes sign, very complex flow patterns emerge, inducing hydrodynamic instability under certain conditions. These in turn would affect the melt pool size (depth) and shape, influencing the resulting microstructure, properties, and performance of a finished product part produced using 3D metal laser printing technologies. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344. Information management release number LLNL-ABS-735908.

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)

Melt water interaction tests. PREMIX tests PM10 and PM11

Energy Technology Data Exchange (ETDEWEB)

Kaiser, A.; Schuetz, W.; Will, H. [Forschungszentrum Karlsruhe Inst. fuer Reaktorsicherheit, Karlsruhe (Germany)

1998-01-01

A series of experiments is being performed in the PREMIX test facility in which the mixing behaviour is investigated of a hot alumina melt discharged into water. The major parameters have been: the melt mass, the number of nozzles, the distance between the nozzle and the water, and the depth of the water. The paper describes the last two tests in which 20 kg of melt were released through one and three nozzles, respectively, directly into the water whose depth was 500 mm. The melt penetration and the associated phenomena of mixing are described by means of high-speed films and various measurements. The steam production and, subsequently, the pressure increased markedly only after the melt had reached the bottom of the pool. Spreading of the melt across the bottom caused violent boiling in both tests. Whereas the boiling lasted for minutes in the single-jet test, a steam explosion occurred in the triple-jet test about one second after the start of melt penetration. (author)

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

Role of crucible partition in improving Czochralski melt conditions

Science.gov (United States)

Jafri, I. H.; Prasad, V.; Anselmo, A. P.; Gupta, K. P.

1995-09-01

Many of the inhomogeneities and defects in the crystal grown from a pool of melt are because of the inherent unsteady growth kinetics and flow instabilities of the process. A scaled up version of the Czochralski process induces oscillatory and turbulent conditions in the melt, thereby resulting in the production of non-uniform silicon crystals. This numerical study reveals that a crucible partition shorter than the melt height can significantly improve the melt conditions. The obstruction at the bottom of the crucible is helpful but the variations in heat flux and flow patterns remain random. However, when the obstruction is introduced at the top of the melt, the flow conditions become much more desirable and oscillations are greatly suppressed. It is also found that a full-melt height partition or a double-crucible may not be a good choice. An optimal size of the blockage and its location to produce the most desirable process conditions will depend on the growth parameters including the melt height and the crucible diameter. These findings should be particularly useful in designing a solid polysilicon pellets-feed continuous Czochralski process for Si crystals.

Laser melt injection of hard ceramic particles into Al and Ti alloys - processing, microstructure and mechanical behavior

NARCIS (Netherlands)

Ocelik, V; Nijman, S.; van Ingen, R; Oliveira, U; De Hosson, J Th M

2003-01-01

The conditions for a successful Laser Melt Injection (LMI) of SiC and WC particles into the melt pool of Al8Si and Ti6Al4V alloys were studied exptl. and theor. by FEM calcns. The laser employed is a high power Nd:YAG. The formation of a relatively thick aluminum oxide layer on the Al melt surface

Protection system for minimizing the consequences of a flow blockage incident at a pool-type research reactor

International Nuclear Information System (INIS)

de Vries, J.W.; van Dam, H.; Gysler, G.

1990-01-01

Safety analysis activities were performed for the HOR, a pool-type research reactor with plate-type fuel elements and a maximum licensed power of 3 MW. Following internationally accepted guidelines, a wide variety of possible process disturbances has been considered. For the HOR the most aggravating accident conditions could result from a sudden flow blockage of cooling channels. If this event occurs in the high power density region of the core, a decrease of the hot channel flow either causes flow reversal or prompts burnout. Unless the reactor is scrammed in time, the fuel plates will heat up rapidly and local melting will occur with possible propagation of voiding and burnout to adjacent channels. In the analysis, melting of the cladding has been considered by using a simplified model approach. The number of voided coolant channels, as well as the propagation rate of fuel plates reaching locally the melting temperature, were calculated for different conditions of operation. In order to reduce the risk of a fuel melt accident occurring at the HOR, the protection system features a special design option. The system recognizes cooling channel voiding by detection of a sudden decrease of neutron flux. In the present work, it has been shown that a flow blockage incident can be detected in the early stages of development. Also, in accordance with the results of experimental tests, it can be concluded that in many cases melting of fuel plates will be effectively prevented. If such an accident occurs on a very fast time scale, at least the radiological consequences are significantly mitigated by preventing propagation, thus limiting the number of molten fuel plates

Microstructure and mechanical behavior of pulsed laser surface melted AISI D2 cold work tool steel

Science.gov (United States)

Yasavol, N.; Abdollah-zadeh, A.; Ganjali, M.; Alidokht, S. A.

2013-01-01

D2 cold work tool steel (CWTS) was subjected to pulse laser surface melting (PLSM) at constant frequency of 20 Hz Nd: YAG laser with different energies, scanning rate and pulse durations radiated to the surface. Characterizing the PLSM, with optical and field emission scanning electron microscopy, electron backscattered diffraction and surface hardness mapping technique was used to evaluate the microhardness and mechanical behavior of different regions of melting pool. Increasing laser energy and reducing the laser scanning rate results in deeper melt pool formation. Moreover, PLSM has led to entirely dissolution of the carbides and re-solidification of cellular/dendritic structure of a fine scale surrounded by a continuous interdendritic network. This caused an increase in surface microhardness, 2-4 times over that of the base metal.

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.

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

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.)

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

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

Phenomenological studies on melt-structure-water interactions (MSWI) during severe accidents

International Nuclear Information System (INIS)

Sehgal, B.R.; Yang, Z.L.; Haraldsson, H.O.; Nourgaliev, R.R.; Konovalikhin, M.; Paladino, D.; Gubaidullin, A.A.; Kolb, G.; Theerthan, A.

2000-05-01

. Experimental results show that a downcomer may increase the dryout heat flux by 50% to 350%. The experimental observations show that stratification and mixing in the pool have profound effect on the heat flux distribution on the pressure vessel lower head. The FOREVER test provides at 1/10th scale new and unique data for high-temperature multi-axial creep deformation of the prototypical reactor pressure vessel, and for natural convection heat transfer in a hemispherical melt pool

Phenomenological studies on melt-structure-water interactions (MSWI) during severe accidents

Energy Technology Data Exchange (ETDEWEB)

Sehgal, B.R.; Yang, Z.L.; Haraldsson, H.O.; Nourgaliev, R.R.; Konovalikhin, M.; Paladino, D.; Gubaidullin, A.A.; Kolb, G.; Theerthan, A. [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety

2000-05-01

. Experimental results show that a downcomer may increase the dryout heat flux by 50% to 350%. The experimental observations show that stratification and mixing in the pool have profound effect on the heat flux distribution on the pressure vessel lower head. The FOREVER test provides at 1/10th scale new and unique data for high-temperature multi-axial creep deformation of the prototypical reactor pressure vessel, and for natural convection heat transfer in a hemispherical melt pool.

Test Area North Pool Stabilization Project: Environmental assessment

International Nuclear Information System (INIS)

1996-05-01

The Test Area North (TAN) Pool is located within the fenced TAN facility boundaries on the Idaho National Engineering Laboratory (INEL). The TAN pool stores 344 canisters of core debris from the March, 1979, Three Mile Island (TMI) Unit 2 reactor accident; fuel assemblies from Loss-of-Fluid Tests (LOFT); and Government-owned commercial fuel rods and assemblies. The LOFT and government owned commercial fuel rods and assemblies are hereafter referred to collectively as open-quotes commercial fuelsclose quotes except where distinction between the two is important to the analysis. DOE proposes to remove the canisters of TMI core debris and commercial fuels from the TAN Pool and transfer them to the Idaho Chemical Processing Plant (ICPP) for interim dry storage until an alternate storage location other than at the INEL, or a permanent federal spent nuclear fuel (SNF) repository is available. The TAN Pool would be drained and placed in an industrially and radiologically safe condition for refurbishment or eventual decommissioning. This environmental assessment (EA) identifies and evaluates environmental impacts associated with (1) constructing an Interim Storage System (ISS) at ICPP; (2) removing the TMI and commercial fuels from the pool and transporting them to ICPP for placement in an ISS, and (3) draining and stabilizing the TAN Pool. Miscellaneous hardware would be removed and decontaminated or disposed of in the INEL Radioactive Waste Management Complex (RWMC). This EA also describes the environmental consequences of the no action alternative

High Power Laser Beam Welding of Thick-walled Ferromagnetic Steels with Electromagnetic Weld Pool Support

Science.gov (United States)

Fritzsche, André; Avilov, Vjaceslav; Gumenyuk, Andrey; Hilgenberg, Kai; Rethmeier, Michael

The development of modern high power laser systems allows single pass welding of thick-walled components with minimal distortion. Besides the high demands on the joint preparation, the hydrostatic pressure in the melt pool increases with higher plate thicknesses. Reaching or exceeding the Laplace pressure, drop-out or melt sagging are caused. A contactless electromagnetic weld support system was used for laser beam welding of thick ferromagnetic steel plates compensating these effects. An oscillating magnetic field induces eddy currents in the weld pool which generate Lorentz forces counteracting the gravity forces. Hysteresis effects of ferromagnetic steels are considered as well as the loss of magnetization in zones exceeding the Curie temperature. These phenomena reduce the effective Lorentz forces within the weld pool. The successful compensation of the hydrostatic pressure was demonstrated on up to 20 mm thick plates of duplex and mild steel by a variation of the electromagnetic power level and the oscillation frequency.

Steam blowdown experiments with the condensation pool test rig

International Nuclear Information System (INIS)

Purhonen, H.; Puustinen, M.; Laine, J.; Raesaenen, A.; Kyrki-Rajamaeki, R.; Vihavainen, J.

