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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

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

Experiment on heat transfer in simulated molten core/concrete interaction

International Nuclear Information System (INIS)

Katsumura, Yukihiro; Hashizume, Hidetoshi; Toda, Saburo; Kawaguchi, Takahiro.

1993-01-01

In order to investigate heat transfer between molten core and concrete in LWR severe accidents, experiments were performed using water as the molten core, paraffin as the concrete, and air as gases from the decomposition of concrete. It was found that the heat transfer on the interface between paraffin and water were promoted strongly by the air gas. (author)

Candidate molten salt investigation for an accelerator driven subcritical core

International Nuclear Information System (INIS)

Sooby, E.; Baty, A.; Beneš, O.; McIntyre, P.; Pogue, N.; Salanne, M.; Sattarov, A.

2013-01-01

Highlights: • Developing accelerator driven subcritical fission to destroy transuranics in SNF. • The core is a vessel containing a molten mixture of NaCl and transuranic chlorides. • Molecular dynamics used to calculate the thermophysical properties of the salt. • Density and molecular structure for actinide salts reported here. • The neutronics of ADS fission in molten salt are presented. -- Abstract: We report a design for accelerator-driven subcritical fission in a molten salt core (ADSMS) that utilizes a fuel salt composed of NaCl and transuranic (TRU) chlorides. The ADSMS core is designed for fast neutronics (28% of neutrons >1 MeV) to optimize TRU destruction. The choice of a NaCl-based salt offers benefits for corrosion, operating temperature, and actinide solubility as compared with LiF-based fuel salts. A molecular dynamics (MD) code has been used to estimate properties of the molten salt system which are important for ADSMS design but have never been measured experimentally. Results from the MD studies are reported. Experimental measurements of fuel salt properties and studies of corrosion and radiation damage on candidate metals for the core vessel are anticipated

Candidate molten salt investigation for an accelerator driven subcritical core

Energy Technology Data Exchange (ETDEWEB)

Sooby, E., E-mail: soobyes@tamu.edu [Texas A and M University, Accelerator Research Laboratory, 3380 University Dr. East, College Station, TX 77845 (United States); Baty, A. [Texas A and M University, Accelerator Research Laboratory, 3380 University Dr. East, College Station, TX 77845 (United States); Beneš, O. [European Commission, DG Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); McIntyre, P.; Pogue, N. [Texas A and M University, Accelerator Research Laboratory, 3380 University Dr. East, College Station, TX 77845 (United States); Salanne, M. [Université Pierre et Marie Curie, CNRS, Laboratoire PECSA, F-75005 Paris (France); Sattarov, A. [Texas A and M University, Accelerator Research Laboratory, 3380 University Dr. East, College Station, TX 77845 (United States)

2013-09-15

Highlights: • Developing accelerator driven subcritical fission to destroy transuranics in SNF. • The core is a vessel containing a molten mixture of NaCl and transuranic chlorides. • Molecular dynamics used to calculate the thermophysical properties of the salt. • Density and molecular structure for actinide salts reported here. • The neutronics of ADS fission in molten salt are presented. -- Abstract: We report a design for accelerator-driven subcritical fission in a molten salt core (ADSMS) that utilizes a fuel salt composed of NaCl and transuranic (TRU) chlorides. The ADSMS core is designed for fast neutronics (28% of neutrons >1 MeV) to optimize TRU destruction. The choice of a NaCl-based salt offers benefits for corrosion, operating temperature, and actinide solubility as compared with LiF-based fuel salts. A molecular dynamics (MD) code has been used to estimate properties of the molten salt system which are important for ADSMS design but have never been measured experimentally. Results from the MD studies are reported. Experimental measurements of fuel salt properties and studies of corrosion and radiation damage on candidate metals for the core vessel are anticipated.

Impact on breeding rate of different Molten Salt reactor core structures

International Nuclear Information System (INIS)

