In-vessel coolability and steam explosion in Nordic BWRs

International Nuclear Information System (INIS)

Ma, W.; Hansson, R.; Li, L.; Kudinov, P.; Cadinu, F.; Tran, C-.T.

2010-05-01

The INCOSE project is to reduce the uncertainty in quantification of steam explosion risk and in-vessel coolability in Nordic BWR plants with the cavity flooding as a severe accident management (SAM) measure. During 2009 substantial advances and new insights into physical mechanisms were gained for studies of: (i) in-vessel corium coolability - development of the methodologies to assess the efficiency of the control rod guide tube (CRGT) cooling as a potential SAM measure; (ii) debris bed coolability - characterization of the effective particle diameter of multi-size particles and qualification of friction law for two-phase flow in the beds packed with multi-size particles; and (iii) steam explosion - investigation of the effect of binary oxides mixtures properties on steam explosion. An approach for coupling of ECM/PECM models with RELAP5 was developed to enhance predictive fidelity for melt pool heat transfer. MELCOR was employed to examine the CRGT cooling efficiency by considering an entire accident scenario, and the simulation results show that the nominal flowrate (∼10kg/s) of CRGT cooling is sufficient to maintain the integrity of the vessel in a BWR of 3900 MWth, if the water injection is activated no later than 1 hour after scram. The POMECO-FL experimental data suggest that for a particulate bed packed with multi-size particles, the effective particle diameter can be represented by the area mean diameter of the particles, while at high velocity (Re>7) the effective particle diameter is closer to the length mean diameter. The pressure drop of two-phase flow through the particulate bed can be predicted by Reed's model. The steam explosion experiments performed at high melt superheat (>200oC) using oxidic mixture of WO3-CaO didn't detect an apparent difference in steam explosion energetics and preconditioning between the eutectic and noneutectic melts. This points out that the next step of MISTEE experiment will be conducted at lower superheat. (author)

Experimental investigation of coolability behaviour of irregularly shaped particulate debris bed

International Nuclear Information System (INIS)

Kulkarni, P.P.; Rashid, M.; Kulenovic, R.; Nayak, A.K.

2010-01-01

In case of a severe nuclear reactor accident, the core can melt and form a particulate debris bed in the lower plenum of the reactor pressure vessel (RPV). Due to the decay heat, the particle bed, if not cooled properly, can cause failure of the RPV. In order to avoid further propagation of the accident, complete coolability of the debris bed is necessary. For that, understanding of various phenomena taking place during the quenching is important. In the frame of the reactor safety research, fundamental experiments on the coolability of debris beds are carried out at IKE with the test facility 'DEBRIS'. In the present paper, the boiling and dry-out experimental results on a particle bed with irregularly shaped particles mixed with stainless steel balls have been reported. The pressure drops and dry-out heat fluxes of the irregular-particle bed are very similar to those for the single-sized 3 mm spheres bed, despite the fact that the irregular-particle bed is composed of particles with equivalent diameters ranging from 2 to 10 mm. Under top-flooding conditions, the pressure gradients are all smaller than the hydrostatic pressure gradient of water, indicating an important role of the counter-current interfacial drag force. For bottom-flooding with a liquid inflow velocity higher than about 2.7 mm/s, the pressure gradient generally increases consistently with the vapour velocity and the fluid-particle drag becomes important. The system pressures (1 and 3 bar) have negligible effects on qualitative behaviour of the pressure gradients. The coolability of debris beds is mainly limited by the counter-current flooding limit (CCFL) even under bottom-flooding conditions with low flow rates. The system pressure and the flow rate are found to have a distinct effect on the dry-out heat flux. Different classical models have been used to predict the pressure drop characteristics and the dry-out heat flux (DHF). Comparisons are made among the models and experimental results for

Coolability of a 3D homogeneous debris bed, experimental and numerical investigations

International Nuclear Information System (INIS)

Atkhen, K.; Berthoud, G.

2001-01-01

Within the framework of nuclear safety analysis, we present here experimental and numerical results in the field of debris bed coolability. Experimental data are provided by the SILFIDE 3D experimental facility in which the debris bed is heated by induction, at Electricite de France (EDF). Numerical computations are obtained with MC3D-REPO which is a 3-phase and 3D code developed by the Commissariat a l'Energie Atomique (CEA). The uniform debris bed consists of 2 and 3,17 mm diameter steel beads contained in a 50 cm x 60 cm x 10 cm vessel. Water is used as a coolant and can be introduced either by the top or the bottom of the bed at a determined temperature. Due to heterogeneous power distribution within the bed, two definitions for the critical heat flux are proposed: the classical mean value and the local flux (much higher). Even in the first case, the measured dryout heat flux is higher than the Lipinsky 1-D flux. Temperature curve analyses show that the dryout phenomenon is very local, therefore one should be careful about the right flux definition to use. As the injected power is being increased stepwise, steady temperature stages above saturation temperature before dryout can be observed. A discussion is proposed. For some very high values of the induction power, some spheres melted together, leading to a bigger non-porous region. Even if the local temperature went over 1300 C, the bed was still coolable and the critical heat flux value was not impacted. Some parametric studies led to the following conclusions: bottom coolant injection leads to a twice time higher critical flux than by top injection, the influence of the height of the water pool above debris bed is negligible, a sub-cooled liquid injection has no influence on the coolability. Fluidization of surface particles is also discussed. The MC3D-REPO model assumes a thermal equilibrium between the three phases, which gives results in agreement with experiments until dryout occurs. (author)

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

International Nuclear Information System (INIS)

Park, H.S.; Nayak, A.K.; Hansson, R.C.; Sehgal, B.R.

2005-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2005-10-01

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

Ex-vessel debris coolability test during severe accident (COTELS project)

International Nuclear Information System (INIS)

Ogasawara, H.

1998-01-01

The objectives of the COTELS project are for severe accident management, to investigate phenomena of ex-vessel fuel-coolant interactions after reactor pressure vessel (RPV) failure and to investigate molten core-concrete interaction when coolant is injected onto molten debris. The project has being cooperated with the National Nuclear Center in the Republic of Kazakstan from 1994 to 1997 under the sponsorship of the Ministry of International Trade and Industry of Japan. Total programs are composed with the following tests. (1) Test 01 was meant to observe flow mode of falling debris. (2) Test A was meant to investigate phenomena of fuel-coolant interactions when molten debris falls into a coolant pool. (3) Test B/C investigated fuel coolant interactions and molten core-concrete interaction when coolant is injected onto debris. Detail data evaluation is underway. The following results were thus for obtained: (1) It was confirmed in Test 01 series that about 60 kg of UO 2 mixture was completely melted and fallen as a continuous jet. (2) No energetic fuel-coolant interaction was observed both in Test A and B series. (3) Debris in which decay heat was simulated was cooled by water injection in Test C series

Multiphase flow in ex-vessel coolability: development of an innovative concept

International Nuclear Information System (INIS)

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 paper provides the background of past experiments as well as key fundamentals that are 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

Experimental study on coolability of particulate core-metal debris bed with oxidization, (2). Fragmentation and enhanced heat transfer in zircaloy debris bed

International Nuclear Information System (INIS)

Su, Guanghui; Sugiyama, Ken-ichiro; Aoki, Hiroomi; Kimura, Iichi

2006-01-01

The oxidization and coolability characteristics of the particulate Zircaloy debris bed, which is deposited under the hard debris and through which first vapor penetrates and then water penetrates, are studied in the present paper. In the vapor penetration experiments, it is found that Zircaloy debris particles are effectively broken into small pieces after making thick oxidized layer with deep clacks by rapid oxidization under the condition that vapor with 20 cm/s penetrates for 30 to 70 min at an initial debris bed temperature of 1,030degC. It is also confirmed in the water penetration experiments that the oxidized particle debris bed has potentially of high coolability when water penetrates through the fully oxidized particle bed because of a high capillary force originating from those particles with deep cracks on their surfaces. Based on the present study, a new scenario for the appearance and disappearance of the hot spot in the TMI-2 accident is possible. The particulate core-metal core-metal debris bed is first heated up by rapid oxidization with heat generation when vapor can penetrate through the debris bed with porosities. This corresponds to the appearance of the hot spot. The resultant oxidized particulate debris bed causes a high coolability due to its high capillary force when the water can touch the debris bed at wet condition. This corresponds to the disappearance of the hot spot. (author)

Development and application of surrogate model for assessment of ex-vessel debris bed dryout probability - 15157

International Nuclear Information System (INIS)

Yakush, S.E.; Lubchenko, N.T.; Kudinov, P.

2015-01-01

In this work we consider a water-cooled power reactor severe accident scenario with pressure vessel failure and subsequent release of molten corium. A surrogate model for prediction of dryout heat flux for ex-vessels debris beds of different shapes is developed. Functional form of dryout heat flux dependence on problem parameters is developed by the analysis of coolability problem in non-dimensional variables. It is shown that for a flat debris bed the dryout heat flux can be represented in terms of three 1-dimensional functions for which approximating formulas are found. For two-dimensional debris beds (cylindrical, conical, Gaussian heap, mound-shaped), an additional function taking into account the bed shape geometry is obtained from numerical simulations using DECOSIM code as a full model. With the surrogate model in hand, risk analysis of debris bed coolability is carried out by Monte Carlo sampling of the input parameters within selected ranges, with assumed distribution functions

The coolability limits of a reactor pressure vessel lower head

Energy Technology Data Exchange (ETDEWEB)

Theofanous, T.G.; Syri, S. [Univ. of California, Santa Barbara, CA (United States)

1995-09-01

Configuration II of the ULPU experimental facility is described, and from a comprehensive set of experiments are provided. The facility affords full-scale simulations of the boiling crisis phenomenon on the hemispherical lower head of a reactor pressure vessel submerged in water, and heated internally. Whereas Configuration I experiments (published previously) established the lower limits of coolability under low submergence, pool-boiling conditions, with Configuration II we investigate coolability under conditions more appropriate to practical interest in severe accident management; that is, heat flux shapes (as functions of angular position) representative of a core melt contained by the lower head, full submergence of the reactor pressure vessel, and natural circulation. Critical heat fluxes as a function of the angular position on the lower head are reported and related the observed two-phase flow regimes.

Investigation of the coolability of a continuous mass of relocated debris to a water-filled lower plenum. Technical report

International Nuclear Information System (INIS)

Rempe, J.L.; Wolf, J.R.; Chavez, S.A.; Condie, K.G.; Hagrman, D.L.; Carmack, W.J.

1994-09-01

This report documents work performed to support the development of an analytical and experimental program to investigate the coolability of a continuous mass of debris that relocates to a water-filled lower plenum. The objective of this program is to provide an adequate data base for developing and validating a model to predict the coolability of a continuous mass of debris relocating to a water-filled lower plenum. The model must address higher pressure scenarios, such as the TMI-2 accident, and lower pressure scenarios, which recent calculations indicate are more likely for most operating LWR plants. The model must also address a range of possible debris compositions

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

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

Quench of molten aluminum oxide associated with in-vessel debris retention by RPV internal water

International Nuclear Information System (INIS)

Maruyama, Yu; Yamano, Norihiro; Moriyama, Kiyofumi; Park, Hyun Sun; Kudo, Tamotsu; Yang, Yanhua; Sugimoto, Jun

1999-01-01

In-vessel debris coolability experiments were performed in ALPHA program at JAERI. Molten aluminum oxide (Al 2 O 3 ) was poured into a pool of water in a lower head experimental vessel. Post-test observation and measurement using an ultrasonic technique indicated the formation of the interfacial gap between the solidified Al 2 O 3 and the vessel wall. Thermal responses of the vessel wall implied that the interfacial gap acted initially as a thermal resistance and water subsequently penetrated into the interfacial gap. The maximum heat flux at the inner surface of the vessel facing to the solidified Al 2 O 3 was roughly evaluated to be ranged from 320 kW/m 2 to 600 kW/m 2 . A post-test analysis was conducted with CAMP code. The influence of the interfacial gap on thermal behavior of Al 2 O 3 and the vessel wall was examined. (authors)

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

International Nuclear Information System (INIS)

Kim, Eunho; Lee, Mooneon; Park, Hyun Sun; Moriyama, Kiyofumi; Park, Jin Ho

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-03-15

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

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

Code development for debris bed coolability problem. Final report for the period 1997-05-01 - 1999-08-14

International Nuclear Information System (INIS)

Loboiko, A.I.

2000-03-01

The study was devoted to the problem of debris bed coolability arising from severe accident at nuclear power reactor. After reactor core melting occurs and subsequent debris bed is formed in the lower plenum of reactor pressure vessel (RPV) it is important to confine this debris bed inside RPV boundary. One of the possible accident scenarios assumes the interaction between coolant and molten core materials resulting from rapid melt quenching, freezing and fragmentation. Particulated fuel and steel may subsequently settle on available surfaces within the reactor vessel, forming debris porous beds which produce radioactive decay heating. In case of severe core degradation, such heat transfer mechanisms as radiation, conduction and natural single-phase convection may appear to be insufficient and coolant boiling may happen on the surface or inside the bed. Depending on rate of heat generation there may be sufficient debris cool down or its 'dryout' which pose a danger for RPV integrity. The study considers development of 2D numerical code capable to predict coolant saturation as a function of different parameters. Analysis of previous activities on one-dimensional and multi-dimensional models was done. On the basis of the analysis it was concluded that the correct prediction of the debris saturation on dryant power requires two-dimensional numerical simulation considering the processes like two-phase convection, capillary effects, different models of permeability, different models of heat transfer between solid debris and coolant, non-homogeneity of parameters porous medium, heat and mass transfer between debris bed and a highly porous gap along the inner RPV surface. Particular attention was given to consideration of boundary conditions for debris bed. Introduction of the analytical model for dependence of gap properties on heat flux from debris bed allowed to create an algorithm for use in numerical calculations and finally to develop a code which allowed for stable

Characteristics of debris in the lower head of a BWR in different severe accident scenarios

International Nuclear Information System (INIS)

Phung, Viet-Anh; Galushin, Sergey; Raub, Sebastian; Goronovski, Andrei; Villanueva, Walter; Kööp, Kaspar; Grishchenko, Dmitry; Kudinov, Pavel

2016-01-01

Highlights: • Station blackout scenario with delayed recovery of safety systems in a Nordic BWR is considered. • Genetic algorithm and random sampling methods are used to explore accident scenario domain. • Main groups of scenarios are identified. • Ranges and distributions of characteristics of debris bed in the lower head are determined. - Abstract: Nordic boiling water reactors (BWRs) adopt ex-vessel debris cooling to terminate severe accident progression. Core melt released from the vessel into a deep pool of water is expected to fragment and form a coolable debris bed. Characteristics of corium melt ejection from the vessel determine conditions for molten fuel–coolant interactions (FCI) and debris bed formation. Non-coolable debris bed or steam explosion can threaten containment integrity. Vessel failure and melt ejection mode are determined by the in-vessel accident progression. Characteristics (such as mass, composition, thermal properties, timing of relocation, and decay heat) of the debris bed formed in the process of core relocation into the vessel lower plenum define conditions for the debris reheating, remelting, melt-vessel structure interactions, vessel failure and melt release. Thus core degradation and relocation are important sources of uncertainty for the success of the ex-vessel accident mitigation strategy. The goal of this work is improve understanding how accident scenario parameters, such as timing of failure and recovery of different safety systems can affect characteristics of the debris in the lower plenum. Station blackout scenario with delayed power recovery in a Nordic BWR is considered using MELCOR code. The recovery timing and capacity of safety systems were varied using genetic algorithm (GA) and random sampling methods to identify two main groups of scenarios: with relatively small ( 100 tons) amount of relocated debris. The domains are separated by the transition regions, in which relatively small variations of the input

Characteristics of debris in the lower head of a BWR in different severe accident scenarios

Energy Technology Data Exchange (ETDEWEB)

Phung, Viet-Anh, E-mail: vaphung@kth.se; Galushin, Sergey, E-mail: galushin@kth.se; Raub, Sebastian, E-mail: raub@kth.se; Goronovski, Andrei, E-mail: andreig@kth.se; Villanueva, Walter, E-mail: walterv@kth.se; Kööp, Kaspar, E-mail: kaspar@safety.sci.kth.se; Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se

2016-08-15

Highlights: • Station blackout scenario with delayed recovery of safety systems in a Nordic BWR is considered. • Genetic algorithm and random sampling methods are used to explore accident scenario domain. • Main groups of scenarios are identified. • Ranges and distributions of characteristics of debris bed in the lower head are determined. - Abstract: Nordic boiling water reactors (BWRs) adopt ex-vessel debris cooling to terminate severe accident progression. Core melt released from the vessel into a deep pool of water is expected to fragment and form a coolable debris bed. Characteristics of corium melt ejection from the vessel determine conditions for molten fuel–coolant interactions (FCI) and debris bed formation. Non-coolable debris bed or steam explosion can threaten containment integrity. Vessel failure and melt ejection mode are determined by the in-vessel accident progression. Characteristics (such as mass, composition, thermal properties, timing of relocation, and decay heat) of the debris bed formed in the process of core relocation into the vessel lower plenum define conditions for the debris reheating, remelting, melt-vessel structure interactions, vessel failure and melt release. Thus core degradation and relocation are important sources of uncertainty for the success of the ex-vessel accident mitigation strategy. The goal of this work is improve understanding how accident scenario parameters, such as timing of failure and recovery of different safety systems can affect characteristics of the debris in the lower plenum. Station blackout scenario with delayed power recovery in a Nordic BWR is considered using MELCOR code. The recovery timing and capacity of safety systems were varied using genetic algorithm (GA) and random sampling methods to identify two main groups of scenarios: with relatively small (<20 tons) and large (>100 tons) amount of relocated debris. The domains are separated by the transition regions, in which relatively small

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

International Nuclear Information System (INIS)

Farmer, M.T.; Basu, S.

2006-01-01

The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) program is conducting reactor material experiments and associated analysis with the objectives of resolving the ex-vessel debris coolability issue, and to address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two objectives will demonstrate the efficacy of severe accident management guidelines for existing plants and provide the technical basis for better containment designs for future plants. Despite years of international research, there are remaining uncertainties in the models that evaluate the lateral vs. axial power split during core-concrete interaction because of a lack of truly two-dimensional experiment data. As a result, there are differences in the 2-D cavity erosion predicted by codes such as MELCOR, WECHSL, and COSACO. In the continuing effort to bridge this data gap, the third in a series of large scale Core-Concrete Interaction experiments (CCI-3) has been conducted as part of the MCCI program. This test involved the interaction of a 375 kg core-oxide melt within a two-dimensional siliceous concrete crucible. The initial phase of the test was conducted under dry conditions. After a predetermined ablation depth was reached, the cavity was flooded to obtain data on the coolability of a core melt after core-concrete interaction has progressed for some time. This paper provides a summary description of the test facility and an overview of test results

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

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

International Nuclear Information System (INIS)

Kim, Eunho; Park, Jin Ho; Moriyama, Kiyofumi; Park, Hyun Sun

2015-01-01

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

Coolability of the melt in the reactor tank. A compilation and evaluation of the state of the art and suggestions for experiments in the area; Smaeltans kylbarhet i reaktortanken. En sammanstaellning och vaerdering av kunskapslaeget och foerslag till experiment inom omraadet

Energy Technology Data Exchange (ETDEWEB)

Mueller, Ferenc [ES-konsult Energi och Saekerhet AB, Stockholm (Sweden)

2002-04-01

This study is limited to experiments about phenomena and mechanisms that affect the coolability of core debris in the reactor tank that may delay the tank rupture. The goal of the study is to get an overview of the phenomena that are important for the in-vessel coolability, and to evaluate the need for new experiments. Both theoretical and experimental projects are suggested.

Debris bed coolability using a 3-D two phase model in a porous medium

Energy Technology Data Exchange (ETDEWEB)

Bechaud, C.; Duval, F.; Fichot, F. [CEA Cadarache, Inst. de Protection et de Surete Nucleaire13 - Saint-Paul-lez-Durance (France); Quintard, M. [Institut de Mecanique des Fluides de Toulouse, 31 (France); Parent, M. [CEA Grenoble, Dept. de Thermohydraulique et de Physique, 38 (France)

2001-07-01

During a severe nuclear accident, a part of the molten corium resulting from the core degradation may relocate in the lower plenum of the reactor vessel. In order to predict the safety margin of the reactor under such conditions, the coolability of this porous heat-generating medium is evaluated in this study and compared with other investigations. In this work, conservation equations derived for debris beds are implemented in the three dimensional thermal-hydraulic module of the CATHARE code. The coolant flow is a two phase flow with phase change. The momentum balance equation for each fluid phase is an extension of Darcy's law. This extension takes into account the capillary effects between the two phases, the relative permeabilities and passabilities of each phase, the interfacial drag force between liquid and gas, and the porous bed configuration (porosity, particle diameter,... ). The model developed is three-dimensional which is important to better predict the flow in configuration such as counter-current flow or to emphasize preferential ways induced by porous geometry. The energy balance equations of the three phases (liquid, gas and solid phase) are obtained by a volume averaging process of the local conservation equations. In this method, the local thermal non-equilibrium between the three phases is considered and the heat exchanges, the phase change rate as well as the thermal dispersion coefficients are calculated as a function of the local geometry of the porous medium. Such a method allows the numerical estimation of these thermal properties which are very difficult to determine experimentally. This feature is a great advantage of this approach. After a brief description of the thermal-hydraulic model, one-dimensional predictions of critical dryout fluxes are presented and compared with results from the literature. Reasonable agreement is obtained. Then a two-dimensional calculation is presented and shows the influence of the porous medium

In-Vessel Coolability. Workshop Proceedings, in collaboration with EC-SARNET

International Nuclear Information System (INIS)

2011-01-01

Severe Accident Management Guidelines increase focus on containment integrity after some progression in the course of a severe accident. This change in priorities is made according to criteria that vary depending on reactor type and specific procedures. Once a water source has been recovered, different accident management strategies can be used: send water into the core and/or cool the reactor pressure vessel (RPV) externally. It should be noticed that, depending on the amount of water available, these strategies might conflict with other uses of water such as for instance activating spray systems in the containment or may have deleterious effects as for instance an increase in the production of hydrogen. Generally, for in-vessel reflooding, the models used for evaluation of accident management measures suffer from a lack of validation. Given this background, the objectives of the workshop were: -) to exchange information on different Severe Accident Management strategies used or contemplated for the in-vessel coolability issue; -) to review recent, ongoing and planned experimental programmes on reflooding; -) to review models used for reflooding in severe accident calculation tools, either simplified or sophisticated; -) to exchange information on the treatment of reflooding in different safety studies such as Probabilistic Safety Assessment; and -) to provide recommendations for future work, as necessary

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Experimental study on in-vessel debris coolability during severe accident

International Nuclear Information System (INIS)

Kim, S. B.; Park, R. J.; Kim, H. D.

2002-05-01

A research program, called SONATA-IV(Simulation of Naturally Arrested Thermal Attack In-Vessel), has been performed to verify the gap cooling mechanism of corium in the lower plenum, and to develop management and mitigation strategies under severe accident conditions. For the proof-of-principles experiment, the LAVA(Lower-plenum Arrested Vessel Attack) experiments have been performed to gather proof of gap formation and to evaluate the gap effect on in-vessel cooling, using Al 2 O 3 /Fe (or Al 2 O 3 only) thermite melt as corium simulant. And also the CHFG(Critical Heat Flux in Gap) experiments have been performed to measure the critical power and to investigate the inherent cooling mechanism in the hemispherical narrow gap. In addition to the experiments, LILAC code was developed to analyze and predict the thermo-hydraulic phenomena of the corium relocated in the reactor lower plenum. It could be found from the LAVA and CHFG experimental results that continuous gap ranged from 1 to 5 mm was formed and that maximum heat removal capacity through a gap is a key factor in determining the potentials of the integrity of the vessel. After all the possibility of IVR(In-Vessel corium Retention) through gap cooling highly depends on the melt relocated into the lower plenum and the gap size. So, feasibility experiments have been performed for the assessment of improved IVR concepts using an internal engineered gap device and a dual strategy of In/Ex-vessel cooling using the LAVA facility. It is preliminarily concluded that these cooling measures lead to an enhanced cooling of the corium in the lower plenum of the reactor vessel. The additional studies will be performed to verify the quantitative heat removal capacity for these cooling measures in the 2nd phase of mid- and long term project period

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Prediction of corium debris characteristics in lower plenum of a nordic BWR in different accident scenarios using MELCOR code - 15367

International Nuclear Information System (INIS)

Phung, V.A.; Galushin, S.; Raub, S.; Goronovski, A.; Villanueva, W.; Koeoep, K; Grishchenko, D.; Kudinov, P.

2015-01-01

Severe accident management strategy in Nordic boiling water reactors (BWRs) relies on ex-vessel core debris coolability. The mode of corium melt release from the vessel determines conditions for ex-vessel accident progression and threats to containment integrity, e.g., formation of a non-coolable debris bed and possibility of energetic steam explosion. In-vessel core degradation and relocation is an important stage which determines characteristics of corium debris in the vessel lower plenum, such as mass, composition, thermal properties, timing of relocation, and decay heat. These properties affect debris reheating and remelting, melt interactions with the vessel structures, and possibly vessel failure and melt ejection mode. Core degradation and relocation is contingent upon the accident scenario parameters such as recovery time and capacity of safety systems. The goal of this work is to obtain a better understanding of the impact of the accident scenarios and timing of the events on core relocation phenomena and resulting properties of the debris bed in the vessel lower plenum of Nordic BWRs. In this study, severe accidents in a Nordic BWR reference plant are initiated by a station black out event, which is the main contributor to core damage frequency of the reactor. The work focuses on identifying ranges of debris bed characteristics in the lower plenum as functions of the accident scenario with different recovery timing and capacity of safety systems. The severe accident analysis code MELCOR coupled with GA-IDPSA is used in this work. GA-IDPSA is a Genetic Algorithm-based Integrated Deterministic Probabilistic Safety Analysis tool, which has been developed to search uncertain input parameter space. The search is guided by different target functions. Scenario grouping and clustering approach is applied in order to estimate the ranges of debris characteristics and identify scenario regions of core relocation that can lead to significantly different debris bed

Physics of coolability of top flooded molten corium

International Nuclear Information System (INIS)

Kulkarni, P.P.; Singh, R.K.; Nayak, A.K.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

2011-01-01

During a postulated severe accident in a nuclear reactor in case of ex-vessel scenario the molten corium can be relocated in the containment cavity forming a melt pool. In order to arrest further progression of severe accident, complete quenching of the molten corium pool is necessary. Most common way to deal with ex-vessel scenario is to flood the melt pool with large quantity of water. However, the mechanism of coolability is much more complex involving multi-component, multiphase heat, mass and momentum transfer. In this paper, a mechanistic model has been presented for the corium coolability under top flooding conditions. The model has been validated with the experimental data of COMECO test facility available in literature. Simulations have been carried out using the model to explore the physics behind the corium coolability with MCCI under top flooding condition. Variations in the thermo-physical properties as a result of MCCI have been considered and its effect on coolability has been studied. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-15

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

Integral experiments on in-vessel coolability and vessel creep: results and analysis of the FOREVER-C1 test

Energy Technology Data Exchange (ETDEWEB)

Sehgal, B.R.; Nourgaliev, R.R.; Dinh, T.N.; Karbojian, A. [Division of Nuclear Power Safety, Royal Institute of Technology, Drottning Kristinas Vaeg., Stockholm (Sweden)

1999-07-01

This paper describes the FOREVER (Failure Of REactor VEssel Retention) experimental program, which is currently underway at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS). The objectives of the FOREVER experiments are to obtain data and develop validated models (i) on the melt coolability process inside the vessel, in the presence of water (in particular, on the efficacy of the postulated gap cooling to preclude vessel failure); and (ii) on the lower head failure due to the creep process in the absence of water inside and/or outside the lower head. The paper presents the experimental results and analysis of the first FOREVER-C1 test. During this experiment, the 1/10th scale pressure vessel, heated to about 900degC and pressurized to 26 bars, was subjected to creep deformation in a non-stop 24-hours test. The vessel wall displacement data clearly shows different stages of the vessel deformation due to thermal expansion, elastic, plastic and creep processes. The maximum displacement was observed at the lowermost region of the vessel lower plenum. Information on the FOREVER-C1 measured thermal characteristics and analysis of the observed thermal and structural behavior is presented. The coupled nature of thermal and mechanical processes, as well as the effect of other system conditions (such as depressurization) on the melt pool and vessel temperature responses are analyzed. (author)

Status Report on Ex-Vessel Coolability and Water Management

Energy Technology Data Exchange (ETDEWEB)

Farmer, M. T. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Robb, K. R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-09-15

Specific to BWR plants, current accident management guidance calls for flooding the drywell to a level of approximately 1.2 m (4 feet) above the drywell floor once vessel breach has been determined. While this action can help to submerge ex-vessel core debris, it can also result in flooding the wetwell and thereby rendering the wetwell vent path unavailable. An alternate strategy is being developed in the industry guidance for responding to the severe accident capable vent Order, EA-13-109. The alternate strategy being proposed would throttle the flooding rate to achieve a stable wetwell water level while preserving the wetwell vent path. The overall objective of this work is to upgrade existing analytical tools (i.e. MELTSPREAD and CORQUENCH - which have been used as part of the DOE-sponsored Fukushima accident analyses) in order to provide flexible, analytically capable, and validated models to support the development of water throttling strategies for BWRs that are aimed at keeping ex-vessel core debris covered with water while preserving the wetwell vent path.

Quantification of the ex-vessel severe accident risks for the Swedish boiling water reactors. A scoping study performed for the APRI project

International Nuclear Information System (INIS)

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

1995-07-01

Results of a scoping study to quantify the ex-vessel severe accident risks for the Swedish BWRs are reported. The study considers that a pool of water is established in the containment prior to vessel failure, as prescribed by the accident management scheme for the newer Swedish BWRs. The integrated methodology developed and employed combines probabilistic and deterministic treatment of the various melt-structure-water interaction processes occurring in sequence. The potential steam explosion, and the melt attack on the containment basemat, are treated with enveloping analyses. Uncertain parameters in the models and the initial conditions are treated with Monte Carlo simulations. Independent models are developed for melt coolability and possible attack on the concrete basemat. It is found that, with current models, the melt discharge scenarios, in which a large amount of accumulated melt may be released from the vessel, could subject the containment to large steam explosion loads. However, the uncertainties are so large that no definite conclusion can be drawn. The assessment of ex-vessel core debris coolability is disturbed by similar phenomenological uncertainties. Presently, coolability of the core debris can not be demonstrated. 133 refs

Quantification of the ex-vessel severe accident risks for the Swedish boiling water reactors. A scoping study performed for the APRI project

Energy Technology Data Exchange (ETDEWEB)

Okkonen, T; Dinh, T N; Bui, V A; Sehgal, B R [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Energy Systems Technology

1995-07-01

Results of a scoping study to quantify the ex-vessel severe accident risks for the Swedish BWRs are reported. The study considers that a pool of water is established in the containment prior to vessel failure, as prescribed by the accident management scheme for the newer Swedish BWRs. The integrated methodology developed and employed combines probabilistic and deterministic treatment of the various melt-structure-water interaction processes occurring in sequence. The potential steam explosion, and the melt attack on the containment basemat, are treated with enveloping analyses. Uncertain parameters in the models and the initial conditions are treated with Monte Carlo simulations. Independent models are developed for melt coolability and possible attack on the concrete basemat. It is found that, with current models, the melt discharge scenarios, in which a large amount of accumulated melt may be released from the vessel, could subject the containment to large steam explosion loads. However, the uncertainties are so large that no definite conclusion can be drawn. The assessment of ex-vessel core debris coolability is disturbed by similar phenomenological uncertainties. Presently, coolability of the core debris can not be demonstrated. 133 refs.

Prediction of mass fraction of agglomerated debris in a LWR severe accident

International Nuclear Information System (INIS)

Kudinov, P.; Davydov, M.

2011-01-01

Ex-vessel termination of accident progression in Swedish type Boiling Water Reactors (BWRs) is contingent upon efficacy of melt fragmentation and solidification in a deep pool of water below reactor vessel. When liquid melt reaches the bottom of the pool it can create agglomerated debris and “cake” regions that increase hydraulic resistance of the bed and affect coolability of the bed. This paper discusses development and application of a conservative-mechanistic approach to quantify mass fractions of agglomerated debris. Experimental data from the DEFOR-A (Debris Bed Formation and Agglomeration) tests with high superheat of binary oxidic simulant material melt is used for validation of the methods. Application of the approach to plant accident analysis suggests that melt superheat has less significant influence on agglomeration of the debris than jet penetration depth. The paper also discusses the impact of the uncertainty in the jet disintegration and penetration behavior on the agglomeration mode map. (author)

In-vessel core debris retention experiments. Final report

International Nuclear Information System (INIS)

1998-10-01

The in-vessel cooling experimental program (Phase 1 and 2) was motivated by the survivability of the TMI lower vessel head during the TMI-2 accident. During that accident, molten debris relocation into the water filled lower head resulted in a localized hot spot in the lower head, but no lower head failure occurred. A postulated set of mechanisms which could be involved in and responsible for the survivability of the TMI lower head were identified and experimentally investigated as part of this program. These mechanisms included: the formation of a gap (contact resistance) between the relocated and frozen debris and the vessel wall was a key aspect of the in-vessel cooling mechanism; wall heatup due to the relocated debris in the presence of wall stress due to a pressure gradient across the vessel wall; gap growth due to a lack of debris adherence to the vessel wall and material creep of the heated vessel wall; and the potential for enhanced wall cooling due to gap growth. Each of these postulated mechanisms was investigated in this experimental program. This report summarizes the several insights and conclusions that were obtained from this experimental program. This report documents the entire set of five experiments completed in Phase 2 of this experimental program. Results from the Phase 1 effort were used to plan and select the Phase 2 test matrix. Conclusions from the Phase 1 and 2 experiments are identified and recommendations for future work are provided

Empirical closures for particulate debris bed spreading induced by gas–liquid flow

Energy Technology Data Exchange (ETDEWEB)

Basso, S., E-mail: simoneb@kth.se; Konovalenko, A.; Kudinov, P.

2016-02-15

Highlights: • Experimental study of the debris bed self-leveling phenomenon. • A scaling approach and a non-dimensional model to describe particle flow rate are proposed. • The model is validated against experiments with particles of different properties and at different gas injection conditions. - Abstract: Efficient removal of decay heat from the nuclear reactor core debris is paramount for termination of severe accident progression. One of the strategies is based on melt fragmentation, quenching and cooling in a deep pool of water under the reactor vessel. Geometrical configuration of the debris bed is among the important factors which determine possibility of removing the decay heat from the debris bed by natural circulation of the coolant. For instance, a tall mound-shape debris bed can be non-coolable, while the same debris can be coolable if spread uniformly. Decay heat generates a significant amount of thermal energy which goes to production of steam inside the debris bed. Two-phase flow escaping through the top layer of the bed becomes a source of mechanical energy which can move the particulate debris along the slope of the bed. The motion of the debris will lead to flattening of the bed. Such process is often called “self-leveling” phenomenon. Spreading of the debris bed by the self-leveling process can take significant time, depending on the initial debris bed configuration and other parameters. There is a competition between the time scales for reaching (i) a coolable configuration of the bed, and (ii) onset of dryout and re-melting of the debris. In the previous work we have demonstrated that the rate of particulate debris spreading is determined by local gas velocity and local slope angle of the bed. In this work we develop a scaling approach and a closure for prediction of debris spreading rate based on generalization of available experimental data. We demonstrate that introduced scaling criteria are universal for particles of different

Analysis methodology for RBMK-1500 core safety and investigations on corium coolability during a LWR severe accident

International Nuclear Information System (INIS)

Jasiulevicius, Audrius

2003-01-01

This thesis presents the work involving two broad aspects within the field of nuclear reactor analysis and safety. These are: - development of a fully independent reactor dynamics and safety analysis methodology of the RBMK-1500 core transient accidents and - experiments on the enhancement of coolability of a particulate bed or a melt pool due to heat removal through the control rod guide tubes. The first part of the thesis focuses on the development of the RBMK-1500 analysis methodology based on the CORETRAN code package. The second part investigates the issue of coolability during severe accidents in LWR type reactors: the coolability of debris bed and melt pool for in-vessel and ex-vessel conditions. The first chapter briefly presents the status of developments in both the RBMK-1500 core analysis and the corium coolability areas. The second chapter describes the generation of the RBMK-1500 neutron cross section data library with the HELIOS code. The cross section library was developed for the whole range of the reactor conditions. The results of the benchmarking with the WIMS-D4 code and validation against the RBMK Critical Facility experiments is also presented here. The HELIOS generated neutron cross section data library provides a close agreement with the WIMS-D4 code results. The validation against the data from the Critical Experiments shows that the HELIOS generated neutron cross section library provides excellent predictions for the criticality, axial and radial power distribution, control rod reactivity worths and coolant reactivity effects, etc. The reactivity effects of voiding for the system, fuel assembly and additional absorber channel are underpredicted in the calculations using the HELIOS code generated neutron cross sections. The underprediction, however, is much less than that obtained when the WIMS-D4 code generated cross sections are employed. The third chapter describes the work, performed towards the accurate prediction, assessment and

Cooling of an internal-heated debris bed with fine particles

International Nuclear Information System (INIS)

Yang, Z.L.; Sehgal, B.R.