2005-01-01

During a possible loss-of-coolant accident (Local) a large amount of non-condensable (nitrogen) and condensable (steam) gas is blown from the upper drywell of the containment to the condensation pool through the blowdown pipes at the boiling water reactors (BWRs). The wet well pool serves as the major heat sink for condensation of steam. The blowdown causes both dynamic and structural loads to the condensation pool. There might also be a risk that the gas discharging to the pool could push its way to the emergency core cooling systems (ECCS) and undermine their performance. (author)

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

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

Developmental assessment of the SCDAP/RELAP5 code

International Nuclear Information System (INIS)

Harvego, E.A.; Slefken, L.J.; Coryell, E.W.

1997-01-01

The development and assessment of the late-phase damage progression models in the current version (designated MOD3.2) of the SCDAP/RELAP5 code are described. The SCDAP/RELAP5 code is being developed at the Idaho National Engineering and Environmental Laboratory under the primary sponsorship of the US Nuclear Regulatory Commission (NRC) to provide best-estimate transient simulations of light water reactor coolant systems (RCS) during severe accident conditions. Recent modeling improvements made to the MOD3.2 version of the code include (1) molten pool formation and heat up, including the transient start-up of natural circulation heat transfer, (2) in-core molten pool thermal-mechanical crust failure, (3) the melting and relocation of upper plenum structures, and (4) improvements in the modeling of lower plenum debris behavior and the potential for failure of the lower head. Finally, to eliminate abrupt transitions between core damage states and provide more realistic predictions of late phase accident progression phenomena, a transition smoothing methodology was developed and implemented that results in the calculation of a gradual transition from an intact core geometry through the different core damage states leading to molten pool formation. A wide range of experiments and modeling tools were used to assess the capabilities of MOD3.2. The results of the SCDAP/RELAP5/MOD3.2 assessment indicate that modeling improvements have significantly enhanced the code capabilities and performance in several areas compared to the earlier code version. New models for transition smoothing between core damage states, and modeling improvements/additions for cladding oxide failure, molten pool behavior, and molten pool crust failure have significantly improved the code usability for a wide range of applications and have significantly improved the prediction of hydrogen production, molten pool melt mass and core melt relocation time

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

Microstructure and mechanical behavior of pulsed laser surface melted AISI D2 cold work tool steel

International Nuclear Information System (INIS)

Yasavol, N.; Abdollah-zadeh, A.; Ganjali, M.; Alidokht, S.A.

2013-01-01

Highlights: ► Melted zone contained fine dendrites in the bottom and equiaxed grains on the top. ► Microstructural refinements of PLSM led to microhardness enhancement. ► Higher scanning rate and lower laser energy were more effective to refine the microstructure. - Abstract: D2 cold work tool steel (CWTS) was subjected to pulse laser surface melting (PLSM) at constant frequency of 20 Hz Nd: YAG laser with different energies, scanning rate and pulse durations radiated to the surface. Characterizing the PLSM, with optical and field emission scanning electron microscopy, electron backscattered diffraction and surface hardness mapping technique was used to evaluate the microhardness and mechanical behavior of different regions of melting pool. Increasing laser energy and reducing the laser scanning rate results in deeper melt pool formation. Moreover, PLSM has led to entirely dissolution of the carbides and re-solidification of cellular/dendritic structure of a fine scale surrounded by a continuous interdendritic network. This caused an increase in surface microhardness, 2–4 times over that of the base metal.

Microstructure and mechanical behavior of pulsed laser surface melted AISI D2 cold work tool steel

Energy Technology Data Exchange (ETDEWEB)

Yasavol, N. [Department of Materials Engineering, Tarbiat Modares University, P.O. Box 14115-143, Tehran (Iran, Islamic Republic of); Abdollah-zadeh, A., E-mail: zadeh@modares.ac.ir [Department of Materials Engineering, Tarbiat Modares University, P.O. Box 14115-143, Tehran (Iran, Islamic Republic of); Ganjali, M. [Materials and Energy Research Center, P.O. Box 14155-4777, Karaj (Iran, Islamic Republic of); Alidokht, S.A. [Department of Materials Engineering, Tarbiat Modares University, P.O. Box 14115-143, Tehran (Iran, Islamic Republic of)

2013-01-15

Highlights: Black-Right-Pointing-Pointer Melted zone contained fine dendrites in the bottom and equiaxed grains on the top. Black-Right-Pointing-Pointer Microstructural refinements of PLSM led to microhardness enhancement. Black-Right-Pointing-Pointer Higher scanning rate and lower laser energy were more effective to refine the microstructure. - Abstract: D2 cold work tool steel (CWTS) was subjected to pulse laser surface melting (PLSM) at constant frequency of 20 Hz Nd: YAG laser with different energies, scanning rate and pulse durations radiated to the surface. Characterizing the PLSM, with optical and field emission scanning electron microscopy, electron backscattered diffraction and surface hardness mapping technique was used to evaluate the microhardness and mechanical behavior of different regions of melting pool. Increasing laser energy and reducing the laser scanning rate results in deeper melt pool formation. Moreover, PLSM has led to entirely dissolution of the carbides and re-solidification of cellular/dendritic structure of a fine scale surrounded by a continuous interdendritic network. This caused an increase in surface microhardness, 2-4 times over that of the base metal.

Molten pool characterization of laser lap welded copper and aluminum

Science.gov (United States)

Xue, Zhiqing; Hu, Shengsun; Zuo, Di; Cai, Wayne; Lee, Dongkyun; Elijah, Kannatey-Asibu, Jr.

2013-12-01

A 3D finite volume simulation model for laser welding of a Cu-Al lap joint was developed using ANSYS FLUENT to predict the weld pool temperature distribution, velocity field, geometry, alloying element distribution and transition layer thickness—all key attributes and performance characteristics for a laser-welded joint. Melting and solidification of the weld pool was simulated with an enthalpy-porosity formulation. Laser welding experiments and metallographic examination by SEM and EDX were performed to investigate the weld pool features and validate the simulated results. A bowl-shaped temperature field and molten pool, and a unique maximum fusion zone width were observed near the Cu-Al interface. Both the numerical simulation and experimental results indicate an arch-shaped intermediate layer of Cu and Al, and a gradual transition of Cu concentration from the aluminum plate to the copper plate with high composition gradient. For the conditions used, welding with Cu on top was found to result in a better weld joint.

Molten pool characterization of laser lap welded copper and aluminum

International Nuclear Information System (INIS)

Xue, Zhiqing; Hu, Shengsun; Zuo, Di; Cai, Wayne; Lee, Dongkyun; Elijah, Kannatey-Asibu Jr

2013-01-01

A 3D finite volume simulation model for laser welding of a Cu–Al lap joint was developed using ANSYS FLUENT to predict the weld pool temperature distribution, velocity field, geometry, alloying element distribution and transition layer thickness—all key attributes and performance characteristics for a laser-welded joint. Melting and solidification of the weld pool was simulated with an enthalpy-porosity formulation. Laser welding experiments and metallographic examination by SEM and EDX were performed to investigate the weld pool features and validate the simulated results. A bowl-shaped temperature field and molten pool, and a unique maximum fusion zone width were observed near the Cu–Al interface. Both the numerical simulation and experimental results indicate an arch-shaped intermediate layer of Cu and Al, and a gradual transition of Cu concentration from the aluminum plate to the copper plate with high composition gradient. For the conditions used, welding with Cu on top was found to result in a better weld joint. (paper)

Quench cooling of superheated debris beds in containment during LWR core meltdown accidents

International Nuclear Information System (INIS)

Ginsberg, T.; Chen, J.C.

1984-01-01

Light water reactor core meltdown accident sequence studies suggest that superheated debris beds may settle on the concrete floor beneath the reactor vessel. A model for the heat transfer processes during quench (removal of stored energy from initial temperature to saturation temperature) of superheated debris beds cooled by an overlying pool of water has been presented in a prior paper. This paper discusses the coolability of decay-heated debris beds from the standpoint of their transient quench characteristics. It is shown that even though a debris bed configuration may be coolable from the point of view of steady-state decay heat removal, the quench behavior from an initially elevated temperature may lead to bed melting prior to quench of the debris

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)

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

Laser Beam Melting of Alumina: Effect of Absorber Additions

Science.gov (United States)

Moniz, Liliana; Colin, Christophe; Bartout, Jean-Dominique; Terki, Karim; Berger, Marie-Hélène

2018-03-01

Ceramic laser beam melting offers new manufacturing possibilities for complex refractory structures. Poor absorptivity in near infra-red wavelengths of oxide ceramics is overcome with absorber addition to ceramic powders. Absorbers affect powder bed densities and geometrical stability of melted tracks. Optimum absorber content is defined for Al2O3 by minimizing powder bed porosity, maximizing melting pool geometrical stability and limiting shrinkage. Widest stability fields are obtained with addition of 0.1 wt.% C and 0.5 wt.% β-SiC. Absorption coefficient values of Beer-Lambert law follow stability trends: they increase with C additions, whereas with β-SiC, a maximum is reached for 0.5 wt.%. Powder particle ejections are also identified. Compared to metallic materials, this ejection phenomenon can no longer be neglected when establishing a three-dimensional manufacturing strategy.

Odonata (Insecta at a wadi Pool near Nizwa, northern Oman

Directory of Open Access Journals (Sweden)

Elaine M. Cowan

2015-07-01

Full Text Available Fourteen damselfly and dragonfly species were recorded in 68 visits to a wadi pool in northern Oman, March 2012 to June 2014.  All identifications were based on photographs.  Apparently the pool has a core community of eight resident species.  Paragomphus sinaiticus, globally Near Threatened, was regularly  recorded. 

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

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

Convective heat transfer the molten metal pool heated from below and cooled by two-phase flow

International Nuclear Information System (INIS)

Cho, J. S.; Suh, K. Y.; Chung, C. H.; Park, R. J.; Kim, S. B.