Wang Haiwei; Mei Longwei; Cai Xiangzhou; Chen Jingen; Guo Wei; Jiang Dazhen

2013-01-01

Background: Molten Salt Reactor (MSR) has several advantages over the other Generation IV reactor. Referred to the French CNRS research and compared to the fast reactor, super epithermal neutron spectrum reactor type is slightly lower and beading rate reaches 1.002. Purpose: The aim is to explore the best conversion zone layout scheme in the super epithermal neutron spectrum reactor. This study can make nuclear fuel as one way to solve the energy problems of mankind in future. Methods: Firstly, SCALE program is used for molten salt reactor graphite channel, molten salt core structure, control rods, graphite reflector and layer cladding structure. And the SMART modules are used to record the important actinides isotopes and their related reaction values of each reaction channel. Secondly, the thorium-uranium conversion rate is calculated. Finally, the better molten salt reactor core optimum layout scheme is studied comparing with various beading rates. Results: Breading zone layout scheme has an important influence on the breading rate of MSR. Central graphite channels in the core can get higher neutron flux irradiation. And more 233 Th can convert to 233 Pa, which then undergoes beta decay to become 233 U. The graphite in the breading zone gets much lower neutron flux irradiation, so the life span of this graphite can be much longer than that of others. Because neutron flux irradiation in the uranium molten salt graphite has nearly 10 times higher than the graphite in the breading zone, it has great impact on the thorium-uranium conversion rates. For the super epithermal neutron spectrum molten salt reactors, double salt design cannot get higher thorium-uranium conversion rates. The single molten salt can get the same thorium-uranium conversion rate, meanwhile it can greatly extend the life of graphite in the core. Conclusions: From the analysis of calculation results, Blanket breeding area in different locations in the core can change the breeding rates of thorium

Core-concrete molten pool dynamics and interfacial heat transfer

International Nuclear Information System (INIS)

Benjamin, A.S.

1980-01-01

Theoretical models are derived for the heat transfer from molten oxide pools to an underlying concrete surface and from molten steel pools to a general concrete containment. To accomplish this, two separate effects models are first developed, one emphasizing the vigorous agitation of the molten pool by gases evolving from the concrete and the other considering the insulating effect of a slag layer produced by concrete melting. The resulting algebraic expressions, combined into a general core-concrete heat transfer representation, are shown to provide very good agreement with experiments involving molten steel pours into concrete crucibles

Ex-vessel molten core debris interactions at CANDU nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Lewis, M J; Oyinloye, J O; Chambers, I [Electrowatt Consulting Engineers and Scientists, Warrington, Cheshire (United Kingdom); Scott, C K [Atlantic Nuclear Services, Fredericton, NB (Canada); Omar, A M [Atomic Energy Control Board, Ottawa, ON (Canada)

1991-12-31

Currently, the Atomic Energy Control Board (AECB) of Canada is sponsoring a project with a long term objective of obtaining an evaluation, independent of the industry, of the consequences to the public and the environment of postulated severe accidents at a Canadian nuclear power plant. Phase 1 of this project is a scoping study conducted to establish the relative consequences of a number of postulated event sequences. The studies in this paper model a multi-unit CANDU reactor at which pre-defined initiating events and their consequences could lead to severe core damage and relocation of the core debris onto the floor of the concrete reactor vault. Depending on the accident sequence assumptions made, an overlying pool of water may or may not be present. The US-NRC computer code CORCON Mod 2.0 was used to calculate the behaviour of the core material interacting with the concrete. The code calculates the decomposition of concrete by the molten core, and also the gases produced, which are released into the containment. The challenges to containment integrity are described, from the viewpoint of foundation decomposition and failure due to overpressure. The containment thermal-hydraulic behaviour is examined using an in-house computer code (CREM) written for this purpose. It is found that the containment envelope, in the absence of mitigating operator actions or design safety features, even for a case involving early core disassembly with the vacuum building unavailable, is unlikely to be failed within the 48 hours time frame examined. The paper identifies several areas for improvement in the models for future studies of core-concrete interactions for CANDU reactor plants. (author). 8 refs., 1 tab., 5 figs.

Ex-vessel molten core debris interactions at CANDU nuclear power plants

International Nuclear Information System (INIS)

Lewis, M.J.; Oyinloye, J.O.; Chambers, I.; Scott, C.K.; Omar, A.M.

1990-01-01

Currently, the Atomic Energy Control Board (AECB) of Canada is sponsoring a project with a long term objective of obtaining an evaluation, independent of the industry, of the consequences to the public and the environment of postulated severe accidents at a Canadian nuclear power plant. Phase 1 of this project is a scoping study conducted to establish the relative consequences of a number of postulated event sequences. The studies in this paper model a multi-unit CANDU reactor at which pre-defined initiating events and their consequences could lead to severe core damage and relocation of the core debris onto the floor of the concrete reactor vault. Depending on the accident sequence assumptions made, an overlying pool of water may or may not be present. The US-NRC computer code CORCON Mod 2.0 was used to calculate the behaviour of the core material interacting with the concrete. The code calculates the decomposition of concrete by the molten core, and also the gases produced, which are released into the containment. The challenges to containment integrity are described, from the viewpoint of foundation decomposition and failure due to overpressure. The containment thermal-hydraulic behaviour is examined using an in-house computer code (CREM) written for this purpose. It is found that the containment envelope, in the absence of mitigating operator actions or design safety features, even for a case involving early core disassembly with the vacuum building unavailable, is unlikely to be failed within the 48 hours time frame examined. The paper identifies several areas for improvement in the models for future studies of core-concrete interactions for CANDU reactor plants. (author). 8 refs., 1 tab., 5 figs

CFD to modeling molten core behavior simultaneously with chemical phenomena

International Nuclear Information System (INIS)

Vladimir V Chudanov; Anna E Aksenova; Valerii A Pervichko

2005-01-01

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

On the chemical constitution of a molten oxide core of a fast breeder reactor

International Nuclear Information System (INIS)

Hodkin, D.J.; Potter, P.E.