2001-01-01

In this paper, an analytical model on dryout heat flux of ex-vessel debris beds with fines particles under top flooding conditions has been developed. The parametric study is performed on the effect of the stratification of the debris beds on the dryout heat flux. The calculated results show that the stratification configuration of the debris beds with smaller particles and lower porosity layer resting on the top of another layer of the beds has profound effect on the dryout heat flux for the debris beds both with and without a downcomer. The enhancement of the dryout heat flux by the downcomer is significant. The efficiency of the single downcomer on the enhancement of the dryout heat flux is also analyzed. This, in general, agrees well with experimental data. The model is also employed to perform the assessment on the coolability of the ex-vessel debris bed under representative accidental conditions. One conservative case is chosen, and it is found that the downcomer could be efficient measure to cool the debris bed and hence terminate the severe accident. (authors)

Level 2 probabilistic event analyses and quantification

International Nuclear Information System (INIS)

Boneham, P.

2003-01-01

In this paper an example of quantification of a severe accident phenomenological event is given. The performed analysis for assessment of the probability that the debris released from the reactor vessel was in a coolable configuration in the lower drywell is presented. It is also analysed the assessment of the type of core/concrete attack that would occur. The coolability of the debris ex-vessel evaluation by an event in the Simplified Boiling Water Reactor (SBWR) Containment Event Tree (CET) and a detailed Decomposition Event Tree (DET) developed to aid in the quantification of this CET event are considered. The headings in the DET selected to represent plant physical states (e.g., reactor vessel pressure at the time of vessel failure) and the uncertainties associated with the occurrence of critical physical phenomena (e.g., debris configuration in the lower drywell) considered important to assessing whether the debris was coolable or not coolable ex-vessel are also discussed

Numerical models for the analysis of thermal behavior and coolability of a particulate debris bed in reactor lower head

Energy Technology Data Exchange (ETDEWEB)

Ahn, Kwang Il; Kim, Sang Baik; Kim, Byung Seok [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-04-01

This report provides three distinctive, but closely related numerical models developed for the analysis of thermal behavior and coolability of a particulate debris bed that is may be formed inside the reactor lower head during severe accident late phases. The first numerical module presented in the report, MELTPRO-DRY, is used to analyze numerically heat-up and melting process of the dry particle bed, downward- and sideward-relocation of the liquid melt under gravity force and capillary force acting among porous particles, and solidification of the liquid melt relocated into colder region. The second module, MELTPROG-WET, is used to simulate numerically the cooling process of the particulate debris bed under the existence of water, which is subjected to two types of numerical models. The first type of WET module utilizes distinctive models that parametrically simulate the water cooling process, that is, quenching region, dryout region, and transition region. The choice of each parametric model depends on temperature gradient between the cooling water and the debris particles. The second type of WET module utilizes two-phase flow model that mechanically simulates the cooling process of the debris bed. For a consistent simulation from the water cooling to the dryout debris bed, on the other hand, the aforementioned two modules, MELTPROG-DRY and MELTPROG-WET, were integrated into a single computer program DBCOOL. Each of computational models was verified through limited applications to a heat-generating particulate bed contained in the rectangular cavity. 22 refs., 5 figs., 2 tabs. (Author)

Dryout heat flux and flooding phenomena in debris beds consisting of homogeneous diameter particles

International Nuclear Information System (INIS)

Maruyama, Yu; Abe, Yutaka; Yamano, Norihiro; Soda, Kunihisa

1988-08-01

Since the TMI-2 accident, which occurred in 1979, necessity of understanding phenomena associated with a severe accident have been recognized and researches have been conducted in many countries. During a severe accident of a light water reactor, a debris bed consisting of the degraded core materials would be formed. Because the debris bed continues to release decay heat, the debris bed would remelt when the coolable geometry is not maintained. Thus the degraded core coolability experiments to investigate the influence of the debris particle diameter and coolant flow conditions on the coolability of the debris bed and the flooding experiments to investigate the dependence of flooding phenomena on the configuration of the debris bed have been conducted in JAERI. From the degraded core coolability experiments, the following conclusions were derived; the coolability of debris beds would be improved by coolant supply into the beds, Lipinski's 1-dimensional model shows good agreement with the measured dryout heat flux for the beds under stagnant and forced flow conditions from the bottom of the beds, and the analytical model used for the case that coolant is fed by natural circulation through the downcomer reproduces the experimental results. And the following conclusions were given from the flooding experiments ; no dependence between bed height and the flooding constant exists for the beds lower than the critical bed height, flooding phenomena of the stratified beds would be dominated by the layer consisting of smaller particles, and the predicted dryout heat flux by the analytical model based on the flooding theory gives underestimation under stagnant condition. (author)

Status of the corquench model for calculation of ex-vessel corium coolability by an overlying water layer

International Nuclear Information System (INIS)

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

2000-01-01

The results of melt attack and coolability experiment (MACE) tests have identified several heat transfer mechanisms which could potentially lead to long term corium coolability. Based on physical observations from these tests, an integrated model of corium quenching (CORQUENCH) behavior is being developed. Aside from modeling of the primary physical processes observed in the tests, considerable effort has also been devoted to modeling of test occurrences which deviate from the behavior expected at reactor scale. In this manner, extrapolation of the models validated against the test data to the reactor case can be done with increased confidence. The integrated model currently addresses early bulk cooling and incipient crust formation heat transfer phases, as well as a follow-on water ingression phase which leads to development of a sustained quench front progressing downwards through the debris. In terms of experiment distortions, the model is also able to mechanistically calculate crust anchoring to the test section sidewalls, as well as the subsequent melt/crust separation phase which arises due to concrete densification upon melting. In this paper, the status of the model development and validation activities are described. In addition, representative calculations for PWR plant conditions are provided in order to illustrate the potential benefits of overlying water on mitigation of the accident sequence. (orig.)

Modes of heat removal from a heat-generating debris bed

International Nuclear Information System (INIS)

Squarer, D.; Hochreiter, L.E.; Piecznski, A.T.

1984-01-01

In the worst hypothetical accident in a light water reactor, when all protection systems fail, the core could be converted into a deep particulate bed either in-vessel or ex-vessel. The containment of such an accident depends on the coolability of a heat-generating debris bed. Some recent experimental and analytical studies that are concerned with heat removal from such a particulate bed are reviewed. Studies have indicated that bed dryout flux and, therefore, the heat removal rate from the particulate bed increases with the particle diameter (i.e., the permeability) for pool boiling conditions and can exceed the critical heat flux of a flat plate. Bed dryout in a large particle bed (i.e., a few millimetres) was found to be closely related to the ''flooding'' limit of the bed. Dryout under forced flow conditions was found to be affected by both forced and natural convection for mass flow rate smaller than m /SUB cr/ , whereas above this mass flow rate, bed dryout is proportional to the mass flow rate. Recent analyses were found to be in agreement with experimental data; however, additional research is needed to assess factors not accounted for in previous studies (e.g., effect of pressure, multidimensionality, stratification, etc.). Based on the expected pressure and particle sizes in a postulated severe accident sequence, a debris bed should be coolable, given a sufficient water supply

Experimental investigation of the coolability of blocked hexagonal bundles

Energy Technology Data Exchange (ETDEWEB)

Hózer, Zoltán, E-mail: zoltan.hozer@energia.mta.hu; Nagy, Imre; Kunstár, Mihály; Szabó, Péter; Vér, Nóra; Farkas, Róbert; Trosztel, István; Vimi, András

2017-06-15

Highlights: • Experiments were performed with electrically heated hexagonal fuel bundles. • Coolability of ballooned VVER-440 type bundle was confirmed up to high blockage rate. • Pellet relocation effect causes delay in the cool-down of the bundle. • The bypass line does not prevent the reflood of ballooned fuel rods. - Abstract: The CODEX-COOL experimental series was carried out in order to evaluate the effect of ballooning and pellet relocation in hexagonal bundles on the coolability of fuel rods after a LOCA event. The effects of blockage geometry, coolant flowrate, initial temperature and axial profile were investigated. The experimental results confirmed that a VVER bundle up to 80% blockage rate remains coolable after a LOCA event under design basis conditions. The ballooned section creates some obstacles for the cooling water during reflood of the bundle, but this effect causes only a short delay in the cooling down of the hot fuel rods. The accumulation of fuel pellet debris in the ballooned volume results in a local power peak, which leads to further slowing down of quench front.

An experimental study on coolability through the external reactor vessel cooling according to RPV insulation design

Energy Technology Data Exchange (ETDEWEB)

Kang, Kyoung Ho; Koo, Kil Mo; Park, Rae Joon; Cho, Young Ro; Kim, Sang Baik

2004-01-01

LAVA-ERVC experiments have been performed to investigate the effect of insulation design features on the water accessibility and coolability in case of the external reactor vessel cooling. Alumina iron thermite melt was used as corium stimulant. And the hemispherical test vessel is linearly scaled-down of RPV lower plenum. 4 tests have been performed varying the melt composition and the configuration of the insulation system. Due to the limited steam venting capacity through the insulation, steam binding occurred inside the annulus in the LAVA- ERVC-1, 2 tests which were performed for simulating the KSNP insulation design. This steam binding brought about incident heat up of the vessel outer surface at the upper part in the LAVA-ERVC-1, 2 tests. On the contrary, in the LAVA-ERVC-3, 4 tests which were performed for simulating the APR1400 insulation design, the temperatures of the vessel outer surface maintained near saturation temperature. Sufficient water ingression and steam venting through the insulation lead to effective cooldown of the vessel characterized by nucleate boiling in the LAVA-ERVC-3, 4 tests. From the LAVA-ERVC experimental results, it could be preliminarily concluded that if pertinent modification of the insulation design focused on the improvement of water ingression and steam venting should be preceded the possibility of in-vessel corium retention through the external vessel cooling could be considerably increased.

Coolability of volumetrically heated particle beds

Energy Technology Data Exchange (ETDEWEB)

Rashid, Muhammad

2017-03-22

In case of a severe nuclear reactor accident, with loss of coolant, a particle bed may be formed from the fragmentation of the molten core in the residual water at different stages of the accident. To avoid further propagation of the accident and maintain the integrity of the reactor pressure vessel, the decay heat of the particle bed must be removed. To better understand the various thermo-hydraulic processes within such heat-generating particle beds, the existing DEBRIS test facility at IKE has been modified to be able to perform novel boiling, dryout and quenching experiments. The essential experimental data includes the pressure gradients measured by 8 differential pressure transducers along the bed height as a function of liquid and vapour superficial velocities, the determination of local dryout heat fluxes for different system pressures as well as the local temperature distribution measured by a set of 51 thermocouples installed inside the particle bed. The experiments were carried out for two different particle beds: a polydispersed particle bed which consisted of stainless steel balls (2 mm, 3 mm and 6 mm diameters) and an irregular particle bed which consisted of a mixture of steel balls (3 mm and 6 mm) and irregularly shaped Al{sub 2}O{sub 3} particles. Additionally, all experiments were carried out for different flow conditions, such as the reference case of passive 1D top-flooding, 1D bottom flooding (driven by external pumps and different downcomer configurations) and 2D top-/bottom-/lateral flooding with a perforated downcomer. In this work, it has been observed that for both particle beds with downcomer configurations an open downcomer leads to the best coolability (dryout heat flux = 1560 kW/m{sup 2}, polydispersed particle bed, psys = 1 bar) of the particle bed, mainly due to bottom-flow with enhanced natural convection. It has also been shown that a potential lateral flow via a perforation of the downcomer does not bring any further improvements

Coolability of volumetrically heated particle beds

International Nuclear Information System (INIS)

Rashid, Muhammad

2017-01-01

In case of a severe nuclear reactor accident, with loss of coolant, a particle bed may be formed from the fragmentation of the molten core in the residual water at different stages of the accident. To avoid further propagation of the accident and maintain the integrity of the reactor pressure vessel, the decay heat of the particle bed must be removed. To better understand the various thermo-hydraulic processes within such heat-generating particle beds, the existing DEBRIS test facility at IKE has been modified to be able to perform novel boiling, dryout and quenching experiments. The essential experimental data includes the pressure gradients measured by 8 differential pressure transducers along the bed height as a function of liquid and vapour superficial velocities, the determination of local dryout heat fluxes for different system pressures as well as the local temperature distribution measured by a set of 51 thermocouples installed inside the particle bed. The experiments were carried out for two different particle beds: a polydispersed particle bed which consisted of stainless steel balls (2 mm, 3 mm and 6 mm diameters) and an irregular particle bed which consisted of a mixture of steel balls (3 mm and 6 mm) and irregularly shaped Al 2 O 3 particles. Additionally, all experiments were carried out for different flow conditions, such as the reference case of passive 1D top-flooding, 1D bottom flooding (driven by external pumps and different downcomer configurations) and 2D top-/bottom-/lateral flooding with a perforated downcomer. In this work, it has been observed that for both particle beds with downcomer configurations an open downcomer leads to the best coolability (dryout heat flux = 1560 kW/m 2 , polydispersed particle bed, psys = 1 bar) of the particle bed, mainly due to bottom-flow with enhanced natural convection. It has also been shown that a potential lateral flow via a perforation of the downcomer does not bring any further improvements in

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

Coupled thermo-mechanical creep analysis for boiling water reactor pressure vessel lower head

International Nuclear Information System (INIS)

Villanueva, Walter; Tran, Chi-Thanh; Kudinov, Pavel

2012-01-01

Highlights: ► We consider a severe accident in a BWR with melt pool formation in the lower head. ► We study the influence of pool depth on vessel failure mode with creep analysis. ► There are two modes of failure; ballooning of vessel bottom and a localized creep. ► External vessel cooling can suppress creep and subsequently prevent vessel failure. - Abstract: In this paper we consider a hypothetical severe accident in a Nordic-type boiling water reactor (BWR) at the stage of relocation of molten core materials to the lower head and subsequent debris bed and then melt pool formation. Nordic BWRs rely on reactor cavity flooding as a means for ex-vessel melt coolability and ultimate termination of the accident progression. However, different modes of vessel failure may result in different regimes of melt release from the vessel, which determine initial conditions for melt coolant interaction and eventually coolability of the debris bed. The goal of this study is to define if retention of decay-heated melt inside the reactor pressure vessel is possible and investigate modes of the vessel wall failure otherwise. The mode of failure is contingent upon the ultimate mechanical strength of the vessel structures under given mechanical and thermal loads and applied cooling measures. The influence of pool depth and respective transient thermal loads on the reactor vessel failure mode is studied with coupled thermo-mechanical creep analysis. Efficacy of control rod guide tube (CRGT) cooling and external vessel wall cooling as potential severe accident management measures is investigated. First, only CRGT cooling is considered in simulations revealing two different modes of vessel failure: (i) a ‘ballooning’ of the vessel bottom and (ii) a ‘localized creep’ concentrated within the vicinity of the top surface of the melt pool. Second, possibility of in-vessel retention with CRGT and external vessel cooling is investigated. We found that the external vessel

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

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.

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

Ex-vessel core catcher design requirements and preliminary concepts evaluation

International Nuclear Information System (INIS)

Friedland, A.J.; Tilbrook, R.W.

1974-01-01

As part of the overall study of the consequences of a hypothetical failure to scram following loss of pumping power, design requirements and preliminary concepts evaluation of an ex-vessel core catcher (EVCC) were performed. EVCC is the term applied to a class of devices whose primary objective is to provide a stable subcritical and coolable configuration within containment following a postulated accident in which it is assumed that core debris has penetrated the Reactor Vessel and Guard Vessel. Under these assumed conditions a set of functional requirements were developed for an EVCC and several concepts were evaluated. The studies were specifically directed toward the FFTF design considering the restraints imposed by the physical design and construction of the FFTF plant

Analysis of dryout behaviour in laterally non-homogeneous debris beds using the MEWA-2D code

International Nuclear Information System (INIS)

Rahman, Saidur; Buerger, Manfred; Buck, Michael; Pohlner, Georg; Kulenovic, Rudi; Nayak, Arun Kumar; Sehgal, Bal Raj

2009-01-01

The present study analyses the impact of lateral non-homogeneities on the coolability of heated, initially water filled debris beds. Debris beds which may be formed in a postulated severe accident in light water reactors can not be expected to have a homogeneous structure. Lateral non-homogeneities are given e.g. already by a variation in height as in a heap of debris. Internally, less porous or more porous region may occur, the latter even as downcomer-like structures are considered to favour supply of water to the bed and thus coolability. In previous work it has been shown that such non-homogeneities are often strongly enhancing coolability, as compared to earlier investigations on laterally homogeneous beds. The present contribution aims at extending the view by analysing further cases of non-homogeneities with the MEWA-2D code. Especially, effects of capillary forces are considered in contrast to earlier analysis. Part of the paper deals with specific experiments performed in the POMECO facility at KTH in which a laterally stratified debris bed has been considered, whereby especially a strong jump of porosity, from 0.26 to 0.38, has been established. Astonishingly, under top as well as bottom flooding, dryout in these experiments occurred first in the lateral layer with higher porosity. Understanding is now provided by the effect of capillary forces: water is drawn from this layer to the less porous one. This effect improves the cooling in the less porous layer while it reduces coolability of the more porous layer. No real loop behaviour of inflow via the higher porosities with subsequent upflow in the less porous layer establishes here, in contrast to expectations. Other cases (different lateral heating in an otherwise homogeneous bed, closed downcomer in a homogeneous bed and heap-like debris) show, on the other hand, strongly improved coolability by such loops establishing due to the lateral differences in void and the corresponding pressure differences

Thermal-hydraulic and characteristic models for packed debris beds

International Nuclear Information System (INIS)

Mueller, G.E.; Sozer, A.

1986-12-01

APRIL is a mechanistic core-wide meltdown and debris relocation computer code for Boiling Water Reactor (BWR) severe accident analyses. The capabilities of the code continue to be increased by the improvement of existing models. This report contains information on theory and models for degraded core packed debris beds. The models, when incorporated into APRIL, will provide new and improved capabilities in predicting BWR debris bed coolability characteristics. These models will allow for a more mechanistic treatment in calculating temperatures in the fluid and solid phases in the debris bed, in determining debris bed dryout, debris bed quenching from either top-flooding or bottom-flooding, single and two-phase pressure drops across the debris bed, debris bed porosity, and in finding the minimum fluidization mass velocity. The inclusion of these models in a debris bed computer module will permit a more accurate prediction of the coolability characteristics of the debris bed and therefore reduce some of the uncertainties in assessing the severe accident characteristics for BWR application. Some of the debris bed theoretical models have been used to develop a FORTRAN 77 subroutine module called DEBRIS. DEBRIS is a driver program that calls other subroutines to analyze the thermal characteristics of a packed debris bed. Fortran 77 listings of each subroutine are provided in the appendix

An experimental study on coolability of a particulate bed with radial stratification or triangular shape

Energy Technology Data Exchange (ETDEWEB)

Thakre, Sachin; Li, Liangxing; Ma, Weimin, E-mail: ma@safety.sci.kth.se

2014-09-15

Highlights: • Dryout heat flux of a particulate bed with radial stratification is obtained. • It was found to be dominated by hydrodynamics in the bigger size of particle layer. • Coolability of a particulate bed with triangular shape is investigated. • The coolability is improved in the triangular bed due to lateral ingression of coolant. • Coolability of both beds is enhanced by a downcomer. - Abstract: This paper deals with the results of an experimental study on the coolability of particulate beds with radial stratification and triangular shape, respectively. The study is intended to get an idea on how the coolability is affected by the different features of a debris bed formed in a severe accident of light water reactors. The experiments were performed on the POMECO-HT facility which was constructed to investigate two-phase flow and heat transfer in particulate beds under either top-flooding or bottom-fed condition. A downcomer is designed to enable investigation of the effectiveness of natural circulation driven coolability. Two homogenous beds were also employed in the present study to compare their dryout power densities with those of the radially stratified bed and the triangular bed. The results show that the dryout heat fluxes of the homogeneous beds at top-flooding condition can be predicted by the Reed model. For the radially stratified bed, the dryout heat flux is dominated by two-phase flow in the columns packed with larger particles, and the dryout occurred initially near the boundary between the middle column and a side column. Given the same volume of particles under top-flooding condition, the dryout power density of the triangular bed is about 69% higher than that of the homogenous bed. The coolability of all the beds is enhanced by bottom-fed coolant driven by either forced injection or downcomer-induced natural circulation.

An experimental study on coolability of a particulate bed with radial stratification or triangular shape

International Nuclear Information System (INIS)

Thakre, Sachin; Li, Liangxing; Ma, Weimin

2014-01-01

Highlights: • Dryout heat flux of a particulate bed with radial stratification is obtained. • It was found to be dominated by hydrodynamics in the bigger size of particle layer. • Coolability of a particulate bed with triangular shape is investigated. • The coolability is improved in the triangular bed due to lateral ingression of coolant. • Coolability of both beds is enhanced by a downcomer. - Abstract: This paper deals with the results of an experimental study on the coolability of particulate beds with radial stratification and triangular shape, respectively. The study is intended to get an idea on how the coolability is affected by the different features of a debris bed formed in a severe accident of light water reactors. The experiments were performed on the POMECO-HT facility which was constructed to investigate two-phase flow and heat transfer in particulate beds under either top-flooding or bottom-fed condition. A downcomer is designed to enable investigation of the effectiveness of natural circulation driven coolability. Two homogenous beds were also employed in the present study to compare their dryout power densities with those of the radially stratified bed and the triangular bed. The results show that the dryout heat fluxes of the homogeneous beds at top-flooding condition can be predicted by the Reed model. For the radially stratified bed, the dryout heat flux is dominated by two-phase flow in the columns packed with larger particles, and the dryout occurred initially near the boundary between the middle column and a side column. Given the same volume of particles under top-flooding condition, the dryout power density of the triangular bed is about 69% higher than that of the homogenous bed. The coolability of all the beds is enhanced by bottom-fed coolant driven by either forced injection or downcomer-induced natural circulation

Numerical simulation for debris bed behavior in sodium cooled fast reactor

International Nuclear Information System (INIS)

Tagami, Hirotaka; Tobita, Yoshiharu

2014-01-01

For safety analysis of SFR, it is necessary to evaluate behavior along with coolability of debris bed in lower plenum which is formed in severe accident. In order to analyze debris behavior, model for dense sediment particles behavior was proposed and installed in SFR safety analysis code SIMMER. SIMMER code could adequately reproduce experimental results simulating the self-leveling phenomena with appropriate model parameters for bed stiffness. In reactor condition, the self-leveling experiment for prototypical debris bed has not been performed. Additionally, the prototypical debris bed consists of non-spherical particles and it is difficult to quantify model parameters. This situation brings sensitivity analysis to investigate effect of model parameters on the self-leveling phenomena of prototypical debris bed in present paper. As initial condition for sensitivity analysis, simple mound-like debris bed in sodium-filled lower plenum in reactor vessel is considered. The bed consists of the mixture of fuel debris of 3,300 kg and steel debris of 1,570 kg. Decay heat is given to this fuel debris. The model parameter is chosen as sensitivity parameter. Sensitivity analysis shows that the model parameters can effect on intensity of self-leveling phenomena and eventual flatness of bed. In all analyses, however, coolant and sodium vapor break the debris bed at mainly center part of bed and the debris is relocated to outside of bed. Through this process, the initial debris bed is almost planarized before re-melting of debris. This result shows that the model parameters affect the self-leveling phenomena, but its effect in the safety analysis of SFRs is limited. (author)

Fundamental study on dynamic behaviors of fuel debris bed. Research report in 2007 (Joint research)

International Nuclear Information System (INIS)

Morita, Koji; Fukuda, Kenji; Matsumoto, Tatsuya; Tobita, Yoshiharu; Suzuki, Tohru; Yamano, Hidemasa

2009-05-01

It is important to make a reasonable evaluation of coolability of debris bed with decay heat source in assessing post accident heat removal of a liquid metal cooled fast reactor. In general, the coolability of fuel debris depends on coolant convection, boiling and debris bed movement. In the present study, to understand fundamental characteristics of debris movement, self-leveling behavior caused by the coolant boiling was investigated experimentally using simulant materials. The present experiments employed depressurization boiling of water to simulate void distribution in a debris bed, which consists of solid particles of alumina. A rough estimation model of self-leveling occurrence was proposed and compared with the experimental results. Its extrapolation to reactor accident conditions was also discussed. In addition, solid-liquid flow experiments, which are relevant to debris bed movement behaviors, were analyzed to verify the validity of multiphase flow models employed in a safety analysis code. In the present verification study, basic validity of the code was demonstrated by analyzing experiments of water-column sloshing with solid particles. (author)

SULTAN test facility for large-scale vessel coolability in natural convection at low pressure

International Nuclear Information System (INIS)

Rouge, S.

1997-01-01

The SULTAN facility (France/CEA/CENG) was designed to study large-scale structure coolability by water in boiling natural convection. The objectives are to measure the main characteristics of two-dimensional, two-phase flow, in order to evaluate the recirculation mass flow in large systems, and the limits of the critical heat flux (CHF) for a wide range of thermo-hydraulic (pressure, 0.1-0.5 MPa; inlet temperature, 50-150 C; mass flow velocity, 5-4400 kg s -1 m -2 ; flux, 100-1000 kW m -2 ) and geometric (gap, 3-15 cm; inclination, 0-90 ) parameters. This paper makes available the experimental data obtained during the first two campaigns (90 , 3 cm; 10 , 15 cm): pressure drop differential pressure (DP) = f(G), CHF limits, local profiles of temperature and void fraction in the gap, visualizations. Other campaigns should confirm these first results, indicating a favourable possibility of the coolability of large surfaces under natural convection. (orig.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-15

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

The evaluation of pressure effects on the ex-vessel cooling for KNGR with MELCOR

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Hwa; Park, Soo Yong; Kim, Dong Ha

2001-03-01

In this report, the effect of external vessel cooling on debris coolability and vessel integrity for the KNGR were examined from the two typical pressure range of high(170 bar) and low(5 bar)case using the lower plenum model in MELCOR1.8.4. As the conditions of these calculations, 80 ton of debris was relocated simultaneously into the lower vessel head and the debris relocation temperature from the core region was 2700 K. The decay heat has been assumed to be that of one hour after reactor shutdown. The creep failure of the vessel wall was simulated with 1-D model, which can consider the rapid temperature gradient over the wall thickness during the ex-vessel cooling. From the calculation results, both the coolant temperature and the total amount of coolant mass injected into the cavity are known to be the important factors in determining the time period to keep the external vessel cool. Therefore, a long-term strategy to keep the coolant temperature subcooled throughout the transient is suggested to sustain or prolong the effect of external vessel cooling. Also, it is expected that to keep the primary side at low pressure and to perform the ex-vessel flooding be the essential conditions to sustain the vessel integrity. From MELCOR, the penetration failure always occurs after relocation regardless of the RCS pressure or availability of the external vessel cooling. Therefore, It is expected that the improvement of the model for the penetration tube failure will be necessary.

The evaluation of pressure effects on the ex-vessel cooling for KNGR with MELCOR

International Nuclear Information System (INIS)

Park, Jong Hwa; Park, Soo Yong; Kim, Dong Ha

2001-03-01

In this report, the effect of external vessel cooling on debris coolability and vessel integrity for the KNGR were examined from the two typical pressure range of high(170 bar) and low(5 bar)case using the lower plenum model in MELCOR1.8.4. As the conditions of these calculations, 80 ton of debris was relocated simultaneously into the lower vessel head and the debris relocation temperature from the core region was 2700 K. The decay heat has been assumed to be that of one hour after reactor shutdown. The creep failure of the vessel wall was simulated with 1-D model, which can consider the rapid temperature gradient over the wall thickness during the ex-vessel cooling. From the calculation results, both the coolant temperature and the total amount of coolant mass injected into the cavity are known to be the important factors in determining the time period to keep the external vessel cool. Therefore, a long-term strategy to keep the coolant temperature subcooled throughout the transient is suggested to sustain or prolong the effect of external vessel cooling. Also, it is expected that to keep the primary side at low pressure and to perform the ex-vessel flooding be the essential conditions to sustain the vessel integrity. From MELCOR, the penetration failure always occurs after relocation regardless of the RCS pressure or availability of the external vessel cooling. Therefore, It is expected that the improvement of the model for the penetration tube failure will be necessary

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.

Analysis methodology for RBMK-1500 core safety and investigations on corium coolability during a LWR severe accident

Energy Technology Data Exchange (ETDEWEB)

Jasiulevicius, Audrius

2003-07-01

This thesis presents the work involving two broad aspects within the field of nuclear reactor analysis and safety. These are: - development of a fully independent reactor dynamics and safety analysis methodology of the RBMK-1500 core transient accidents and - experiments on the enhancement of coolability of a particulate bed or a melt pool due to heat removal through the control rod guide tubes. The first part of the thesis focuses on the development of the RBMK-1500 analysis methodology based on the CORETRAN code package. The second part investigates the issue of coolability during severe accidents in LWR type reactors: the coolability of debris bed and melt pool for in-vessel and ex-vessel conditions. The first chapter briefly presents the status of developments in both the RBMK-1500 core analysis and the corium coolability areas. The second chapter describes the generation of the RBMK-1500 neutron cross section data library with the HELIOS code. The cross section library was developed for the whole range of the reactor conditions. The results of the benchmarking with the WIMS-D4 code and validation against the RBMK Critical Facility experiments is also presented here. The HELIOS generated neutron cross section data library provides a close agreement with the WIMS-D4 code results. The validation against the data from the Critical Experiments shows that the HELIOS generated neutron cross section library provides excellent predictions for the criticality, axial and radial power distribution, control rod reactivity worths and coolant reactivity effects, etc. The reactivity effects of voiding for the system, fuel assembly and additional absorber channel are underpredicted in the calculations using the HELIOS code generated neutron cross sections. The underprediction, however, is much less than that obtained when the WIMS-D4 code generated cross sections are employed. The third chapter describes the work, performed towards the accurate prediction, assessment and

Assessment of capability of models for prediction of pressure drop and dryout heat flux in a heat generating particulate debris bed

International Nuclear Information System (INIS)

Kulkarni, P.P.; Nayak, A.K.; Rashid, M.; Kulenovic, R.

2009-01-01

During a severe accident in a light water reactor, the core can melt and be relocated to the lower plenum of the reactor pressure vessel. There it can form a particulate debris bed due to the possible presence of water. This bed, if not quenched in time, can lead to the failure of the pressure vessel because of the insufficient heat removal of decay heat in the debris bed. Therefore, addressing the issue of coolability behaviour of heat generating particulate debris bed is of prime importance in the framework of severe accident management strategies, particularly in case of above mentioned late phase scenario of an accident. In order to investigate the coolability behaviour of particulate debris bed, experiments were carried out at IKE test facility 'DEBRIS' on particle beds of irregularly shaped particles mixed with spheres under top- and bottom-flooding condition. The pressure drop and dryout heat flux (DHF) were measured for top- and bottom-flooding conditions. For top-flooding conditions, it was found that the pressure gradients are all smaller than the hydrostatic pressure gradient of water, indicating an important role of the counter-current interfacial shear stress of the two-phase flow. For bottom-flooding with a relatively high liquid inflow velocity, the pressure gradient increases consistently with the vapour velocity and the fluid-particle drags become important. Also, with additional forced liquid inflow from the bottom, the DHF increases dramatically. In all the cases, it was found that the DHF is significantly larger with bottom-flooding condition compared to top-flooding condition. Different models such as Lipinski, Reed, Tung and Dhir, Hu and Theophanous, and Schulenberg and Mueller have been used to predict the pressure drop characteristics and the DHF of heat generating particulate debris beds. Comparison is made among above mentioned models and experimental results for DHF and pressure drop characteristics. Considering the overall trend in

An experimental simulation study of debris quenching in a radially stratified porous bed

International Nuclear Information System (INIS)

Sehgal, B.R.; Nayak, A.K.; Stepanyan, A.

2004-01-01

During a severe accident condition in a nuclear power plant, the core melt can fail the reactor vessel and relocate into the containment basement. In some accident management schemes, the vessel cavity is flooded with water. For these a particulate debris bed is likely to form on the cavity floor due to melt break-up in water. . In this situation, the coolability of debris bed on the containment floor is a crucial issue. This is because the debris bed still generates the decay heat and if it is uncoolable, it can eventually remelt and react with concrete basement generating a lot of noncondensable gases and pressurising the containment. Hence, it is important to cool the debris bed as an accident management programme. The main parameters affecting the coolability of the debris bed are its porosity which is a function of the size and shape of the particles which constitute the debris bed, the operating condition such as water flooding from the top or bottom of debris bed, water temperature and non-condensable gas generated during bed-concrete interactions. It is found from previous studies that the debris bed has a non-uniform particle distribution or a porosity stratification. This can happen both in radial and axial plane. For example, the bed can have a lower porosity at the centre and higher porosity at the periphery. It is of interest to investigate the quenching phenomena in such configurations so as to find an effective means of quenching the heat generating bed. While most of the previous investigations mainly concentrate on quenching of a homogenous or axially stratified particulate bed with volumetric heat generation, there are almost no studies on the above phenomena in a radially stratified porous bed. So the objective of this paper is to investigate the quenching phenomena in a radially stratified bed. To simulate the phenomena, we conducted experiments in an experimental facility named as POMECO (POrous MEdia COolability). The facility has a square

Two-Phase Flow Effect on the Ex-Vessel Corium Debris Bed Formation in Severe Accident

International Nuclear Information System (INIS)

Kim, Eunho; Park, Jin Ho; Kim, Moo Hwan; Park, Hyun Sun; Ma, Weimin; Bechta, Sevostian V.

2014-01-01

In Korean IVR-ERVC(In-Vessel Retention of molten corium through External Reactor Vessel Cooling) strategy, if the situation degenerates into insufficient external vessel cooling, the molten core mixture can directly erupt into the flooded cavity pool from the weakest point of the vessel. Then, FCI (molten Fuel Coolant Interaction) will fragment the corium jet into small particulates settling down to make porous debris bed on the cavity basemat. To secure the containment integrity against the MCCI (Molten Core - Concrete Interaction), cooling of the heat generating porous corium debris bed is essential and it depends on the characteristics of the bed itself. For the characteristics of corium debris bed, many previous experimental studies with simulant melts reported the heap-like shape mostly. There were also following experiments to develop the correlation for the heap-like shaped debris bed. However, recent studies started to consider the effect of the decay heat and reported some noticeable results with the two-phase flow effect on the debris bed formation. The Kyushu University and JAEA group reported the experimental studies on the 'self-leveling' effect which is the flattening effect of the particulate bed by the inside gas generation. The DECOSIM simulation study of RIT (Royal Institute of Technology, Sweden) with Russian researchers showed the 'large cavity pool convection' effect, which is driven by the up-rising gas bubble flow from the pre-settled debris bed, on the particle settling trajectories and ultimately final bed shape. The objective of this study is verification of the two-phase flow effect on the ex-vessel corium debris bed formation in the severe accident. From the analysis on the test movie and resultant particle beds, the two-phase flow effect on the debris bed formation, which has been reported in the previous studies, was verified and the additional findings were also suggested. For the first, in quiescent pool the

Review of the TMI-2 accident evaluation and vessel investigation projects

Energy Technology Data Exchange (ETDEWEB)

Ladekarl Thomsen, Knud

1998-03-01

The results of the TMI-2 Accident Evaluation Programme and the Vessel Investigation Project have been reviewed as part of a literature study on core meltdown and in-vessel coolability. The emphasis is placed on the late phase melt progression, which is of special relevance to the NKS-sponsored RAK-2.1 project on Severe Accident Phenomenology. The body of the report comprises three main sections, The TMI-2 Accident Scenario, Core Region and Relocation Path Investigations, and Lower Head Investigations. In the final discussion, the lower head gap formation mechanism is explained in terms of thermal contraction and fracturing of the debris crust. This model seems more plausible than the MAAP model based on creep expansion of the lower head. (au) 1 tab., 33 ills., 31 refs.