1998-01-01

During a hypothetical servere accident in the nuclear power plant, a molten core material may form stratified fluid layers. These layers may be composed of high temperature molten debris pool and water coolant in the lower plenum of the reactor vessel or in the reactor cavity. This study is concerned with the experimental test and numerical analysis on the heat transfer and solidification of the molten metal pool with overlying coolant with boiling. This work examines the crust formation and the heat transfer characteristics of the molten metal pool immersed in the boiling coolant. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. The simulant molten pool material is tin (Sn) with the melting temperature of 232 .deg. C. Demineralized water is used as the working coolant. Tests were performed under the condition of the bottom surface heating in the test section and the forced convection of the coolant being injected onto the molten metal pool. The constant temperature and constant heat flux conditions are adopted for the bottom heating. The test parameters included the heated bottom surface temperature of the molten metal pool, the input power to the heated bottom surface of the test section, and the coolant injection rate. Numerical analyses were simultaneously performed in a two-dimensional rectangular domain of the molten metal pool to check on the measured data. The numerical program has been developed using the enthalpy method, the finite volume method and the SIMPLER algorithm. The experimental results of the heat transfer show general agreement with the calculated values. In this study, the relationship between the Nusselt number and Rayleigh number in the molten metal pool region was estimated and compared with the dry experiment without coolant nor solidification of the molten metal pool, and with the crust formation experiment with subcooled coolant, and against other correlations. In the experiments, the

Estimation of reactor pool water temperature after shutdown in JRR-3M

International Nuclear Information System (INIS)

Yagi, Masahiro; Sato, Mitsugu; Kakefuda, Kazuhiro

1999-01-01

The reactor pool water temperature increasing by the decay heat was estimated by calculation. The reactor pool water temperature was calculated by increased enthalpy that was estimated by the reactor decay heat, the heat released from the reactor biological shielding concrete, reactor pool water surface, the heat conduction from the canal and the core inlet piping. These results of calculation were compared with the past measured data. As the results of estimation, after the JRR-3M shutdown, the calculated reactor pool temperature first increased sharply. This is because the decay heat was the major contribution. And then, rate of increased reactor pool temperature decreased. This is because the ratio of heat released from reactor biological shielding concrete and core inlet piping to the decay heat increased. Besides, the calculated reactor pool water temperature agreed with the past measured data in consequence of correcting the decay heat and the released heat. The corrected coefficient k 1 of decay heat was 0.74 - 0.80. And the corrected coefficient k 2 of heat released from the reactor biological shielding concrete was 3.5 - 4.5. (author)

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

Laser melt injection of hard ceramic particles into Al and Ti alloys - processing, microstructure and mechanical behaviour

NARCIS (Netherlands)

Ocelik, V.; Nijman, S.; van Ingen, R.; Oliveira, U.; de Hosson, J.T.M.; Brebbia, CA; DeHosson, JTM; Nishida, SI

2003-01-01

The conditions for a successful Laser Melt Injection (LMI) of SiC and WC particles into the melt pool of Al8Si and Ti6A14V alloys were studied experimentally and theoretically by FEM calculations. The laser employed is a high power Nd:YAG. The formation of a relatively thick aluminium oxide layer on

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.

Simulation of a pool type research reactor

International Nuclear Information System (INIS)

Oliveira, Andre Felipe da Silva de; Moreira, Maria de Lourdes

2011-01-01

Computational fluid dynamic is used to simulate natural circulation condition after a research reactor shutdown. A benchmark problem was used to test the viability of usage such code to simulate the reactor model. A model which contains the core, the pool, the reflector tank, the circulation pipes and chimney was simulated. The reactor core contained in the full scale model was represented by a porous media. The parameters of porous media were obtained from a separate CFD analysis of the full core model. Results demonstrate that such studies can be carried out for research and test of reactors design. (author)

Horizontal above-rack pool storage

International Nuclear Information System (INIS)

Moscardini, R.L.

1993-08-01

This report describes a unique method for storing spent, six year out of core, fuel at a prototypical PWR nuclear power station. The study describes a conceptual design, with favorable structural, thermal and criticality technical evaluations. However, economic considerations and licensing risks are judged to be less favorable. The concept study prescribes a fuel over fuel arrangement in an existing Spent Fuel Pool (SFP) with full maintenance of ALARA principles. This concept study is specific to a prototypical pool design, but may easily be projected to other nuclear facilities with other SFP conditions. For the prototypical PWR, the conceptual fuel bridge design will store over 200 additional fuel assemblies without significant facility modifications and for an indefinite time period

The WECHSL-Mod3 code: A computer program for the interaction of a core melt with concrete including the long term behavior. Model description and user's manual

International Nuclear Information System (INIS)

Foit, J.J.; Adroguer, B.; Cenerino, G.; Stiefel, S.

1995-02-01

The WECHSL-Mod3 code is a mechanistic computer code developed for the analysis of the thermal and chemical interaction of initially molten reactor materials with concrete in a two-dimensional as well as in a one-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. It is assumed that an underlying metallic layer exists covered by an oxidic layer or that only one oxidic layer is present which can contain a homogeneously dispersed metallic phase. 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 CALTHER code (developed at CEA, France) which models the radiative heat transfer from the upper surface of the corium melt to the surrounding cavity is implemented in the present WECHSL version. The WECHSL code in its present version was validated by the BETA, ACE and SURC experiments. The test samples include a BETA and the SURC2 post test calculations and a WECHSL application to a reactor accident. (orig.) [de

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.

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

Convection-driven melting in an n-octane pool fire bounded by an ice wall

Science.gov (United States)

Farmahini Farahani, Hamed; Alva, Ulises; Rangwala, Ali; Jomaas, Grunde

2017-11-01

Burning of the liquid fuels adjacent to ice bodies creates a lateral cavity due to melting of the ice. The formation of lateral cavities are noticed recently and only a few experimental studies have addressed them. One study has shown lateral cavity formation with length of 12 cm for 5 minutes burning of oil. Based on the hypothesis that melting is facilitated by the convection in the liquid fuel, a series of PIV tests were conducted on burning of n-octane in a square glass tray with a 3 cm thick ice wall placed on one side of the tray. Marangoni generates a flow below the surface of the fuel and near the ice from hot to cold regions. The flow measurements by a 2D PIV system indicated the existence of different flow regimes. Before ignition, combined surface tension and buoyancy effects led to a one roll structure. After ignition the flow field began transitioning toward an unstable regime with an increase in velocity magnitude. Unfortunately, the PIV quality declined in the unstable regime, but indications of a multi-roll structure separating from a primary horizontal flow on the top driven by Marangoni convection were observed. The knowledge gained from these experiments will help determine the influential parameters in ice melting during burning of oil in ice-infested waters.

Assessment of structural materials inside the reactor pool of the Dalat research reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Luong Ba Vien; Nguyen Minh Tuan; Trang Cao Su

2010-01-01

Originally the Dalat Nuclear Research Reactor (DNRR) was a 250-kW TRIGA MARK II reactor, started building from early 1960s and achieved the first criticality on February 26, 1963. During the 1982-1984 period, the reactor was reconstructed and upgraded to 500kW, and restarted operation on March 20, 1984. From the original TRIGA reactor, only the pool liner, beam ports, thermal columns, and graphite reflector have been remained. The structural materials of pool liner and other components of TRIGA were made of aluminum alloy 6061 and aluminum cladding fuel assemblies. Some other parts, such as reactor core, irradiation rotary rack around the core, vertical irradiation facilities, etc. were replaced by the former Soviet Union's design with structural materials of aluminum alloy CAV-1. WWR-M2 fuel assemblies of U-Al alloy 36% and 19.75% 235 U enrichment and aluminum cladding have been used. In its original version, the reactor was setting upon an all-welded aluminum frame supported by four legs attached to the bottom of the pool. After the modification made, the new core is now suspended from the top of the pool liner by means of three aluminum concentric cylindrical shells. The upper one has a diameter of 1.9m, a length of 3.5m and a thickness of 10mm. This shell prevents from any visual access to the upper part of the pool liner, but is provided with some holes to facilitate water circulation in the 4cm gap between itself and the reactor pool liner. The lower cylindrical shells act as an extracting well for water circulation. As reactor has been operated at low power of 500 kW, it was no any problem with degradation of core structural materials due to neutron irradiation and thermal heat, but there are some ageing issues with aluminum liner and other structures (for example, corrosion of tightening-up steel bolt in the fourth beam port and flood of neutron detector housing) inside the reactor pool. In this report, the authors give an overview and assessment of

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.

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

Pool scrubbing models for iodine components

Energy Technology Data Exchange (ETDEWEB)

Fischer, K [Battelle Ingenieurtechnik GmbH, Eschborn (Germany)

1996-12-01

Pool scrubbing is an important mechanism to retain radioactive fission products from being carried into the containment atmosphere or into the secondary piping system. A number of models and computer codes has been developed to predict the retention of aerosols and fission product vapours that are released from the core and injected into water pools of BWR and PWR type reactors during severe accidents. Important codes in this field are BUSCA, SPARC and SUPRA. The present paper summarizes the models for scrubbing of gaseous Iodine components in these codes, discusses the experimental validation, and gives an assessment of the state of knowledge reached and the open questions which persist. The retention of gaseous Iodine components is modelled by the various codes in a very heterogeneous manner. Differences show up in the chemical species considered, the treatment of mass transfer boundary layers on the gaseous and liquid sides, the gas-liquid interface geometry, calculation of equilibrium concentrations and numerical procedures. Especially important is the determination of the pool water pH value. This value is affected by basic aerosols deposited in the water, e.g. Cesium and Rubidium compounds. A consistent model requires a mass balance of these compounds in the pool, thus effectively coupling the pool scrubbing phenomena of aerosols and gaseous Iodine species. Since the water pool conditions are also affected by drainage flow of condensate water from different regions in the containment, and desorption of dissolved gases on the pool surface is determined by the gas concentrations above the pool, some basic limitations of specialized pool scrubbing codes are given. The paper draws conclusions about the necessity of coupling between containment thermal-hydraulics and pool scrubbing models, and proposes ways of further simulation model development in order to improve source term predictions. (author) 2 tabs., refs.