Debris interactions in reactor vessel lower plena during a severe accident. II. Integral analysis

International Nuclear Information System (INIS)

Suh, K.Y.; Henry, R.E.

1996-01-01

For pt.I see ibid., p.147-63, 1996. The integral physico-numerical model for the reactor vessel lower head response has been exercised for the TMI-2 accident and possible severe accident scenarios in PWR and BWR designs. The proposed inherent cooling mechanism of the reactor material creep and subsequent water ingression implemented in this predictive model provides a consistent representation of how the debris was finally cooled in the TMI-2 accident and how the reactor lower head integrity was maintained during the course of the incident. It should be recalled that in order for this strain to occur, the vessel lower head had to achieve temperatures in excess of 1000 C. This is certainly in agreement with the temperatures determined by metallographic examinations during the TMI-2 vessel inspection program. The integral model was also applied to typical PWR and BWR lower plena with and without structures under pressurized conditions spanning the first relocation of core material to the reactor vessel failure due to creep without recovery actions. The design application results are presented with particular attention being focused on water ingression into the debris bed through the gap formed between the debris and the vessel wall. As an illustration of the accident management application, the lower plenum with structures was recovered after an extensive amount of creep had damaged the vessel wall. The computed lower head temperatures were found to be significantly lower (by more than 300 K in this particular example) with recovery relative to the case without recovery. This clearly demonstrates the potential for in-vessel cooling of the reactor vessel without a need to externally submerge the lower head should such a severe accident occur as core melting and relocation. (orig.)

Flow characteristics of counter-current flow in debris bed

International Nuclear Information System (INIS)

Abe, Yutaka; Adachi, Hiromichi

2004-01-01

In the course of a severe accident, a damaged core would form a debris bed consisting of once-molten and fragmented fuel elements. It is necessary to evaluate the dryout heat flux for the judgment of the coolability of the debris bed during the severe accident. The dryout phenomena in the debris bed is dominated by the counter-current flow limitation (CCFL) in the debris bed. In this study, air-water counter-current flow behavior in the debris bed is experimentally investigated with glass particles simulating the debris beds. In this experiment, falling water flow rate and axial pressure distributions were experimentally measured. As the results, it is clarified that falling water flow rate becomes larger with the debris bed height and the pressure gradient in the upper region of the debris bed is different from that in the lower region of the debris bed. These results indicate that the dominant region for CCFL in the debris bed is identified near the top of the debris bed. Analytical results with annular flow model indicates that interfacial shear stress in the upper region of the debris bed is larger than that in the lower region of the debris bed. (author)

Study on effective particle diameters and coolability of particulate beds packed with irregular multi-size particles

Energy Technology Data Exchange (ETDEWEB)

Thakre, S.; Ma, W.; Kudinov, P.; Bechta, S. [Royal Institute of Technology, KTH. Div. of Nuclear Power Safety, Stockholm (Sweden)

2013-08-15

One of the key questions in severe accident research is the coolability of the debris bed, i.e., whether decay heat can be completely removed by the coolant flow into the debris bed. Extensive experimental and analytical work has been done to substantiate the coolability research. Most of the available experimental data is related to the beds packed with single size (mostly spherical) particles, and less data is available for multi-size/irregular-shape particles. There are several analytical models available, which rely on the mean particle diameter and porosity of the bed in their predictions. Two different types of particles were used to investigate coolability of particulate beds at VTT, Finland. The first type is irregular-shape Aluminum Oxide gravel particles whose sizes vary from 0.25 mm to 10 mm, which were employed in the STYX experiment programme (2001-2008). The second type is spherical beads of Zirconium silicate whose sizes vary between 0.8 mm to 1 mm, which were used in the COOLOCE tests (Takasuo et al., 2012) to study the effect of multi-dimensional flooding on coolability. In the present work, the two types of particles are used in the POMECO-FL and POMECO-HT test facility to obtain their effective particle diameters and dryout heat flux of the beds, respectively. The main idea is to check how the heaters' orientations (vertical in COOLOCE vs. horizontal in POMECO-HT) and diameters (6 mm in COOLOCE vs. 3 mm in POMECO-HT) affect the coolability (dryout heat flux) of the test beds. The tests carried out on the POMECO-FL facility using a bed packed with aluminum oxide gravel particles show the effective particle diameter of the gravel particles is 0.65 mm, by which the frictional pressure gradient can be predicted by the Ergun equation. After the water superficial velocity is higher than 0.0025 m/s, the pressure gradient is underestimated. The effective particle diameter of the zirconium particles is found as 0.8 mm. The dryout heat flux is measured on

Study on effective particle diameters and coolability of particulate beds packed with irregular multi-size particles

International Nuclear Information System (INIS)

Thakre, S.; Ma, W.; Kudinov, P.; Bechta, S.

2013-08-01

One of the key questions in severe accident research is the coolability of the debris bed, i.e., whether decay heat can be completely removed by the coolant flow into the debris bed. Extensive experimental and analytical work has been done to substantiate the coolability research. Most of the available experimental data is related to the beds packed with single size (mostly spherical) particles, and less data is available for multi-size/irregular-shape particles. There are several analytical models available, which rely on the mean particle diameter and porosity of the bed in their predictions. Two different types of particles were used to investigate coolability of particulate beds at VTT, Finland. The first type is irregular-shape Aluminum Oxide gravel particles whose sizes vary from 0.25 mm to 10 mm, which were employed in the STYX experiment programme (2001-2008). The second type is spherical beads of Zirconium silicate whose sizes vary between 0.8 mm to 1 mm, which were used in the COOLOCE tests (Takasuo et al., 2012) to study the effect of multi-dimensional flooding on coolability. In the present work, the two types of particles are used in the POMECO-FL and POMECO-HT test facility to obtain their effective particle diameters and dryout heat flux of the beds, respectively. The main idea is to check how the heaters' orientations (vertical in COOLOCE vs. horizontal in POMECO-HT) and diameters (6 mm in COOLOCE vs. 3 mm in POMECO-HT) affect the coolability (dryout heat flux) of the test beds. The tests carried out on the POMECO-FL facility using a bed packed with aluminum oxide gravel particles show the effective particle diameter of the gravel particles is 0.65 mm, by which the frictional pressure gradient can be predicted by the Ergun equation. After the water superficial velocity is higher than 0.0025 m/s, the pressure gradient is underestimated. The effective particle diameter of the zirconium particles is found as 0.8 mm. The dryout heat flux is measured on

Structural failure analysis of reactor vessels due to molten core debris

International Nuclear Information System (INIS)

Pfeiffer, P.A.

1993-01-01

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

A review of dryout heat fluxes and coolability of particle beds. APRI 4, Stage 2 Report

International Nuclear Information System (INIS)

Lindholm, Ilona

2002-04-01

were studied. Significant amount of data with prototypic material tests exists. All of the tests show significant fragmentation in case of deep subcooled pool. An additional observation is that no energetic melt coolant interaction (steam explosion) has been reported for prototypic materials. A set of most relevant data for reactor applications have been chosen. Based on this, a general particle size distribution has been constructed. The average particle size obtained by this way was about 3.5 mm. Information from fragmentation and dryout tests and the Lipinski 0-D correlation have been utilised to assess the debris bed coolability for the Olkiluoto severe accident scenario. The calculation shows that for well-mixed beds with 3.5 mm particles the dryout heat flux would be close to 1 MW/m 2 , well above the estimated heat flux due to decay heat. Stratification of finer particles on top of the bed due to e.g. a steam explosion would reduce the dryout heat flux to 50-200 kW/m 2 . This would be below heat fluxes produced by decay heat in Nordic BWRs. The key uncertainty considering particle bed coolability is due to the particle size distribution and stratification. If the possibility of a thick fine particle layer on top of the bed can be ruled out, the particulate debris bed in Nordic BWRs will be coolable. A rough estimate of melt pool coolability in Nordic BWRs has also been conducted. The MACE and COTELS experimental data have been summarised. Based on the data, the melt pools in the pedestal are slowly coolable. The concrete erosion does not threaten the containment failure margins, except maybe at Forsmark 1 and 2 units. Release of non-condensable gases during MCCI may cause an earlier start of filtered venting in Olkiluoto, Forsmark and Oskarshamn 3 plants

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2005-04-01

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

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

On the influence of water subcooling and melt jet parameters on debris formation

Energy Technology Data Exchange (ETDEWEB)

Manickam, Louis, E-mail: louis@safety.sci.kth.se; Kudinov, Pavel; Ma, Weimin; Bechta, Sevostian; Grishchenko, Dmitry

2016-12-01

Highlights: • Melt and water configuration effects on debris formation is studied experimentally. • Melt superheat and water subcooling are most influential compared to jet size. • Melt-water configuration and material properties influence particle fracture rate. • Results are compared with large scale experiments to study effect of spatial scales. - Abstract: Breakup of melt jet and formation of a porous debris bed at the base-mat of a flooded reactor cavity is expected during the late stages of a severe accident in light water reactors. Debris bed coolability is determined by the bed properties including particle size, morphology, bed height and shape as well as decay heat. Therefore understanding of the debris formation phenomena is important for assessment of debris bed coolability. A series of experiments was conducted in MISTEE-Jet facility by discharging binary-oxide mixtures of WO{sub 3}–Bi{sub 2}O{sub 3} and WO{sub 3}–ZrO{sub 2} into water in order to investigate properties of resulting debris. The effect of water subcooling, nozzle diameter and melt superheat was addressed in the tests. Experimental results reveal significant influence of water subcooling and melt superheat on debris size and morphology. Significant differences in size and morphology of the debris at different melt release conditions is attributed to the competition between hydrodynamic fragmentation of liquid melt and thermal fracture of the solidifying melt droplets. The particle fracture rate increases with increased subcooling. Further the results are compared with the data from larger scale experiments to discern the effects of spatial scales. The present work provides data that can be useful for validation of the codes used for the prediction of debris formation phenomena.

Rates of chemical reaction and atmospheric heating during core debris expulsion from a pressurized vessel

International Nuclear Information System (INIS)

Powers, D.A.; Tarbell, W.W.; Brockman, J.E.; Pilch, M.

1986-01-01

Core debris may be expelled from a pressurized reactor vessel during a severe nuclear reactor accident. Experimental studies of core debris expulsion from pressurized vessels have established that the expelled material can be lofted into the atmosphere of the reactor containment as particulate 0.4 to 2 mm in diameter. These particles will vigorously react with steam and oxygen in the containment atmosphere. Data on such reactions during tests with 80 kg of expelled melt will be reported. A model of the reaction rates based on gas phase mass transport will be described and shown to account for atmospheric heating and aerosol generation observed in the tests

The influence of selected containment structures on debris dispersal and transport following high pressure melt ejection from the reactor vessel

International Nuclear Information System (INIS)

Pilch, M.; Tarbell, W.W.; Brockmann, J.E.

1988-09-01

High pressure expulsion of molten core debris from the reactor pressure vessel may result in dispersal of the debris from the reactor cavity. In most plants, the cavity exits into the containment such that the debris impinges on structures. Retention of the debris on the structures may affect the further transport of the debris throughout the containment. Two tests were done with scaled structural shapes placed at the exit of 1:10 linear scale models of the Zion cavity. The results show that the debris does not adhere significantly to structures. The lack of retention is attributed to splashing from the surface and reentrainment in the gas flowing over the surface. These processes are shown to be applicable to reactor scale. A third experiment was done to simulate the annular gap between the reactor vessel and cavity wall. Debris collection showed that the fraction of debris exiting through the gap was greater than the gap-to-total flow area ratio. Film records indicate that dispersal was primarily by entrainment of the molten debris in the cavity. 29 refs., 36 figs., 11 tabs

Safety Strategy of JSFR establishing In-Vessel Retention of Core Disruptive Accident

International Nuclear Information System (INIS)

Tobita, Yoshiharu

2013-01-01

Coolability of debris bed was confirmed by debris bed temperature analysis coupled with the cooling system, according to the following material relocation scenario. → Case 1: Upward ejection in Transition Phase to cause shutdown. → Case 2: Early downward ejection of fuel through CRGT. → Case 3: Whole fuel accumulates on the core catcher (bounding). The flow reversal of a primary coolant loop of the two loop system of the JSFR which is caused by possible imbalance between two DHRS loops increase the flow in RV. Helpful for long-term cooling

Experimental investigations on the coolability of prototypical particle beds with respect to reactor safety; Experimentelle Untersuchungen der Kuehlbarkeit prototypischer Schuettungskonfigurationen unter dem Aspekt der Reaktorsicherheit

Energy Technology Data Exchange (ETDEWEB)

Leininger, Simon

2017-02-22

In case of a severe accident in a light water reactor, continuous unavailability of cooling water to the reactor core may result in overheating of the fuel elements and finally the loss of core integrity. Under such conditions, a structure of heat-releasing particles of different size and shape may be formed by fragmentation of molten core material in several stages of the accident. The long-term coolability of such beds is of prime im-portance to avoid any damage to the reactor pressure vessel or even a release of fission products to the environment. In the frame of this work, specific experiments were con-ducted under prototypical conditions employing the existing DEBRIS test facility in order to gain further knowledge about the thermohydraulic behavior of such beds. In steady state boiling experiments, the pressure gradients in particle beds were meas-ured both for one- and multi-dimensional cooling water flow conditions and compared with one another in order to assess the flow behavior inside the bed. For these different flow conditions as well as for stratified bed configurations, the maximum removable heat flux densities were determined in the dryout experiments. E. g., it was found that an axial stratification of the permeability can significantly reduce the bed's coolability. For the first time, the quenching behavior of dry, superheated beds was investigated at elevated system pressure up to 0.5 MPa. In these experiments, the effect of system pressure on the coolability was quantified by means of the quenching time (time period to cool down the bed to saturation temperature). The investigated particle beds mainly consisted of non-spherical particles with well-defined geometry (cylinders and screws). It was shown that the effect of the particles geometry on the flow in a particle bed can be best estimated by using an equivalent particle diameter calculated for monodisperse particle beds from the product of the Sauter diameter and a shape factor and for

An experimental study on in-vessel debris retention through gap cooling

International Nuclear Information System (INIS)

Kang, Kyung Ho; Kim, Jong Whan; Cho, Young Ro; Chang, Young Cho; Park, Rae Jun; Gu, Kil Mo; Kim, Sang Baik; Kim, Hee Dong

1999-04-01

LAVA experiments have been performed using high temperature molten material to be relocated into the 1/8 linear scaled vessel of a reactor lower plenum filled with water. An Al 2 O 3 /Fe tehrmite melt (Al 2 O 3 only) was used as a corium simulant. In this study, the influence of various initial conditions, such as internal pressure load across the vessel wall, the material composition of the melt simulant, water subcooling and depth on gap formation were investigated. As well, the thermal and mechanical behaviors of the vessel were examined. In case the internal pressure load was imposed, the gap formation between the continuous solidified debris and the vessel wall was clearly shown with post-test examination. On the other hand, in case the internal pressure load was not imposed, the iron welded to the inner surface of the vessel and the vessel experienced ablation to about 5 mm. The cooling rate of the vessel was very slow in the tests using Al 2 O 3 /Fe thermite melt but it was rather fast using Al 2 O 3 melt. It is postulated that in the Al 2 O 3 /Fe thermite tests, the iron melt layer is so dense that water ingression into the gap is difficult due to the high pressure of escaping steam. On the other hand, in the porosity of an Al 2 O 3 melt layer could enhance water ingression into the gap by giving the flow path of the evaporated steam through the porous media. The water height and subcooling could affect the melt pool formation and the initial thermal attack to the vessel. However, at present stage, the effect of water subcooling on the thermal behavior of the vessel couldn't be generalized. For clear confirmation of the effect of water subcooling, the tests will be performed at the saturated water condition. (author). 19 refs., 3 tabs., 36 figs

An assessment of ex-vessel fuel-coolant interaction energetics for advanced light water reactors

International Nuclear Information System (INIS)

Murphy, J.G.; Corradini, M.L.

1997-01-01

The occurrence of an energetic fuel/coolant interaction (FCI) below the reactor pressure vessel in the cavity of advanced light water reactors (ALWRs) are analyzed to determine the possible hazard to structural walls as a result of dynamic liquid phase pressures. Such analyses are important to demonstrate that these cavity walls will maintain their integrity so that ex-vessel core debris coolability is possible. Past studies that have examined this or related issues are reviewed, and a methodology is proposed to analyze the occurrence of this physical event using the IFCI and TEXAS models for the FCI as well as dynamic shock wave propagation estimates using hand calculations as well as the CTH hydro model. Scenarios for the ALWRs are reviewed, and one severe accident scenario is used as an example to demonstrate the methodology. Such methodologies are recommended for consideration in future safety studies. These methodologies should be verified with direct comparison to energetic FCI data such as that being produced in KROTOS at the Joint Research Centre, Ispra

Refined model for the coolability of core debris with flow entry from the bottom

International Nuclear Information System (INIS)

Schulenberg, T.; Mueller, U.

1986-01-01

Within the context of a hypothetical severe accident in light water reactors also heat generating debris beds of a coarse particle size are discussed. A refined model for two-phase flow in particle beds is presented. Compared to previous models this model takes into account the effect of interfacial drag forces between liquid and vapor. These effects are important in coarse debris beds. The model is based on the momentum equations for separated flow, which are closed by empirical relations for the wall shear stress and the interfacial drag. When the refined model is applied to LWR severe accident scenarios an increased dryout heat flux is predicted for debris beds with flow entry from the bottom driven by a moderate downcomer head

Development of a method for detecting nuclear fuel debris and water leaks at a nuclear reactor/containment vessel by flow visualization

International Nuclear Information System (INIS)

Umezawa, Shuichi; Tanaka, Katsuhiko

2013-01-01

It is the important issue to fill up each nuclear reactor/containment vessel with water and to take out debris of damaged fuel from them for decommissioning of Fukushima Daiichi nuclear power plants. It is necessary to detect the debris and water leaks at a nuclear reactor/containment vessel for the purpose. However, the method is not completely developed in the present stage. Accordingly, we have developed a method for detecting debris and water leaks at a nuclear reactor/containment vessel by flow visualization. Experiments of the flow visualization were conducted using two types of water tanks. An optical fiber and a collimator lens were employed for modifying a straight laser beam into a sheet projection. Some visualized images were obtained through the experiments. Particle Image Velocimetry, i.e. PIV, analysis was applied to the images for quantitative flow rate analysis. Consequently, it is considered that the flow visualization method has a possibility for the practical use. (author)

Effects of thermohydraulics on clad ballooning, flow blockage and coolability in a LOCA

International Nuclear Information System (INIS)

Erbacher, F.J.; Neitzel, H.J.; Wiehr, K.

1983-01-01

Thermohydraulic boundary conditions have a dominating effect on clad ballooning, flow blockage and coolability: Increasing heat transfer to the fluid decreases the total circumferential strain; Countercurrent flow in a combined injection leads to a relatively small flow blockage; Burst claddings exhibit premature quenching. Differences in the test results obtained in several countries are mainly due to different thermohydraulic test conditions; all test data are consistent with the understanding elaborated within the REBEKA program. Core coolability in a LOCA can be maintained. (author)

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)

Current position on severe accident phenomena

Energy Technology Data Exchange (ETDEWEB)

Henry, Robert E [Fauske and Associates, Inc., Burr Ridge, IL (United States)

2004-07-01

The phenomena addressed in this lecture are: in-vessel and ex-vessel hydrogen generation; in-vessel and in-containment natural circulation, steam explosions, direct containment heating, core-concrete interaction; debris coolability, containment strength/failure. The following events were modeled: axial and radial power distribution, two-phase level in the core, steam generation in covered section, decay heat generation, convection to gas, cladding oxidation, cold ballooning and rupture, natural circulation between the core and upper plenum, hydrogen generation, core meltdown, reflooding. Differences between PWR and BWR type reactors.

Current position on severe accident phenomena

International Nuclear Information System (INIS)

Henry, Robert E.

2004-01-01

The phenomena addressed in this lecture are: in-vessel and ex-vessel hydrogen generation; in-vessel and in-containment natural circulation, steam explosions, direct containment heating, core-concrete interaction; debris coolability, containment strength/failure. The following events were modeled: axial and radial power distribution, two-phase level in the core, steam generation in covered section, decay heat generation, convection to gas, cladding oxidation, cold ballooning and rupture, natural circulation between the core and upper plenum, hydrogen generation, core meltdown, reflooding. Differences between PWR and BWR type reactors

PIRAMID 1: a prime European CEC experiment in BR2 at Mol, Belgium

International Nuclear Information System (INIS)

Joly, C.; Simoni, O.

1987-01-01

In the event of a core disruptive accident in a Liquid Metal Fast Breeder Reactor (LMFBR), molten core materials interacting with liquid sodium may form debris beds to settle on the retention structures or on the reactor vessel. The decay heat of retained fission products can induce high temperatures or high thermal loads on the retention structures or the reactor vessel with consequent fission product release. To assess the long-term coolability of core debris beds the Commission of the European Communities (CEC) decided to coordinate and fund the European PAHR (Post-Accident Heat Removal) program. The first work carried out for the program resulted in PIRAMID 1 (Pahr IRradiation According to a Mol Integrated Device), a unique irradiation device, designed, constructed and tested at Mol

Thermal Load Analysis of Multilayered Corium in the Lower Head of Reactor Pressure Vessel during Severe Accident

Energy Technology Data Exchange (ETDEWEB)

Whang, Seok Won; Park, Hyun Sun [POSTECH, Pohang (Korea, Republic of); Hwang, Tae Suk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2014-05-15

In-Vessel Retention (IVR) is one of the severe accident management strategies to terminate or mitigate the severe accident which is also called 'core-melt accident'. The reactor vessel would be cooled by flooding the cavity with water. The molten core mixture is divided into two or three layers due to the density difference. Light metal layer which contains Fe and Zr is on the oxide layer which is consist of UO{sub 2} and ZrO{sub 2}. Heavy metal layer which contains U, Fe and Zr is located under the oxide layer. In oxide layer, the crust which is solidified material is formed along the boundary. The assessment of IVR for nuclear power plant has been conducted with lumped parameter method by Theofanous, Rempe and Esmaili. In this paper, the numerical analysis was performed and verified with the Esmaili's work to analyze thermal load of multilayered corium in pressurized reactor vessel and also to examine the condition of in-vessel corium characteristic before the vessel failure that lead to ex-vessel severe accident progression for example, ex-vessel debris bed cooling. The in-vessel coolability analysis for several scenarios is conducted for the plant which has higher power than AP1000. Two sensitivity analyses are conducted, the first is emissivity of light metal layer and the second is the heat transfer coefficient correlations of oxide layer. The effect of three layered system also investigated. In this paper, the numerical analysis was performed and verified with Esmaili's model to analyze thermal load of multilayered corium in pressurized reactor vessel. For two layered system, thermal load was analyzed according to the severe accident scenarios, emissivity of the light metal layer and heat transfer correlations of the.

On the air coolability of TRIGA reactors following a loss-of-coolant accident

International Nuclear Information System (INIS)

El-Genk, Mohamed S.; Kim, Sung-Ho; Zaki, Galal M.; Foushee, Fabian; Philbin, Jeffrey S.; Schulze, James

1986-01-01

This paper describes the experiments on the air-coolability of a heated rod in a vertical open annulus at near atmospheric pressure. This data can be applied to the coolability of reactor fuel rods that are totally uncovered in a Loss-of-Coolant Accident (LOCA). As a prelude to measuring air coolability of specific core geometries (bundles), heat transfer data was collected for natural convection of atmospheric air in open vertical annuli with an isoflux inner wall and an insulated outer wall (diameter ratios, annulus ratio, of 1.155, 1.33, 1.63, and 12). Although the inner heated tube had the same overall dimensions as the fuel rod in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories (3.81 cm o.d. and 55.5 cm long), the heated length was only 36.0 cm rather than the entire 50.5 cm for the ACRR's rods. The test assembly was operated at heat fluxes up to 1.38 W/cm 2 with a corresponding surface temperature of 852 K. The annulus data was extrapolated to an equilibrium surface temperature of 1200 K (as a coolability limit of TRIGA reactors) to provide a qualitative estimate of the coolability of multirod bundles by free convection of atmospheric air. The results suggest that for a typical pitch-to-diameter ratio of 1.12 in the ACRR the decay heat removal level is about 1.0 kW/m. This corresponds to an initial decay power following sustained operations at about 12.5 kW/m in the ACRR. However, because of the uncertainties in duplicating the actual thermal-hydraulic conditions in a multirod bundle using a single rod annulus, the actual coolability of open pool reactors could be different from those suggested in this paper. (author)

Suggestion of typical phases of in-vessel fuel-debris by thermodynamic calculation for decommissioning technology of Fukushima-Daiichi nuclear power station

Energy Technology Data Exchange (ETDEWEB)

Ikeuchi, Hirotomo; Yano, Kimihiko; Kaji, Naoya; Washiya, Tadahiro [Japan Atomic Energy Agency, 4-33 Muramatsu, Tokai-mura, Ibaraki-ken, 319-1194 (Japan); Kondo, Yoshikazu; Noguchi, Yoshikazu [PESCO Co.Ltd. (Korea, Republic of)

2013-07-01

For the decommissioning of the Fukushima-Daiichi Nuclear Power Station (1F), the characterization of fuel-debris in cores of Units 1-3 is necessary. In this study, typical phases of the in-vessel fuel-debris were estimated using a thermodynamic equilibrium (TDE) calculation. The FactSage program and NUCLEA database were applied to estimate the phase equilibria of debris. It was confirmed that the TDE calculation using the database can reproduce the phase separation behavior of debris observed in the Three Mile Island accident. In the TDE calculation of 1F, the oxygen potential [G(O{sub 2})] was assumed to be a variable. At low G(O{sub 2}) where metallic zirconium remains, (U,Zr)O{sub 2}, UO{sub 2}, and ZrO{sub 2} were found as oxides, and oxygen-dispersed Zr, Fe{sub 2}(Zr,U), and Fe{sub 3}UZr{sub 2} were found as metals. With an increase in zirconium oxidation, the mass of those metals, especially Fe{sub 3}UZr{sub 2}, decreased, but the other phases of metals hardly changed qualitatively. Consequently, (U,Zr)O{sub 2} is suggested as a typical phase of oxide, and Fe{sub 2}(Zr,U) is suggested as that of metal. However, a more detailed estimation is necessary to consider the distribution of Fe in the reactor pressure vessel through core-melt progression. (authors)

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.

Molten core debris-sodium interactions: M-Series experiments

International Nuclear Information System (INIS)

Sowa, E.S.; Gabor, J.D.; Pavlik, J.R.; Cassulo, J.C.; Cook, C.J.; Baker, L. Jr.

1979-01-01

Five new kilogram-scale experiments have been carried out. Four of the experiments simulated the situation where molten core debris flows from a breached reactor vessel into a dry reactor cavity and is followed by a flow of sodium (Ex-vessel case) and one experiment simulated the flow of core debris into an existing pool of sodium (In-vessel case). The core debris was closely simulated by a thermite reaction which produced a molten mixture of UO 2 , ZrO 2 , and stainless steel. There was efficient fragmentation of the debris in all experiments with no explosive interactions observed

Fuel-coolant interaction (FCI) phenomena in reactor safety. Current understanding and future research needs

Energy Technology Data Exchange (ETDEWEB)

Speis, T.P. [Maryland Univ., College Park, MD (United States); Basu, S.

1998-01-01

This paper gives an account of the current understanding of fuel-coolant interaction (FCI) phenomena in the context of reactor safety. With increased emphasis on accident management and with emerging in-vessel core melt retention strategies for advanced light water reactor (ALWR) designs, recent interest in FCI has broadened to include an evaluation of potential threats to the integrity of reactor vessel lower head and ex-vessel structural support, as well as the role of FCI in debris quenching and coolability. The current understanding of FCI with regard to these issues is discussed, and future research needs to address the issues from a risk perspective are identified. (author)

Experimental modelling of core debris dispersion from the vault under a PWR pressure vessel: Part 1

International Nuclear Information System (INIS)

Macbeth, R.V.; Trenberth, R.

1987-12-01

Modelling experiments have been done on a 1/25 scale model in Perspex of the vault under a PWR pressure vessel. Various liquids have been used to simulate molten core debris assumed to have fallen on to the vault floor from a breach at the bottom of the pressure vessel. High pressure air and helium have been used to simulate the discharge of steam and gas from the breach. The dispersion of liquid via the vault access shafts has been measured. Photographs have been taken of fluid flow patterns and velocity profiles have been obtained. The requirements for further experiments are indicated. (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

Second OECD (NEA) CSNI specialist meeting on molten core debris-concrete interactions

International Nuclear Information System (INIS)

Alsmeyer, H.

1992-11-01

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

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

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)

The TMI-2 core relocation: Heat transfer and mechanism

International Nuclear Information System (INIS)

Epstein, M.; Fauske, H.K.

1987-07-01

It is postulated that the collapse of the upper debris bed was the main cause of core failure and core material relocation during the TMI-2 accident. It is shown that this mechanism of core relocation can account for the timescale(s) and energy transfer rate inferred from plant instrumentation. Additional analysis suggests that the water in the lower half of the reactor vessel was subcooled at the onset of relocation, as subcooling serves to explain the final coolable configuration at the bottom of the TMI vessel

Sensitivity analysis for maximum heat removal from debris in the lower head

International Nuclear Information System (INIS)

Kim, Yong Hoon; Suh, Kune Y.

2000-01-01

Sensitivity analyses were performed to determine the maximum heat removal capability from the debris and the reactor pressure vessel (RPV) wall through the gap that may be formed during a core melt relocation accident. Cases studied included four different nuclear power plant (TMI-2,KORI-2,YGN 3and4 and KNGR) per the thermal opower output. Results of the analysis show that the heat removal through gap cooling relative to flooding is efficacious as much as about 40% of the core material accumulated in the lower plenum in case of the TMI-2 reactor. In excess of 40%, however, the gap cooling alone was found not to be enough for heat removal from the core debris. There being uncertaainties aoboout the assumptions made in the present study,the analyses yield consistent results. If different cooling effects are considered, heat removal may be greatly enhanced. The LAVA experiements were performed at the Korea Atomic Energy Research Institute (KAERI) using al 2 O 3 /Fe thermite melt relocating down to the scaled vessel of a reactor lower head filled with preheated water. Test results indicated a cooling effect of water ingression through the debris-to-vessel gap and the intra-debris pores and crevices. If the cooling capacity of the intra-debris pores and crevices is comparable to debris-to-vessel heat removal capability, heat removal from the debris will be greatly augmented than heat removal by the gap cooling alone. The three nuclear reactor (KORI-2, YGN 3and4 and KNGR) calculation results for heat removal through the debris-to-vessel gap size of about 1mm were compared with the TMI-2 reactor calculation results for the case of gap cooling alone. (author)

Experimental modelling of core debris dispersion from the vault under a PWR pressure vessel. Pt. 2

International Nuclear Information System (INIS)

Rose, P.W.

1987-12-01

In previous experiments, done on a 1/25 scale model in Perspex of the vault under a PWR pressure vessel, the instrument tubes support structure built into the vault was not included. It consists of a number of grids made up of fairly massive steel girders. These have now been added to the model and experiments performed using water to simulate molten core debris assumed to have fallen on to the vault floor and high-pressure air to simulate the discharge of steam or gas from the assumed breach at the bottom of the pressure vessel. The results show that the tubes support structure considerably reduces the carry-over of liquid via the vault access shafts. (author)

Effect of a blockage length on the coolability during reflood in a 2 × 2 rod bundle with a 90% partially blocked region

Energy Technology Data Exchange (ETDEWEB)

Kim, Kihwan, E-mail: kihwankim@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of); Kim, Byung-Jae, E-mail: byoungjae@kaeri.re.kr [School of Mechanical Engineering, Chungnam National University, 99 Daehak-ro, Yuseoung-Gu, Daejeon 34134 (Korea, Republic of); Choi, Hae-Seob, E-mail: hschoi@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of); Moon, Sang-Ki, E-mail: skmoon@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of); Song, Chul-Hwa, E-mail: chsong@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of)

2017-02-15

Highlights: • This test was conducted to understand the effect of blockage length on the coolability. • Reflood tests were conducted with blockage simulators for various reflood rates. • The coolability in the downstream of the blockage region is significantly enhanced. - Abstract: If fuel rods are ballooned or rearranged during the reflood phase of a large break loss-of-coolant accident (LBLOCA) in a pressurized-water reactor (PWR), the transient heat transfer behavior is entirely different with those of the intact fuel rods owing to the deformed blockage region. The coolability in the blocked region depends on a complex two-phase heat transfer with various thermal hydraulic conditions. In addition, the blockage characteristics, such as the blockage ratio, length, shape, and configurations, are also significant factors affecting the coolability. In the present study, reflood experiments were carried out to understand the effect of the blockage length upon the coolability by varying the reflooding rates. The experiments were performed in electrically heated 2 × 2 rod bundles with blockage simulators having the same blockage ratio but different blockage lengths. The characteristics of quenching and heat transfer were evaluated to investigate the influence of the blockage region on the coolability. The droplet behaviors were also observed by measuring the droplets velocity and size near the blockage region. The coolability in the downstream region of the blockage was significantly enhanced, owing to the reduced flow area of the sub-channel, intensification of turbulence, and the entrained droplets in the blockage region.

Experimental studies on the coolability of packed beds. Flooding of hot dry packed beds

International Nuclear Information System (INIS)

Leininger, S.; Kulenovic, R.; Laurien, E.

2013-01-01

In case of a severe accident in a nuclear power plant meltdown of the reactor core can occur and form a packed bed in the lower plenum of the reactor pressure vessel (RPV) after solidification due to contact with water. The removal of after-heat and the long-term coolability is of essential interest. The efficient injection of cooling water into the packed bed has to be assured without endangering the structural integrity of the reactor pressure vessel. The experiments performed aimed to study the dry-out and the quenching (flooding) of hot dry packed beds. Two different inflow variants, bottom- and top-flooding including the variation of the starting temperature of the packed bed and the injection rate were studied. In case of bottom flooding the quenching time increases with increasing packed bed temperature and decreasing injection rate. In case of top flooding the flow pattern is more complex, in a first phase the water flows preferentially toward the RPV wall, the flow paths conduct the water downwards. The flow resistance of the packed bed increases with increasing bed temperatures. The quenching temperatures increase significantly above average.

Assessment of reactor vessel integrity (ARVI)

Energy Technology Data Exchange (ETDEWEB)

Sehgal, B.R. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden)]. E-mail: sehgal@ne.kth.se; Karbojian, A. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden); Giri, A. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden); Kymaelaeinen, O. [FortumEngNP (Finland); Bonnet, J.M. [CEA (France); Ikkonen, K. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden); Sairanen, R. [VTT (Finland); Bhandari, S. [FRAMATOME (France); Buerger, M. [USTUTT (Germany); Dienstbier, J. [NRI Rez (Czech Republic); Techy, Z. [VEIKI (Hungary); Theofanous, T. [UCSB (United States)

2005-02-01

The assessment of reactor vessel integrity (ARVI) project involved a total of nine organizations from Europe and USA. The work consisted of experiments and analysis development. The modeling activities in the area of structural analyses were focused on the support of EC-FOREVER experiments as well as on the exploitation of the data obtained from those experiments for modeling of creep deformation and the validation of the industry structural codes. Work was also performed for extension of melt natural convection analyses to consideration of stratification, and mixing (in the CFD codes). Other modeling activities were for (1) gap cooling CHF and (2) developing simple models for system code. Finally, the methodology and data was applied for the design of IVMR severe accident management scheme for VVER-440/213 plants. The work was broken up into five packages. They were divided into tasks, which were performed by different partners. The major experimental project continued was EC-FOREVER in which data was obtained on in-vessel melt pool coolability. In previous EC-FOREVER experiments data was obtained on melt pool natural convection and lower head creep failure and rupture. Those results obtained were related to the following issues: (1) multiaxial creep laws for different vessel steels (2) effects of penetrations, and (3) mode and location of lower head failure. The two EC-FOREVER tests reported here are related to (a) the effectiveness of gap cooling and (b) water ingression for in vessel melt coolability. Two other experimental projects were also conducted. One was the COPO experiments, which was concerned with the effects of stratification and metal layer on the thermal loads on the lower head wall during melt pool convection. The second experimental project was conducted at ULPU facility, which provided data and correlations of CHF due to the external cooling of the lower head.