Pool scrubbing models for iodine components

International Nuclear Information System (INIS)

Fischer, K.

1996-01-01

Pool scrubbing is an important mechanism to retain radioactive fission products from being carried into the containment atmosphere or into the secondary piping system. A number of models and computer codes has been developed to predict the retention of aerosols and fission product vapours that are released from the core and injected into water pools of BWR and PWR type reactors during severe accidents. Important codes in this field are BUSCA, SPARC and SUPRA. The present paper summarizes the models for scrubbing of gaseous Iodine components in these codes, discusses the experimental validation, and gives an assessment of the state of knowledge reached and the open questions which persist. The retention of gaseous Iodine components is modelled by the various codes in a very heterogeneous manner. Differences show up in the chemical species considered, the treatment of mass transfer boundary layers on the gaseous and liquid sides, the gas-liquid interface geometry, calculation of equilibrium concentrations and numerical procedures. Especially important is the determination of the pool water pH value. This value is affected by basic aerosols deposited in the water, e.g. Cesium and Rubidium compounds. A consistent model requires a mass balance of these compounds in the pool, thus effectively coupling the pool scrubbing phenomena of aerosols and gaseous Iodine species. Since the water pool conditions are also affected by drainage flow of condensate water from different regions in the containment, and desorption of dissolved gases on the pool surface is determined by the gas concentrations above the pool, some basic limitations of specialized pool scrubbing codes are given. The paper draws conclusions about the necessity of coupling between containment thermal-hydraulics and pool scrubbing models, and proposes ways of further simulation model development in order to improve source term predictions. (author) 2 tabs., refs

Single track and single layer formation in selective laser melting of niobium solid solution alloy

Directory of Open Access Journals (Sweden)

Yueling GUO

2018-04-01

Full Text Available Selective laser melting (SLM was employed to fabricate Nb-37Ti-13Cr-2Al-1Si (at% alloy, using pre-alloyed powders prepared by plasma rotating electrode processing (PREP. A series of single tracks and single layers under different processing parameters was manufactured to evaluate the processing feasibility by SLM, including laser power, scanning speed, and hatch distance. Results showed that continuous single tracks could be fabricated using proper laser powers and scanning velocities. Both the width of a single track and its penetration depth into a substrate increased with an increase of the linear laser beam energy density (LED, i.e., an increase of the laser power and a decrease of the scanning speed. Nb, Ti, Si, Cr, and Al elements distributed heterogeneously over the melt pool in the form of swirl-like patterns. An excess of the hatch distance was not able to interconnect neighboring tracks. Under improper processing parameters, a balling phenomenon occurred, but could be eliminated with an increased LED. This work testified the SLM-processing feasibility of Nb-based alloy and promoted the application of SLM to the manufacture of niobium-based alloys. Keywords: Additive manufacturing, Melt pool, Niobium alloy, Powder metallurgy, Selective laser melting

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

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

Grain structure evolution in Inconel 718 during selective electron beam melting

Energy Technology Data Exchange (ETDEWEB)

Helmer, H.; Bauereiß, A., E-mail: Andreas.Bauereiss@fau.de; Singer, R.F.; Körner, C.

2016-06-21

Selective electron beam melting (SEBM) is an additive manufacturing method where complex parts are built from metal powders in layers of typically 50 µm. An electron beam is used for heating (about 900 °C building temperature) and selective melting of the material. The grain structure evolution is a result of the complex thermal and hydrodynamic conditions in the melt pool. We show how different scanning strategies can be used to produce either a columnar grain structure with a high texture in building direction or an equiaxed fine grained structure. Numerical simulations of the selective melting process are applied to study the fundamental mechanisms responsible for differing grain structures. It is shown, that the direction of the thermal gradient during solidification can be altered by scanning strategies to acquire either epitaxial growth or stray grains. We show that it is possible to locally alter the grain structure of a part, thus allowing tailoring of the mechanical properties.

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

Seismic analysis and design of spent subassembly storage bay (SSSB) pool

International Nuclear Information System (INIS)

Abdul Gani, H.I.; Ramanjaneyulu, K.V.S.; Pillai, C.S.; Chetal, S.C.

2003-01-01

Fuel bundles, after their specified stay in reactor core, are replaced by fresh fuel for sustaining power generation at rated levels. The irradiated fuel subassembly, removed fresh from core, known as spent fuel sub assembly, is radioactive and decay heat generating. It needs to be cooled before it becomes amenable for handling, either for reprocessing or for immobilisation. For this purpose, it is immersed in a pool of water, retained in a concrete structure referred as Spent Subassembly Storage Bay (SSSB) pool. The height of water column above fuel bundles is arrived from shielding considerations. SSSB pool is one of the nuclear safety related structures and warrants rigorous analysis and design. The SSSB pool, in case of PFBR 500 MW(e) is located in fuel building. It is a stainless steel lined. water retaining rectangular R.C.C. open tank of size 7.5 X 29.0 m, with a height of 11.0 m. This structure is analysed for two levels of site specific earthquakes taking in to account liquid structure interactions as per ASCE-4, 1998. The design of walls and bottom slab is carried out satisfying the AERB code for nuclear safety related structures. Analysis and design of SSSB pool of PFBR is presented in the following paper. (author)

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

Atlantic Warm Pool Trigger for the Younger Dryas Climate Event

Science.gov (United States)

Abdul, N. A.; Mortlock, R. A.; Wright, J. D.; Fairbanks, R. G.; Teneva, L. T.

2011-12-01

There is growing evidence that variability in the size and heat content of the tropical Atlantic Warm Pool impacts circum-North Atlantic climate via the Atlantic Multi-decadal Oscillation mode (Wang et al., 2008). The Atlantic Warm Pool spans the Gulf of Mexico, Caribbean Sea and the western tropical North Atlantic. Barbados is located near the center of the tropical Atlantic Warm Pool and coupled ocean models suggest that Barbados remains near the center of the tropical Atlantic Warm Pool under varying wind stress simulations. Measurements of the oxygen isotope paleothermometer in Acropora palmata coral species recovered from cores offshore Barbados, show a 3oC monotonic decrease in sea surface temperature from 13106 ± 83 to 12744 ± 61 years before present (errors given as 2 sigma). This interval corresponds to a sea level rise from 71.4 meters to 67.1 meters below present levels at Barbados. The 3oC temperature decrease is captured in eight A. palmata specimens that are in stratigraphic sequence, 230Th/234U dated, and analyzed for oxygen isotopes. All measurements are replicated. We are confident that this is the warm pool equivalent of the Younger Dryas climate event. The initiation of this temperature drop in the Atlantic Warm Pool predates the Younger Dryas start in Greenland ice cores, reported to start at 12896 ± 138 years (relative to AD 2000) (Rasmussen et al., 2006), while few other Younger Dryas climate records are dated with similar accuracy to make the comparison. Rasmussen, S.O., Andersen, K.K., Svensson, A.M., Steffensen, J.P., Vinther, B.M., Clausen, H.B., Siggaard-Andersen, M.L., Johnsen, S.J., Larsen, L.B., Dahl-Jensen, D., Bigler, M., Röthlisberger, R., Fischer, H., Goto-Azuma, K., Hansson, M.E., and Ruth, U., 2006, A new Greenland ice core chronology for the last glacial termination: J. Geophys. Res., v. 111, p. D06102. Wang, C., Lee, S.-K., and Enfield, D.B., 2008, Atlantic Warm Pool acting as a link between Atlantic Multidecadal

Influence of the arc plasma parameters on the weld pool profile in TIG welding

Science.gov (United States)

Toropchin, A.; Frolov, V.; Pipa, A. V.; Kozakov, R.; Uhrlandt, D.

2014-11-01

Magneto-hydrodynamic simulations of the arc and fluid simulations of the weld pool can be beneficial in the analysis and further development of arc welding processes and welding machines. However, the appropriate coupling of arc and weld pool simulations needs further improvement. The tungsten inert gas (TIG) welding process is investigated by simulations including the weld pool. Experiments with optical diagnostics are used for the validation. A coupled computational model of the arc and the weld pool is developed using the software ANSYS CFX. The weld pool model considers the forces acting on the motion of the melt inside and on the surface of the pool, such as Marangoni, drag, electromagnetic forces and buoyancy. The experimental work includes analysis of cross-sections of the workpieces, highspeed video images and spectroscopic measurements. Experiments and calculations have been performed for various currents, distances between electrode and workpiece and nozzle diameters. The studies show the significant impact of material properties like surface tension dependence on temperature as well as of the arc structure on the weld pool behaviour and finally the weld seam depth. The experimental weld pool profiles and plasma temperatures are in good agreement with computational results.