Evaluation of molten lead mixing in sodium coolant by diffusion for application to PAHR

International Nuclear Information System (INIS)

Chawla, T.C.; Pedersen, D.R.; Leaf, G.; Minkowycz, W.J.

1983-01-01

In post-accident heat removal (PAHR) applications the use of a lead slab is being considered for protecting a porous bed of steel shots in ex-vessel cavity from direct impingement of molten steel or fuel upon vessel failure following a hypothetical core dissembly accident in an LMFBR. The porous bed is provided to increase coolability of the fuel debris by the sodium coolant. The objectives of the present study are (1) to determine melting rates of lead slabs of various thicknesses in contact with sodium coolant and (2) to evaluate the extent of penetration and mixing rates of molten lead into sodium coolant by molecular diffusion alone

Analysis for the coolability of the reactor cavity in a Korean 1000 MWe PWR using MELCOR 1.8.3 computer code

International Nuclear Information System (INIS)

Lee, Byung Chul; Kim, Ju Yeul; Chung, Chang Hyun; Park, Soo Yong

1996-01-01

The analysis for the coolability of the reactor cavity in typical Korean 1000 MWe Nuclear Unit under severe accidents is performed using MELCOR 1.8.3 code. The key parameters molten core-concrete interaction (MCCI) such as melt temperature, concrete ablation history and gas generation are investigated. Total twenty cases are selected according to ejected debris fraction and coolant mass. The ablation rate of concrete decreases as mass of the melt decreases and coolant mass increases. Heat loss from molten pool to coolant is comparable to total decay heat, so concrete ablation is delayed until water is absent and crust begins to remove. Also, overpressurization due to non-condensible gases generated during corium and concrete interacts can cause to additional risk of containment failure. It is concluded that flooded reactor cavity condition is very important to minimize the cavity ablation and pressure load by non-condensible gases on containment

Proceedings of the Second OECD (NEA) CSNI Specialist Meeting on Molten Core Debris-Concrete Interactions

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-07-01

The Second CSNI Specialist Meeting on Molten Core Debris-Concrete Interactions was held at Kernforschungszentrum Karlsruhe, Germany on April 1-3, 1992. The status and progress in this field of severe reactor accidents were discussed from researchers around the world including participants from Russia and the Czech and Slovak Federal Republic. The contributions concentrated on two main topics. The first topic is the 'classical' core debris-concrete interaction, both experimental and theoretical. Integral effects and separate effects were addressed in thermal hydraulics and heat transfer, material interaction, and aerosol release during concrete erosion, with some applications to prototypical nuclear power plants. The second topic gaining more and more interest is the possibility of controlling and ending the erosion of the concrete by spreading of the core melt, and/or achieving coolability by the addition of water. In the final session it was concluded that considerable progress has been made in understanding and modelling the important phenomena. For the first topic a broad and generally sufficient experimental data base is existing, allowing further improvement qualification of the theoretical models which at present give reasonable agreement with the most important experimental data. A validation matrix is recommended for final validation of the codes. With respect to fission product release during MCCI measurements show that the releases are significantly less than previously estimated. The relatively new topic of melt coolability deserves further investigations which are already underway at different places or international coordinated efforts.

Proceedings of the Second OECD (NEA) CSNI Specialist Meeting on Molten Core Debris-Concrete Interactions

International Nuclear Information System (INIS)

1992-01-01

The Second CSNI Specialist Meeting on Molten Core Debris-Concrete Interactions was held at Kernforschungszentrum Karlsruhe, Germany on April 1-3, 1992. The status and progress in this field of severe reactor accidents were discussed from researchers around the world including participants from Russia and the Czech and Slovak Federal Republic. The contributions concentrated on two main topics. The first topic is the 'classical' core debris-concrete interaction, both experimental and theoretical. Integral effects and separate effects were addressed in thermal hydraulics and heat transfer, material interaction, and aerosol release during concrete erosion, with some applications to prototypical nuclear power plants. The second topic gaining more and more interest is the possibility of controlling and ending the erosion of the concrete by spreading of the core melt, and/or achieving coolability by the addition of water. In the final session it was concluded that considerable progress has been made in understanding and modelling the important phenomena. For the first topic a broad and generally sufficient experimental data base is existing, allowing further improvement qualification of the theoretical models which at present give reasonable agreement with the most important experimental data. A validation matrix is recommended for final validation of the codes. With respect to fission product release during MCCI measurements show that the releases are significantly less than previously estimated. The relatively new topic of melt coolability deserves further investigations which are already underway at different places or international coordinated efforts

FARO tests corium-melt cooling in water pool: Roles of melt superheat and sintering in sediment

Energy Technology Data Exchange (ETDEWEB)

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

2016-08-15

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

In-vessel coolability and retention of a core melt. Volume 1

Energy Technology Data Exchange (ETDEWEB)

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

1996-10-01

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

In-vessel coolability and retention of a core melt. Volume 2

International Nuclear Information System (INIS)

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

1996-10-01

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

In-vessel coolability and retention of a core melt. Volume 1

International Nuclear Information System (INIS)

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

1996-10-01

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

Experimental investigation on single-phase pressure losses in nuclear debris beds: Identification of flow regimes and effective diameter

Energy Technology Data Exchange (ETDEWEB)

Clavier, R., E-mail: remi.clavier@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN) – PSN-RES/SEREX/LE2M, Cadarache bât. 327, 13115 St Paul-lez-Durance (France); Chikhi, N., E-mail: nourdine.chikhi@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN) – PSN-RES/SEREX/LE2M, Cadarache bât. 327, 13115 St Paul-lez-Durance (France); Fichot, F. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN) – PSN-RES/SAG/LEPC, Cadarache bât. 700, 13115 St Paul-lez-Durance (France); Quintard, M. [Université de Toulouse – INPT – UPS – Institut de Mécanique des Fluides de Toulouse (IMFT), Allée Camille Soula, F-31400 Toulouse (France); CNRS – IMFT, F-31400 Toulouse (France)

2015-10-15

Highlights: • Single-phase pressure drops versus flow rates in particle beds are measured. • Conditions are representative of the reflooding of a nuclear fuel debris bed. • Darcy, weak inertial, strong inertial and weak turbulent regimes are observed. • A Darcy–Forchheimer law is found to be a good approximation in this domain. • A predictive correlation is derived from new experimental data. - Abstract: During a severe nuclear power plant accident, the degradation of the reactor core can lead to the formation of debris beds. The main accident management procedure consists in injecting water inside the reactor vessel. Nevertheless, large uncertainties remain regarding the coolability of such debris beds. Motivated by the reduction of these uncertainties, experiments have been conducted on the CALIDE facility in order to investigate single-phase pressure losses in representative debris beds. In this paper, these results are presented and analyzed in order to identify a simple single-phase flow pressure loss correlation for debris-bed-like particle beds in reflooding conditions, which cover Darcean to Weakly Turbulent flow regimes. The first part of this work is dedicated to study macro-scale pressure losses generated by debris-bed-like particle beds, i.e., high sphericity (>80%) particle beds with relatively small size dispersion (from 1 mm to 10 mm). A Darcy–Forchheimer law, involving the sum of a linear term and a quadratic deviation, with respect to filtration velocity, has been found to be relevant to describe this behavior in Darcy, Strong Inertial and Weak Turbulent regimes. It has also been observed that, in a restricted domain (Re = 15 to Re = 30) between Darcy and Weak Inertial regimes, deviation is better described by a cubic term, which corresponds to the so-called Weak Inertial regime. The second part of this work aims at identifying expressions for coefficients of linear and quadratic terms in Darcy–Forchheimer law, in order to obtain a

Review of the Technical Status on the Debris Bed Cooling Model

Energy Technology Data Exchange (ETDEWEB)

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

2007-09-15

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

Review of the Technical Status on the Debris Bed Cooling Model

International Nuclear Information System (INIS)

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

2007-09-01

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

Fission-product releases from a PHWR terminal debris bed

Energy Technology Data Exchange (ETDEWEB)

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

2016-06-15

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

Analysis of heat transfer mechanism on in-vessel corium coolability in severe accidents

International Nuclear Information System (INIS)

Park, Rae Joon; Jeong, Ji Whan; Kim, Sang Baik; Kang, Kyung Ho; Kim, Jong Whan

1998-04-01

When the molten core material relocates to the lower plenum of the reactor vessel, the cooling process of corium and the related heat transfer mechanism have been analyzed. The critical heat flux in gap (CHFG) test is being performed as a part of simulation of naturally arrested thermal attack in (SONATA-IV) project and the state of art on CHF has been reviewed. A series of complex heat transfer mechanism of molten pool formation, natural convection in the molten pool, solidification and remelting of the corium, conduction in the solidified crust, and boiling heat transfer to surroundings can be occurred in the lower plenum. Many studies are needed to investigate the complex heat transfer mechanism in the lower plenum, because these phenomena have not been clearly understand until now. The SONATA-IV/CHFG experiments are being carried out to develop CHF correlation in a hemispherical gap, which is the upper limit of heat transfer. There is no experimental or analytical CHF correlation applicable to a hemispherical gap. So lots of analytical and experimental correlations developed using the similar experimental condition were gathered and compared with each other. According to the experimental work that was carried out with pool boiling condition, CHF in a parallel gap was reduced by 1/30 compared with the value measured without gap. A basic form of a CHF correlation has been developed to correlate measurements that will be made in the SONATA-IV/CHFG experiments. That correlation is based on the fact that the CHF in a hemispherical gap is enhanced by CCFL and a Kutateladze type CCFL correlation develops CCFL date will in geometry like this. The experimental facility consists of a heater, a pressure vessel, a heat exchanger and lots of sensors. The heater capacity is 40 kw and the maximum heat flux at the surface is 100 kw/m 2 . The experiments will be carried out in the range of 1 to 10 atm and the gap size of 0.5, 1, 2 mm. The CHF will be detected using 66 type

In-vessel coolability and retention of a core melt

International Nuclear Information System (INIS)

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

1997-01-01

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

Integral analysis of debris material and heat transport in reactor vessel lower plenum

International Nuclear Information System (INIS)

Suh, K.Y.; Henry, R.E.

1994-01-01

An integral, fast-running, two-region model has been developed to characterize the debris material and heat transport in the reactor lower plenum under severe accident conditions. The debris bed is segregated into the oxidic pool and an overlying metallic layer. Debris crusts can develop on three surfaces: the top of the molten pool, the RPV wall, and the internal structures. To account for the decay heat generation, the crust temperature profile is assumed to be parabolic. The oxidic debris pool is homogeneously mixed and has the same material composition, and hence the same thermophysical properties, as the crusts, while the metallic constituents are assumed to rise to the top of the debris pool. Steady-state relationships are used to describe the heat transfer rates, with the assessment of solid or liquid state, and the liquid superheat in the pool being based on the average debris temperature. Natural convection heat transfer from the molten debris pool to the upper, lower and embedded crusts is calculated based on the pool Rayleigh number with the conduction heat transfer from the crusts being determined by the crust temperature profile. The downward heat flux is transferred to the lowest part of the RPV lower head through a crust-to-RPV contact resistance. The sideward heat flux is transferred to the upper regions of the RPV lower head as well as to the internal structures. The upward heat flux goes to the metal layer, water, or available heat sink structures above. Quenching due to water ingression is modeled separately from the energy transfer through the crust. The RPV wall temperature distribution and the primary system pressure are utilized to estimate challenges to the RPV integrity. ((orig.))

A probabilistic approach for debris impact risk with numerical simulations of debris behaviors

International Nuclear Information System (INIS)

Kihara, Naoto; Matsuyama, Masafumi; Fujii, Naoki

2013-01-01

We propose a probabilistic approach for evaluating the impact risk of tsunami debris through Monte Carlo simulations with a combined system comprising a depth-averaged two-dimensional shallow water model and a discrete element model customized to simulate the motions of floating objects such as vessels. In the proposed method, first, probabilistic tsunami hazard analysis is carried out, and the exceedance probability of tsunami height and numerous tsunami time series for various hazard levels on the offshore side of a target site are estimated. Second, a characteristic tsunami time series for each hazard level is created by cluster analysis. Third, using the Monte Carlo simulation model the debris impact probability with the buildings of interest and the exceedance probability of debris impact speed are evaluated. (author)

Crust formation and its effect on the molten pool coolability

Energy Technology Data Exchange (ETDEWEB)

Park, R.J.; Lee, S.J.; Sim, S.K. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-09-01

Experimental and analytical studies of the crust formation and its effect on the molten pool coolability have been performed to examine the crust formation process as a function of boundary temperatures as well as to investigate heat transfer characteristics between molten pool and overlying water in order to evaluate coolability of the molten pool. The experimental test results have shown that the surface temperature of the bottom plate is a dominant parameter in the crust formation process of the molten pool. It is also found that the crust thickness of the case with direct coolant injection into the molten pool is greater than that of the case with a heat exchanger. Increasing mass flow rate of direct coolant injection to the molten pool does not affect the temperature of molten pool after the crust has been formed in the molten pool because the crust behaves as a thermal barrier. The Nusselt number between the molten pool and the coolant of the case with no crust formation is greater than that of the case with crust formation. The results of FLOW-3D analyses have shown that the temperature distribution contributes to the crust formation process due to Rayleigh-Benard natural convection flow.

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-05-23

reactor material database for dry cavity conditions is solely one-dimensional. Although the MACE Scoping Test was carried out with a two-dimensional concrete cavity, the interaction was flooded soon after ablation was initiated to investigate debris coolability. Moreover, due to the scoping nature of this test, the apparatus was minimally instrumented and therefore the results are of limited value from the code validation viewpoint. Aside from the MACE program, the COTELS test series also investigated 2-D CCI under flooded cavity conditions. However, the input power density for these tests was quite high relative to the prototypic case. Finally, the BETA test series provided valuable data on 2-D core concrete interaction under dry cavity conditions, but these tests focused on investigating the interaction of the metallic (steel) phase with concrete. Due to these limitations, there is significant uncertainty in the partition of energy dissipated for the ablation of concrete in the lateral and axial directions under dry cavity conditions for the case of a core oxide melt. Accurate knowledge of this 'power split' is important in the evaluation of the consequences of an ex-vessel severe accident; e.g., lateral erosion can undermine containment structures, while axial erosion can penetrate the basemat, leading to ground contamination and/or possible containment bypass. As a result of this uncertainty, there are still substantial differences among computer codes in the prediction of 2-D cavity erosion behavior under both wet and dry cavity conditions. In light of the above issues, the OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) program was initiated at Argonne National Laboratory. The project conducted reactor materials experiments and associated analysis to achieve the following technical objectives: (1) resolve the ex-vessel debris coolability issue through a program that focused on providing both confirmatory evidence and test data for the

Improved Design of PECS to reduce Flow Instability for EU-APR1400

International Nuclear Information System (INIS)

Hwang, Do Hyun; Lee, Keunsung

2014-01-01

For EU-APR1400, PECS (Passive Ex-vessel corium retaining and Cooling System), so-called core catcher, was adopted to keep the integrity of basemat in containment by preventing MCCI (Molten Core Concrete Interaction) through retaining core debris and cooling corium outside the reactor vessel. In this paper, the improved design of PECS is presented to increase coolability by reducing flow instability in the region of cooling channel. In this paper, flow instability analysis was carried out using CFD code to find out the most improved design of PECS, which is to increase coolability by reducing bubble entrainment in the region of cooling channel. The reduction of bubble entrainment in the downcomer facilitates higher mass flow rates in the downcomer. Among presented four designed for the downcomer of PECS, the superstep design shows the highest mass flow rate and the lowest gas holdup in the downcomer as well as in the cooling channel. Compared with the existing design, the elimination of the horizontal part and the addition of an extra space above the vertical entrance to the downcomer seem to help the separation of the vapor

Melt coolability modeling and comparison to MACE test results

International Nuclear Information System (INIS)

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

1992-01-01

An important question in the assessment of severe accidents in light water nuclear reactors is the ability of water to quench a molten corium-concrete interaction and thereby terminate the accident progression. As part of the Melt Attack and Coolability Experiment (MACE) Program, phenomenological models of the corium quenching process are under development. The modeling approach considers both bulk cooldown and crust-limited heat transfer regimes, as well as criteria for the pool thermal hydraulic conditions which separate the two regimes. The model is then compared with results of the MACE experiments

Experimental study on in-vessel debris coolability; experiments on heat transfer in downward-facing hemicircular gap

Energy Technology Data Exchange (ETDEWEB)

Suh, Kune Yull; Kim, Yong Hoon; Kim, Seong Joong; Lee, Seung Dong [Seoul National University, Seoul (Korea)

2002-03-01

Experiments were performed to measure the CHF and the critical power and to investigate the heat transfer mechanism in the narrow gap with 2D slice test sections. Test parameters in this study included the gap size of 1, 2, 5, and 10mm and the system pressure from 0.1 to 1.0 MPa. The CHF was measured for the distilled water. Results of this study may be summarized as follows. 1) In the narrow gap size of 1 and 2mm, the CHF occurs at the bottom and propagates upwards as the inclination angle relative to gravitational force increases. 2) Dryout is the limiting heat transfer mechanism in the 2D sliced experiments, and the CHF reaches 80-90% of the 3D CHFG value. 3) In the narrow gap size of 1 and 2mm the CHF values lie on the order of 104 kW/m2, while in the gap sizes of 5 and 10mm the CHF values have an order of 105 kW/m2. 4) The flow visualization study revealed that the shape of bubble is elliptic and the hydrodynamic bubble diameter exceeds the size of 20mm. 5) In analyzing the CHF with the inclination angle of the heated surface in the narrow gap size of 1 and 2mm, there exists a transition angle in the vicinity of 20-30 .deg.. From the region of the inclination angle of 0 .deg. C to 20 .deg. C or 30 .deg. C, the dryout mechanism may be directly applicable. On the contrary the transition angle is not conspicuously seen in the gap sizes of 5 and 10mm in which dominant mechanism is closer to pool boiling. 27 refs., 29 figs., 8 tabs. (Author)

Study on coolability of melt pool with different strategies

International Nuclear Information System (INIS)

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

2014-01-01

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

Development of in-vessel type control rod drive mechanism for a innovative small reactor (Contract research)

Energy Technology Data Exchange (ETDEWEB)

Yoritsune, Tsutomu; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-03-01

Although the control rod drive mechanism of an existing large scale light water reactor is generally installed outside the reactor vessel, an in-vessel type control rod drive mechanism (INV-CRDM) is installed inside the reactor vessel. The INV-CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. Therefore, INV-CRDM is an important technology adopted in an innovative small reactor. Japan Atomic Energy Research Institute (JAERI) has developed this type of CRDM driven by an electric motor, which can work under high temperature and high pressure water for the advanced marine reactor. On the basis of this research result, a driving motor coil and a bearing were developed to be used under the high temperature steam, severe condition for an innovative small reactor. About the driving motor, we manufactured the driving motor available for high temperature steam and carried out performance test under room temperature atmosphere to confirm the electric characteristic and coolability of the driving coil. With these test results and the past test results under high temperature water, we analyzed and evaluated the electric performance and coolability of the driving coil under high temperature steam. Concerning bearing, we manufactured the test pieces using some candidate material for material characteristic test and carried out the rolling wear test under high temperature steam to select the material. Consequently, we confirmed that performance of the driving coil for the advanced type driving motor, is enough to be used under high temperature steam. And, we evaluated the performance of the bearing and selected the material of the bearing, which can be used under high temperature steam. From these results, we have obtained the prospect that the INV-CRDM can be used for an innovative small reactor under steam atmosphere could be developed. (author)

TMI defueling project fuel debris removal system

International Nuclear Information System (INIS)

Burdge, B.

1992-01-01

The three mile Island Unit 2 (TMI-2) pressurized water reactor loss-of-coolant accident on March 28, 1979, presented the nuclear community with many challenging remediation problems; most importantly, the removal of the fission products within the reactor containment vessel. To meet this removal problem, an air-lift system (ALS) can be used to employ compressed air to produce the motive force for transporting debris. Debris is separated from the transport stream by gravity separation. The entire method does not rely on any moving parts. Full-scale testing of the ALS at the Idaho National Engineering Laboratory (INEL) has demonstrated the capability of transporting fuel debris from beneath the LCSA into a standard fuel debris bucket at a minimum rate of 230 kg/min

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

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

Preliminary investigation for the development of surrogate debris from nuclear detonations in marine-urban environments

International Nuclear Information System (INIS)

Seybert, A.G.; Auxier II, J.D.; University of Tennessee, Knoxville, TN; Hall, H.L.; University of Tennessee, Knoxville, TN; University of Tennessee, Knoxville, TN

2017-01-01

Since no nuclear weapon surface detonations have occurred in urban harbor environments, the nuclear forensic community has no actual debris from which to develop and validate analytical methods for radiochemistry analysis, making the development of surrogate debris representative of this a marine-urban detonation a vital undertaking. This work seeks to build a robust model that accounts for natural and manmade environmental variations in harbor environments and vessel compositions to statistically define the elemental composition of vaporized debris from a marine-urban nuclear detonation. This initial work is necessary for follow-on neutron-activation and debris formation analysis. (author)

Preliminary results from initial in-pile debris bed experiments

International Nuclear Information System (INIS)

Rivard, J.B.

1977-01-01

An accident in a liquid metal fast breeder reactor (LMFBR) in which molten core material is suddenly quenched with subcooled liquid sodium could result in extensive fragmentation and dispersal of fuel as subcritical beds of frozen particulate debris within the reactor vessel. Since this debris will continue to generate power due to decay of retained fission products, containment of the debris is threatened if the generated heat is not removed. Therefore, the initial safety question is the capacity which debris beds may have for transfer of the decay heat to overlying liquid sodium by natural processes--i.e., without the aid of forced circulation of the coolant. Up to the present time, all experiments on debris bed behavior either have used substitute materials (e.g., sand and water) or have employed actual materials, but atypical heating methods. Increased confidence in the applicability of debris bed simulations is afforded if the heat is generated within the fuel component of the appropriate fast reactor materials. The initial series of in-pile tests reported on herein constitutes the first experiments in which the internal heating mode has been produced in particulate oxide fuel immersed in liquid sodium. Fission heating of the fully-enriched UO 2 in the experiment while it is contained within Sandia Laboratories Annular Core Pulse Reactor (ACPR), operating in its steady-state mode, approximates the decay heating of debris. Preliminary results are discussed

Marine debris surveys at Volunteer Beach, Falkland Islands, during the summer of 2001/02.

Science.gov (United States)

Otley, Helen; Ingham, Rebecca

2003-12-01

This survey evaluated the monthly accumulation rate of marine debris and the types of objects washed ashore at Volunteer Beach on East Falkland between October 2001 and March 2002. The mean (+/-SD) accumulation rate of marine debris was 77+/-25 items/km/month, of a mean weight of 17.3+/-12 kg. Forty different objects were collected and the five most frequent items were cotton fabric, string, polystyrene packing sheet, plastic packing tape and broken plastic pieces. The debris on Volunteer Beach was dominated by fishing debris; 42% of the items were discarded fishing equipment, while 39% of the items were of a packaging or associated nature. The mostly likely source of this household waste was fishing vessels, with Falkland Islands Government (FIG) fisheries observers seeing 27 of the 40 items of debris collected from Volunteer Beach being discarded from fishing vessels. It is suggested that, although further marine debris research is warranted, more effective at-sea ship waste disposal regulations are required in Falkland waters to reduce environmental and economic threats both at the local and international level.

Marine debris surveys at Volunteer Beach, Falkland Islands, during the summer of 2001/02

International Nuclear Information System (INIS)

Otley, Helen; Ingham, Rebecca

2003-01-01

This survey evaluated the monthly accumulation rate of marine debris and the types of objects washed ashore at Volunteer Beach on East Falkland between October 2001 and March 2002. The mean (±SD) accumulation rate of marine debris was 77 ± 25 items/km/month, of a mean weight of 17.3 ± 12 kg. Forty different objects were collected and the five most frequent items were cotton fabric, string, polystyrene packing sheet, plastic packing tape and broken plastic pieces. The debris on Volunteer Beach was dominated by fishing debris; 42% of the items were discarded fishing equipment, while 39% of the items were of a packaging or associated nature. The mostly likely source of this household waste was fishing vessels, with Falkland Islands Government (FIG) fisheries observers seeing 27 of the 40 items of debris collected from Volunteer Beach being discarded from fishing vessels. It is suggested that, although further marine debris research is warranted, more effective at-sea ship waste disposal regulations are required in Falkland waters to reduce environmental and economic threats both at the local and international level

Experimental study of self-leveling behavior in debris bed

International Nuclear Information System (INIS)

Zhang, Bin; Harada, Tetsushi; Hirahara, Daisuke; Matsumoto, Tatsuya; Morita, Koji; Fukuda, Kenji; Yamano, Hidemasa; Suzuki, Tohru; Tobita, Yoshiharu

2008-01-01

After a core disruptive accident in a sodium-cooled fast reactor, core debris may settle on locations such as within the core-support structure or in the lower inlet plenum of the reactor vessel as debris beds, as a consequence of rapid quenching and fragmentation of core materials in subcooled sodium. The particle beds that are initially of varying depth have been observed to undergo a process of self-leveling when sodium boiling occurs within the beds. The boiling is believed to provide the driven force with debris needed to overcome resisting forces. Self-leveling ability has much effect on heat-removal capability of debris beds. In the present study, characteristics of self-leveling behaviors were investigated experimentally with simulant materials. Although the decay heat from fuel debris drives the coolant boiling in reactor accident conditions, the present experiments employed depressurization boiling of water to simulate axially increasing void distribution in a debris bed, which consists of solid particles of alumina or lead with different density. The particle size (from 0.5 mm to 6 mm in diameter) and shape (spherical or non-spherical particles) were also taken as experimental parameters. A rough criteria for self-leveling occurrence is proposed and compared with the experimental results. Characteristics of the self-leveling behaviors observed are analyzed and extrapolate to reactor accident conditions. (author)

In-vessel core debris retention through external flooding of the reactor pressure vessel. State-of-the-art report

Energy Technology Data Exchange (ETDEWEB)

Heel, A.M.J.M. van

1995-07-01

An overview of the state-of-the-art knowledge on the ex-vessel flooding accident management strategy for severe accidents in a NPP has been given. The feasibility has been discussed, as well as the in- and ex-vessel phenomena, which influence the structural integrity of the vessel. Finally, some computer codes with the ability to model the phenomena involved in ex-vessel flooding have been discussed. (orig./HP).

In-vessel core debris retention through external flooding of the reactor pressure vessel. State-of-the-art report

International Nuclear Information System (INIS)

Heel, A.M.J.M. van.

1995-07-01

An overview of the state-of-the-art knowledge on the ex-vessel flooding accident management strategy for severe accidents in a NPP has been given. The feasibility has been discussed, as well as the in- and ex-vessel phenomena, which influence the structural integrity of the vessel. Finally, some computer codes with the ability to model the phenomena involved in ex-vessel flooding have been discussed. (orig./HP)

Performance experiments on the in-vessel core catcher during severe accidents

International Nuclear Information System (INIS)

Kang, Kyoung Ho; Park, Rae Joon; Cho, Young Rho; Kim, Sang Baik

2004-01-01

A US-Korean International Nuclear Energy Research Initiative (INERI) project has been initiated by the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korean Atomic Energy Research Institute (KAERI) to determine if IVR is feasible for high power reactors up to 1500 MWe by investigating the performance of enhanced ERVC and in-vessel core catcher. This program is initially focusing on the Korean Advanced Power Reactor 1400 MWe (APR1400) design. As for the enhancement of the coolability through the ERVC, boiling tests are conducted by using appropriate coating material on the vessel outer surface to promote downward facing boiling and selecting an improved vessel/insulation design to facilitate water flow and steam venting through the insulation in this program. Another approach for successful IVR are investigated by applying the in-vessel core catcher to provide an 'engineered gap' between the relocated core materials and the water-filled reactor vessel and a preliminary design for an in-vessel core catcher was developed during the first year of this program. Feasibility experiments using the LAVA facility, named LAVA-GAP experiments, are in progress to investigate the core catcher performance based on the conceptual design of the in-vessel core catcher proposed in this INERI project. The experiments were performed using 60kg of Al 2 O 3 thermite melt as a core material simulant with a 1/8 linear scale mock-up of the reactor vessel lower plenum. The hemispherical in-vessel core catcher was installed inside the lower head vessel maintaining a uniform gap of 10mm from the inner surface of the lower head vessel. Two types of the core catchers were used in these experiments. The first one was a single layered in-vessel core catcher without internal coating and the second one was a two layered in-vessel core catcher with an internal coating of 0.5mm-thick ZrO 2 via the plasma

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.

Experimental study and modelling of pressure losses during reflooding of a debris beds

International Nuclear Information System (INIS)

Clavier, Remi

2015-01-01

This work deals with single and two-phase flow pressure losses in porous media. The aim is to improve understanding and modeling of momentum transfer inside particle beds, in relation with nuclear safety issues concerning the reflooding of debris beds during severe nuclear accidents. Indeed, the degradation of the core during such accidents can lead to the collapse of the fuel assemblies, and to the formation of a debris bed, which can be described as a hot porous medium. This thesis is included in a nuclear safety research project on coolability of debris beds during reflooding sequences. An experimental study of single and two-phase cold-flow pressure losses in particle beds is proposed. The geometrical characteristics of the debris and the hydrodynamic conditions are representative of the real case, in terms of granulometry, particle shapes, and flow velocities. The new data constitute an important contribution. In particular, they contain pressure losses and void fraction measurements in two-phase air-water flows with non-zero liquid Reynolds numbers, which did not exist before. Predictive models for pressure losses in single and two-phase flow through particle beds have been established from experimental data. Their structures are based on macroscopic equations obtained from the volume averaging of local conservation equations. Single-phase flow pressure losses can be described by a Darcy-Forchheimer law with a quadratic correction, in terms of filtration velocity, with a better-than-10 % precision. Numerical study of single-phase flows through porous media shows that this correlation is valid for disordered smooth particle beds. Two-phase flow pressure losses are described using a generalized Darcy-Forchheimer structure, involving inertial and cross flow terms. A new model is proposed and compared to the experimental data and to the usual models used in severe accident simulation codes. (author)

Drone Use in Monioring Open Ocean Surface Debris, Including Paired Manta and Tucker Trawls for Relateing Sea State to Vertical Debris Distribution

Science.gov (United States)

Lattin, G.

2016-02-01

Monitoring debris at sea presents challenges not found in beach or riverine habitats, and is typically done with trawl nets of various apertures and mesh sizes, which limits the size of debris captured and the area surveyed. To partially overcome these limitations in monitoring floating debris, a Quadcopter drone with video transmitting and recording capabilities was deployed at the beginning and the end of manta trawl transects within the North Pacific Subtropical Gyre's eastern convergence zone. Subsurface tucker trawls at 10 meters were conducted at the same time as the manta trawls, in order to assess the effect of sea state on debris dispersal. Trawls were conducted on an 11 station grid used repeatedly since 1999. For drone observations, the operator and observer were stationed on the mother ship while two researchers collected observed debris using a rigid inflatable boat (RIB). The drone was flown to a distance of approximately 100 meters from the vessel in a zigzag or circular search pattern. Here we examine issues arising from drone deployment during the survey: 1) relation of area surveyed by drone to volume of water passing through trawl; 2) retrieval of drone-spotted and associated RIB spotted debris. 3) integrating post- flight image analysis into retrieved debris quantification; and 4) factors limiting drone effectiveness at sea. During the survey, debris too large for the manta trawl was spotted by the drone, and significant debris not observed using the drone was recovered by the RIB. The combination of drone sightings, RIB retrieval, and post flight image analysis leads to improved monitoring of debris at sea. We also examine the issue of the distribution of floating debris during sea states varying from 0-5 by comparing quantities from surface manta trawls to the tucker trawls at a nominal depth of 10 meters.

Coolability of degraded core under reflooding conditions in Nordic boiling water reactors

Energy Technology Data Exchange (ETDEWEB)

Lindholm, I; Pekkarinen, E [VTT Energy, Espoo (Finland); Nilsson, L [Studsvik EcoSafe AB, Nykoeping (Sweden); Sjoevall, H [Teollisuuden Voima Oy, Olkiluoto (Finland)

1995-09-01

Present work is part of the first phase of subproject RAK-2.1 of the new Nordic Co-operative Reactor Safety Program, NKS. The first phase comprises reflooding calculations for the boiling water reactors (BWRs) TVO I/II in Finland and Forsmark 3 in Sweden, as a continuation of earlier severe accident analyses which were made in the SIK-2 project. The objective of the core reflooding studies is to evaluate when and how the core is still coolable with water and what are the probable consequences of water cooling. In the following phase of the RAK-2.1 project, recriticality studies will be performed. Conditions for recriticality might occur if control rods have melted away with the fuel rods intact in a shape that critical conditions can be created in reflooding with insufficiently borated water. Core coolability was investigated for two reference plants, TVO I/II and Forsmark 3. The selected accident cases were anticipated station blackout with or without successful depressurization of reactor coolant system (RCS). The effects of the recovery of emergency core cooling (ECC) were studied by varying the starting time of core reflooding. The start of ECC systems were assigned to reaching a maximum cladding temperature: 1400 K, 1600 K, 1800 K and 2000 K in the core. Cases with coolant injection through the downcomer were studied for TVO I/II and both downcomer injection and core top spray were investigated for Forsmark 3. Calculations with three different computer codes: MAAP 4, MELCOR 1.8.3 and SCDA/RELAP5/MOD 3.1 for the basis for the presented reflooding studies. Presently, and experimental programme on core reflooding phenomena has been started in Kernforschungszentrum Karlsruhe in QUENCH test facility. (EG) 17 refs.

Vulnerability analysis of a pressurized aluminum composite vessel against hypervelocity impacts

Directory of Open Access Journals (Sweden)

Hereil Pierre-Louis

2015-01-01

Full Text Available Vulnerability of high pressure vessels subjected to high velocity impact of space debris is analyzed with the response of pressurized vessels to hypervelocity impact of aluminum sphere. Investigated tanks are CFRP (carbon fiber reinforced plastics overwrapped Al vessels. Explored internal pressure of nitrogen ranges from 1 bar to 300 bar and impact velocity are around 4400 m/s. Data obtained from Xrays radiographies and particle velocity measurements show the evolution of debris cloud and shock wave propagation in pressurized nitrogen. Observation of recovered vessels leads to the damage pattern and to its evolution as a function of the internal pressure. It is shown that the rupture mode is not a bursting mode but rather a catastrophic damage of the external carbon composite part of the vessel.

Apparatus for controlling molten core debris

International Nuclear Information System (INIS)

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

1972-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-04-15

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

Identification of gap cooling phenomena from LAVA-4 experiment using MELCOR

International Nuclear Information System (INIS)

Park, Jong-Hwa; Kim, Dong-Ha; Kim, See-Darl; Kim, Sang-Baik; Kim, Hee-Dong

2000-01-01

During the severe accident, whether the hot debris in. lower head will be cool-down or not is the important issue concerning the plant safety. KAERI has launched the 'LAVA' experimental program to examine the existence of initial gap and its effect on the cooling of hot debris. The objective of this study is to identify the gap cooling phenomena from the analysis of simulation results on LAVA-4 experiment using MELCOR1.8.4 code. Three parameters on the debris coolability in MELCOR are the quenching heat transfer coefficient for the interaction between molten Al 2 O 3 and water, the heat transfer coefficient from debris to wall and the diameter of the particulate debris for calculating the available heat transfer area with water. The sensitivity study was performed with these three parameters. However it was believed that there must be a gap between debris and inside wall during the transient. MELCOR1.8.4 does not consider these gap-cooling phenomena. Therefore a conceptual gap-cooling model has been developed and implemented into the lower plenum model in MELCOR to take into account the gap effect in the lower plenum. When the 'gap model' is implemented, the peak temperature of the vessel wall was reduced and its cooling rate was increased. (author)

Mapping coastal marine debris using aerial imagery and spatial analysis.

Science.gov (United States)

Moy, Kirsten; Neilson, Brian; Chung, Anne; Meadows, Amber; Castrence, Miguel; Ambagis, Stephen; Davidson, Kristine

2017-12-19

This study is the first to systematically quantify, categorize, and map marine macro-debris across the main Hawaiian Islands (MHI), including remote areas (e.g., Niihau, Kahoolawe, and northern Molokai). Aerial surveys were conducted over each island to collect high resolution photos, which were processed into orthorectified imagery and visually analyzed in GIS. The technique provided precise measurements of the quantity, location, type, and size of macro-debris (>0.05m 2 ), identifying 20,658 total debris items. Northeastern (windward) shorelines had the highest density of debris. Plastics, including nets, lines, buoys, floats, and foam, comprised 83% of the total count. In addition, the study located six vessels from the 2011 Tōhoku tsunami. These results created a baseline of the location, distribution, and composition of marine macro-debris across the MHI. Resource managers and communities may target high priority areas, particularly along remote coastlines where macro-debris counts were largely undocumented. Copyright © 2017 Elsevier Ltd. All rights reserved.

Experimental study of pressure drops through LOCA-generated debris deposited on a fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Suh, Jeong Kwan, E-mail: jksuh@khnp.co.kr [KHNP Central Research Institute, 1312-70 Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of); Kim, Jae Won; Kwon, Sun Guk; Lee, Jae Yong [KHNP Central Research Institute, 1312-70 Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of); Cho, Hyoung Kyu; Park, Goon Cherl [Department of Nuclear Engineering, Seoul National University, Seoul 151-742 (Korea, Republic of)

2015-08-15

Highlights: • In-vessel downstream effect tests were performed in the presence of LOCA-generated debris. • Available driving heads under each LOCA scenario were verified using experimental data. • Fibrous debris was prepared to satisfy the length distribution obtained from the bypass test. • Limiting test conditions were identified through sensitivity studies. - Abstract: Under post loss-of-coolant accident (LOCA) conditions, it is postulated that debris can be generated and transported to the containment sump strainer. Some of the debris may pass through the strainer and could challenge the long-term core cooling capability of the plant. To address this safety issue, in-vessel downstream effect tests for the advanced power reactor (APR) 1400 were performed. Fibrous debris is the most crucial material in terms of causing pressure drops, and was prepared in this study to satisfy the fiber length distribution obtained through a strainer bypass test. Sensitivity studies on pressure drops through LOCA-generated debris deposited on a fuel assembly were performed to evaluate the effects of water chemistry and fiber length distribution. The pressure drops with debris laden pure water were substantially less than those with debris laden ordinary tap water. The experiment with fiber length distribution suggested by WCAP-16793 showed lower pressure drops than those with the APR1400 specific fiber length distribution. All the experimental results showed that the pressure drops in the mock-up fuel assembly were less than the available driving head at each LOCA scenario.

Examinations of fuel debris samples from Three Mile Island Unit 2

International Nuclear Information System (INIS)

Nagase, Fumihisa

2012-01-01

In the accident at the Fukushima-Daiichi nuclear power plants, fuels were molten due to loss of coolant and heat-up of the reactor core. Information on properties of molten fuels (debris) is important to analyze progress of the accident, estimate the status inside the damaged reactors and work on a plan for debris removal. Extensive examinations for properties of debris have been conducted after the accident at the Three Mile Island Unit 2 in 1979. The Japan Atomic Energy Agency conducted a part of the examinations in the frame of the OECD/NEA Three Mile Island Vessel Investigation Program. This issue report outline and main results of the TMI-2 debris examination programs. (author)

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

satisfy these PRG recommendations. Specifically, the revised plan focuses on providing data on the extent of crust growth and melt eruptions as a function of gas sparging rate under well-controlled experiment conditions, including a floating crust boundary condition. The overall objective of MET is to determine to what extent core debris is rendered coolable by eruptive-type processes that breach the crust that rests upon the melt. The specific objectives of this test are as follows: (1) Evaluate the augmentation in surface heat flux during periods of melt eruption; (2) Evaluate the melt entrainment coefficient from the heat flux and gas flow rate data for input into models that calculate ex-vessel debris coolability; (3) Characterize the morphology and coolability of debris resulting from eruptive processes that transport melt into overlying water; and (4) Discriminate between periods when eruptions take the form of particle ejections into overlying water, leading to a porous particle bed, and single-phase extrusions, which lead to volcano-type structures.

Self-leveling onset criteria in debris beds

International Nuclear Information System (INIS)

Zhang, Bin; Harada, Tetsushi; Hirahara, Daisuke; Matsumoto, Tatsuya; Morita, Koji; Fukuda, Kenji; Yamano, Hidemasa; Suzuki, Tohru; Tobita, Yoshiharu

2010-01-01

In a core-disruptive accident of a sodium-cooled fast breeder reactor, core debris may settle on the core-support structure and/or in the lower inlet plenum of the reactor vessel because of rapid quenching and fragmentation of molten core materials in the subcooled sodium plenum. Coolant boiling is the mechanism driving the self-leveling of a debris bed that causes significant changes in the heat-removal capability of the beds. In the present study, we develop criteria establishing the onset of this self-leveling behavior that we base on a force balance model assuming a debris bed with a single-sized spherical particle. The model considers drag, buoyancy, and gravity acting on each particle. A series of experiments with simulant materials verified the applicability of this description of self-leveling. Particle size (between 0.5-6 mm), shape (spherical and nonspherical), density (namely of alumina, zirconia, lead, and stainless steel), along with boiling intensity, bed volume, and even experimental methods were taken into consideration to obtain general characteristics of the self-leveling process. We decided to use depressurization boiling to simulate an axially increasing void distribution in the debris bed, although bottom heating was also used to validate the use of the depressurization method. On the self-leveling onset issues, we obtained good agreement between model predictions and experimental results. Extrapolation of our model to actual reactor conditions is discussed. (author)

Feature test report for the Small Debris Collection and Packaging System

International Nuclear Information System (INIS)

Brisbin, S.A.

1995-01-01

The Spent Nuclear Fuel Equipment Engineering group performed feature testing of the Small Debris Collection and Packaging System (SDCPS) in the 305 Cold Test Facility from January 30, 1995, to February 1, 1995. Feature testing of the Small Debris Collection and Packaging System (SDCPS) was performed for the following reasons: To assess the feasibility of using ''drop-out'' vessels to collect small debris (<2.5 cm) in MK-II fuel canisters while transferring sludge to the Weasel Pit. To evaluate system performance under conditions similar to those in the K-Basins (e.g. submerged under 4.9 meters of water and operated with long handled tools) while using a surrogate sludge mixed with debris. To determine if canister weight could be used to predict the volume of sludge and/or debris contained within the canisters during system operation

Generic BWR-4 degraded core in-vessel study. Status report

International Nuclear Information System (INIS)

1984-11-01

Original intent of this project was to produce a phenomenological study of the in-vessel degradation which occurs during the TQUX and TQUV sequences for a generic BWR-4 from the initiation of the FSAR Chapter 15 operational transient through core debris bed formation to the failure of the primary pressure boundary. Bounding calculations were to be performed for the two high pressure and low pressure non-LOCA scenarios to assess the uncertainties in the current state of knowledge regarding the source terms for containment integrity studies. Source terms as such were defined in terms of hydrogen generation, unreacted metal, and coolant inventroy, and in terms of the form, sequencing and mode of dispersal through the primary vessel boundary. Fission product release was not to be considered as part of this study. Premature termination of the project, however, led to the dicontinuation of work on an as is basis. Work on the in-core phase from the point of scram to core debris bed formation was largely completed. A preliminary scoping calculation on the debris bed phase had been initiated. This report documents the status of the study at termination

Fuel-coolant interaction visualization test for in-vessel corium retention external reactor vessel cooling (IVR-ERVC) condition

Energy Technology Data Exchange (ETDEWEB)

Na, Young Su; Hong, Seong Ho; Song, Jin Ho; Hong, Seong Wan [Severe Accident and PHWR Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-12-15

A visualization test of the fuel-coolant interaction in the Test for Real cOrium Interaction with water (TROI) test facility was carried out. To experimentally simulate the In-Vessel corium Retention (IVR)- External Reactor Vessel Cooling (ERVC) conditions, prototypic corium was released directly into the coolant water without a free fall in a gas phase before making contact with the coolant. Corium (34.39 kg) consisting of uranium oxide and zirconium oxide with a weight ratio of 8:2 was superheated, and 22.54 kg of the 34.39 kg corium was passed through water contained in a transparent interaction vessel. An image of the corium jet behavior in the coolant was taken by a high-speed camera every millisecond. Thermocouple junctions installed in the vertical direction of the coolant were cut sequentially by the falling corium jet. It was clearly observed that the visualization image of the corium jet taken during the fuel-coolant interaction corresponded with the temperature variations in the direction of the falling melt. The corium penetrated through the coolant, and the jet leading edge velocity was 2.0 m/s. Debris smaller than 1 mm was 15% of the total weight of the debris collected after a fuel-coolant interaction test, and the mass median diameter was 2.9 mm.

Debris thickness patterns on debris-covered glaciers

Science.gov (United States)

Anderson, Leif S.; Anderson, Robert S.

2018-06-01

Many debris-covered glaciers have broadly similar debris thickness patterns: surface debris thickens and tends to transition from convex- to concave-up-down glacier. We explain this pattern using theory (analytical and numerical models) paired with empirical observations. Down glacier debris thickening results from the conveyor-belt-like nature of the glacier surface in the ablation zone (debris can typically only be added but not removed) and from the inevitable decline in ice surface velocity toward the terminus. Down-glacier thickening of debris leads to the reduction of sub-debris melt and debris emergence toward the terminus. Convex-up debris thickness patterns occur near the up-glacier end of debris covers where debris emergence dominates (ablation controlled). Concave-up debris thickness patterns occur toward glacier termini where declining surface velocities dominate (velocity controlled). A convex-concave debris thickness profile inevitably results from the transition between ablation-control and velocity-control down-glacier. Debris thickness patterns deviating from this longitudinal shape are most likely caused by changes in hillslope debris supply through time. By establishing this expected debris thickness pattern, the effects of climate change on debris cover can be better identified.

Simulation of the arrival and evolution of debris in a PWR lower head with the SFD ICARE2 code

International Nuclear Information System (INIS)

Fichot, F.; Babik, F.; Zabiego, M.; Barrachin, M.; Chatelard, P.; Lefevre, B.

1999-01-01

In a severe accident scenario, the prediction of vessel failure is related to the prediction of the behaviour of solid and liquid debris which have fallen into the lower plenum. One of the difficulties is to define the initial debris bed conditions. They will depend on the presence of water or not in the lower plenum. They will also depend on the rate and composition of the falling debris. In this context, IPSN performs calculations using the ICARE2 code in order to predict the core degradation in a scenario similar to TMI-2, and to estimate the so-called initial conditions of the debris bed, i.e. the history of the debris falling into the lower plenum and the evolution of this debris bed. The latest version of ICARE2 (V3mod0) which deals with molten pools and debris bed allows to follow the materials from their early melting in the core region to their later relocation into the lower plenum. A description of the modeling of the debris bed and molten pool formation is provided in this paper. The debris bed modeling is based on a porous medium approach. Mass and energy conservation equations for each of the three phases and momentum conservation equations for the liquid and gas phases are solved. Up to now, due to the lack of knowledge, no model has been developed to estimate the debris size distribution. It is chosen by the user at the beginning of the calculation. As the temperature increases in the debris bed, a molten pool appears and starts to grow. Thermal effects of the natural convection movements in the pool are taken into account using classical correlations. The present study corresponds to a first application of the new capabilities of ICARE2. However it shows the interest of such an approach for a problem where the behaviour in the vessel is closely related to the previous events that occurred in the core region. Improvements are foreseen, especially for the natural convection, crust formation and interaction with water. (authors)

Overview of in-vessel retention concept involving level of passivity: with application to evolutionary pressurized water reactor design

International Nuclear Information System (INIS)

Ghyym, Seong H.

1998-01-01

In this work, one strategy of severe accident management, the applicability of the in-vessel retention (IVR) concept, which has been incorporated in passive type reactor designs, to evolutionary type reactor designs, is examined with emphasis on the method of external reactor vessel cooling (ERVC) to realize the IVR concept in view of two aspects: for the regulatory aspect, it is addressed in the context of the resolution of the issue of corium coolability; for the technical one, the reliance on and the effectiveness of the IVR concept are mentioned. Additionally, for the ERVC method to be better applied to designs of the evolutionary type reactor, the conditions to be met are pointed out in view of the technical aspect. Concerning the issue of corium coolability/quenchability, based on results of the review, plausible alternative strategies are proposed. According to the decision maker's risk behavior, these would help materialize the conceptual design for evolutionary type reactors, especially Korea Next Generation Reactors (KNGRs), which have been developing at the Korea Electric Power Research Institute (KEPRI): (A1) Strategy 1A: strategy based on the global approach using the reliance on the wet cavity method; (A2) Strategy 1B: strategy based on the combined approach using both the reliance on the wet cavity method and the counter-measures for preserving containment integrity; (A3) Strategy 2A: strategy based on the global approach to the reliance on the ERVC method; (A4) Strategy 2B: strategy based on the balanced approach using both the reliance on the ERVC method and the countermeasures for preserving containment integrity. Finally, in application to an advanced pressurized water reactor (PWR) design, several recommendations are made in focusing on both monitoring the status of approaches and preparing countermeasures in regard to the regulatory and the technical aspects

Neutron Assay System for Confinement Vessel Disposition

International Nuclear Information System (INIS)

Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Mayo, Douglas R.; Miko, David K.; Salazar, William R.; Stange, Sy; Valdez, Jose I.; Vigil, Georgiana M.

2012-01-01

Los Alamos National Laboratory has a number of spherical confinement vessels (CVs) remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1-inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the CVs. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of special nuclear material (SNM) in CVs before and after cleanout. Prior to cleanout, the system will be used to perform a verification measurement of each vessel. After cleanout, the system will be used to perform safeguards-quality assays of (le)100-g 239 Pu equivalent in a vessel for safeguards termination. The CVAS has been tested and calibrated in preparation for verification and safeguards measurements.

Attenuation of airborne debris from LMFBR accidents

International Nuclear Information System (INIS)

Morewitz, H.A.; Johnson, R.P.; Nelson, C.T.; Vaughan, E.U.; Guderjahn, C.A.; Hilliard, R.K.; McCormack, J.D.; Postma, A.K.

1978-01-01

Experimental and theoretical studies have been performed to characterize the behavior of airborne particulates (aerosols) expected to be produced by hypothetical core disassembly accidents (HCDA's) in liquid metal fast breeder reactors (LMFBR's). These aerosol studies include work on aerosol transport in a 20-m high, 850-m 3 closed vessel at moderate concentrations; aerosol transport in a small vessel under conditions of high concentration (approximately 1,000 g/m 3 ), high turbulence, and high temperature (approximately 2000 0 C); and aerosol transport through various leak paths. These studies have shown that tittle, if any, airborne debris from LMFBR HCDA's would reach the atmosphere exterior to an intact reactor containment building. (author)

Severe accident research and management in Nordic Countries - A status report

International Nuclear Information System (INIS)

Frid, W.

2002-01-01

The report describes the status of severe accident research and accident management development in Finland, Sweden, Norway and Denmark. The emphasis is on severe accident phenomena and issues of special importance for the severe accident management strategies implemented in Sweden and in Finland. The main objective of the research has been to verify the protection provided by the accident mitigation measures and to reduce the uncertainties in risk dominant accident phenomena. Another objective has been to support validation and improvements of accident management strategies and procedures as well as to contribute to the development of level 2 PSA, computerised operator aids for accident management and certain aspects of emergency preparedness. Severe accident research addresses both the in-vessel and the ex-vessel accident progression phenomena and issues. Even though there are differences between Sweden and Finland as to the scope and content of the research programs, the focus of the research in both countries is on in-vessel coolability, integrity of the reactor vessel lower head and core melt behaviour in the containment, in particular the issues of core debris coolability and steam explosions. Notwithstanding that our understanding of these issues has significantly improved, and that experimental data base has been largely expanded, there are still important uncertainties which motivate continued research. Other important areas are thermal-hydraulic phenomena during reflooding of an overheated partially degraded core, fission product chemistry, in particular formation of organic iodine, and hydrogen transport and combustion phenomena. The development of severe accident management has embraced, among other things, improvements of accident mitigating procedures and strategies, further work at IFE Halden on Computerised Accident Management Support (CAMS) system, as well as plant modifications, including new instrumentation. Recent efforts in Sweden in this area

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

International Nuclear Information System (INIS)

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

2001-03-01

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

The effect of bed non-uniformities and porosity of particles on dryout in boiling particle beds

International Nuclear Information System (INIS)

Macbeth, R.V.; Mogford, D.J.; Willshire, S.J.

1988-03-01

This report relates to an on-going experimental programme concerned with the coolability of beds of reactor core debris or rubble immersed in a liquid coolant, as might occur in an accident situation. The objectives are to develop experimental techniques, improve the understanding of bed cooling mechanisms, determine dry-out limitations of various bed configurations and particle shapes and sizes and devise ways of improving bed coolability. The report concentrates on a recently discovered effect on bed coolability of particle porosity, such as exists in fragmented UO 2 fuel pellets. It is shown that porosity can lower bed dry-out powers by a factor of 4 or 5. A mechanism which explains the effect is presented. The report also gives results of bed non-uniformities obtained by mixing glass particles with the dielectrically heated 'ferrite' particles used in the experiments. (author)

Electrical discharge machining for vessel sample removal

International Nuclear Information System (INIS)

Litka, T.J.

1993-01-01

Due to aging-related problems or essential metallurgy information (plant-life extension or decommissioning) of nuclear plants, sample removal from vessels may be required as part of an examination. Vessel or cladding samples with cracks may be removed to determine the cause of cracking. Vessel weld samples may be removed to determine the weld metallurgy. In all cases, an engineering analysis must be done prior to sample removal to determine the vessel's integrity upon sample removal. Electrical discharge machining (EDM) is being used for in-vessel nuclear power plant vessel sampling. Machining operations in reactor coolant system (RCS) components must be accomplished while collecting machining chips that could cause damage if they become part of the flow stream. The debris from EDM is a fine talclike particulate (no chips), which can be collected by flushing and filtration

Experimental investigations on friction laws and dryout heat flux of particulate beds packed with multi-size spheres and irregular particles

International Nuclear Information System (INIS)

Li, Liangxing; Ma, Weimin

2011-01-01

This paper is concerned with reducing uncertainty in quantification of debris bed coolability in a hypothetical severe accident of light water reactors (LWRs). A test facility named POMECO-FL is constructed to investigate the friction laws of adiabatic single and two-phase flow in a particulate bed packed with multi-size spheres or irregular particles. The same types of particles were then loaded in the test section of the POMECO-HT facility to obtain the dryout heat flux of the volumetrically heated particulate bed. The POMECO-HT facility features a high power capacity (up to 2.1 MW/m 2 ) which enables coolability study on particulate bed with broad variations in porosity and particle diameters under both top-flooding and bottom-injection conditions. The results show that given the effective particle diameter obtained from single-phase flow through the packed bed with multi-size spheres or irregular particles, both the pressure drop and the dryout heat flux of two-phase flow through the bed can be predicted by the Reed model. The bottom injection of coolant increases the dryout heat flux significantly. Meanwhile, the elevation of the dryout position is moving upwards with increasing bottom-injection flowrate. The experimental data provides insights for interpretation of debris bed coolability, as well as high-quality data for validation of the coolability analysis models and codes. (author)

Post-accident heat removal research: A state of the art review

International Nuclear Information System (INIS)

Mueller, U.; Schulenberg, T.

1983-11-01

For a realistic assessment of the consequence of extremely unlikely reactor accidents resulting in core degradation or core meltdown key questions are how to remove the decay heat from the reactor system and how to retain the radioactive core debris within the containment. Usually, this complex of questions is referred to as Post-Accident Heat Removal (PAHR). In this article the research work on PAHR performed by various institutions during the last decade has been reviewed. The main results have been summarized under the chapter headings ''Accident Scenarios,'' - ''Core Debris Accommodation Concepts,'' and ''PAHR Topics.'' Particular emphasis has been placed on the presentation of the following problems: characteristics and coolability of solid core debris in the vector vessel, heat removal from molten pools of core material, and core-melt interaction with structural materials. Some unresolved or insufficiently answered questions relating to special ''PAHR Topics'' have been mentioned or discussed at the end of the particular Chapter. Problem areas of major uncertainty have been identified and listed at the end of the review article. They include the following subjects: formation of debris beds and bed characteristics, post dryout behaviour of particle beds, long-term availability and proper location of heat sinks, creep rupture of structures under high thermal loads. (orig.) [de

Seasonal trends in abundance and composition of marine debris in selected public beaches in Peninsular Malaysia

Science.gov (United States)

Mobilik, Julyus-Melvin; Ling, Teck-Yee; Husain, Mohd-Lokman Bin; Hassan, Ruhana

2015-09-01

The abundance and composition of marine debris were investigated at Saujana (in the state of Negeri Sembilan) and Batu Rakit (in the state of Terengganu) beaches during surveys conducted in December 2012 (northeast monsoon), May 2013 (intermediate monsoon) and July 2013 (southwest monsoon). A total of 4,682 items of debris weighing 231.4 kg were collected and sorted. Batu Rakit received substantially greater quantities of debris (815±717 items/km or 40.4±13.0 kg/km) compared to Saujana (745±444 items/km or 36.7±18.0 kg/km). Total debris item was more abundant during the southwest monsoon (SWM) (1,122±737 items/km) compared to the northeast monsoon (NEM) (825±593 items/ km) and the intermediate monsoon (IM) (394±4 items/km) seasons. Plastic category (88%) was the most numerous items collected and object items contributed 44.18% includes packaging, plastic fragments, cups, plastic shopping bags, plastic food wrapper, clear plastic bottles from the total debris items collected. Object items associated with common source (47%) were the highest debris accumulated, followed by terrestrial (30%) and marine (23%) sources. The high percentage of common and terrestrial sources during SWM season requires immediate action by marine environment stakeholders to develop and introduce strategies to reduce if not totally eliminates the marine debris in the marine environment. Awareness should be continued and focused on beach users and vessels' crew to alert them on the alarming accumulation rate of marine debris and its pathways into the marine environment.

Radionuclide release and aerosol generation during core debris interactions with concrete

International Nuclear Information System (INIS)

Powers, D.A.

1986-01-01

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

SCDAP/RELAP5 Modeling of Heat Transfer and Flow Losses in Lower Head Porous Debris

International Nuclear Information System (INIS)

Coryell, E.W.; Siefken, L.J.; Paik, S.

1998-01-01

Designs are described for implementing models for calculating the heat transfer and flow losses in porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and non-porous debris that results from core material slumping into the lower head. Currently, the COUPLE model has the capability to model convective and radiative heat transfer from the surfaces of non-porous debris in a detailed manner and to model only in a simplistic manner the heat transfer from porous debris. In order to advance beyond the simplistic modeling for porous debris, designs are developed for detailed calculations of heat transfer and flow losses in porous debris. Correlations are identified for convective heat transfer in porous debris for the following modes of heat transfer; (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, and (5) film boiling. Interphase heat transfer is modeled in an approximate manner. A design is also described for implementing a model of heat transfer by radiation from debris to the interstitial fluid. A design is described for implementation of models for flow losses and interphase drag in porous debris. Since the models for heat transfer and flow losses in porous debris in the lower head are designed for general application, a design is also described for implementation of these models to the analysis of porous debris in the core region. A test matrix is proposed for assessing the capability of the implemented models to calculate the heat transfer and flow losses in porous debris. The implementation of the models described in this report is expected to improve the COUPLE code calculation of the temperature distribution in porous debris and in the lower head that supports the debris. The implementation of these models is also expected to improve the calculation of the temperature and flow distribution in porous debris in the core region

Research on removal technologies of fuel debris and in-vessel structures using laser light (1). Research plan and research activities on FY2012

International Nuclear Information System (INIS)

Muramatsu, Toshiharu; Yamada, Tomonori; Hanari, Toshihide; Takebe, Toshihiko; Matsunaga, Yukihiro

2013-08-01

In decommissioning works of the Fukushima Daiichi nuclear power plants, it is required that fuel debris solidifying mixed materials of fuels and in-vessel structures should be removed. The fuel debris is considered to have characteristics, such as indefinite shapes, porous bodies, multi-compositions, higher hardness, etc. from the knowledge in decommissioning process of the Three Mile Island nuclear power plant. Laser lights are characterized by higher power density, local processability, remote controllability, etc. and can be performed thermal cutting and crushing-up for various materials which does not depend on fracture toughness. This report describes a research program and research activities in FY2012 aiming at developing removal system of fuel debris by the use of laser lights. Main results obtained from research activities in FY2012 are as follows: (1) Improvements of experimental infrastructures. A beam switching unit for an existing fiber laser system, an x-y-z tri-axes robot system to investigate remote control performances, and a particle image velocimetry (PIV) system for quantitation of assist gas flow characteristics were introduced to the experimental laboratory of our Applied Laser Technology Institute in Tsuruga. (2) Laser cutting performances for thick metal plates. To quantify laser cutting performance for thick metal plates of in-vessel structures, after the evaluation of the relationship between the kerf depth and amount of laser irradiation energy to the metal test piece, we evaluated for heat transfer behavior due to temperature measurement of thick metal plate on the laser cutting process. It is suggested that the heat diffusion into the cutting object can affect the heat input efficiency of the laser irradiation energy to kerf front. On the viewpoint of suppressing this thermal diffusion, it was found that it is important in improving the laser cutting performance to increase the ejection of molten metal by the assist gas, and to optimize

Assessment of damage potential to the TMI-2 lower head due to thermal attack by core debris

International Nuclear Information System (INIS)

Cronenberg, A.W.; Behling, S.R.; Broughton, J.M.

1986-06-01

Camera inspection of the Three Mile Island Unit 2 (TMI-2) inlet plenum region has shown that approximately 10 to 20 percent of the core material loading may have relocated to the lower plenum. Although vessel integrity was maintained, a question of primary concern is ''how close to vessel failure'' did this accident come. This report summarizes the results of thermal analyses aimed at assessing damage potential to the TMI-2 lower head and attached instrument penetration tubes due to thermal attack by hot core debris. Results indicate that the instrument penetration nozzles could have experienced melt failure at localized hot spot regions, with attendant debris drainage and plugging of the instrument lead tubes. However, only minor direct thermal attack of the vessel liner is predicted

Prediction of thermal margin for external cooling of reactor vessel lower head during a severe accident

International Nuclear Information System (INIS)

Yoon, Ho Jun; Suh, Kune Y.

1998-01-01

In the TMI-2 accident, approximately nineteen (19) tons of molten core material drained into the lower plenum. One of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100 .deg. C for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel. Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident management strategies. As an advanced in-vessel design concept, the COrium Attak Syndrome Immunization Structures (COASIS) are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in -vessel (COASISI) and ex-vessel (COASISO) were demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the TMI-2 and the Korean Standard Nuclear Power Plant (KSNPP) reactors. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options. In studying the in-vessel severe accident phenomena, one of the main goals is to verify the cooling mechanism in the reactor vessel lower plenum and thereby to prevent the vessel failure from thermal attack by the molten debris. This paper presents the first-principle calculation results for the thermal margin for the case of external cooling of the reactor vessel lower head. Adopting the method presented by F.B. Cheung, et al., we calculated the departure from nucleate boiling ratio (DNBR) for the three cases of pool boiling, flow boiling

SCDAP/RELAP5 Modeling of Movement of Melted Material Through Porous Debris in Lower Head

International Nuclear Information System (INIS)

Siefken, L. J.

1998-01-01

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

FOREVER Experiments on Thermal and Mechanical Behavior of a Reactor Pressure Vessel during a Severe Accident

International Nuclear Information System (INIS)

Sehgal, B.R.; Nourgaliev, R.R.; Dinh, T.N.; Karbojian, A.; Green, J.A.; Bui, V.A.

1999-01-01

This paper describes the FOREVER (Failure Of Reactor Vessel Retention) experimental program, which is currently underway at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS). The objectives of the FOREVER experiments are to obtain data and develop validated models (i) on the melt coolability process inside the vessel, in the presence of water (in particular, on the efficacy of the postulated gap cooling to preclude vessel failure); and (ii) on the lower head failure due to the creep process in the absence of water inside and/or outside the lower head. The facility employs 1/10.-scale carbon steel vessels of 0.4 m diameter, 15 mm thickness and 600 mm height. Up to 20 liters of binary-oxide melts with 100-300 K superheat are employed, as a simulant for the prototypic corium melt, and internal heating is provided by electrical heaters of up to 20 kW power in order to maintain the vessel wall temperatures at 1100-1200 K. Auxiliary systems are designed to provide an overpressure up to 4 MPa in the test vessel. Thus, severe accident scenarios with RCS depressurization are modeled. Creep behavior of the three-dimensional vessel, formation of the gap between the melt pool crust and the creeping vessel, and mechanisms of the gap cooling by water ingression will be the subjects of study and measurements in the FOREVER experimental program. Scaling rationale as well as pre-test analyses of the thermal and mechanical behavior of the FOREVER test vessels are presented. (authors)

Reactor Core Coolability Analysis during Hypothesized Severe Accidents of OPR1000

International Nuclear Information System (INIS)

Lee, Yongjae; Seo, Seungwon; Kim, Sung Joong; Ha, Kwang Soon; Kim, Hwan-Yeol

2014-01-01

Assessment of the safety features over the hypothesized severe accidents may be performed experimentally or numerically. Due to the considerable time and expenditures, experimental assessment is implemented only to the limited cases. Therefore numerical assessment has played a major role in revisiting severe accident analysis of the existing or newly designed power plants. Computer codes for the numerical analysis of severe accidents are categorized as the fast running integral code and detailed code. Fast running integral codes are characterized by a well-balanced combination of detailed and simplified models for the simulation of the relevant phenomena within an NPP in the case of a severe accident. MAAP, MELCOR and ASTEC belong to the examples of fast running integral codes. Detailed code is to model as far as possible all relevant phenomena in detail by mechanistic models. The examples of detailed code is SCDAP/RELAP5. Using the MELCOR, Carbajo. investigated sensitivity studies of Station Black Out (SBO) using the MELCOR for Peach Bottom BWR. Park et al. conduct regulatory research of the PWR severe accident. Ahn et al. research sensitivity analysis of the severe accident for APR1400 with MELCOR 1.8.4. Lee et al. investigated RCS depressurization strategy and developed a core coolability map for independent scenarios of Small Break Loss-of-Coolant Accident (SBLOCA), SBO, and Total Loss of Feed Water (TLOFW). In this study, three initiating cases were selected, which are SBLOCA without SI, SBO, and TLOFW. The initiating cases exhibit the highest probability of transitioning into core damage according to PSA 1 of OPR 1000. The objective of this study is to investigate the reactor core coolability during hypothesized severe accidents of OPR1000. As a representative indicator, we have employed Jakob number and developed JaCET and JaMCT using the MELCOR simulation. Although the RCS pressures for the respective accident scenarios were different, the JaMCT and Ja

SCDAP/RELAP5 Modeling of Heat Transfer and Flow Losses in Lower Head Porous Debris

International Nuclear Information System (INIS)

Siefken, Larry James; Coryell, Eric Wesley; Paik, Seungho; Kuo, Han Hsiung

1999-01-01

Designs are described for implementing models for calculating the heat transfer and flow losses in porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and nonporous debris that results from core material slumping into the lower head. Currently, the COUPLE model has the capability to model convective and radiative heat transfer from the surfaces of nonporous debris in a detailed manner and to model only in a simplistic manner the heat transfer from porous debris. In order to advance beyond the simplistic modeling for porous debris, designs are developed for detailed calculations of heat transfer and flow losses in porous debris. Correlations are identified for convective heat transfer in porous debris for the following modes of heat transfer; (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, and (5) film boiling. Interphase heat transfer is modeled in an approximate manner. Designs are described for models to calculate the flow losses and interphase drag of fluid flowing through the interstices of the porous debris, and to apply these variables in the momentum equations in the RELAP5 part of the code. Since the models for heat transfer and flow losses in porous debris in the lower head are designed for general application, a design is also described for implementation of these models to the analysis of porous debris in the core region. A test matrix is proposed for assessing the capability of the implemented models to calculate the heat transfer and flow losses in porous debris. The implementation of the models described in this report is expected to improve the COUPLE code calculation of the temperature distribution in porous debris and in the lower head that supports the debris. The implementation of these models is also expected to improve the calculation of the temperature and flow distribution in porous debris in the core region

GAREC analyses in Support of In-Vessel Retention Concept

International Nuclear Information System (INIS)

Azarian, G.; Gandrille, P.; Dumontet, A.; Grange; Barbier, F; Bellon, M.; Bordier, G.; Boulanger, F.; Cognet, G.; Gatt, J.M.; Humbert, J.M.; Laporte, T.; Lepareux, M.; Richard, P.; Robert, G.; Seiler, J.M.; Szabo, I.; Tourasse, M.; Valin, F.; Van Dorsselaere, J.P.

1999-01-01

The authors describe the analyses of the in-vessel retention capability which the GAREC group has performed for present and future French PWR designs. They present the reactor characteristics which are considered, describe the physical situations which are analysed and the relocation processes initiated by a corium flow, discuss the jet impacts, the debris formation and behaviour in the vessel lower head in a dry situation with absence of cooling, in wet situations in absence of external cooling, in wet situation with external cooling, in dry situation with external cooling. In this last case, they discuss the power dissipated in the corium, the molten salt behaviour, the heat flux distribution from the pool, the residual wall thickness, the heat flux distribution from the metal layer, the thermal-hydraulic aspects of water injection in the pool, the effects of crust instabilities, the external cooling, and the vessel mechanical behaviour. Then, they address the vapour explosion which may occur: mechanical loads leading to vessel failure in the cases of an eroded or non-eroded vessel, corium masses participating to the interaction (corium jets to the lower head, reflooding of corium pools with water). They finally briefly discuss the possible design improvements for in-vessel retention

Numerical module for debris behavior under severe accident conditions

International Nuclear Information System (INIS)

Kisselev, A.E.; Kobelev, G.V.; Strizhov, V.F.; Vasiliev, A.D.

2005-01-01

The late phase of a hypothetical severe accident in a nuclear reactor is characterized by the appearance of porous debris and liquid pools in core region and lower head of the reactor vessel. Thermal hydraulics and heat transfer in these regions are very important for adequate analysis of severe accident dynamics. The purpose of this work is to develop a universal module which is able to model above-mentioned phenomena on the basis of modern physical concepts. The original approach for debris evolution is developed from classical principles using a set of parameters including debris porosity; average particle diameter; temperatures and mass fractions of solid, liquid and gas phases; specific interface areas between different phases; effective thermal conductivity of each phase, including radiative heat conductivity; mass and energy fluxes through the interfaces. The calculation results of several tests on modeling of porous debris behavior, including the MP-1 experiment, are presented in comparison with experimental data. The results are obtained using this module implemented into the Russian best estimate code, RATEG/SVECHA/HEFEST, which was developed for modeling severe accident thermal hydraulics and late phase phenomena in VVER nuclear power plants. (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

Neutron Assay System for Con?nement Vessel Disposition

International Nuclear Information System (INIS)

Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Mayo, Douglas R.; Miko, David K.; Salazar, William R.; Stange, Sy; Valdez, Jose I.; Vigil, Georgiana M.