Influence of the arc plasma parameters on the weld pool profile in TIG welding

International Nuclear Information System (INIS)

Toropchin, A; Frolov, V; Pipa, A V; Kozakov, R; Uhrlandt, D

2014-01-01

Magneto-hydrodynamic simulations of the arc and fluid simulations of the weld pool can be beneficial in the analysis and further development of arc welding processes and welding machines. However, the appropriate coupling of arc and weld pool simulations needs further improvement. The tungsten inert gas (TIG) welding process is investigated by simulations including the weld pool. Experiments with optical diagnostics are used for the validation. A coupled computational model of the arc and the weld pool is developed using the software ANSYS CFX. The weld pool model considers the forces acting on the motion of the melt inside and on the surface of the pool, such as Marangoni, drag, electromagnetic forces and buoyancy. The experimental work includes analysis of cross-sections of the workpieces, highspeed video images and spectroscopic measurements. Experiments and calculations have been performed for various currents, distances between electrode and workpiece and nozzle diameters. The studies show the significant impact of material properties like surface tension dependence on temperature as well as of the arc structure on the weld pool behaviour and finally the weld seam depth. The experimental weld pool profiles and plasma temperatures are in good agreement with computational results

Establishment and validation of the model of molten pool in fast reactor

International Nuclear Information System (INIS)

Zhou Shufeng; Luo Rui; Wang Zhou; Shi Xiaobo; Yang Xianyong

2007-01-01

Running under the beyond design base accidental condition, sodium boiling and dry-out will soon be brought about in LMFBR. If not stopped timely, the fuel pins of the subassembly will be melt and broken to form a molten pool at the bottom of the subassembly. to present a reasonable analysis about the molten pool accident, a method of establishing model according to the mechanism is selected, by which an integral model of the molten pool is established. Validated on the three power groups of BF1 experiments which belong to the France SCARABEE series experimenters, the model shows good results. After compared with the models of GEYSER and BF2 experiments which had been validated before, some conclusions about mechanism of molten pool are derived. Moreover, through comparing the relative parameters such as the discharged heat and the increment of temperature etc., a reasonable analysis about the type of heat transfer is present, on the basis of which some conclusions are derived as well. (authors)

Swimming-pool piles; Piles piscines

Energy Technology Data Exchange (ETDEWEB)

Trioulaire, M [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1959-07-01

In France two swimming-pool piles, Melusine and Triton, have just been set in operation. The swimming-pool pile is the ideal research tool for neutron fluxes of the order of 10{sup 13}. This type of pile can be of immediate interest to many research centres, but its cost must be reduced and a break with tradition should be observed in its design. It would be an advantage: - to bury the swimming-pool; - to reject the experimental channel; - to concentrate the cooling circuit in the swimming-pool; - to carry out all manipulations in the water; - to double the core. (author) [French] En France, deux piles piscines, Melusine et Triton, viennent d'entrer en service. La pile piscine est l'outil de recherche ideal pour des flux de neutrons de l'ordre de 10{sup 13}. Ce type de pile peut interesser des maintenant de nombreux centres de recherches mais il faut reduire son prix de revient et rompre avec le conformisme de sa conception. Il y a avantage: - a enterrer la piscine; - a supprimer les canaux experimentaux; - a concentrer le circuit de refrigeration dans la piscine; - a effectuer toutes les manipulations dans l'eau; - a doubler le coeur. (auteur)

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 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

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.

The hydrometeor partitioning and microphysical processes over the Pacific Warm Pool in numerical modeling

Science.gov (United States)

Huang, Yi-Chih; Wang, Pao K.

2017-01-01

Numerical modeling is conducted to study the hydrometeor partitioning and microphysical source and sink processes during a quasi-steady state of thunderstorms over the Pacific Warm Pool by utilizing the microphysical model WISCDYMM to simulate selected storm cases. The results show that liquid-phase hydrometeors dominate thunderstorm evolution over the Pacific Warm Pool. The ratio of ice-phase mass to liquid-phase mass is about 41%: 59%, indicating that ice-phase water is not as significant over the Pacific Warm Pool as the liquid water compared to the larger than 50% in the subtropics and 80% in the US High Plains in a previous study. Sensitivity tests support the dominance of liquid-phase hydrometeors over the Pacific Warm Pool. The major rain sources are the key hail sinks: melting of hail and shedding from hail; whereas the crucial rain sinks are evaporation and accretion by hail. The major snow sources are Bergeron-Findeisen process, transfer of cloud ice to snow and accretion of cloud water; whereas the foremost sink of snow is accretion by hail. The essential hail sources are accretions of rain, cloud water, and snow; whereas the critical hail sinks are melting of hail and shedding from hail. The contribution and ranking of sources and sinks of these precipitates are compared with the previous study. Hydrometeors have their own special microphysical processes in the development and depletion over the Pacific Warm Pool. Microphysical budgets depend on atmospheric dynamical and thermodynamical conditions which determine the partitioning of hydrometeors. This knowledge would benefit the microphysics parameterization in cloud models and cumulus parameterization in global circulation models.

Complement of existing ASAMPSA2 guidance for Level 2 PSA for shutdown states of reactors, Spent Fuel Pool and recent R and D results

International Nuclear Information System (INIS)

Kumar, M.; Olsson, A.; Loeffler, H.; Morandi, S.; Gumenyuk, D.; Dejardin, P.; Yu, S.; Jan, P.; Kubicek, J.; Serrano, C.; Raimond, E.; Dirksen, G.; Ivanov, I.; Groudev, P.; Kowal, K.; Prosek, Andrej; Nitoi, M.; Vitazkova, J.; Hirata, K.; Burgazzi, L.

2016-01-01

This report can be considered as an addendum to the existing ASAMPSA2 guidance for Level 2 PSA. It provides complementary guidance for Level 2 PSA for accident in the NPP shutdown states and on spent fuel pool and comments on the importance of these accidents on nuclear safety. It includes also information on recent research and development useful for Level 2 PSA developments. The conclusions of the ASAMPSA-E end-users survey and of technical meetings of WP10, WP21, WP22, and WP30 at Vienna University in September 2014 which are relevant for Level 2 PSA have been reflected and are taken into account as much as it is possible with the current status of knowledge. For Level 2 PSA in shutdown states, two plant conditions are to be distinguished: - accident sequences with RPV head closed, - accident sequences with RPV head open. When the RPV head is closed, core melt accident phenomena are very similar to the sequences going on in full power mode. Therefore, the large body of guidance which is available for full power mode is basically applicable to shutdown mode with RPV closed as well. When the RPV is open, some of the L2 PSA issues become irrelevant compared to full power mode, while others come into existence. The situation is different for aspects which do not exist or which are less pronounced in sequences with RPV closed. The report also covers containment issues in shutdown states and discusses the applicability of existing guidance, potential gaps and deficiencies and recommendations are provided. For spent fuel pool accidents in Level 2 PSA, a set of issues is identified and addressed. If the spent fuel pool is located inside the containment, the potential release paths to the environment are almost the same as for core melt accidents in the RPV. If the spent fuel pool is located outside the containment, the potential release paths to the environment depend very much on plant specific properties, e.g. ventilation systems, building doors, roof under thermal

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

Simulation of the Gamma Dose Rate in Loss of Pool Water Accident of the Second Egyptian Research Reactor ETRR-2

International Nuclear Information System (INIS)

Amin, E.; Saleh, H.; Ashoub, N.

2000-01-01

The Second Egyptian Research Reactor ETRR-2, is a pool type reactor, a sudden loss of pool water resulting of leaving the core region un-covered. The reactor core is surrounded by chimney chambers whose water is isolated from pool water. This accident would lead to significant external dose. A model is developed and is used to calculate the dose rates for key access and traffic plans from indirect line of sight of the core have a maximum dose rate. The model developed uses the discrete ordinate method as implemented in the code DOT 3.5

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)

New set of convective heat transfer coefficients established for pools and validated against CLARA experiments for application to corium pools

Energy Technology Data Exchange (ETDEWEB)

Michel, B., E-mail: benedicte.michel@irsn.fr

2015-05-15

Highlights: • A new set of 2D convective heat transfer correlations is proposed. • It takes into account different horizontal and lateral superficial velocities. • It is based on previously established correlations. • It is validated against recent CLARA experiments. • It has to be implemented in a 0D MCCI (molten core concrete interaction) code. - Abstract: During an hypothetical Pressurized Water Reactor (PWR) or Boiling Water Reactor (BWR) severe accident with core meltdown and vessel failure, corium would fall directly on the concrete reactor pit basemat if no water is present. The high temperature of the corium pool maintained by the residual power would lead to the erosion of the concrete walls and basemat of this reactor pit. The thermal decomposition of concrete will lead to the release of a significant amount of gases that will modify the corium pool thermal hydraulics. In particular, it will affect heat transfers between the corium pool and the concrete which determine the reactor pit ablation kinetics. A new set of convective heat transfer coefficients in a pool with different lateral and horizontal superficial gas velocities is modeled and validated against the recent CLARA experimental program. 155 tests of this program, in two size configurations and a high range of investigated viscosity, have been used to validate the model. Then, a method to define different lateral and horizontal superficial gas velocities in a 0D code is proposed together with a discussion about the possible viscosity in the reactor case when the pool is semi-solid. This model is going to be implemented in the 0D ASTEC/MEDICIS code in order to determine the impact of the convective heat transfer in the concrete ablation by corium.

Simulation of Thermal Stratification in BWR Suppression Pools with One Dimensional Modeling Method

Energy Technology Data Exchange (ETDEWEB)

Haihua Zhao; Ling Zou; Hongbin Zhang

2014-01-01

The suppression pool in a boiling water reactor (BWR) plant not only is the major heat sink within the containment system, but also provides the major emergency cooling water for the reactor core. In several accident scenarios, such as a loss-of-coolant accident and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; the pool temperature distribution also affects the NPSHa (available net positive suction head) and therefore the performance of the Emergency Core Cooling System and Reactor Core Isolation Cooling System pumps that draw cooling water back to the core. Current safety analysis codes use zero dimensional (0-D) lumped parameter models to calculate the energy and mass balance in the pool; therefore, they have large uncertainties in the prediction of scenarios in which stratification and mixing are important. While three-dimensional (3-D) computational fluid dynamics (CFD) methods can be used to analyze realistic 3-D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, resulting in a long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code (Berkeley mechanistic MIXing code in C++) has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by one-dimensional (1-D) transient partial differential equations and substructures (such as free or wall jets) are modeled with 1-D integral models. This allows very large reductions in computational effort compared to multi-dimensional CFD modeling. One heat-up experiment performed at the Finland POOLEX facility, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, is used for

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)

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

Governing equations for heat and mass transfer in heat-generating porous beds-II. Particulate melting and substrate penetration by dissolution

Energy Technology Data Exchange (ETDEWEB)

Chawla, T C; Minkowycz, W J; Pedersen, D R

1985-11-01

Upon dryout of the bed, the dominant modes of heat transfer are conduction and radiation. Radiation is modeled through the Rosseland approximation. The melting of stainless-steel particulate imbedded in the fuel is modeled by assuming the bed to be a continuum with conduction and radiatio as the dominant modes of heat transfer. The molten steel, after it drains to the bottom of the bed, is assumed to disappear into cracks and mortar joints of the MgO bricks. The melting of fuel in the interior of the bed is modeled identically to the steel particulate, except for the bed settling which is more pronounced in the case of fuel melting and is assumed to be instantaneous owing to the significant weight of overlying bed and sodium pool. The molten layer of fuel, as it collects at the bottom of the bed, causes the heatup of the MgO lining to the eutectic temperature (2280/sup 0/C), and the MgO lining begins to dissolve. The density gradient caused by the dissolution of MgO leads to natural convection and mixing in the molten layer. The submerged fuel particulate also begins to dissolve in the molten solution and ultimately leads to the conversion of debris to a molten pool of fuel and MgO. The process of penetration of the MgO lining continues until the mixing process lowers the concentration of fuel in the volume of the pool to the level where the internal heat rate per unit volume is not enough to keep the body of the pool molten and leads to freezing in the cooler part of the pool. A the molten pool reaches a frozen or a quiescent state, the MgO brick lining thickness provided is deemed ''safe'' for a given bed loading and the external rate of cooling.

Equipment for the conditioning of core components in the fuel element storage pool with particular respect to the design required by the conditions for nuclear facilities in operation and the surveillance in accordance with atomic rules and regulations

International Nuclear Information System (INIS)

Dumpe, J.; Schwiertz, V.; Geiser, C.; Prucker, E.

2001-01-01

In nuclear power plants worn out and activated parts from the reactor core (core components) which are placed into the fuel element storage pool arise on a regular basis during the technical maintenance and the review. The disposal of these core components due to radiation protection aspects is only feasible within the fuel element storage pool during the operation of the NPP using techniques of the under water conditioning. Therefore, special GNS equipment is used for the conditioning, using under water conditioning equipment, such as UWS, BZ, and ZVA, a number of lifting and auxiliary equipment for mounting and dismantling purposes and the handling of the core components and the waste casks within the fuel element storage pool. These components must meet particular safety requirements with regard to their integrity and reliability. They are designed according to the requirements on nuclear components (KTA). The manipulating equipment must be partly redundant and the protection goals for nuclear accidents must be met. The Bavarian Ministry for Development and Environment tasked the TUeV Sueddeutschland with the surveillance and control. The conditioning equipment of GNS is therefore designed in co-ordination with the examiner of the Governmental Regulating Agency, in particular respect to all safety aspects. Furthermore the examiners perform reviews of the construction and the documentation during the design and construction phase. (orig.)

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.

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

Energy Technology Data Exchange (ETDEWEB)

An, Sang Mo, E-mail: sangmoan@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yueong-gu, Daejeon, 305-353 (Korea, Republic of); Song, Jin Ho [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yueong-gu, Daejeon, 305-353 (Korea, Republic of); Kim, Jong-Yun [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yueong-gu, Daejeon, 305-353 (Korea, Republic of); Radiochemistry & Nuclear Nonproliferation, University of Science & Technology, Gajeong-ro 217, Yuseong-gu, Daejeon, 34113 (Korea, Republic of); Kim, HwanYeol [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yueong-gu, Daejeon, 305-353 (Korea, Republic of); Naitoh, Masanori [The Institute of Applied Energy, 1-14-2 Nishi-shimbashi, 1-Chome, Minato-ku, Tokyo, 105-0003 (Japan)

2016-09-15

A configuration of molten core in the Fukushima Daiichi NPP (nuclear power plant) was investigated by a melting and solidification experiment. About 5 kg of a mixture, whose composition in terms of weight is UO{sub 2} (60%), Zr + ZrO{sub 2} (25%), stainless steel (14%), B{sub 4}C (1%), was melted in a cold crucible using an induction heating technique. It was shown that the solidified melt consists of upper crust and lower solidified ingot. The solidified ingot was separated into two layers. A physical and chemical analysis was performed for the samples taken from the solidified melt to investigate the morphology and chemical characteristics. It was found that the solidified ingot consists of a metal-rich layer on the top and an oxide-rich layer at the bottom. In addition, the oxide layer at the bottom has composition close to the initial charge composition and surrounded by a thin crust layer. It turned out that B{sub 4}C was more concentrated in the upper metal-rich layer. These findings provide important insights for understanding the core melt progression and taking proper post-accident recovery actions for the Fukushima Daiichi NPP. - Highlights: • A configuration of molten core in the Fukushima Daiich NPP unit 1 is investigated. • Corium ingot consists of metallic layer on the top and oxidic layer at the bottom. • Boron carbide was more concentrated in the upper metallic layer. • Two layered configuration would contribute to the post-accident recovery actions.

Untitled

Indian Academy of Sciences (India)

particularly easy for pool type reactors where the possibility of loss of sodium by pipe ... and secondary pumps followed by a failure to shut down the reactor does not ... collapse of the core assuming gross core melting due to loss of flow of.

Natural convection heat transfer in SIGMA experiment

International Nuclear Information System (INIS)

Lee, Seung Dong; Lee, Gang Hee; Suh, Kune Yull

2004-01-01

A loss-of-coolant accident (LOCA) results in core melt formation and relocation at various locations within the reactor core over a considerable period of time. If there is no effective cooling mechanism, the core debris may heat up and commence natural circulation. The high temperature pool of molten core material will threaten the structural integrity of the reactor vessel. The extent and urgency of this threat depend primarily upon the intensity of the internal heat sources and upon the consequent distribution of the heat fluxes on the vessel walls in contact with the molten core material pools. In such a steady molten pool convection state, the thermal loads against the vessel would be determined by the in-vessel heat transfer distribution involving convective and conductive heat transfer from the decay-heated core material pool to the lower head wall in contact with the core material. In this study, upward and downward heat transfer fraction ratio is focused on

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.

In vitro biocompatibility of CoCrMo dental alloys fabricated by selective laser melting.

Science.gov (United States)

Hedberg, Yolanda S; Qian, Bin; Shen, Zhijian; Virtanen, Sannakaisa; Wallinder, Inger Odnevall

2014-05-01

Selective laser melting (SLM) is increasingly used for the fabrication of customized dental components made of metal alloys such as CoCrMo. The main aim of the present study is to elucidate the influence of the non-equilibrium microstructure obtained by SLM on corrosion susceptibility and extent of metal release (measure of biocompatibility). A multi-analytical approach has been employed by combining microscopic and bulk compositional tools with electrochemical techniques and chemical analyses of metals in biologically relevant fluids for three differently SLM fabricated CoCrMo alloys and one cast CoCrMo alloy used for comparison. Rapid cooling and strong temperature gradients during laser melting resulted in the formation of a fine cellular structure with cell boundaries enriched in Mo (Co depleted), and suppression of carbide precipitation and formation of a martensitic ɛ (hcp) phase at the surface. These features were shown to decrease the corrosion and metal release susceptibility of the SLM alloys compared with the cast alloy. Unique textures formed in the pattern of the melting pools of the three different laser melted CoCrMo alloys predominantly explain observed small, though significant, differences. The susceptibility for corrosion and metal release increased with an increased number (area) of laser melt pool boundaries. This study shows that integrative and interdisciplinary studies of microstructural characteristics, corrosion, and metal release are essential to assess and consider during the design and fabrication of CoCrMo dental components of optimal biocompatibility. The reason is that the extent of metal release from CoCrMo is dependent on fabrication procedures. Copyright © 2014 Academy of Dental Materials. Published by Elsevier Ltd. All rights reserved.

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

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.)

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

Post-accident cooling capacity analysis of the AP1000 passive spent fuel pool cooling system

International Nuclear Information System (INIS)

Su Xia

2013-01-01

The passive design is used in AP1000 spent fuel pool cooling system. The decay heat of the spent fuel is removed by heating-boiling method, and makeup water is provided passively and continuously to ensure the safety of the spent fuel. Based on the analysis of the post-accident cooling capacity of the spent fuel cooling system, it is found that post-accident first 72-hour cooling under normal refueling condition and emergency full-core offload condition can be maintained by passive makeup from safety water source; 56 hours have to be waited under full core refueling condition to ensure the safety of the core and the spent fuel pool. Long-term cooling could be conducted through reserved safety interface. Makeup measure is available after accident and limited operation is needed. Makeup under control could maintain the spent fuel at sub-critical condition. Compared with traditional spent fuel pool cooling system design, the AP1000 design respond more effectively to LOCA accidents. (authors)

In-service inspection of pool type research reactors

International Nuclear Information System (INIS)

Rajamani, K.

2002-01-01

In the case of Apsara Reactor, it has been proposed to carry out major modifications in the near future. It is planned to modify the core suitably with a heavy water reflector tank to demonstrate the Multiple Purpose Research Reactor concept. The core structure will be a stationary one and will be located at the 'B' position of the pool. The modified reactor will be operated at 1 MW power level. Suitable methodologies are evolved for carrying out a planned ISI for this modified reactor

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

Dynamic behavior of a solid particle bed in a liquid pool

International Nuclear Information System (INIS)

Liu Ping; Yasunaka, Satoshi; Matsumoto, Tatsuya; Morita, Koji; Fukuda, Kenji; Yamano, Hidemasa; Tobita, Yoshiharu

2007-01-01

Dynamic behavior of solid particle beds in a liquid pool against pressure transients was investigated to model the mobility of core materials in a postulated disrupted core of a liquid metal fast reactor. A series of experiments was performed with a particle bed of different bed heights, comprising different monotype solid particles, where variable initial pressures of the originally pressurized nitrogen gas were adopted as the pressure sources. Computational simulations of the experiments were performed using SIMMER-III, a fast reactor safety analysis code. Comparisons between simulated and experimental results show that the physical model for multiphase flows used in the SIMMER-III code can reasonably represent the transient behaviors of pool multiphase flows with rich solid phases, as observed in the current experiments. This demonstrates the basic validity of the SIMMER-III code on simulating the dynamic behaviors induced by pressure transients in a low-energy disrupted core of a liquid metal fast reactor with rich solid phases

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

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).