2012-01-01

Waste will be removed from confinement vessels remaining from 1970s-era experiments. Los Alamos has 9+ spherical confinement vessels remaining from experiments. Each vessel contains ∼ 500 lbs of radioactive debris such as actinide metals and oxides, metals, powdered silica, graphite, and wires and hardware. In order to dispose of the vessels, debris and contamination must be removed. Neutron assay system was designed to assay vessels before and after cleanout. System requirements are: (1) Modular and moveable; (2) Capable of detecting ∼100g 239 Pu equivalent in a 2-inch thick steel sphere with 6 foot diameter; and (3) Capable of safeguards-quality assays. Initial design parameters arethe use of 4-atm 3 He tubes with length of 6 feet, and 3 He tubes embedded in polyethelene for moderation. This paper describes the calibration of the Confinement Vessel Assay System (CVAS) and quantification of its uncertainties. Assay uncertainty depends on five factors: (1) Statistical uncertainty in the assay measurement; (2) Statistical uncertainty in the background measurement; (3) Statistical uncertainty in the isotopics determination - This should be much smaller than the other uncertainties; (4) Systematic uncertainty due to position bias; and (5) Systematic uncertainty due to fluctuations in cosmic ray spallation. This one can be virtually eliminated by performing the background measurement with an empty vessel - but that may not be possible. We used modeling and experiments to quantify the systematic uncertainties. The calibration assumes a uniform distribution of material, but reality will be different. MCNPX modeling was used to quantify the positional bias. The model was benchmarked to build confidence in its results. Material at top of vessel is 44% greater than amount assayed, according to singles. Material near 19-tube detector is 38% less than amount assayed, according to singles. Cosmic ray spallation contributes significantly to the background. Comparing rates

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

In-vessel natural circulation during a hypothetical loss-of-heat-sink accident in the Fast Flux Test Facility

International Nuclear Information System (INIS)

Perkins, K.R.; Bari, R.A.; Pratt, W.T.

1979-05-01

The capability to remove decay heat from the FFTF core via in-vessel natural circulation has been analyzed for the preboiling phase using a lumped parameter model. The results indicate that boiling will occur in the average fuel assembly for a wide spectrum of initial conditions which appear to be representative of the hypothetical loss-of-heat-sink accident. Two-phase pressure drop calculations indicate that, once the saturation temperature is reached, coolability can only be assured for decay heat levels which are less than 0.5% of the operating power. A review of the limited sodium boiling data indicates that boiling-induced natural circulation may support up to 4% of the operating power, but geometric atypicalities and a large degree of inlet subcooling for the existing data limit the applicability to the loss-of-heat-sink accident in FFTF

Large-scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

International Nuclear Information System (INIS)

Chu, T.Y.; Bentz, J.H.; Bergeron, K.D.; Slezak, S.E.; Simpson, R.B.

1994-01-01

The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the ''flooded cavity'', is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array can deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications

Large-Scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

International Nuclear Information System (INIS)

Chu, T.Y.; Bentz, J.H.; Bergeron, K.D.; Slezak, S.E.; Simpson, R.B.

1994-01-01

The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the open-quotes flooded cavityclose quotes, is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array can deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications

Reduced clot debris size using standing waves formed via high intensity focused ultrasound

Science.gov (United States)

Guo, Shifang; Du, Xuan; Wang, Xin; Lu, Shukuan; Shi, Aiwei; Xu, Shanshan; Bouakaz, Ayache; Wan, Mingxi

2017-09-01

The feasibility of utilizing high intensity focused ultrasound (HIFU) to induce thrombolysis has been demonstrated previously. However, clinical concerns still remain related to the clot debris produced via fragmentation of the original clot potentially being too large and hence occluding downstream vessels, causing hazardous emboli. This study investigates the use of standing wave fields formed via HIFU to disintegrate the thrombus while achieving a reduced clot debris size in vitro. The results showed that the average diameter of the clot debris calculated by volume percentage was smaller in the standing wave mode than in the travelling wave mode at identical ultrasound thrombolysis settings. Furthermore, the inertial cavitation dose was shown to be lower in the standing wave mode, while the estimated cavitation bubble size distribution was similar in both modes. These results show that a reduction of the clot debris size with standing waves may be attributed to the particle trapping of the acoustic potential well which contributed to particle fragmentation.

SCDAP/RELAP5 modeling of heat transfer and flow losses in lower head porous debris. Rev. 1

International Nuclear Information System (INIS)

Siefken, L.J.; Coryell, E.W.; Paik, S.; Kuo, H.

1999-01-01

Designs are described for implementing models for calculating the heat transfer and flow losses in porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and nonporous debris that results from core material slumping into the lower head. Currently, the COUPLE model has the capability to model convective and radiative heat transfer from the surfaces of nonporous debris in a detailed manner and to model only in a simplistic manner the heat transfer from porous debris. In order to advance beyond the simplistic modeling for porous debris, designs are developed for detailed calculations of heat transfer and flow losses in porous debris. Correlations are identified for convective heat transfer in porous debris for the following modes of heat transfer; (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, and (5) film boiling. Interphase heat transfer is modeled in an approximate ma nner. Designs are described for models to calculate the flow losses and interphase drag of fluid flowing through the interstices of the porous debris, and to apply these variables in the momentum equations in the RELAP5 part of the code. Since the models for heat transfer and flow losses in porous debris in the lower head are designed for general application, a design is also described for implementation of these models to the analysis of porous debris in the core region. A test matrix is proposed for assessing the capability of the implemented models to calculate the heat transfer and flow losses in porous debris. The implementation of the models described in this report is expected to improve the COUPLE code calculation of the temperature distribution in porous debris and in the lower head that supports the debris. The implementation of these models is also expected to improve the calculation of the temperature and flow distribution in porous debris in the core region

Validation of ASTEC V2 models for the behaviour of corium in the vessel lower head

International Nuclear Information System (INIS)

Carénini, L.; Fleurot, J.; Fichot, F.

2014-01-01

The paper is devoted to the presentation of validation cases carried out for the models describing the corium behaviour in the “lower plenum” of the reactor vessel implemented in the V2.0 version of the ASTEC integral code, jointly developed by IRSN (France) and GRS (Germany). In the ASTEC architecture, these models are grouped within the single ICARE module and they are all activated in typical accident scenarios. Therefore, it is important to check the validity of each individual model, as long as experiments are available for which a single physical process is involved. The results of ASTEC applications against the following experiments are presented: FARO (corium jet fragmentation), LIVE (heat transfer between a molten pool and the vessel), MASCA (separation and stratification of corium non miscible phases) and OLHF (mechanical failure of the vessel). Compared to the previous ASTEC V1.3 version, the validation matrix is extended. This work allows determining recommended values for some model parameters (e.g. debris particle size in the fragmentation model and criterion for debris bed liquefaction). Almost all the processes governing the corium behaviour, its thermal interaction with the vessel wall and the vessel failure are modelled in ASTEC and these models have been assessed individually with satisfactory results. The main uncertainties appear to be related to the calculation of transient evolutions

Vic-Maui Race Observed Marine Debris Map Service, Pacific Ocean, 2012, US EPA Region 9

Data.gov (United States)

U.S. Environmental Protection Agency — Location and descriptions of marine debris observed by the Sailing Vessel (S/V) Family Affair yacht during the Victoria, British Columbia to Maui, Hawaii Yacht Race...

Test Data for USEPR Severe Accident Code Validation

Energy Technology Data Exchange (ETDEWEB)

J. L. Rempe

2007-05-01

This document identifies data that can be used for assessing various models embodied in severe accident analysis codes. Phenomena considered in this document, which were limited to those anticipated to be of interest in assessing severe accidents in the USEPR developed by AREVA, include: • Fuel Heatup and Melt Progression • Reactor Coolant System (RCS) Thermal Hydraulics • In-Vessel Molten Pool Formation and Heat Transfer • Fuel/Coolant Interactions during Relocation • Debris Heat Loads to the Vessel • Vessel Failure • Molten Core Concrete Interaction (MCCI) and Reactor Cavity Plug Failure • Melt Spreading and Coolability • Hydrogen Control Each section of this report discusses one phenomenon of interest to the USEPR. Within each section, an effort is made to describe the phenomenon and identify what data are available modeling it. As noted in this document, models in US accident analysis codes (MAAP, MELCOR, and SCDAP/RELAP5) differ. Where possible, this report identifies previous assessments that illustrate the impact of modeling differences on predicting various phenomena. Finally, recommendations regarding the status of data available for modeling USEPR severe accident phenomena are summarized.

Advanced in-vessel retention design for next generation risk management

Energy Technology Data Exchange (ETDEWEB)

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

1998-12-31

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

Advanced in-vessel retention design for next generation risk management

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

Experimental study on in-vessel debris coolability during severe accident - Experimental and analytical model study on gap cooling in gap

Energy Technology Data Exchange (ETDEWEB)

Chang, Soon Heung; Baek, Won Pil; Yang, Soo Hyung; Kim, Soo Hyoung; Lee, Yong Ho; Chung, Yong Hun [Korea Advanced Institute of Science and Technology, Taejon (Korea)

1999-04-01

To understand the flooding and mechanism in gap geometry, research was conducted. Final objectives of research are as follows: 1) Literature survey of the flooding and heat transfer mechanism in gap geometry 2) Performing CHF experiments using bottom closed rectangular channels test section 3) Development of flooding correlation using flooding data 4) Derive instability wave length in narrow gap. The major results of research are as follows: 1) Gap size and span of channel are important parameter for flooding. 2) Kutateladze number used for analysing flooding data, is appropriate to the analysis of the flooding using non-circular narrow gap channel. 3) Flooding correlation was developed using collected flooding data and it predicts flooding data lower than 10%. 4) CHF correlation derived from developed flooding collreation overpredict CHF. 5) Instability wave length is increased as gap size is decreased. 26 refs., 46 figs., 2 tabs. (Author)

CHARACTERISTICS OF SELF-LEVELING BEHAVIOR OF DEBRIS BEDS IN A SERIES OF EXPERIMENTS

Directory of Open Access Journals (Sweden)

SONGBAI CHENG

2013-06-01

Full Text Available During a hypothetical core-disruptive accident (CDA in a sodium-cooled fast reactor (SFR, degraded core materials can form roughly conically-shaped debris beds over the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of the core material pool. However, coolant boiling may ultimately lead to leveling of the debris bed, which is crucial to the relocation of the molten core and heat-removal capability of the debris bed. To clarify the mechanisms underlying this self-leveling behavior, a large number of experiments were performed within a variety of conditions in recent years, under the constructive collaboration between the Japan Atomic Energy Agency (JAEA and Kyushu University (Japan. The present contribution synthesizes and gives detailed comparative analyses of those experiments. Effects of various experimental parameters that may have potential influence on the leveling process, such as boiling mode, particle size, particle density, particle shape, bubbling rate, water depth and column geometry, were investigated, thus giving a large palette of favorable data for the better understanding of CDAs, and improved verifications of computer models developed in advanced fast reactor safety analysis codes.

Characteristics of Self-Leveling Behavior of Debris Beds in A Series of Experiments

Energy Technology Data Exchange (ETDEWEB)

Cheng, Songbai; Yamano, Hidemasa; Suzuki, Tohru; Tobita, Yoshiharu [Japan Atomic Energy Agency, Ibaraki (Japan); Yuya, Nakamura; Bin, Zhang; Tatsuya, Matsumoto; Koji, Morita [Kyushu Univ., Fukuoka (Japan)

2013-06-15

During a hypothetical core-disruptive accident (CDA) in a sodium-cooled fast reactor (SFR), degraded core materials can form roughly conically-shaped debris beds over the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of the core material pool. However, coolant boiling may ultimately lead to leveling of the debris bed, which is crucial to the relocation of the molten core and heat-removal capability of the debris bed. To clarify the mechanisms underlying this self-leveling behavior, a large number of experiments were performed within a variety of conditions in recent years, under the constructive collaboration between the Japan Atomic Energy Agency (JAEA) and Kyushu University (Japan). The present contribution synthesizes and gives detailed comparative analyses of those experiments. Effects of various experimental parameters that may have potential influence on the leveling process, such as boiling mode, particle size, particle density, particle shape, bubbling rate, water depth and column geometry, were investigated, thus giving a large palette of favorable data for the better understanding of CDAs, and improved verifications of computer models developed in advanced fast reactor safety analysis codes.

Characteristics of Self-Leveling Behavior of Debris Beds in A Series of Experiments

International Nuclear Information System (INIS)

Cheng, Songbai; Yamano, Hidemasa; Suzuki, Tohru; Tobita, Yoshiharu; Yuya, Nakamura; Bin, Zhang; Tatsuya, Matsumoto; Koji, Morita

2013-01-01

During a hypothetical core-disruptive accident (CDA) in a sodium-cooled fast reactor (SFR), degraded core materials can form roughly conically-shaped debris beds over the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of the core material pool. However, coolant boiling may ultimately lead to leveling of the debris bed, which is crucial to the relocation of the molten core and heat-removal capability of the debris bed. To clarify the mechanisms underlying this self-leveling behavior, a large number of experiments were performed within a variety of conditions in recent years, under the constructive collaboration between the Japan Atomic Energy Agency (JAEA) and Kyushu University (Japan). The present contribution synthesizes and gives detailed comparative analyses of those experiments. Effects of various experimental parameters that may have potential influence on the leveling process, such as boiling mode, particle size, particle density, particle shape, bubbling rate, water depth and column geometry, were investigated, thus giving a large palette of favorable data for the better understanding of CDAs, and improved verifications of computer models developed in advanced fast reactor safety analysis codes

Simulation and uncertainties of the heat transfer from a heat-generating DEBRIS bed in the lower plenum

International Nuclear Information System (INIS)

Schaaf, K.; Trambauer, K.

1999-01-01

The findings of the TMI-2 post-accident analyses indicated that internal cooling mechanisms may have a considerable potential to sustain the vessel integrity after a relocation of core material to the lower plenum, provided that water is continuously available in the RPV. Numerous analytical and experimental research activities are currently underway in this respect. This paper illustrates some major findings of the experimental work on internal cooling mechanisms and describes the limitations and the uncertainties in the simulation of the heat transfer processes. Reference is made especially to the joint German DEBRIS/ RPV research program, which encompasses the experimental investigation of the thermal-hydraulics in gaps, of the heat transfer within a particulate debris bed, and of the high temperature performance of vessel steel, as well as the development of simulation models for the heat transfer in the lower head and the structural response of the RPV. In particular, the results of uncertainty and sensitivity analyses are presented, which have been carried out at GRS using an integral model that describes the major phenomena governing the long-term integrity of the reactor vessel. The investigation of a large-scale relocation indicated that the verification of a gap cooling mechanism as an inherent mechanism is questionable in terms of a stringent probabilistic uncertainty criterion, as long as the formation of a large molten pool cannot be excluded. (author)

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

Thermal and stress analyses of the reactor pressure vessel lower head of the Three Mile Island Unit 2

International Nuclear Information System (INIS)

Hashimoto, K.; Onizawa, K.; Kurihara, R.; Kawasaki, S.; Soda, K.

1992-01-01

Thermal and stress analyses were performed using the finite element analysis code ABAQUS to clarify the factors which caused tears in the stainless steel liner of the reactor pressure vessel lower head of the Three Mile Island Unit 2 (TMI-2) reactor pressure vessel during the accident on 28 March 1979. The present analyses covered the events which occurred after approximately 20 tons of molten core material were relocated to the lower head of the reactor pressure vessel. They showed that the tensile stress was highest in the case where the relocated core material consisting of homogeneous UO 2 debris was assumed to attack the lower head and the debris was then quenched. The peak tensile stress was in the vicinity of the welded zone of the penetration nozzle. This result agrees with the findings from the examination of the TMI-2 reactor pressure vessel that major tears in the stainless steel liner were observed around two penetration nozzles of the lower head. (author)

Studies on core melt behaviour in a BWR pressure vessel lower head

International Nuclear Information System (INIS)

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

1999-01-01

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

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

This is the annual report for the work performed in 1999 in the research project Melt-Structure-Water Interactions During Severe Accidents in LWRs, under the auspices of the APRI Project, jointly funded by SKI, HSK, USNRC and the Swedish and Finnish power companies. The emphasis of the work is placed on phenomena and properties which govern the fragmentation and breakup of melt jets and droplets, melt spreading and coolability, and thermal and mechanical loadings of a pressure vessel during melt-vessel interaction. We believe that significant technical advances have been achieved during the course of these studies. It was found that: The coolant temperature has significant influence on the characteristics of debris fragments produced from the breakup of an oxidic melt jet. At low subcooling the fragments are relatively large and irregular compared to the smaller particles produced at high subcooling. The melt jet density has considerable effect on the fragment size produced. As the melt density increases the fragment size becomes smaller. The mass mean size of the debris changes proportionally to the square root of the coolant to melt density ratio. The melt superheat has little effect on the debris particle size distribution produced during the melt jet fragmentation. The impingement velocity of the jet has significant impact on the fragmentation process. At lower jet velocity the melt fragments agglomerate and form a cake of large size debris. When the jet velocity is increased more complete fragmentation is obtained. The scaling methodology for melt spreading, developed during 1998, has been further validated against almost all of the spreading experimental data available so far. Experimental results for the dryout heat flux of homogeneous particulate debris beds with top flooding compare well with the Lipinski correlation. For the stratified particle beds, the fine particle layer resting on the top of another particle layer dominates the dryout processes

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

This is the annual report for the work performed in 1999 in the research project Melt-Structure-Water Interactions During Severe Accidents in LWRs, under the auspices of the APRI Project, jointly funded by SKI, HSK, USNRC and the Swedish and Finnish power companies. The emphasis of the work is placed on phenomena and properties which govern the fragmentation and breakup of melt jets and droplets, melt spreading and coolability, and thermal and mechanical loadings of a pressure vessel during melt-vessel interaction. We believe that significant technical advances have been achieved during the course of these studies. It was found that: The coolant temperature has significant influence on the characteristics of debris fragments produced from the breakup of an oxidic melt jet. At low subcooling the fragments are relatively large and irregular compared to the smaller particles produced at high subcooling. The melt jet density has considerable effect on the fragment size produced. As the melt density increases the fragment size becomes smaller. The mass mean size of the debris changes proportionally to the square root of the coolant to melt density ratio. The melt superheat has little effect on the debris particle size distribution produced during the melt jet fragmentation. The impingement velocity of the jet has significant impact on the fragmentation process. At lower jet velocity the melt fragments agglomerate and form a cake of large size debris. When the jet velocity is increased more complete fragmentation is obtained. The scaling methodology for melt spreading, developed during 1998, has been further validated against almost all of the spreading experimental data available so far. Experimental results for the dryout heat flux of homogeneous particulate debris beds with top flooding compare well with the Lipinski correlation. For the stratified particle beds, the fine particle layer resting on the top of another particle layer dominates the dryout processes

Experimental and theoretical study of large scale debris bed reflood in the PEARL facility

Energy Technology Data Exchange (ETDEWEB)

Chikhi, Nourdine, E-mail: nourdine.chikhi@irsn.fr; Fichot, F.

2017-02-15

Highlights: • Five reflooding tests have been carried out with an experimental bed, 500 mm in height and 500 mm in diameter, made of 4 mm stainless steel balls. • For the first time, such a large bed was heated practically homogenously. • The quench front velocity was determined according to thermocouple measurements inside the bed. • An analytical model, assuming a quasi-steady progression of the quench front, allows to predict the conversion ratio in most cases. • It appears that the efficiency of cooling can be increased only up to a certain limit when increasing the inlet water flow rate. - Abstract: During a severe accident in a nuclear power plant, the degradation of fuel rods and melting of materials lead to the accumulation of core materials, which are commonly, called “debris beds”. To stop core degradation and avoid the reactor vessel rupture, the main accident management procedure consists in injecting water. In the case of debris bed, the reflooding models used for Loss of Coolant Accident are not applicable. The IRSN has launched an experimental program on debris bed reflooding to develop new models and to validate severe accident codes. The PEARL facility has been designed to perform, for the first time, the reflooding of large scale debris bed (Ø540 mm, h = 500 mm and 500 kg of steel debris) in a pressurized containment. The bed is heated by means of an induction system. A specific instrumentation has been developed to measure the debris bed temperature, pressure drop inside the bed and the steam flow rate during the reflooding. In this paper, the results of the first integral reflooding tests performed in the PEARL facility at atmospheric pressure up to 700 °C are presented. Focus is made on the quench front propagation and on the steam flow rate during reflooding. The effect of water injection flow rate, debris initial temperature and residual power are also discussed. Finally, an analytical model providing the steam flow rate and

An interior vessel viewing system for DIII-D

International Nuclear Information System (INIS)

Senior, R.

1989-11-01

It was anticipated that there could be damage to the interior walls of the vacuum vessel during operations of the DIII-D tokamak. A method of viewing the inside of the vessel from the outside was required, that would allow the interior walls to be inspected visually for damage and to locate any debris resulting from operations. A miniature closed circuit television color camera system was developed which could be inserted into one of several ports of the vessel during a 'clean' vent, i.e., vented to inert gas. The system has pan, tilt and zoom capability and carries its own lighting. The use of this system allows a quick assessment of the condition of the vessel to be made under 'clean' vent conditions. This precludes the need for the permit process and manned entry into the vessel which would allow air inside the vessel. A permanent record of the inspection can then be made on video tape. The design and configuration of this camera system is presented and its use as a diagnostic tool discussed. 2 refs., 5 figs

Confinement Vessel Assay System: Calibration and Certification Report

Energy Technology Data Exchange (ETDEWEB)

Frame, Katherine C. [Los Alamos National Laboratory; Bourne, Mark M. [Los Alamos National Laboratory; Crooks, William J. [Los Alamos National Laboratory; Evans, Louise [Los Alamos National Laboratory; Gomez, Cipriano [Retired CMR-OPS: OPERATIONS; Mayo, Douglas R. [Los Alamos National Laboratory; Miko, David K. [Los Alamos National Laboratory; Salazar, William R. [Los Alamos National Laboratory; Stange, Sy [Los Alamos National Laboratory; Vigil, Georgiana M. [Los Alamos National Laboratory

2012-07-17

Los Alamos National Laboratory has a number of spherical confinement vessels (CVs) remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1 to 2 inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the vessels. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of SNM in CVs before and after cleanout. Prior to cleanout, the system will be used to perform a verification measurement of each vessel. After cleanout, the system will be used to perform safeguards-quality assays of {le} 100-g {sup 239}Pu equivalent in a vessel for safeguards termination. The system was calibrated in three different mass regions (low, medium, and high) to cover the entire plutonium mass range that will be assayed. The low mass calibration and medium mass calibration were verified for material positioned in the center of an empty vessel. The systematic uncertainty due to position bias was estimated using an MCNPX model to simulate the response of the system to material localized at various points along the inner surface of the vessel. The background component due to cosmic ray spallation was determined by performing measurements of an empty vessel and comparing to measurements in the same location with no vessel present. The CVAS has been tested and calibrated in preparation for verification and safeguards measurements of CVs before and after cleanout.

Confinement Vessel Assay System: Calibration and Certification Report

International Nuclear Information System (INIS)

Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Gomez, Cipriano; Mayo, Douglas R.; Miko, David K.; Salazar, William R.; Stange, Sy; Vigil, Georgiana M.

2012-01-01

Los Alamos National Laboratory has a number of spherical confinement vessels (CVs) remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1 to 2 inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the vessels. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of SNM in CVs before and after cleanout. Prior to cleanout, the system will be used to perform a verification measurement of each vessel. After cleanout, the system will be used to perform safeguards-quality assays of (le) 100-g 239 Pu equivalent in a vessel for safeguards termination. The system was calibrated in three different mass regions (low, medium, and high) to cover the entire plutonium mass range that will be assayed. The low mass calibration and medium mass calibration were verified for material positioned in the center of an empty vessel. The systematic uncertainty due to position bias was estimated using an MCNPX model to simulate the response of the system to material localized at various points along the inner surface of the vessel. The background component due to cosmic ray spallation was determined by performing measurements of an empty vessel and comparing to measurements in the same location with no vessel present. The CVAS has been tested and calibrated in preparation for verification and safeguards measurements of CVs before and after cleanout.

In-vessel core debris retention through external flooding of the reactor pressure vessel. SCDAP/RELAP5 assessment for the SBWR lower head

International Nuclear Information System (INIS)

Heel, A.M.J.M. van.

1995-09-01

In this report the results are discussed from various analyses on the feasibility and phenomenology of the External Flooding (EF) concept for an SBWR lower head, filled with a large heat generating corium mass. In applying External Flooding as an accident management strategy after or during core melt down, the lower drywell is filled with water up to a level where a large portion of the Reactor Pressure Vessel (RPV) is flooded. The purpose of this method is to establish cooling of the vessel wall, that is challenged by the heat load resulting from the corium, in such a way that its structural integrity if not endangered. The analysis discussed in this report focus on the thermal response of the vessel wall and the ex-vessel boiling processes under the conditions described above. For these analyses the SCDAP/REALP5 MOD 3.1 code was used. The major outcome of the calculations is, that a major part of the vessel wall remains well below themelting temperature of carbon steel, as long as flooding of the external surface of the lower head is established. The SCDAP/RELAP5 analyses indicated that low-quality Critical Heat Flux (CHF) was not exceeded, under all the conditions that had been tested. However, a comaprison of the heat fluxes, as calculated in RELAP5, with the CHF values obtained from the Zuber correlation and the Vishnev correction factor (for boiling at inclined surfaces) proved that CHF values, based on these criteria, were exceeded in several surface points of the lower head mesh. The correlations, as resident in the current version of RELAP5 MOD 3.1, might lead to over-estimation of CHF for the EF analyses discussed in this report. The use of the more conservative Zuber correlation with the Vishnev correction factor is recommended for EF analyses. (orig.)

[Research progress in post-fire debris flow].

Science.gov (United States)

Di, Xue-ying; Tao, Yu-zhu

2013-08-01

The occurrence of the secondary disasters of forest fire has significant impacts on the environment quality and human health and safety. Post-fire debris flow is one of the most hazardous secondary disasters of forest fire. To understand the occurrence conditions of post-fire debris flow and to master its occurrence situation are the critical elements in post-fire hazard assessment. From the viewpoints of vegetation, precipitation threshold and debris flow material sources, this paper elaborated the impacts of forest fire on the debris flow, analyzed the geologic and geomorphic conditions, precipitation and slope condition that caused the post-fire debris flow as well as the primary mechanisms of debris-flow initiation caused by shallow landslide or surface runoff, and reviewed the research progress in the prediction and forecast of post-fire debris flow and the related control measures. In the future research, four aspects to be focused on were proposed, i. e., the quantification of the relationships between the fire behaviors and environmental factors and the post-fire debris flow, the quantitative research on the post-fire debris flow initiation and movement processes, the mechanistic model of post-fire debris flow, and the rapid and efficient control countermeasures of post-fire debris flow.

Numerical investigation of debris materials prior to debris flow hazards using satellite images

Science.gov (United States)

Zhang, N.; Matsushima, T.

2018-05-01

The volume of debris flows occurred in mountainous areas is mainly affected by the volume of debris materials deposited at the valley bottom. Quantitative evaluation of debris materials prior to debris flow hazards is important to predict and prevent hazards. At midnight on 7th August 2010, two catastrophic debris flows were triggered by the torrential rain from two valleys in the northern part of Zhouqu City, NW China, resulting in 1765 fatalities and huge economic losses. In the present study, a depth-integrated particle method is adopted to simulate the debris materials, based on 2.5 m resolution satellite images. In the simulation scheme, the materials are modeled as dry granular solids, and they travel down from the slopes and are deposited at the valley bottom. The spatial distributions of the debris materials are investigated in terms of location, volume and thickness. Simulation results show good agreement with post-disaster satellite images and field observation data. Additionally, the effect of the spatial distributions of the debris materials on subsequent debris flows is also evaluated. It is found that the spatial distributions of the debris materials strongly influence affected area, runout distance and flow discharge. This study might be useful in hazard assessments prior to debris flow hazards by investigating diverse scenarios in which the debris materials are unknown.

Investigation of safety measures to severe accident of Fast Breeder Reactor

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

So as to plan the accident management to severe accident of Fast Breeder Reactor (FBR), it is primary important to understand the progression of severe accident (SA) precisely. In this study, it has been aimed to reveal two items that work as keys in the evaluation of SA in sodium cooled FBR. One is the cool-ability of degraded core on the core support plate by sodium natural circulation in the post accident heat removal (PAHR) phase. An obstacle that hinders the smooth heat transfer from fuel debris to coolant is the formation of sodium-uranate by chemical reaction between sodium and fuel. Following the measurement of physical values of sodium-uranate in FY 2011, experiments has been performed to reveal the conditions for sodium-uranate formation on fuel debris in sodium pool simulating the actual situation of the degraded core. The cool-ability of the debris bed was analyzed using the Lipinski 1-D model. Another research performed in this study is the measurement of fission product (cesium and antimony) evaporation rates from FBR fuel as a function of temperature, because presently the fission product evaporation rates data for LWR is also temporarily used for FBR SA analysis. The measurement was performed using the irradiated fuels in the Test Reactor JOYO. (author)

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

Identification and initial assessment of candidate BWR late-phase in-vessel accident management strategies

International Nuclear Information System (INIS)

Hodge, S.A.

1991-01-01

Work sponsored by the United States Nuclear Regulatory Commission (USNRC) to identify and perform preliminary assessments of candidate BWR [boiling water reactor] in-vessel accident management strategies was completed at Oak Ridge National Laboratory (ORNL) during fiscal year 1990. Mitigative strategies for containment events have been the subject of a companion study at Brookhaven National Laboratory. The focus of this Oak Ridge effort was the development of new strategies for mitigation of the late phase events, that is, the events that would occur in-vessel after the onset of significant core damage. The work began with an investigation of the current status of BWR in-vessel accident management procedures and proceeded through a preliminary evaluation of several candidate new strategies. The steps leading to the identification of the candidate strategies are described. The four new candidate late-phase (in-vessel) accident mitigation strategies identified by this study and discussed in the report are: (1) keep the reactor vessel depressurized; (2) restore injection in a controlled manner; (3) inject boron if control blade damage has occurred; and (4) containment flooding to maintain core and structural debris in-vessel. Additional assessments of these strategies are proposed

In-vessel retention modeling capabilities in MAAP5

International Nuclear Information System (INIS)

Paik, Chan Y.; Lee, Sung Jin; Zhou, Quan; Luangdilok, W.; Reeves, R.W.; Henry, R.E.; Plys, M.; Scobel, J.H.

2012-01-01

Modular Accident Analysis Program (MAAP) is an integrated severe accident analysis code for both light water and heavy water reactors. New and improved models to address the complex phenomena associated with in-vessel retention (IVR) were incorporated into MAAP5.01. They include: -a) time-dependent volatile and non-volatile decay heat, -b) material properties at high temperatures, -c) finer vessel wall nodalization, -d) new correlations for natural convection heat transfer in the oxidic pool, -e) refined metal layer heat transfer to the reactor vessel wall and surroundings, -f) formation of a heavy metal layer, and -g) insulation cooling channel model and associated ex-vessel heat transfer and critical heat flux correlations. In this paper, the new and improved models in MAAP5.01 are described and sample calculation results are presented for the AP1000 passive plant. For the IVR evaluation, a transient calculation is useful because the timing of corium relocation, decaying heat load, and formation of separate layers in the lower plenum all affect integrity of the lower head. The key parameters affecting the IVR success are the metal layer emissivity and thickness of the top metal layer, which depends on the amount of steel in the oxidic pool and in the heavy metal layer. With the best estimate inputs for the debris mixing parameters in a conservative IVR scenario, the AP1000 plant results show that the maximum ex-vessel heat flux to CHF ratio is about 0.7, which occurs before 10.000 seconds when the decay heat is high. The AP1000 plant results demonstrate how MAAP5.01 can be used to evaluate IVR and to gain insight into responses of the lower head during a severe accident

Liquid entrainment through orifices by sparging gas

International Nuclear Information System (INIS)

Bonnet, J.M.; Malara, M.; Amblard, M.; Seiler, J.M.

2001-01-01

Corium Coolability by water flood during an MCCI (Molten Corium Concrete Interaction) is still an open problem. Several physical mechanisms have been identified which may reduce significantly and finally stop the ablation of concrete. Among these mechanisms, corium ejection by sparging gas into the overlying water may represent an important contribution. This mechanism was at the origin of a large and coolable debris bed and volcano formation in the MACE M3B test. This mechanism has also been observed in simulant material tests performed at UCSB and at FZK. The objective of the work, which is described in the present paper, is to model this mechanism and to quantify the liquid entrainment rate by sparging gas. (author)

Nonlinear response of vessel walls due to short-time thermomechanical loading

International Nuclear Information System (INIS)

Pfeiffer, P.A.; Kulak, R.F.

1994-01-01

Maintaining structural integrity of the reactor pressure vessel (RPV) during a postulated core melt accident is an important safety consideration in the design of the vessel. This study addresses the failure predictions of the vessel due to thermal and pressure loadings fro the molten core debris depositing on the lower head of the vessel. Different loading combinations were considered based on the dead load, yield stress assumptions, material response and internal pressurization. The analyses considered only short term failure (quasi static) modes, long term failure modes were not considered. Short term failure modes include plastic instabilities of the structure and failure due to exceeding the failure strain. Long term failure odes would be caused by creep rupture that leads to plastic instability of the structure. Due to the sort time durations analyzed, creep was not considered in the analyses presented

Confinement Vessel Assay System: Design and Implementation Report

International Nuclear Information System (INIS)

Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Mayo, Douglas R.; Gomez, Cipriano D.; Miko, David K.; Salazar, William R.; Stange, Sy; Vigil, Georgiana M.

2012-01-01

Los Alamos National Laboratory has a number of spherical confinement vessels remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1- to 2-inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the vessels. We have developed a neutron assay system for the purposes of Materials Control and Accountability (MC and A) measurements of the vessel prior to and after cleanout. We present our approach to confronting the challenges in designing, building, and testing such a system. The system was designed to meet a set of functional and operational requirements. A Monte Carlo model was developed to aid in optimizing the detector design as well as to predict the systematic uncertainty associated with confinement vessel measurements. Initial testing was performed to optimize and determine various measurement parameters, and then the system was characterized using 252 Cf placed a various locations throughout the measurement system. Measurements were also performed with a 252 Cf source placed inside of small steel and HDPE shells to study the effect of moderation. These measurements compare favorably with their MCNPX model equivalent, making us confident that we can rely on the Monte Carlo simulation to predict the systematic uncertainty due to variations in response to material that may be localized at different points within a vessel.

Comparison of an Inductance In-Line Oil Debris Sensor and Magnetic Plug Oil Debris Sensor

Science.gov (United States)

Dempsey, Paula J.; Tuck, Roger; Showalter, Stephen

2012-01-01

The objective of this research was to compare the performance of an inductance in-line oil debris sensor and magnetic plug oil debris sensor when detecting transmission component health in the same system under the same operating conditions. Both sensors were installed in series in the NASA Glenn Spiral Bevel Gear Fatigue Rig during tests performed on 5 gear sets (pinion/gear) when different levels of damage occurred on the gear teeth. Results of this analysis found both the inductance in-line oil debris sensor and magnetic plug oil debris sensor have benefits and limitations when detecting gearbox component damage.

Heat transfer between relocated materials and the RPV lower head

International Nuclear Information System (INIS)

Rempe, J.L.; Knudson, D.L.; Kohriyama, T.

2001-01-01

Questions about the coolability of a continuous mass of relocated corium were raised during the Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP) Post-accident examinations indicate that nearly half of the material that relocated to the vessel lower head during the TMI-2 accident formed a cohesive or ''continuous'' layer. TMI-2 VIP results and other evidence suggest that conduction through this continuous layer of solidified corium materials was assisted by other cooling mechanisms. Because increased knowledge about in-vessel coolability of corium materials may assist reactor designers in demonstrating that their concepts are passively safe, there is international interest in this topic. However, data are needed to identify what cooling mechanism(s) occurred and to develop a validated model for predicting this cooling. Corium cooling models significantly impact predictions for subsequent accident progression, such as the estimated time and mode of vessel failure. Hence, improved cooling models will provide a much needed, missing component of severe accident analyses. This paper provides a critical review of research investigating the coolability of corium relocating to a water-filled lower head. Where possible, existing models and data for predicting cooling are quantitatively compared; and governing relationships are identified. Key phenomena that should be incorporated into models for predicting this heat transfer are discussed, and deficiencies in current models and available data for predicting cooling are noted. Recommendations for improving these models and for obtaining data to validate these models are also provided. (author)

Hydraulic Simulation of In-vessel Downstream Effect Test Using MARS-KS Code

Energy Technology Data Exchange (ETDEWEB)

Bang, Young Seok; Lee, Joon Soo; Ryu, Seung Hoon [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2015-05-15

In-vessel downstream effect test (IDET) has been required to evaluate the effect of debris on long term core cooling following a loss of coolant accident (LOCA) in support of resolution of Generic Safety Issue (GSI) 191. Head loss induced by debris (fiber and particle) accumulated on prototypical fuel assembly (FA) should be compared with the available driving head to the core for the various combinations of LOCA and Emergency Core Cooling System (ECCS) injection. The actual simulation was conducted using MARS-KS code. Also the influence of small difference in gap size which was found in the actual experiment is evaluated using the present model. A simple model to determine the form loss factors of FA and gap in clean state and the debris laden state is discussed based on basic fluid mechanics. Those form loss factors were applied to the hydraulic simulation of a selected IDET using MARS-KS code. The result indicated that the present model can be applied to IDET simulation. The pressure drop influenced by small difference in gap size can be evaluated by the present model with practical assumption.