Results of the Electron-Beam Button Melting of very clean Ni-base superalloys for the identification of nonmetallic inclusions

Energy Technology Data Exchange (ETDEWEB)

Hauner, F.; Stephan, H.; Stumpp, H.

1986-02-01

The reliability of components made of high strength materials is substantially influenced by their cleanliness. For example, the ductility, the fatigue-characteristics and the stress resistance of high strength alloys can be improved by increasing the cleanliness along with decreasing the inclusion size to below 25 ..mu..m. For the analysis of such high clean alloys with decreasing size of nonmetallic inclusions, the metallographic texting methods become troublesome and inexact for a dependable quality control. The Electron-Beam Button Melt Test offers a possibility for the examination and qualification of the small amounts of different inclusions in the high clean materials. During a process-controlled melting procedure, inclusions of high density sink to the bottom of a water-cooled copper crucible. Low density inclusions float to the pool surface and are concentrated in the upper center of the button by means of a controlled solidification of the melting pool. For the utilization of the process in the production quality control, development and research, we have developed the Electron-Beam Button Melting Furnace ES 1/07/30 B. In this paper we will present results of the application of the ES1/07/30 B. In this paper we will present results of the application of the ES 1/07/30 B to the EB-Button melting of the Ni-Base Superalloys IN718 and Astroloy. (orig.).

Core-concrete interactions using molten urania with zirconium on a limestone concrete basemat

International Nuclear Information System (INIS)

Copus, E.R.; Brockmann, J.E.; Simpson, R.B.; Lucero, D.A.; Blose, R.E.

1992-09-01

An inductively heated experiment SURC-1, using UO 2 -ZrO 2 material, was executed to measure and assess the thermal, gas, and aerosol source terms produced during core debris/concrete interactions. The SURC-1 experiment eroded a total of 27 cm of limestone concrete during 130 minutes of sustained interaction using 204.2 kg of molten prototypic UO 2 -ZrO 2 core debris material that included 18 kg of zr metal and 3.4 kg of fission product simulants. The melt pool temperature ranged from 2100 to 2400 degrees C during the first 50 minutes of the test, followed by steady temperatures of 2000 to 2100 degrees C during the middle portion of the test and temperatures of 1800 to 2000 degrees C during the final 50 minutes of testing. The total erosion during the first 50 minutes was 16 cm with an additional 2 cm during the middle part of the test and 9 cm of ablation during the final 50 minutes. Aerosols were continuously released in concentrations ranging from 30 to 200 g/m 3 . Comprehensive gas flow rates, gas compositions, and aerosol compositions were also measured during the SURC-1 test

Analysis of radiation shields of BNPP spent fuel pool

International Nuclear Information System (INIS)

Ayoobian, N.; Hadad, K.; Nematollahi, M. R.

2007-01-01

Radioactive protection is one of the most important subjects in nuclear power plants safety. Analysis of BNPP spent fuel pool shielding , as a main source of energetic γ-rays was the main goal of this project. Firstly, we simulated the reactor core using WIMSD-4 neutronic code and the amount of fission product in the fuel assembly (FA) was calculated during the reactor operation. Then, by obtaining the results from the previous calculation and by using MCNP4C nuclear code , the intensity of γ-rays was obtained in layers of spent fuel pool shields. The results have shown that no significant γ-rays passed through these shields. Finally, an accident and resulting exposure dose above the pool was analyzed

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

Simulation of the gamma dose rate in a loss of pool water accident of the second Egyptian research reactor ET-RR-2

International Nuclear Information System (INIS)

Amin, E.; Saleh, H.G.; Ashoub, N.

2002-01-01

The second Egyptian research reactor ET-RR-2, is a pool type reactor. A sudden loss of pool water would leave the core region uncovered. The reactor core is surrounded by chimney chambers with water isolated from the pool water. This accident would lead to significant external doses. A model is developed and used to calculate the dose rates for key access-areas and traffic plans from indirect line of sight of the core which have a maximum dose rate. The model developed uses the discrete ordinate method as implemented in the code DOT3.5. (orig.) [de

Corium quench in deep pool mixing experiments

International Nuclear Information System (INIS)

Spencer, B.W.; McUmber, L.; Gregorash, D.; Aeschlimann, R.; Sienicki, J.J.

1985-01-01

The results of two recent corium-water thermal interaction (CWTI) tests are described in which a stream of molten corium was poured into a deep pool of water in order to determine the mixing behavior, the corium-to-water heat transfer rates, and the characteristic sizes of the quenched debris. The corium composition was 60% UO 2 , 16% ZrO 2 , and 24% stainless steel by weight; its initial temperature was 3080 K, approx.160 K above the oxide phase liquidus temperature. The corium pour stream was a single-phase 2.2 cm dia liquid column which entered the water pool in film boiling at approx.4 m/s. The water subcooling was 6 and 75C in the two tests. Test results showed that with low subcooling, rapid steam generation caused the pool to boil up into a high void fraction regime. In contrast, with large subcooling no net steam generation occurred, and the pool remained relatively quiescent. Breakup of the jet appeared to occur by surface stripping. In neither test was the breakup complete during transit through the 32 cm deep water pool, and molten corium channeled to the base where it formed a melt layer. The characteristic heat transfer rates measured 3.5 MJ/s and 2.7 MJ/s during the fall stage for small and large subcooling, respectively; during the initial stage of bed quench, the surface heat fluxes measured 2.4 MW/m 2 and 3.7 MW/m 2 , respectively. A small mass of particles was formed in each test, measuring typically 0.1 to 1 mm and 1 to 5 mm dia for the large and small subcooling conditions, respectively. 9 refs., 13 figs., 1 tab

Experimental simulation of fragmentation and stratification of core debris on the core catcher of a fast breeder reactor

Energy Technology Data Exchange (ETDEWEB)

Pillai, Dipin S.; Vignesh, R. [Indian Institute of Technology, Chennai, Tamil Nadu (India); Sudha, A. Jasmin, E-mail: jasmin@igcar.gov.in [Safety Engineering Division, Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India); Pushpavanam, S.; Sundararajan, T. [Indian Institute of Technology, Chennai, Tamil Nadu (India); Nashine, B.K.; Selvaraj, P. [Safety Engineering Division, Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India)

2016-05-15

Highlights: • Fragmentation of two simultaneous metals jets in a bulk coolant analysed. • Particle size from experiments compared with theoretical analysis. • Jet breakup modes explained using dimensionless numbers. • Settling aspects of aluminium and lead debris on collector plate studied. • Results analysed in light of core debris settling on core catcher in a FBR. - Abstract: The complex and coupled phenomena of two simultaneous molten metal jets fragmenting inside a quiescent liquid pool and settling on a collector plate are experimentally analysed in the context of safety analysis of a fast breeder reactor (FBR) in the post accident heat removal phase. Following a hypothetical core melt down accident in a FBR, a major portion of molten nuclear fuel and clad/structural material which are collectively termed as ‘corium’ undergoes fragmentation in the bulk coolant sodium in the lower plenum of the reactor main vessel and settles on the core catcher plate. The coolability of this decay heat generating debris bed is dependent on the particle size distribution and its layering i.e., stratification. Experiments have been conducted with two immiscible molten metals of different densities poured inside a coolant medium to understand their fragmentation behaviour and to assess the possibility of formation of a stratified debris bed. Molten aluminium and lead have been used as simulants in place of molten stainless steel and nuclear fuel to facilitate easy handling. This paper summarizes the major findings from these experiments. The fragmentation of the two molten metals are explained in the light of relevant dimensionless numbers such as Reynolds number and Weber Number. The mass median diameter of the fragmented debris is predicted from nonlinear stability analysis of slender jets for lead jet and using Rayleigh's classical theory of jet breakup for aluminium jet. The agreement of the predicted values with the experimental results is good. These

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.

Assessment of the integral code ASTEC with respect to the late in-vessel phase of core degradation

International Nuclear Information System (INIS)

D'Alessandro, Christophe; Starflinger, Joerg

2014-01-01

The integral code ASTEC is being developed jointly by GRS and IRSN as the European reference code for severe accidents. In the EU project CESAM it is foreseen to assess the capabilities of ASTEC to deal with a broad range of reactor designs (PWR, BWR, VVER, CANDU, Gen III+, etc.) and especially to model and capture the effect of severe accident mitigation measures. This requires a physically sound and sufficiently accurate modelling of the processes and phenomena that govern the course of the accident, and the modelling has to be validated to a sufficient extent. Concentrating on the in-vessel aspects of severe accidents, the present paper addresses these requirements by presenting results of ASTEC calculations for relevant experiments that cover the major physical phenomena during core degradation (melting and relocation of the fuel, oxidation, molten corium pool formation and its coolability in the lower plenum once it slumped from the core region). The assessment of models for bundle degradation is based on CORA (13 and W2). CORA represented a bundle of non-irradiated, electrically heated UO 2 -rods. Melt progression in strongly degraded geometry is addressed in the PHEBUS-FTP4 experiment carried out with irradiated fuel in debris bed configuration. The validation of molten pool modelling is based on BALI and RASPLAV-Salt experiments. The BALI-facility consists of a full-scale slice of lower plenum (allowing experiments at prototypical Rayleigh numbers) and utilizes uniformly heated water as simulant for corium. The RASPLAV experiments use a scaled-down slice of the lower head. Use of non-eutectic molten salt as simulant should address the effect of a significant solidification range typical for real corium. Calculation results of ASTEC are discussed in comparison with experimental measurements. Further, questions concerning the extrapolation of findings from validation to reactor application are critically discussed, concerning e.g. choice of model parameters

Investigations on sump cooling after core melt down

Energy Technology Data Exchange (ETDEWEB)

Knebel, J.U. [Forschungeszentrum Karlsruhe - Technik und Umwelt Institut fuer Angewandte Thermo- und Fluiddynamik, Karlsruhe (Germany)

1995-09-01

This article presents the basic physical phenomena and scaling criteria of decay heat removal from a large coolant pool by single-phase and two-phase natural circulation flow. The physical significance of the dimensionless similarity groups derived is evaluated. The above results are applied to the SUCO program that is performed at the Forschungszentrum Karlsruhe. The SUCO program is a three-step series of scaled model experiments investigating the possibility of an optional sump cooling concept for the European Pressurized Water Reactor EPR. This concept is entirely based on passive safety features within the containment. The work is supported by the German utilities and the Siemens dimensional SUCOS-2D test facility. The experimental results of the model geometry are transformed to prototypic conditions.

Investigations on sump cooling after core melt down

International Nuclear Information System (INIS)

Knebel, J.U.

1995-01-01

This article presents the basic physical phenomena and scaling criteria of decay heat removal from a large coolant pool by single-phase and two-phase natural circulation flow. The physical significance of the dimensionless similarity groups derived is evaluated. The above results are applied to the SUCO program that is performed at the Forschungszentrum Karlsruhe. The SUCO program is a three-step series of scaled model experiments investigating the possibility of an optional sump cooling concept for the European Pressurized Water Reactor EPR. This concept is entirely based on passive safety features within the containment. The work is supported by the German utilities and the Siemens dimensional SUCOS-2D test facility. The experimental results of the model geometry are transformed to prototypic conditions

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...

Reflection of illumination laser from gas metal arc weld pool surface

International Nuclear Information System (INIS)

Ma, Xiaoji; Zhang, YuMing

2009-01-01

The weld pool is the core of the welding process where complex welding phenomena originate. Skilled welders acquire their process feedback primarily from the weld pool. Observation and measurement of the three-dimensional weld pool surface thus play a fundamental role in understanding and future control of complex welding processes. To this end, a laser line is projected onto the weld pool surface in pulsed gas metal arc welding (GMAW) and an imaging plane is used to intercept its reflection from the weld pool surface. Resultant images of the reflected laser are analyzed and it is found that the weld pool surface in GMAW does specularly reflect the projected laser as in gas tungsten arc welding (GTAW). Hence, the weld pool surface in GMAW is also specular and it is in principle possible that it may be observed and measured by projecting a laser pattern and then intercepting and imaging the reflection from it. Due to high frequencies of surface fluctuations, GMAW requires a relatively short time to image the reflected laser

Siloe, Osiris, and the future perspective of swimming-pool reactors

International Nuclear Information System (INIS)

Chatoux, J.; Denielou, G.; Lerouge, B.

1964-01-01

Siloe and Osiris are two new general purpose research reactors of the 'Commissariat a l'energie Atomique'. Siloe, located within the 'Centre d'Etudes Nucleaires' of Grenoble is a swimming pool reactor of the same type as Melusine and Triton. It operates, at a nominal power of 15 MW thermal and has reached the peak power of 20 MW thermal with two thirds of its cooling system working. The fast flux above 1 MeV, which is maximum at the center of the core at 15 MW thermal is 1,2. 10 14 . The core, quite open, is downward cooled. Average specific power is 159 kW/l. Osiris is under construction at Saclay. Designed for 50 MW thermal, this reactor is upward cooled. The fast flux at the center of the core above 1 MeV is calculated to be 2, 5.10 14 . The average designed specific power is 280 kW/l. A fixed zircaloy gamma shield makes a box round the core. Future perspectives open to non-pressurised swimming-pool reactors are examined. Ways are suggested for neutronic; thermal and shielding modifications which make possible further improvements in the performances and economy of these devices. (authors) [fr

Velocity Fields Measurement of Natural Circulation Flow inside a Pool Using PIV Technique

International Nuclear Information System (INIS)

Kim, Seok; Kim, Dong Eok; Youn, Young Jung; Euh, Dong Jin; Song, Chul Hwa

2012-01-01

Thermal stratification is encountered in large pool of water increasingly being used as heat sink in new generation of advanced reactors. These large pools at near atmospheric pressure provide a heat sink for heat removal from the reactor or steam generator, and the containment by natural circulation as well as a source of water for core cooling. For examples, the PAFS (passive auxiliary feedwater system) is one of the advanced safety features adopted in the APR+ (Advanced Power Reactor Plus), which is intended to completely replace the conventional active auxiliary feedwater system. The PAFS cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by adopting a natural convection mechanism. In a pool, the heat transfer from the PCHX (passive condensation heat exchanger) contributed to increase the pool temperature up to the saturation condition and induce the natural circulation flow of the PCCT (passive condensate cooling tank) pool water. When a heat rod is placed horizontally in a pool of water, the fluid adjacent to the heat rod gets heated up. In the process, its density reduces and by virtue of the buoyancy force, the fluid in this region moves up. After reaching the top free surface, the heated water moves towards the other side wall of the pool along the free surface. Since this heated water is cooling, it goes downward along the wall at the other side wall. Above heater rod, a natural circulation flow is formed. However, there is no flow below heater rod until pool water temperature increases to saturation temperature. In this study, velocity measurement was conducted to reveal a natural circulation flow structure in a small pool using PIV (particle image velocimetry) measurement technique

Stability and Scalability of the CMS Global Pool: Pushing HTCondor and GlideinWMS to New Limits

Energy Technology Data Exchange (ETDEWEB)

Balcas, J. [Caltech; Bockelman, B. [Nebraska U.; Hufnagel, D. [Fermilab; Hurtado Anampa, K. [Notre Dame U.; Aftab Khan, F. [NCP, Islamabad; Larson, K. [Fermilab; Letts, J. [UC, San Diego; Marra da Silva, J. [Sao Paulo, IFT; Mascheroni, M. [Fermilab; Mason, D. [Fermilab; Perez-Calero Yzquierdo, A. [Madrid, CIEMAT; Tiradani, A. [Fermilab

2017-11-22

The CMS Global Pool, based on HTCondor and glideinWMS, is the main computing resource provisioning system for all CMS workflows, including analysis, Monte Carlo production, and detector data reprocessing activities. The total resources at Tier-1 and Tier-2 grid sites pledged to CMS exceed 100,000 CPU cores, while another 50,000 to 100,000 CPU cores are available opportunistically, pushing the needs of the Global Pool to higher scales each year. These resources are becoming more diverse in their accessibility and configuration over time. Furthermore, the challenge of stably running at higher and higher scales while introducing new modes of operation such as multi-core pilots, as well as the chaotic nature of physics analysis workflows, places huge strains on the submission infrastructure. This paper details some of the most important challenges to scalability and stability that the CMS Global Pool has faced since the beginning of the LHC Run II and how they were overcome.

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

3-dimensional thermohydraulic analysis of KALIMER reactor pool during unprotected accidents

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Bum; Hahn Do Hee

2003-01-01

During a normal reactor scram, the heat generation is reduced almost instantaneously while the coolant flow rate follows the pump coastdown. This mismatch between power and flow results in a situation where the core flow entering the hot pool is at a lower temperature than the temperature of the bulk pool sodium. This temperature difference leads to thermal stratification. Thermal stratification can occur in the hot pool region if the entering coolant is colder than the existing hot pool coolant and the flow momentum is not large enough to overcome the negative buoyancy force. Since the fluid of hot pool enters IHXs, the temperature distribution of hot pool can alter the overall system response. Hence, it is necessary to predict the pool coolant temperature distribution with sufficient accuracy to determine the inlet temperature conditions for the IHXs and its contribution to the net buoyancy head. Therefore, two-dimensional hot pool thermohydraulic model named HP2D has been developed. In this report code-to-code comparison analysis between HP2D and COMMIX-1AR/P has been performed in the case of steady-state and UTOP.

Numerical investigation on turbulent natural convection in partially connected cylindrical enclosures for analysing SFR safety under core meltdown scenario

International Nuclear Information System (INIS)

David, Dijo K.; Mangarjuna Rao, P.; Nashine, B.K.; Selvaraj, P.

2015-01-01

Under the unlikely event of severe core meltdown accident in pool type SFR, the molten core materials may rupture the grid plate which supports the fuel subassemblies and it can get relocated in to the lower pool. These debris may eventually settle on the debris collector (i.e., core catcher) installed above the bottom wall of the lower pool. The bed thus formed generates heat due to radioactive decay which has to be passively removed for maintaining the structural integrity of main vessel. By means of natural convection, the heat generated in the debris bed will be transferred to the top pool where the heat sink (i.e., Decay heat exchanger (DHX)) is installed. Heat transfer to the DHX (which is a part of safety grade decay heat removal system) can take place through the opening created in the grid plate which connects the two liquid pools (i.e., the top pool and the lower pool). Heat transfer can also take place through the lateral wall of the lower cylindrical pool to the side pool and eventually to the top pool, and thus to the DHX. This study numerically investigates the effectiveness of heat transfer between lower pool and top pool during PARR by considering them as partially connected cylindrical enclosures. The governing equations have been numerically solved using finite volume method in cylindrical co-ordinates using SIMPLE algorithm. Turbulence has been modeled using k-ω model and the model is validated against benchmark problems of natural convection found in literature. The effect of parameters such as the heat generation rate in the bed and the size of the grid plate opening are evaluated. Also PAHR in SFR pool is modeled using an axi-symmetric model to fund out the influence of grid plate opening on heat removal from core catcher. The results obtained are useful for improving the cooling capability of in-vessel tray type core catcher for handling the whole core meltdown scenarios in SFR. (author)