Cooling of particulate debris beds: analysis of the initial D-series experiments

International Nuclear Information System (INIS)

Rivard, J.B.

1978-01-01

In an effort to provide basic data on the cooling of fast reactor debris, three in-pile experiments employing oxide fuel particulate in liquid sodium were completed in late 1977. Preliminary results from these experiments were reported shortly after their completion at the Third Post-Accident Heat Removal Information Exchange, at Argonne National Laboratory. In these experiments, a distribution of 100 μm to 1000 μm-sized particles of enriched UO 2 was fission-heated to simulate decay-heated debris. In each experiment, the UO 2 particles were contained in a closed, flat-bottomed vessel 012 mm in diameter which was insulated on the diameter and bottom. Sufficient sodium was included to saturate the bed of particles and to provide a volume of bulk sodium above the bed at a controlled temperature. Parameters of interest in the experiments are given

Simulations of ex-vessel fuel coolant interactions in a Nordic BWR using MC3D code

International Nuclear Information System (INIS)

Thakre, S.; Ma, W.

2013-08-01

Nordic Boiling Water Reactors (BWRs) employ a drywell cavity flooding technique as a nuclear severe accident management strategy. In case of core melt accident where the reactor pressure vessel will fail and the melt will eject from the lower head and fall into a water pool, may be in the form of a continuous jet. It is assumed that the melt jet will fragment, quench and form a coolable debris bed into the water pool. The melt interaction with a water pool may cause an energetic steam explosion which creates a potential risk towards the integrity of containment, leading to fission products release into the atmosphere. The results of the APRI-7 project suggest that the significant damage to containment structures by steam explosion cannot be ruled according to the state-of-the-art knowledge about corresponding accident scenario. In the follow-up project APRI-8 (2012-2016) one of the goals of the KTH research is to resolve the steam explosion energetics (SEE) issue, developing a risk-oriented framework for quantifying conditional threats to containment integrity for a Nordic type BWR. The present study deals with the premixing and explosion phase calculations of a Nordic BWR dry cavity, using MC3D, a multiphase CFD code for fuel coolant interactions. The main goal of the study is the assessment of pressure buildup in the cavity and the impact loading on the side walls. The conditions for the calculations are used from the SERENA-II BWR case exercise. The other objective was to do the sensitivity analysis of the parameters in modeling of fuel coolant interactions, which can help to reduce uncertainty in assessment of steam explosion energetics. The results show that the amount of liquid melt droplets in the water (region of void<0.6) is maximum even before reaching the jet at the bottom. In the explosion phase, maximum pressure is attained at the bottom and the maximum impulse on the wall is at the bottom of the wall. The analysis is carried out using two different

Simulations of ex-vessel fuel coolant interactions in a Nordic BWR using MC3D code

Energy Technology Data Exchange (ETDEWEB)

Thakre, S.; Ma, W. [Royal Institute of Technology, KTH. Div. of Nuclear Power Safety, Stockholm (Sweden)

2013-08-15

Nordic Boiling Water Reactors (BWRs) employ a drywell cavity flooding technique as a nuclear severe accident management strategy. In case of core melt accident where the reactor pressure vessel will fail and the melt will eject from the lower head and fall into a water pool, may be in the form of a continuous jet. It is assumed that the melt jet will fragment, quench and form a coolable debris bed into the water pool. The melt interaction with a water pool may cause an energetic steam explosion which creates a potential risk towards the integrity of containment, leading to fission products release into the atmosphere. The results of the APRI-7 project suggest that the significant damage to containment structures by steam explosion cannot be ruled according to the state-of-the-art knowledge about corresponding accident scenario. In the follow-up project APRI-8 (2012-2016) one of the goals of the KTH research is to resolve the steam explosion energetics (SEE) issue, developing a risk-oriented framework for quantifying conditional threats to containment integrity for a Nordic type BWR. The present study deals with the premixing and explosion phase calculations of a Nordic BWR dry cavity, using MC3D, a multiphase CFD code for fuel coolant interactions. The main goal of the study is the assessment of pressure buildup in the cavity and the impact loading on the side walls. The conditions for the calculations are used from the SERENA-II BWR case exercise. The other objective was to do the sensitivity analysis of the parameters in modeling of fuel coolant interactions, which can help to reduce uncertainty in assessment of steam explosion energetics. The results show that the amount of liquid melt droplets in the water (region of void<0.6) is maximum even before reaching the jet at the bottom. In the explosion phase, maximum pressure is attained at the bottom and the maximum impulse on the wall is at the bottom of the wall. The analysis is carried out using two different

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.

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

United States Nuclear Regulatory Commission research program on core debris/concrete interactions and ex-vessel fission-product release

International Nuclear Information System (INIS)

Burson, S.B.

1987-01-01

The study of core debris/concrete interaction phenomena has been a significant element of the NRC's Severe Accident Research Program for a number of years. The CORCON and VANESA codes used to predict the consequences of high-temperature debris attack on concrete and fission-product aerosol release are state-of-the-art computational tools. The major thrust of current NRC sponsored research focuses on the refinement, verification, and validation of these codes. An overview of the analytical and experimental aspects of the NRC research program is presented

Percolation cooling of the Three Mile Island Unit 2 lower head by way of thermal cracking and gap formation

DEFF Research Database (Denmark)

Thomsen, K.L.

2002-01-01

Two partial models have been developed to elucidate the Three Mile Island Unit 2 lower head coolability by water percolation from above into the thermally cracking debris bed and into a gap between the debris and the wall The bulk permeability of the cracked top crust is estimated based on simple...... fracture mechanics and application of Poiseuille's law to the fractures. The gap is considered as an abstraction representing an initially rugged interface, which probably expanded by thermal deformation and cracking in connection with the water ingress. The coupled flow and heat conduction problem...

Identification and assessment of BWR in-vessel severe accident mitigation strategies

International Nuclear Information System (INIS)

Hodge, S.A.; Kress, T.S.; Cleveland, J.C.; Petek, M.

1992-01-01

This paper briefly describes the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to evaluate the effectiveness and feasibility of current and proposed strategies for BWR severe accident management. These results are described in detail in the just-released report Identification and Assessment of BWR In-Vessel Severe Accident Mitigation Strategies, NUREG/CR-5869, which comprises three categories of findings. First, an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences is combined with a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, two of the four candidate strategies identified by this effort are assessed in detail. These are (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

On the influence of debris in glacier melt modelling: a new temperature-index model accounting for the debris thickness feedback

Science.gov (United States)

Carenzo, Marco; Mabillard, Johan; Pellicciotti, Francesca; Reid, Tim; Brock, Ben; Burlando, Paolo

2013-04-01

The increase of rockfalls from the surrounding slopes and of englacial melt-out material has led to an increase of the debris cover extent on Alpine glaciers. In recent years, distributed debris energy-balance models have been developed to account for the melt rate enhancing/reduction due to a thin/thick debris layer, respectively. However, such models require a large amount of input data that are not often available, especially in remote mountain areas such as the Himalaya. Some of the input data such as wind or temperature are also of difficult extrapolation from station measurements. Due to their lower data requirement, empirical models have been used in glacier melt modelling. However, they generally simplify the debris effect by using a single melt-reduction factor which does not account for the influence of debris thickness on melt. In this paper, we present a new temperature-index model accounting for the debris thickness feedback in the computation of melt rates at the debris-ice interface. The empirical parameters (temperature factor, shortwave radiation factor, and lag factor accounting for the energy transfer through the debris layer) are optimized at the point scale for several debris thicknesses against melt rates simulated by a physically-based debris energy balance model. The latter has been validated against ablation stake readings and surface temperature measurements. Each parameter is then related to a plausible set of debris thickness values to provide a general and transferable parameterization. The new model is developed on Miage Glacier, Italy, a debris cover glacier in which the ablation area is mantled in near-continuous layer of rock. Subsequently, its transferability is tested on Haut Glacier d'Arolla, Switzerland, where debris is thinner and its extension has been seen to expand in the last decades. The results show that the performance of the new debris temperature-index model (DETI) in simulating the glacier melt rate at the point scale

SCDAP/RELAP5 modeling of fluid heat transfer and flow losses through porous debris in a light water reactor

International Nuclear Information System (INIS)

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

2000-01-01

The SCDAP/RELAP5 code is being developed at the Idaho National Engineering and Environmental Laboratory under the primary sponsorship of the U.S. Nuclear Regulatory Commission (NRC) to provide best-estimate transient simulations of light water reactor coolant systems during severe accidents. This paper describes the modeling approach used in the SCDAP/RELAP5 code to calculate fluid heat transfer and flow losses through porous debris that has accumulated in the vessel lower head and core regions during the latter stages of a severe accident. The implementation of heat transfer and flow loss correlations into the code is discussed, and calculations performed to assess the validity of the modeling approach are described. The different modes of heat transfer in porous debris include: (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, (5) film boiling, and (6) transition from film boiling to convection to vapor. The correlations for flow losses in porous debris include frictional and form losses. The correlations for flow losses were integrated into the momentum equations in the RELAP5 part of the code. Since RELAP5 is a very general non-homogeneous non-equilibrium thermal-hydraulics code, the resulting modeling methodology is applicable to a wide range of debris thermal-hydraulic conditions. Assessment of the SCDAP/RELAP5 debris bed thermal-hydraulic models included comparisons with experimental measurements and other models available in the open literature. The assessment calculations, described in the paper, showed that SCDAP/RELAP5 is capable of calculating the heat transfer and flow losses occurring in porous debris regions that may develop in a light water reactor during a severe accident

Safety against releases in severe accidents. Final report

International Nuclear Information System (INIS)

Lindholm, I.; Berg, Oe.; Nonboel, E.

1997-12-01

The work scope of the RAK-2 project has involved research on quantification of the effects of selected severe accident phenomena for Nordic nuclear power plants, development and testing of a computerised accident management support system and data collection and description of various mobile reactors and of different reactor types existing in the UK. The investigations of severe accident phenomena focused mainly on in-vessel melt progression, covering a numerical assessment of coolability of a degraded BWR core, the possibility and consequences of a BWR reactor to become critical during reflooding and the core melt behavior in the reactor vessel lower plenum. Simulant experiments were carried out to investigate lower head hole ablation induced by debris discharge. In addition to the in-vessel phenomena, a limited study on containment response to high pressure melt ejection in a BWR and a comparative study on fission product source term behaviour in a Swedish PWR were performed. An existing computerised accident management support system (CAMS) was further developed in the area of tracking and predictive simulation, signal validation, state identification and user interface. The first version of a probabilistic safety analysis module was developed and implemented in the system. CAMS was tested in practice with Barsebaeck data in a safety exercise with the Swedish nuclear authority. The descriptions of the key features of British reactor types, AGR, Magnox, FBR and PWR were published as data reports. Separate reports were issued also on accidents in nuclear ships and on description of key features of satellite reactors. The collected data were implemented in a common Nordic database. (au)

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.

Heat transfer between relocated materials and the RPV lower head

Energy Technology Data Exchange (ETDEWEB)

Rempe, J.L.; Knudson, D.L. [Idaho National Engineering and Environmental Lab., Idaho Falls, ID (United States); Kohriyama, T. [INSS, Fukui (Japan)

2001-07-01

Questions about the coolability of a continuous mass of relocated corium were raised during the Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP) Post-accident examinations indicate that nearly half of the material that relocated to the vessel lower head during the TMI-2 accident formed a cohesive or ''continuous'' layer. TMI-2 VIP results and other evidence suggest that conduction through this continuous layer of solidified corium materials was assisted by other cooling mechanisms. Because increased knowledge about in-vessel coolability of corium materials may assist reactor designers in demonstrating that their concepts are passively safe, there is international interest in this topic. However, data are needed to identify what cooling mechanism(s) occurred and to develop a validated model for predicting this cooling. Corium cooling models significantly impact predictions for subsequent accident progression, such as the estimated time and mode of vessel failure. Hence, improved cooling models will provide a much needed, missing component of severe accident analyses. This paper provides a critical review of research investigating the coolability of corium relocating to a water-filled lower head. Where possible, existing models and data for predicting cooling are quantitatively compared; and governing relationships are identified. Key phenomena that should be incorporated into models for predicting this heat transfer are discussed, and deficiencies in current models and available data for predicting cooling are noted. Recommendations for improving these models and for obtaining data to validate these models are also provided. (author)

Experimental study of heat transfer in the slotted channels at CTF facility

International Nuclear Information System (INIS)

Asmolov, V.; Kobzar, L.; Nickulshin, V.; Strizhov, V.

1999-01-01

During core melt accident significant amount of core may relocate in the reactor pressure vessel lower head. During its cooling it may form cracks inside the corium and gap between corium and reactor vessel. Gap also may appear due to deformation of the lower head if its temperature exceed creep limit. Slotted channels ensure ingress of the cooling water into the corium, and exit of the generated steam. Study of the cool-down mechanism of the solid core debris in the lower head of the reactor vessel through gap and cracks is the objective of experimental work on the CTF facility. Thermal hydraulics in the heated channels closed from the bottom and flooded with the saturated water from the top of the channel, is characterized by the counterflow of the steam and water, attended by such specific phenomena as the dry out when boiling, flooding and overturning of the coming down flow of water at the certain flow rates of the steam going up, partial dry out of the channel, and reflooding from the top of the heated channel with the saturated water. The above phenomena may reveal independently or in different combinations depending on geometric parameters of the channel, heat release, and coolant parameters. Interchange of these processes with a certain cyclic sequence is possible. Experimental study was performed at the CTF (Coolability Test Facility) facility, which is a part of the thermohydraulic KC test facility in the RRC 'Kurchatov Institute'. Presented results are obtained at the CTF-1 test section which represents a vertical flat channel modeling a single crack in the solidified corium or the gap between the corium and reactor vessel

Wholesale debris removal from LEO

Science.gov (United States)

Levin, Eugene; Pearson, Jerome; Carroll, Joseph

2012-04-01

Recent advances in electrodynamic propulsion make it possible to seriously consider wholesale removal of large debris from LEO for the first time since the beginning of the space era. Cumulative ranking of large groups of the LEO debris population and general limitations of passive drag devices and rocket-based removal systems are analyzed. A candidate electrodynamic debris removal system is discussed that can affordably remove all debris objects over 2 kg from LEO in 7 years. That means removing more than 99% of the collision-generated debris potential in LEO. Removal is performed by a dozen 100-kg propellantless vehicles that react against the Earth's magnetic field. The debris objects are dragged down and released into short-lived orbits below ISS. As an alternative to deorbit, some of them can be collected for storage and possible in-orbit recycling. The estimated cost per kilogram of debris removed is a small fraction of typical launch costs per kilogram. These rates are low enough to open commercial opportunities and create a governing framework for wholesale removal of large debris objects from LEO.

Measurement of critical heat flux in narrow gap with two-dimensional slices

International Nuclear Information System (INIS)

Kim, Yong Hoon; Kim, Sung Joong; Noh, Sang Woo; Suh, Kune Y.

2002-01-01

A cooling mechanism due to boiling in a gap between the debris crust and the reactor pressure vessel (RPV) wall was proposed for the TMI-2 reactor accident analysis. If there is enough heat transfer through the gap to cool the outer surface of the debris and the inner surface of the wall, the RPV wall may preserve its integrity during a severe core melt accident. If the heat removal through gap cooling relative to the counter-current flow limitation (CCFL) is pronounced, the safety margin of the reactor can be far greater than what had been previously known in the severe accident management arena. Should a severe accident take place, the RPV integrity will be maintained because of the inherent nature of degraded core coolability inside the lower head due to boiling in a narrow gap between the debris crust and the RPV wall. As a defense-in-depth measure, the heat removal capability by gap cooling coupled with external cooling can be examined for the Korean Standard Nuclear Power Plant (KSNPP) and the Advanced Power Reactor 1400MWe (APR1400) in light of the TMI-2 vessel survival. A number of studies were carried out to investigate the complex heat transfer mechanisms for the debris cooling in the lower plenum. However, these heat transfer mechanisms have not been clearly understood yet. The CHFG (Critical Heat Flux in Gap) experiments at KAERI were carried out to develop the critical heat flux (CHF) correlation in a hemispherical gap, which is the upper limit of the heat transfer. According to the CHFG experiments performed with a pool boiling condition, the CHF in a parallel gap was reduced by 1/30 compared with the value measured in the open pool boiling condition. The correlation developed from the CHFG experiment is based on the fact that the CHF in a hemispherical gap is governed by the CCFL and a Kutateladze type CCFL parameter correlates CCFL data well in hemispherical gap geometry. However, the results of the CHFG experiments appear to be limited in their

The modeling and analysis of in-vessel corium/structure interaction in boiling water reactors

International Nuclear Information System (INIS)

Podowski, M.Z.; Kurul, N.; Kim, S.-W.; Baltyn, W.; Frid, W.

1997-01-01

A complete stand-alone state-of-the-art model has been developed of the interaction between corium debris in the lower plenum and the RPV walls and internal structures, including the vessel failure mechanisms. This new model has been formulated as a set of consistent computer modules which could be linked with other existing models and/or computer codes. The combined lower head and lower plenum modules were parametrically tested and applied to predict the consequences of a hypothetical station blackout in a Swedish BWR. (author)

Woody debris

Science.gov (United States)

Donna B. Scheungrab; Carl C. Trettin; Russ Lea; Martin F. Jurgensen

2000-01-01

Woody debris can be defined as any dead, woody plant material, including logs, branches, standing dead trees, and root wads. Woody debris is an important part of forest and stream ecosystems because it has a role in carbon budgets and nutrient cycling, is a source of energy for aquatic ecosystems, provides habitat for terrestrial and aquatic organisms, and contributes...

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

Plastic debris in the coastal environment: The invincible threat? Abundance of buried plastic debris on Malaysian beaches.

Science.gov (United States)

Fauziah, S H; Liyana, I A; Agamuthu, P

2015-09-01

Studies on marine debris have gained worldwide attention since many types of debris have found their way into the food chain of higher organisms. Thus, it is crucial that more focus is given to this area in order to curb contaminations in sea food. This study was conducted to quantify plastic debris buried in sand at selected beaches in Malaysia. Marine debris was identified according to size range and distribution, and this information was related to preventive actions to improve marine waste issues. For the purpose of this study, comparison of plastic waste abundance between a recreational beach and fish-landing beaches was also carried out, since the different beach types represent different activities that produce debris. Six beaches along the Malaysian coastline were selected for this study. The plastic types in this study were related to the functions of the beach. While recreational beaches have abundant quantities of plastic film, foamed plastic including polystyrene, and plastic fragment, fish-landing beaches accumulated line and foamed plastic. A total of 2542 pieces (265.30 g m(-2)) of small plastic debris were collected from all six beaches, with the highest number from Kuala Terengganu, at 879 items m(-2) on Seberang Takir Beach, followed by Batu Burok Beach with 780 items m(-2). Findings from studies of Malaysian beaches have provided a clearer understanding of the distribution of plastic debris. This demonstrates that commitments and actions, such as practices of the 'reduce, reuse, recycle' (3R) approach, supporting public awareness programmes and beach clean-up activities, are essential in order to reduce and prevent plastic debris pollution. © The Author(s) 2015.

Effectiveness of In-Vessel Retention Strategies and Minimum Safety Injection Flow over Postulated Severe Accidents of OPR1000

International Nuclear Information System (INIS)

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

2013-01-01

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

Effectiveness of In-Vessel Retention Strategies and Minimum Safety Injection Flow over Postulated Severe Accidents of OPR1000

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

Evaluation of materials for retention of sodium and core debris in reactor systems. Annual progress report, September 1977-December 1978

International Nuclear Information System (INIS)

Swanson, D.G.; Zehms, E.H.; McClelland, J.D.; Meyer, R.A.; van Paassen, H.L.L.

1978-12-01

This report considers some of the consequences of a hypothetical core disruptive accident in a nuclear reactor. The interactions expected between molten core debris, liquid sodium, and materials that might be employed in an ex-vessel sacrificial-bed or in the reactor building are discussed. Experimental work performed for NRC by Sandia Laboratories and Hanford Engineering Development Laboratory on the interactions between liquid sodium and basalt concrete is reviewed. Studies of molten steel interactions with concrete at Sandia Laboratories and molten UO 2 interactions with concrete at The Aerospace Corporation are also discussed. The potential of MgO for use in core containment is discussed and refractory materials other than MgO are reviewed. Finally, results from earlier experiments with molten core debris and various materials performed at The Aerospace Corporation are presented

Orbital debris: a technical assessment

National Research Council Canada - National Science Library

Committee on Space Debris, National Research Council

..., and other debris created as a byproduct of space operations. Orbital Debris examines the methods we can use to characterize orbital debris, estimates the magnitude of the debris population, and assesses the hazard that this population poses to spacecraft...

Investigation of flow regime in debris bed formation behavior with nonspherical particles

Directory of Open Access Journals (Sweden)

Songbai Cheng

2018-02-01

Full Text Available It is important to clarify the characteristics of flow regimes underlying the debris bed formation behavior that might be encountered in core disruptive accidents of sodium-cooled fast reactors. Although in our previous publications, by applying dimensional analysis technique, an empirical model, with its reasonability confirmed over a variety of parametric conditions, has been successfully developed to predict the regime transition and final bed geometry formed, so far this model is restricted to predictions of debris mixtures composed of spherical particles. Focusing on this aspect, in this study a new series of experiments using nonspherical particles have been conducted. Based on the knowledge and data obtained, an extension scheme is suggested with the purpose of extending the base model to cover the particle-shape influence. Through detailed analyses and given our current range of experimental conditions, it is found that, by coupling the base model with this scheme, respectable agreement between experiments and model predictions for the regime transition can be achieved for both spherical and nonspherical particles. Knowledge and evidence from our work might be utilized for the future improvement of design of an in-vessel core catcher as well as the development and verification of sodium-cooled fast reactor severe accident analysis codes in China.

Space Debris & its Mitigation

Science.gov (United States)

Kaushal, Sourabh; Arora, Nishant

2012-07-01

Space debris has become a growing concern in recent years, since collisions at orbital velocities can be highly damaging to functioning satellites and can also produce even more space debris in the process. Some spacecraft, like the International Space Station, are now armored to deal with this hazard but armor and mitigation measures can be prohibitively costly when trying to protect satellites or human spaceflight vehicles like the shuttle. This paper describes the current orbital debris environment, outline its main sources, and identify mitigation measures to reduce orbital debris growth by controlling these sources. We studied the literature on the topic Space Debris. We have proposed some methods to solve this problem of space debris. We have also highlighted the shortcomings of already proposed methods by space experts and we have proposed some modification in those methods. Some of them can be very effective in the process of mitigation of space debris, but some of them need some modification. Recently proposed methods by space experts are maneuver, shielding of space elevator with the foil, vaporizing or redirecting of space debris back to earth with the help of laser, use of aerogel as a protective layer, construction of large junkyards around international space station, use of electrodynamics tether & the latest method proposed is the use of nano satellites in the clearing of the space debris. Limitations of the already proposed methods are as follows: - Maneuvering can't be the final solution to our problem as it is the act of self-defence. - Shielding can't be done on the parts like solar panels and optical devices. - Vaporizing or redirecting of space debris can affect the human life on earth if it is not done in proper manner. - Aerogel has a threshold limit up to which it can bear (resist) the impact of collision. - Large junkyards can be effective only for large sized debris. In this paper we propose: A. The Use of Nano Tubes by creating a mesh

Importance of the in and ex-vessel corium coolability in case of severe accident for the French PWRs. Some views from L2 PSA and perspectives

International Nuclear Information System (INIS)

Raimond, E.; Caroli, C.; Meignen, R.; Rahni, N.; Laurent, B.

2011-01-01

In the case of a severe accident on a NPP leading to core degradation after a default in the core cooling as during the accident of Three Mile Island (TMI2), the most efficient way to stop the accident progression would be the in-vessel water injection if a specific mean is available. The TMI2 accident has shown that the accident can be stopped and that the corium, even if highly degraded, can be cooled, but no one can generalize the TMI2 accident termination to all situations. The present paper aims at presenting the situation for the French operated PWRs and is mainly based on the IRSN experience in level 2 probabilistic safety assessment (L2 PSA) development for this type of reactor. It tries to highlight the benefit that could be obtained from a better understanding of the corium cooling phenomenology, including both possible positive and negative effects. Three main negative effects of in-vessel flooding have to be taken into account in a L2 PSA for a PWR: an increase of the hydrogen production rate, a risk of in-vessel pressure increase and the development of conditions for steam explosion. L2 PSAs in France have now reached a certain maturity allowing raising some more precise issues, but for the issues presented in this paper, some progress from the research-development and the simulation tools (mainly the ASTEC integral code) are still necessary to support decision-making

Testing plan for critical heat flux measurement during in-vessel retention

International Nuclear Information System (INIS)

Aoki, Kazuyoshi; Iwaki, Chikako; Sato, Hisaki; Mimura, Satoshi; Kanamori, Daisuke

2015-01-01

In-Vessel Retention (IVR) is a method to maintain molten debris in a reactor vessel (RV) by RV outer surface cooling. Structural integrity of RV and cooling capacity on RV outer surface are important to verify IVR strategy. Critical Heat Flux (CHF) data is necessary to estimate cooling capacity on the RV outer surface. And there are some CHF data to estimate cooling capacity on the RV outer surface. However, these data were obtained for specific plants. Thus, the objective of this study is developing a CHF correlation for various PWR plants. The objectives of this paper are developing test equipment and testing plan for the CHF correlation. Firstly, plant conditions during severe accidents were organized. Then, ranges of testing parameters were estimated with the plant conditions. And specifications of the test equipment were set to cover the range of parameters. Secondly, testing cases were set based on design of experiments. The test cases are suitable to develop experimental correlations. (author)

JSC Orbital Debris Website Description

Science.gov (United States)

Johnson, Nicholas L.

2006-01-01

Purpose: The website provides information about the NASA Orbital Debris Program Office at JSC, which is the lead NASA center for orbital debris research. It is recognized world-wide for its leadership in addressing orbital debris issues. The NASA Orbital Debris Program Office has taken the international lead in conducting measurements of the environment and in developing the technical consensus for adopting mitigation measures to protect users of the orbital environment. Work at the center continues with developing an improved understanding of the orbital debris environment and measures that can be taken to control its growth. Major Contents: Orbital Debris research is divided into the following five broad efforts. Each area of research contains specific information as follows: 1) Modeling - NASA scientists continue to develop and upgrade orbital debris models to describe and characterize the current and future debris environment. Evolutionary and engineering models are described in detail. Downloadable items include a document in PDF format and executable software. 2) Measurements - Measurements of near-Earth orbital debris are accomplished by conducting ground-based and space-based observations of the orbital debris environment. The data from these sources provide validation of the environment models and identify the presence of new sources. Radar, optical and surface examinations are described. External links to related topics are provided. 3) Protection - Orbital debris protection involves conducting hypervelocity impact measurements to assess the risk presented by orbital debris to operating spacecraft and developing new materials and new designs to provide better protection from the environment with less weight penalty. The data from this work provides the link between the environment defined by the models and the risk presented by that environment to operating spacecraft and provides recommendations on design and operations procedures to reduce the risk as

Debris flow-induced topographic changes: effects of recurrent debris flow initiation.

Science.gov (United States)

Chen, Chien-Yuan; Wang, Qun

2017-08-12

Chushui Creek in Shengmu Village, Nantou County, Taiwan, was analyzed for recurrent debris flow using numerical modeling and geographic information system (GIS) spatial analysis. The two-dimensional water flood and mudflow simulation program FLO-2D were used to simulate debris flow induced by rainfall during typhoon Herb in 1996 and Mindulle in 2004. Changes in topographic characteristics after the debris flows were simulated for the initiation of hydrological characteristics, magnitude, and affected area. Changes in topographic characteristics included those in elevation, slope, aspect, stream power index (SPI), topographic wetness index (TWI), and hypsometric curve integral (HI), all of which were analyzed using GIS spatial analysis. The results show that the SPI and peak discharge in the basin increased after a recurrence of debris flow. The TWI was higher in 2003 than in 2004 and indicated higher potential of landslide initiation when the slope of the basin was steeper. The HI revealed that the basin was in its mature stage and was shifting toward the old stage. Numerical simulation demonstrated that the parameters' mean depth, maximum depth, affected area, mean flow rate, maximum flow rate, and peak flow discharge were increased after recurrent debris flow, and peak discharge occurred quickly.

Quantitative assessment of apical debris extrusion and intracanal debris in the apical third, using hand instrumentation and three rotary instrumentation systems.

Science.gov (United States)

H K, Sowmya; T S, Subhash; Goel, Beena Rani; T N, Nandini; Bhandi, Shilpa H

2014-02-01

Decreased apical extrusion of debris and apical one third debris have strong implications for decreased incidence of postoperative inflammation and pain. Thus, the aim of this study was to assess quantitatively the apical extrusion of debris and intracanal debris in the apical third during root canal instrumentation using hand and three different types of rotary instruments. Sixty freshly extracted single rooted human teeth were randomly divided into four groups. Canal preparation was done using step-back with hand instrumentation, crown-down technique with respect to ProTaper and K3, and hybrid technique with LightSpeed LSX. Irrigation was done with NaOCl, EDTA, and normal saline and for final irrigation, EndoVac system was used. The apically extruded debris was collected on the pre-weighed Millipore plastic filter disk and weighed using microbalance. The teeth were submitted to the histological processing. Sections from the apical third were analyzed by a trinocular research microscope that was coupled to a computer where the images were captured and analyzed using image proplus V4.1.0.0 software. The mean weight of extruded debris for each group and intracanal debris in the root canal was statistically analyzed by a Kruskal-Wallis one-way analysis of variance and Mann-Whitney U test. The result showed that, hand instrumentation using K files showed the highest amount of debris extrusion apically when compared to ProTaper, K3 and LightSpeed LSX. The result also showed that there was no statistically significant difference between the groups in relation to presence of intracanal debris in the apical one third. Based on the results, all instrumentation techniques produced debris extrusion. The engine driven Ni-Ti systems extruded significantly less apical debris than hand instrumentation. There was no statistically significant difference between the groups in relation to presence of intracanal debris in the apical one third.

DebriSat Hypervelocity Impact Test

Science.gov (United States)

2015-08-01

public release; distribution unlimited.  Targets: Scaled Multishock Shield, DebrisLV, and DebriSat  500-600 g hollow aluminum and nylon projectile... insulation . DebriSat’s internal components were structurally similar to real flight hardware but were nonfunctional. AEDC-TR-15-S-2 6...structures with an AL 5052 honeycomb core and M55J carbon fiber face sheets. The basic system characteristics of the DebriSat are given in Table 1

A globally complete map of supraglacial debris cover and a new toolkit for debris cover research

Science.gov (United States)

Herreid, Sam; Pellicciotti, Francesca

2017-04-01

A growing canon of literature is focused on resolving the processes and implications of debris cover on glaciers. However, this work is often confined to a handful of glaciers that were likely selected based on criteria optimizing their suitability to test a specific hypothesis or logistical ease. The role of debris cover in a glacier system is likely to not go overlooked in forthcoming research, yet the magnitude of this role at a global scale has not yet been fully described. Here, we present a map of debris cover for all glacierized regions on Earth including the Greenland Ice Sheet using 30 m Landsat data. This dataset will begin to open a wider context to the high quality, localized findings from the debris-covered glacier research community and help inform large-scale modeling efforts. A global map of debris cover also facilitates analysis attempting to isolate first order geomorphological and climate controls of supraglacial debris production. Furthering the objective of expanding the inclusion of debris cover in forthcoming research, we also present an under development suite of open-source, Python based tools. Requiring minimal and often freely available input data, we have automated the mapping of: i) debris cover, ii) ice cliffs, iii) debris cover evolution over the Landsat era and iv) glacier flow instabilities from altered debris structures. At the present time, debris extent is the only globally complete quantity but with the expanding repository of high quality global datasets and further tool development minimizing manual tasks and computational cost, we foresee all of these tools being applied globally in the near future.

The physics of debris flows

Science.gov (United States)

Iverson, Richard M.

1997-08-01

Recent advances in theory and experimentation motivate a thorough reassessment of the physics of debris flows. Analyses of flows of dry, granular solids and solid-fluid mixtures provide a foundation for a comprehensive debris flow theory, and experiments provide data that reveal the strengths and limitations of theoretical models. Both debris flow materials and dry granular materials can sustain shear stresses while remaining static; both can deform in a slow, tranquil mode characterized by enduring, frictional grain contacts; and both can flow in a more rapid, agitated mode characterized by brief, inelastic grain collisions. In debris flows, however, pore fluid that is highly viscous and nearly incompressible, composed of water with suspended silt and clay, can strongly mediate intergranular friction and collisions. Grain friction, grain collisions, and viscous fluid flow may transfer significant momentum simultaneously. Both the vibrational kinetic energy of solid grains (measured by a quantity termed the granular temperature) and the pressure of the intervening pore fluid facilitate motion of grains past one another, thereby enhancing debris flow mobility. Granular temperature arises from conversion of flow translational energy to grain vibrational energy, a process that depends on shear rates, grain properties, boundary conditions, and the ambient fluid viscosity and pressure. Pore fluid pressures that exceed static equilibrium pressures result from local or global debris contraction. Like larger, natural debris flows, experimental debris flows of ˜10 m³ of poorly sorted, water-saturated sediment invariably move as an unsteady surge or series of surges. Measurements at the base of experimental flows show that coarse-grained surge fronts have little or no pore fluid pressure. In contrast, finer-grained, thoroughly saturated debris behind surge fronts is nearly liquefied by high pore pressure, which persists owing to the great compressibility and moderate

The physics of debris flows

Science.gov (United States)

Iverson, R.M.

1997-01-01

Recent advances in theory and experimentation motivate a thorough reassessment of the physics of debris flows. Analyses of flows of dry, granular solids and solid-fluid mixtures provide a foundation for a comprehensive debris flow theory, and experiments provide data that reveal the strengths and limitations of theoretical models. Both debris flow materials and dry granular materials can sustain shear stresses while remaining static; both can deform in a slow, tranquil mode characterized by enduring, frictional grain contacts; and both can flow in a more rapid, agitated mode characterized by brief, inelastic grain collisions. In debris flows, however, pore fluid that is highly viscous and nearly incompressible, composed of water with suspended silt and clay, can strongly mediate intergranular friction and collisions. Grain friction, grain collisions, and viscous fluid flow may transfer significant momentum simultaneously. Both the vibrational kinetic energy of solid grains (measured by a quantity termed the granular temperature) and the pressure of the intervening pore fluid facilitate motion of grains past one another, thereby enhancing debris flow mobility. Granular temperature arises from conversion of flow translational energy to grain vibrational energy, a process that depends on shear rates, grain properties, boundary conditions, and the ambient fluid viscosity and pressure. Pore fluid pressures that exceed static equilibrium pressures result from local or global debris contraction. Like larger, natural debris flows, experimental debris flows of ???10 m3 of poorly sorted, water-saturated sediment invariably move as an unsteady surge or series of surges. Measurements at the base of experimental flows show that coarse-grained surge fronts have little or no pore fluid pressure. In contrast, finer-grained, thoroughly saturated debris behind surge fronts is nearly liquefied by high pore pressure, which persists owing to the great compressibility and moderate

Long term integrity of reactor pressure vessel and primary containment vessel after the severe accidents in Fukushima Daiichi Nuclear Power Station. Leaching property of spent oxide fuel segment and corrosion property of a carbon steel under artificial seawater immersion

International Nuclear Information System (INIS)

2014-06-01

Primary containment vessel (PCV), reactor pressure vessel and pedestal in Fukushima Daiichi Nuclear power station units 1 through 3 have been exposed to severe thermal, chemical and mechanical conditions due to core meltdown events and seawater injections for emergent core cooling. These components will be immersed in diluted seawater with dissolved fission products under irradiation until the end of debris removal. Fresh water injected into the cores contacts with debris to cool, dissolves or erodes their constituents, mixed with retained water, and becomes 'accumulated water' with radioactive nuclides. We have focused the leaching of fission products into the accumulated water under lower temperature (323 K). FUGEN spent oxide fuel segments were immersed to determine the leaching factor of fission product and actinide elements. Since PCV made from carbon steel is one of the most important boundaries to prevent from fission products release, corrosion behavior has been paid attention to evaluate their integrity. Carbon steel specimens were immersion- and electrochemical-tested in diluted seawater with simulants of the accumulated water at 323 K in order to evaluate the effect of fission products in particular cesium and radiation. (author)

Particulate metallic debris in cemented total hip arthroplasty.

Science.gov (United States)

Salvati, E A; Betts, F; Doty, S B

1993-08-01

Several studies conducted by the authors in the last six years demonstrate that the generation of metallic debris is more severe with titanium alloy than with cobalt-chrome alloy femoral components in cemented total hip arthroplasty (THA). The debris is generated from the articulating surface, particularly if entrapped acrylic debris produces three-body wear, and from the stem surface when the component loosens and abrades against fragmented cement. In selected cases in which the titanium metallic debris is copious, premature failure and severe progressive bone loss occurs. Electron microscopy demonstrates that the particles of metallic debris can be extremely small (a few hundredths of 1 micron). They are phagocytized by the macrophages and transported to the phagolysosomes. In this highly corrosive environment, the very high surface area of the particles may release toxic concentrations of the constituents of the alloy intracellularly, probably leading to progressive cell degeneration and death, with subsequent release of intracellular enzymes and ingested metallic debris. This cycle most likely repeats itself, leading to tissue necrosis. The results presented do not support the use of titanium alloy femoral components for cemented THA, particularly for the articulating surface.

Identification and assessment of BWR in-vessel severe accident mitigation strategies

Energy Technology Data Exchange (ETDEWEB)

Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M. [Oak Ridge National Lab., TN (United States)

1992-10-01

This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners` Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored.

Identification and assessment of BWR in-vessel severe accident mitigation strategies

International Nuclear Information System (INIS)

Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M.

1992-10-01

This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

Alternative fuels in fire debris analysis: biodiesel basics.

Science.gov (United States)

Stauffer, Eric; Byron, Doug

2007-03-01

Alternative fuels are becoming more prominent on the market today and, soon, fire debris analysts will start seeing them in liquid samples or in fire debris samples. Biodiesel fuel is one of the most common alternative fuels and is now readily available in many parts of the United States and around the world. This article introduces biodiesel to fire debris analysts. Biodiesel fuel is manufactured from vegetable oils and/or animal oils/fats. It is composed of fatty acid methyl esters (FAMEs) and is sold pure or as a blend with diesel fuel. When present in fire debris samples, it is recommended to extract the debris using passive headspace concentration on activated charcoal, possibly followed by a solvent extraction. The gas chromatographic analysis of the extract is first carried out with the same program as for regular ignitable liquid residues, and second with a program adapted to the analysis of FAMEs.

Safety against releases in severe accidents. Final report

Energy Technology Data Exchange (ETDEWEB)

Lindholm, I.; Berg, Oe.; Nonboel, E. [eds.

1997-12-01

The work scope of the RAK-2 project has involved research on quantification of the effects of selected severe accident phenomena for Nordic nuclear power plants, development and testing of a computerised accident management support system and data collection and description of various mobile reactors and of different reactor types existing in the UK. The investigations of severe accident phenomena focused mainly on in-vessel melt progression, covering a numerical assessment of coolability of a degraded BWR core, the possibility and consequences of a BWR reactor to become critical during reflooding and the core melt behavior in the reactor vessel lower plenum. Simulant experiments were carried out to investigate lower head hole ablation induced by debris discharge. In addition to the in-vessel phenomena, a limited study on containment response to high pressure melt ejection in a BWR and a comparative study on fission product source term behaviour in a Swedish PWR were performed. An existing computerised accident management support system (CAMS) was further developed in the area of tracking and predictive simulation, signal validation, state identification and user interface. The first version of a probabilistic safety analysis module was developed and implemented in the system. CAMS was tested in practice with Barsebaeck data in a safety exercise with the Swedish nuclear authority. The descriptions of the key features of British reactor types, AGR, Magnox, FBR and PWR were published as data reports. Separate reports were issued also on accidents in nuclear ships and on description of key features of satellite reactors. The collected data were implemented in a common Nordic database. (au) 39 refs.

Coupled thermo-mechanical analysis of corium-loaded lower head of pressure vessel

International Nuclear Information System (INIS)

Mishra, J.; Balasubramaniyan, V.

2016-01-01

A severe accident in the pressurised water reactor may lead to the relocation of core materials to the lower head of Reactor Pressure Vessel (RPV). The core debris at the bottom of RPV forms a melt pool of corium due to decay heat. The understanding of behaviour of pressure vessel, characterised by failure mode and time to failure, in this scenario is one of the important steps in predicting the accident progression. The most predominant failure mode is multi-axial creep deformation of the vessel with a non-uniform temperature field. Towards this, a numerical analysis methodology is developed for the prediction of pressure vessel deformation during the severe accidents. The methodology involves 2-D finite element modelling under multi-physics environment, which account the creep phenomena using Norton-Bailey creep law with a typical damage model of RPV material. The validation of the methodology is carried out using the results from OLHF experiment carried out in Sandia National Laboratory (SNL), USA, within the framework of an OECD. (author)

Detecting debris flows using ground vibrations

Science.gov (United States)

LaHusen, Richard G.

1998-01-01

Debris flows are rapidly flowing mixtures of rock debris, mud, and water that originate on steep slopes. During and following volcanic eruptions, debris flows are among the most destructive and persistent hazards. Debris flows threaten lives and property not only on volcanoes but far downstream in valleys that drain volcanoes where they arrive suddenly and inundate entire valley bottoms. Debris flows can destroy vegetation and structures in their path, including bridges and buildings. Their deposits can cover roads and railways, smother crops, and fill stream channels, thereby reducing their flood-carrying capacity and navigability.

Autogenic dynamics of debris-flow fans

Science.gov (United States)

van den Berg, Wilco; de Haas, Tjalling; Braat, Lisanne; Kleinhans, Maarten

2015-04-01

Alluvial fans develop their semi-conical shape by cyclic avulsion of their geomorphologically active sector from a fixed fan apex. These cyclic avulsions have been attributed to both allogenic and autogenic forcings and processes. Autogenic dynamics have been extensively studied on fluvial fans through physical scale experiments, and are governed by cyclic alternations of aggradation by unconfined sheet flow, fanhead incision leading to channelized flow, channel backfilling and avulsion. On debris-flow fans, however, autogenic dynamics have not yet been directly observed. We experimentally created debris-flow fans under constant extrinsic forcings, and show that autogenic dynamics are a fundamental intrinsic process on debris-flow fans. We found that autogenic cycles on debris-flow fans are driven by sequences of backfilling, avulsion and channelization, similar to the cycles on fluvial fans. However, the processes that govern these sequences are unique for debris-flow fans, and differ fundamentally from the processes that govern autogenic dynamics on fluvial fans. We experimentally observed that backfilling commenced after the debris flows reached their maximum possible extent. The next debris flows then progressively became shorter, driven by feedbacks on fan morphology and flow-dynamics. The progressively decreasing debris-flow length caused in-channel sedimentation, which led to increasing channel overflow and wider debris flows. This reduced the impulse of the liquefied flow body to the flow front, which then further reduced flow velocity and runout length, and induced further in-channel sedimentation. This commenced a positive feedback wherein debris flows became increasingly short and wide, until the channel was completely filled and the apex cross-profile was plano-convex. At this point, there was no preferential transport direction by channelization, and the debris flows progressively avulsed towards the steepest, preferential, flow path. Simultaneously

Space debris: modeling and detectability

Science.gov (United States)

Wiedemann, C.; Lorenz, J.; Radtke, J.; Kebschull, C.; Horstmann, A.; Stoll, E.

2017-01-01

High precision orbit determination is required for the detection and removal of space debris. Knowledge of the distribution of debris objects in orbit is necessary for orbit determination by active or passive sensors. The results can be used to investigate the orbits on which objects of a certain size at a certain frequency can be found. The knowledge of the orbital distribution of the objects as well as their properties in accordance with sensor performance models provide the basis for estimating the expected detection rates. Comprehensive modeling of the space debris environment is required for this. This paper provides an overview of the current state of knowledge about the space debris environment. In particular non-cataloged small objects are evaluated. Furthermore, improvements concerning the update of the current space debris model are addressed. The model of the space debris environment is based on the simulation of historical events, such as fragmentations due to explosions and collisions that actually occurred in Earth orbits. The orbital distribution of debris is simulated by propagating the orbits considering all perturbing forces up to a reference epoch. The modeled object population is compared with measured data and validated. The model provides a statistical distribution of space objects, according to their size and number. This distribution is based on the correct consideration of orbital mechanics. This allows for a realistic description of the space debris environment. Subsequently, a realistic prediction can be provided concerning the question, how many pieces of debris can be expected on certain orbits. To validate the model, a software tool has been developed which allows the simulation of the observation behavior of ground-based or space-based sensors. Thus, it is possible to compare the results of published measurement data with simulated detections. This tool can also be used for the simulation of sensor measurement campaigns. It is

Anthropogenic debris in the nests of kelp gulls in South Africa.

Science.gov (United States)

Witteveen, Minke; Brown, Mark; Ryan, Peter G

2017-01-30

Anthropogenic debris results in detrimental interactions with many marine species. Several seabirds include debris items in their nests, which can lead to entanglement of chicks and adults, resulting in injury or death. Anthropogenic debris was found in 4-67% of kelp gull Larus dominicanus nests in seven colonies in the Western Cape, South Africa. Nests contained two types of litter: items included in the nest structure during construction (mainly ropes and straps), and regurgitated items (mainly bags and food wrappers) that probably accumulate primarily during the chick-rearing period. Debris used in nest construction was more likely to injure gulls, and was found mainly at coastal sites where there was little natural vegetation for construction. Distance to the nearest urban waste landfill significantly affected the occurrence of debris items in nests, especially dietary-derived items. The amount of debris in kelp gull nests highlights the need for improved debris management in South Africa. Copyright © 2016 Elsevier Ltd. All rights reserved.

Debris Avalanches and Debris Flows Transformed from Collapses in the Trans-Mexican Volcanic Belt, México.

Science.gov (United States)

Capra, L.; Macias, J.; Scott, K.; Abrams, M.; Garduño, V.

2001-12-01

Volcanoes of the Trans-Mexican Volcanic Belt (TMVB) have yielded numerous sector and flank collapses during Pleistocene and Holocene time. Sector collapses associated with magmatic activity have yielded debris avalanches with generally limited runout extent (e.g. Popocatépetl, Jocotitlán, and Colima volcanoes). In contrast, flank collapses (smaller failures not involving the volcano summit), both associated and unassociated with magmatic activity and correlated with intense hydrothermal alteration in ice-capped volcanoes, commonly have yielded highly mobile cohesive debris flows (e.g. Pico de Orizaba and Nevado de Toluca volcanoes). Collapse orientation in the TMVB is preferentially to the south and north-east, probably reflecting the tectonic regime of active E-W and NNW faults. The different mobilities of the flows transformed from collapses have important implications for hazard assessment. Both sector and flank collapse can yield highly mobile debris flows, but this transformation is more common in the case of the smaller failures. High mobility is related to factors such as water and clay content of the failed material, the paleotopography, and the extent of entrainment of sediment during flow (bulking). Both debris-avalanches and debris-flows are volcanic hazards that occur from both active volcanoes, as well as those that are inactive or dormant volcanoes, and may by triggered by earthquakes, precipitation, or simple gravity. There will be no precursory warning in such non-volcanic cases.

Debris flows associated with the 2015 Gorkha Earthquake in Nepal

Science.gov (United States)

Dahlquist, M. P.; West, A. J.; Martinez, J.

2017-12-01

Debris flows are a primary driver of erosion and a major geologic hazard in many steep landscapes, particularly near the headwaters of rivers, and are generated in large numbers by extreme events. The 2015 Mw 7.8 Gorkha Earthquake triggered 25,000 coseismic landslides in central Nepal. During the ensuing monsoon, sediment delivered to channels by landslides was mobilized in the heavy rains, and new postseismic landslides were triggered in rock weakened by the shaking. These coseismic and postseismic landslide-generated debris flows form a useful dataset for studying the impact and behavior of debris flows on one of the most active landscapes on Earth. Debris flow-dominated channel reaches are generally understood to have a topographic signature recognizable in slope-area plots and distinct from fluvial channels, but in examining debris flows associated with the Gorkha earthquake we find they frequently extend into reaches with geometry typically associated with fluvial systems. We examine a dataset of these debris flows, considering whether they are generated by coseismic or postseismic landslides, whether they are likely to be driving active incision into bedrock, and whether their channels correspond with those typically associated with debris flows. Preliminary analysis of debris flow channels in Nepal suggests there may be systematic differences in the geometry of channels containing debris flows triggered by coseismic versus postseismic landslides, which potentially holds implications for hazard analyses and the mechanics behind the different debris flow types.

Safe disposal and recycling of water disaster debris in pakistan

International Nuclear Information System (INIS)

Latif, A.

2014-01-01

Depending upon the nature, the disaster may produce large masses of debris. Waste masses from single disaster integrate to larger magnitude annually. This will ultimately causes the extra work load on personnel and reflects the poor existing debris management facilities. Besides, it will take longer time to rehabilitate the debris exaggerated regions. The study focuses on 2 main cases of disaster i.e. earthquake of 2005 and flood of 2010 in Pakistan. Complete analysis involve two stages: the first stage involve development of disaster and disaster debris effects guidance whereas the second stage involves the development of set of criteria to make efficient environment and positive impacts of successful debris managing scheme. Such principles were employed to evaluate efficiency of debris managing scheme for detailed analysis. The discussion of the detailed analysis depicts methodology which assists the disaster managers, planners and researcher to simply multitude of work. Moreover, the disaster and disaster debris influence direction, the effect evaluation criterion and managing criteria have been established having the effect they can be virtually put into service for prospect debris managing scheme, planning and retort. With respect to character and strictness, calamity may make high magnitude of waste. By keeping in view the precedent calamities in the United States (US), concluded that in few situations produced waste masses approximately five to fifteen times more than yearly waste production rate from a single occasion. Same results were revealed by subsequent tsunami of Indian Ocean. Such kind of large masses may effects the existing solid debris management system and human resources. Major disaster yields large masses of debris in few hours or sometimes even in minutes. The volume of disaster debris depends upon the magnitude of trees ball up, indemnity to houses, business, services etc. The disaster remaining may be equally large in metropolitan and non

Analyses of fluid flow and heat transfer inside calandria vessel of CANDU-6 reactor using CFD

International Nuclear Information System (INIS)

Yu, Seon Oh; Kim, Man Woong; Kim, Hho Jung

2005-01-01

In a CANDU (CANada Deuterium Uranium) reactor, fuel channel integrity depends on the coolability of the moderator as an ultimate heat sink under transient conditions such as a Loss Of Coolant Accident (LOCA) with coincident Loss Of Emergency Core Cooling (LOECC). as well as normal operating conditions. This study presents assessments of moderator thermal-hydraulic characteristics in the normal operating conditions and one transient condition for CANDU-6 reactors, using a general purpose three-dimensional computational fluid dynamics code. First, an optimized calculation scheme is obtained by many-sided comparisons of the predicted results with the related experimental data, and by evaluating the fluid flow and temperature distributions. Then, using the optimized scheme, analyses of real CANDU-6 in normal operating conditions and the transition condition have been performed. The present model successfully predicted the experimental results and also reasonably assessed the thermal-hydraulic characteristics of a real CANDU-6 with 380 fuel channels. A flow regime map with major parameters representing the flow pattern inside a calandria vessel has also proposed to be used as operational and/or regulatory guidelines

Colisional Cloud Debris and Propelled Evasive Maneuvers

Science.gov (United States)

Ferreira, L. S.; Jesus, A. D. C.; Carvalho, T. C. F.; Sousa, R. R.

2017-10-01

Space debris clouds exist at various altitudes in the environment outside the Earth. Fragmentation of debris and/or collision between the debris of a cloud increases the amount of debris, producing smaller debris. This event also increases significantly the chances of collision with operational vehicles in orbit. In this work we study clouds of debris that are close to a spacecraft in relation to its distance from the center of the Earth. The results show several layers of colliding debris depending on their size over time of evasive maneuvers of the vehicle. In addition, we have tested such maneuvers for propulsion systems with a linear and exponential mass variation model. The results show that the linear propulsion system is more efficient.

Marine debris removal: one year of effort by the Georgia Sea Turtle-Center-Marine Debris Initiative.

Science.gov (United States)

Martin, Jeannie Miller

2013-09-15

Once in the marine environment, debris poses a significant threat to marine life that can be prevented through the help of citizen science. Marine debris is any manufactured item that enters the ocean regardless of source, commonly plastics, metal, wood, glass, foam, cloth, or rubber. Citizen science is an effective way to engage volunteers in conservation initiatives and provide education and skill development. The Georgia Sea Turtle Center Marine Debris Initiative (GSTC-MDI) is a grant funded program developed to engage citizens in the removal of marine debris from the beaches of Jekyll Island, GA, USA and the surrounding areas. During the first year of effort, more than 200 volunteers donated over 460 h of service to the removal of marine debris. Of the debris removed, approximately 89% were plastics, with a significant portion being cigarette materials. Given the successful first year, the GSTC-MDI was funded again for a second year. Copyright © 2013 Elsevier Ltd. All rights reserved.

Severe accident management: a summary of the VAHTI and ROIMA projects

International Nuclear Information System (INIS)

Sairanen, R.

1998-01-01

Two severe accident research projects: 'Severe Accident Management' (VAHTI), 1994-96 and 'Reactor Accidents' Phenomena and Simulation (ROIMA) 1997-98. have been conducted at VTT Energy within the RETU research programme. The main objective was to assist the severe accident management programmes of the Finnish nuclear power plants. The projects had several subtopics. These included thermal hydraulic validation of the APROS code, studies of failure mode of the BWR pressure vessel, investigation of core melt progression within a BWR pressure vessel, containment phenomena, development of a computerised severe accident training tool, and aerosol behaviour experiments. The last topic is summarised by another paper in the seminar. The projects have met the objectives set at the project commencement. Calculation tools have been developed and validated suitable for analyses of questions specific for the Finnish plants. Experimental fission product data have been produced that can be used to validate containment aerosol codes. The tools and results have been utilised in plant assessments. One of the main achievements has been the computer code PASULA for analysis of interactions between core melt and pressure vessel. The code has been applied to pressure vessel penetration analysis. The results have shown the importance of the nozzle construction. Modelling possibilities have recently improved by addition of a creep and porous debris models. Cooling of a degraded BWR core has been systematically studied as joint Nordic projects with a set of severe accident codes. Estimates for coolable conditions have been provided. Recriticality due to reflooding of a damaged core has been evaluated. (orig.)

Backwater development by woody debris

Science.gov (United States)

Geertsema, Tjitske; Torfs, Paul; Teuling, Ryan; Hoitink, Ton

2017-04-01

Placement of woody debris is a common method for increasing ecological values in river and stream restoration, and is thus widely used in natural environments. Water managers, however, are afraid to introduce wood in channels draining agricultural and urban areas. Upstream, it may create backwater, depending on hydrodynamic characteristics including the obstruction ratio, the Froude number and the surface level gradient. Patches of wood may trigger or counter morphological activity, both laterally, through bank erosion and protection, and vertically, with pool and riffle formation. Also, a permeable construction composed of wood will weather over time. Both morphodynamic activity and weathering cause backwater effects to change in time. The purpose of this study is to quantify the time development of backwater effects caused by woody debris. Hourly water levels gauged upstream and downstream of patches and discharge are collected for five streams in the Netherlands. The water level drop over the woody debris patch relates to discharge in the streams. This relation is characterized by an increasing water level difference for an increasing discharge, up to a maximum. If the discharge increases beyond this level, the water level difference reduces to the value that may represent the situation without woody debris. This reduction depends primarily on the obstruction ratio of the woody debris in the channel cross-section. Morphologic adjustments in the stream and reorientation of the woody material reduce the water level drop over the patches in time. Our results demonstrate that backwater effects can be reduced by optimizing the location where woody debris is placed and manipulating the obstruction ratio. Current efforts are focussed on representing woody debris in a one-dimensional numerical model, aiming to obtain a generic tool to achieve a stream design with woody debris that minimizes backwater.

Demonstration of laser processing technique combined with water jet technique for retrieval of fuel debris at Fukushima Daiichi Nuclear Power Station

International Nuclear Information System (INIS)

Hanari, Toshihide; Takebe, Toshihiko; Yamada, Tomonori; Daido, Hiroyuki; Ishizuka, Ippei; Ohmori, Shinya; Kurosawa, Koichi; Sasaki, Go; Nakada, Masahiro; Sakai, Hideaki

2017-01-01

In decommissioning of Fukushima Daiichi Nuclear Power Station, a retrieval process of fuel debris in the Primary Containment Vessel by a remote operation is one of the key issues. In this process, prevention of spreading radioactive materials is one of the important considerations. Furthermore, an applicable technique to the process requires keeping of reasonable processing-efficiency. We propose to use the combined technique including a laser light and a water jet as a retrieval technique of the fuel debris. The laser processing technique combined with a repetitive pulsed water jet could perform an efficient retrieval processing. Our experimental result encourages us to promote further development of the technique towards a real application at Fukushima Daiichi Nuclear Power Station. (author)

An Ontological Architecture for Orbital Debris Data

OpenAIRE

Rovetto, Robert J.

2017-01-01

The orbital debris problem presents an opportunity for inter-agency and international cooperation toward the mutually beneficial goals of debris prevention, mitigation, remediation, and improved space situational awareness (SSA). Achieving these goals requires sharing orbital debris and other SSA data. Toward this, I present an ontological architecture for the orbital debris domain, taking steps in the creation of an orbital debris ontology (ODO). The purpose of this ontological system is to ...

Active Space Debris Removal System

Directory of Open Access Journals (Sweden)

Gabriele GUERRA

2017-06-01

Full Text Available Since the start of the space era, more than 5000 launches have been carried out, each carrying satellites for many disparate uses, such as Earth observation or communication. Thus, the space environment has become congested and the problem of space debris is now generating some concerns in the space community due to our long-lived belief that “space is big”. In the last few years, solutions to this problem have been proposed, one of those is Active Space Debris Removal: this method will reduce the increasing debris growth and permit future sustainable space activities. The main idea of the method proposed below is a drag augmentation system: use a system capable of putting an expanded foam on a debris which will increase the area-to-mass ratio to increase the natural atmospheric drag and solar pressure. The drag augmentation system proposed here requires a docking system; the debris will be pushed to its release height and then, after un-docking, an uncontrolled re-entry takes place ending with a burn up of the object and the foam in the atmosphere within a given time frame. The method requires an efficient way to change the orbit between two debris. The present paper analyses such a system in combination with an Electric Propulsion system, and emphasizes the choice of using two satellites to remove five effective rockets bodies debris within a year.

Spiders (Araneae of stony debris in North Bohemia

Directory of Open Access Journals (Sweden)

Růžička, Vlastimil

1996-12-01

Full Text Available The arachnofauna was studied at five stony debris sites in northern Bohemia. In Central Europe, the northern and montane species inhabiting cold places live not only on mountain tops and peat bogs but also on the lower edges of boulder debris, where air streaming through the system of inner compartments gives rise to an exceedingly cold microclimate. At such cold sites, spiders can live either on bare stones (Bathyphantes simillimus, Wubanoides uralensis, or in the rich layers of moss and lichen (Diplocentria bidentata. Kratochviliella bicapitata exhibits a diplostenoecious occurence in stony debris and on the tree bark. Latithorax faustus and Theonoe minutissima display diplostenoecious occurence in stony debris and on peat bogs. The occurence of the species Scotina celans in the Czech Republic was documented for the first time.

Debris filtering effectiveness and pressure drop tests of debris resistance-bottom end piece

International Nuclear Information System (INIS)

Chung, Moon Ki; Song, Chul Hwa; Chung, Heung June; Won, Soon Yeun; Cho, Young Ro; Kim, Bok Deuk

1992-03-01

In this final report, described are the test conditions and test procedures for the debris filtering effectiveness and pressure drop tests for developing the Debris Resistance-Bottom End Piece (DR-BEP). And the test results are tabulated for later evaluation. (Author)

Summary of Disposable Debris Shields (DDS) Analysis for Development of Solid Debris Collection at NIF

International Nuclear Information System (INIS)

Shaughnessy, D.A.; Moody, K.J.; Grant, P.M.; Lewis, L.A.; Hutcheon, I.D.; Lindvall, R.; Gostic, J.M.

2011-01-01

Collection of solid debris from the National Ignition Facility (NIF) is being developed both as a diagnostic tool and as a means for measuring nuclear reaction cross sections relevant to the Stockpile Stewardship Program and nuclear astrophysics. The concept is straightforward; following a NIF shot, the debris that is produced as a result of the capsule and hohlraum explosion would be collected and subsequently extracted from the chamber. The number of nuclear activations that occurred in the capsule would then be measured through a combination of radiation detection and radiochemical processing followed by mass spectrometry. Development of the catcher is challenging due to the complex environment of the NIF target chamber. The collector surface is first exposed to a large photon flux, followed by the debris wind that is produced. The material used in the catcher must be mechanically strong in order to withstand the large amount of energy it is exposed to, as well as be chemically compatible with the form and composition of the debris. In addition, the location of the catcher is equally important. If it is positioned too close to the center of the target chamber, it will be significantly ablated, which could interfere with the ability of the debris to reach the surface and stick. If it is too far away, the fraction of the debris cloud collected will be too small to result in a statistically significant measurement. Material, geometric configuration, and location must all be tested in order to design the optimal debris collection system for NIF. One of the first ideas regarding solid debris collection at NIF was to use the disposable debris shields (DDS), which are fielded over the final optics assemblies (FOA) 7 m away from the center of the target chamber. The DDS are meant to be replaced after a certain number of shots, and if the shields could be subsequently analyzed after removal, it would serve as a mechanism for fielding a relatively large collection area

Phase shifting-based debris effect detection in USV-assisted AFM nanomachining

Energy Technology Data Exchange (ETDEWEB)

Shi, Jialin [State Key Laboratory of Robotics, Shenyang Institute of Automation, Chinese Academy of Sciences (CAS), Shenyang, Liaoning 110016 (China); University of the Chinese Academy of Sciences, Beijing 100049 (China); Beijing Advanced Innovation Center for Imaging Technology, Capital Normal University, Beijing 100049 (China); Liu, Lianqing, E-mail: lianqingliu@sia.cn [State Key Laboratory of Robotics, Shenyang Institute of Automation, Chinese Academy of Sciences (CAS), Shenyang, Liaoning 110016 (China); Yu, Peng; Cong, Yang [State Key Laboratory of Robotics, Shenyang Institute of Automation, Chinese Academy of Sciences (CAS), Shenyang, Liaoning 110016 (China); Li, Guangyong [Department of Electrical and Computer Engineering, University of Pittsburgh, Pittsburgh, PA 15213 (United States)

2017-08-15

Highlights: • The mechanism of the debris effect on machining depth in force control mode operation is analyzed. • The relationship between phase shifting and pile-up of debris is investigated. • The phase shifting-based method is hardly affected by the pile-up of debris. • Debris effect detection by phase shifting-based method is achived. - Abstract: Atomic force microscopy (AFM) mechanical-based lithography attracts much attention in nanomanufacturing due to its advantages of low cost, high precision and high resolution. However, debris effects during mechanical lithography often lead to an unstable machining process and inaccurate results, which limits further applications of AFM-based lithography. There is a lack of a real-time debris detection approach, which is the prerequisite to eventually eliminating the influence of the debris, and of a method that can solve the above problems well. The ultrasonic vibration (USV)-assisted AFM has the ability to sense the machining depth in real time by detecting the phase shifting of cantilever. However, whether the pile-up of debris affect the phase response of cantilever is still lack of investigation. Therefore, we analyzed the mechanism of the debris effect on force control mode and investigated the relationship between phase shifting and pile-up of debris. Theoretical analysis and experimental results reveal that the pile-up of debris have negligible effect on phase shifting of cantilever. Therefore, the phase shifting-based method can detect the debris effect on machining depth in force control mode of AFM machining.

NASA's New Orbital Debris Engineering Model, ORDEM2010

Science.gov (United States)

Krisko, Paula H.

2010-01-01

This paper describes the functionality and use of ORDEM2010, which replaces ORDEM2000, as the NASA Orbital Debris Program Office (ODPO) debris engineering model. Like its predecessor, ORDEM2010 serves the ODPO mission of providing spacecraft designers/operators and debris observers with a publicly available model to calculate orbital debris flux by current-state-of-knowledge methods. The key advance in ORDEM2010 is the input file structure of the yearly debris populations from 1995-2035 of sizes 10 micron - 1 m. These files include debris from low-Earth orbits (LEO) through geosynchronous orbits (GEO). Stable orbital elements (i.e., those that do not randomize on a sub-year timescale) are included in the files as are debris size, debris number, material density, random error and population error. Material density is implemented from ground-test data into the NASA breakup model and assigned to debris fragments accordingly. The random and population errors are due to machine error and uncertainties in debris sizes. These high-fidelity population files call for a much higher-level model analysis than what was possible with the populations of ORDEM2000. Population analysis in the ORDEM2010 model consists of mapping matrices that convert the debris population elements to debris fluxes. One output mode results in a spacecraft encompassing 3-D igloo of debris flux, compartmentalized by debris size, velocity, pitch, and yaw with respect to spacecraft ram direction. The second output mode provides debris flux through an Earth-based telescope/radar beam from LEO through GEO. This paper compares the new ORDEM2010 with ORDEM2000 in terms of processes and results with examples of specific orbits.

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

Small satellites and space debris issues

Science.gov (United States)

Yakovlev, M.; Kulik, S.; Agapov, V.

2001-10-01

The objective of this report is the analysis of the tendencies in designing of small satellites (SS) and the effect of small satellites on space debris population. It is shown that SS to include nano- and pico-satellites should be considered as a particularly dangerous source of space debris when elaborating international standards and legal documents concerning the space debris problem, in particular "International Space Debris Mitigation Standard". These issues are in accordance with the IADC goals in its main activity areas and should be carefully considered within the IADC framework.

Monitoring the abundance of plastic debris in the marine environment.

Science.gov (United States)

Ryan, Peter G; Moore, Charles J; van Franeker, Jan A; Moloney, Coleen L

2009-07-27

Plastic debris has significant environmental and economic impacts in marine systems. Monitoring is crucial to assess the efficacy of measures implemented to reduce the abundance of plastic debris, but it is complicated by large spatial and temporal heterogeneity in the amounts of plastic debris and by our limited understanding of the pathways followed by plastic debris and its long-term fate. To date, most monitoring has focused on beach surveys of stranded plastics and other litter. Infrequent surveys of the standing stock of litter on beaches provide crude estimates of debris types and abundance, but are biased by differential removal of litter items by beachcombing, cleanups and beach dynamics. Monitoring the accumulation of stranded debris provides an index of debris trends in adjacent waters, but is costly to undertake. At-sea sampling requires large sample sizes for statistical power to detect changes in abundance, given the high spatial and temporal heterogeneity. Another approach is to monitor the impacts of plastics. Seabirds and other marine organisms that accumulate plastics in their stomachs offer a cost-effective way to monitor the abundance and composition of small plastic litter. Changes in entanglement rates are harder to interpret, as they are sensitive to changes in population sizes of affected species. Monitoring waste disposal on ships and plastic debris levels in rivers and storm-water runoff is useful because it identifies the main sources of plastic debris entering the sea and can direct mitigation efforts. Different monitoring approaches are required to answer different questions, but attempts should be made to standardize approaches internationally.

Investigation of Orbital Debris: Mitigation, Removal, and Modeling the Debris Population

Science.gov (United States)

Slotten, Joel

The population of objects in orbit around Earth has grown since the late 1950s. Today there are over 21,000 objects over 10 cm in length in orbit, and an estimated 500,000 more between 1 and 10 cm. Only a small fraction of these objects are operational satellites. The rest are debris: old derelict spacecraft or rocket bodies, fragments created as the result of explosions or collisions, discarded objects, slag from solid rockets, or even flaked off paint. Traveling at up to 7 km/s, a collision with even a 1 cm piece of debris could severely damage or destroy a satellite. This dissertation examines three aspects of orbital debris. First, the concept of a self-consuming satellite is explored. This nanosatellite would use its own external structure as propellant to execute a deorbit maneuver at the end of its operational life, thus allowing it to meet current debris mitigation standards. Results from lab experiments examining potential materials for this concept have shown favorable results. Second, Particle in Cell techniques are modified and used to model the plasma plume from a micro-cathode arc thruster. This model is then applied to the concept of an ion beam shepherd satellite. This satellite would use its plasma plume to deorbit another derelict satellite. Results from these simulations indicate the micro-cathode arc thruster could potentially deorbit a derelict CubeSat in a matter of a few weeks. Finally, the orbital debris population at geosynchronous orbit is examined, focusing on variations in the density of the population as a function of longitude. New insights are revealed demonstrating that the variation in population density is slightly less than previously reported.

Space debris mitigation - engineering strategies

Science.gov (United States)

Taylor, E.; Hammond, M.

The problem of space debris pollution is acknowledged to be of growing concern by space agencies, leading to recent activities in the field of space debris mitigation. A review of the current (and near-future) mitigation guidelines, handbooks, standards and licensing procedures has identified a number of areas where further work is required. In order for space debris mitigation to be implemented in spacecraft manufacture and operation, the authors suggest that debris-related criteria need to become design parameters (following the same process as applied to reliability and radiation). To meet these parameters, spacecraft manufacturers and operators will need processes (supported by design tools and databases and implementation standards). A particular aspect of debris mitigation, as compared with conventional requirements (e.g. radiation and reliability) is the current and near-future national and international regulatory framework and associated liability aspects. A framework for these implementation standards is presented, in addition to results of in-house research and development on design tools and databases (including collision avoidance in GTO and SSTO and evaluation of failure criteria on composite and aluminium structures).

Development of debris-resistant bottom end piece

International Nuclear Information System (INIS)

Sohn, Dong Seong; Lee, Jae Kyung; Hwang, Dae Hyun; Yim, Jung Sik; Song, Kee Nam; Oh, Dong Seok; Im, Hyun Tae

1993-01-01

Debris-related fuel failures has been identified to be one of the major causes of fuel failures recently occured in nuclear power plants. In order to reduce the possibility of debris-related fuel failures, it is necessary to prevent the debris from reaching to fuel rods. In this regard, it is important to develop Debris-Resistant Bottom End Piece. (Author)

Classification of debris flow phenomena in the Faroe Islands

DEFF Research Database (Denmark)

Dahl, Mads-Peter Jakob; E. Mortensen, Lis; Jensen, Niels H.

2012-01-01

Landslides and debris flow phenomena in particular constitute a threat to human activities in the Faroe Islands. As a contribution to ongoing landslide risk management research, this paper proposes a classification scheme for debris flow phenomena in the Faroe Islands. The scheme, produced through...... a multidisciplinary study involving geomorphological fieldwork and qualitative collection of indigenous landslide knowledge, presents physical characteristics to classify debris flow phenomena into groups named with Faroese terms. The following landslide definitions are proposed. Brekku-skriðulop (English translation...... with international landslide classification systems, significantly increases the knowledge of debris flow phenomena and promotes a consistent terminology of these within the Faroe Islands....