1980-01-01

A knowledge of the chemical constitution of a molten oxide fast reactor core is of great importance in the assessment of heat transfer from a cooling molten pool of debris and in the selection of materials for the construction of sacrificial beds for core containment. In this paper we describe some thermodynamic assessments of the likely chemical constitution of a molten oxide core, and then support our assessments by experimental observations

Candidate molten salt investigation for an accelerator driven subcritical core

Science.gov (United States)

Sooby, E.; Baty, A.; Beneš, O.; McIntyre, P.; Pogue, N.; Salanne, M.; Sattarov, A.

2013-09-01

We report a design for accelerator-driven subcritical fission in a molten salt core (ADSMS) that utilizes a fuel salt composed of NaCl and transuranic (TRU) chlorides. The ADSMS core is designed for fast neutronics (28% of neutrons >1 MeV) to optimize TRU destruction. The choice of a NaCl-based salt offers benefits for corrosion, operating temperature, and actinide solubility as compared with LiF-based fuel salts. A molecular dynamics (MD) code has been used to estimate properties of the molten salt system which are important for ADSMS design but have never been measured experimentally. Results from the MD studies are reported. Experimental measurements of fuel salt properties and studies of corrosion and radiation damage on candidate metals for the core vessel are anticipated. A special thanks is due to Prof. Paul Madden for introducing the ADSMS group to the concept of using the molten salt as the spallation target, rather than a conventional heavy metal spallation target. This feature helps to optimize this core as a Pu/TRU burner.

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

Thermal-hydraulic studies on molten core-concrete interactions

International Nuclear Information System (INIS)

Greene, G.A.

1986-10-01

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

Post-accident fuel relocation and heat removal in the LMFBR

International Nuclear Information System (INIS)

Kazimi, M.S.; Tsai, S.S.; Gasser, R.D.

1976-08-01

Assessment of the dynamics of post-accident fuel relocation and heat removal is an important aspect of the evaluation of the consequences of a hypothetical accident in an LMFBR. Such an assessment is of particular importance in the evaluation of the post-accident radiological doses around the reactor site. In the present evaluation particular attention is given to the design features of the Clinch River Breeder Reactor Plant (CRBR). Fuel relocation and heat removal, assuming certain conditions have resulted in core disruption, are discussed. The discussion of events and phenomena involved in the relocation processes is centered around the resulting patterns of heat source distribution. The factors influencing fuel relocation and distribution in the inlet and outlet plena of the reactor vessel are discussed. The current technology of in-vessel heat removal is applied to the design of the CRBR reactor. Both fuel debris cooling limits and overall coolant flow in the reactor under natural convection conditions are explored. Some of the uncertainties in ex-vessel fuel behavior are addressed. In particular, the effect of melting the cavity bed on the rate of growth of a molten fuel pool is investigated

Model for melt blockage (slug) relocation and physico-chemical interactions during core degradation under severe accident conditions

International Nuclear Information System (INIS)

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

2008-01-01

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

The effects of core zoning on optimization of design analysis of molten salt reactor

International Nuclear Information System (INIS)

Guo, Zhangpeng; Wang, Chenglong; Zhang, Dalin; Chaudri, Khurrum Saleem; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

2013-01-01

Highlights: • 1/8 of core is simulated by MCNP and thermal-hydraulic code simultaneously. • Effects of core zoning are studied by dividing the core into two regions. • Both the neutronics and thermal-hydraulic behavior are investigated. • The flat flux distribution is achieved in the optimization analysis. • The flat flux can lead to worse thermal-hydraulic behavior occasionally. - Abstract: The molten salt reactor (MSR) is one of six advanced reactor types in the frame of the Generation 4 International Forum. In this study, a multiple-channel analysis code (MAC) is developed to analyze thermal-hydraulics behavior and MCNP4c is used to study the neutronics behavior of Molten Salt Reactor Experiment (MSRE). The MAC calculates thermal-hydraulic parameters, namely temperature distribution, flow distribution and pressure drop. The MCNP4c performs the analysis of effective multiplication factor, neutron flux, power distribution and conversion ratio. In this work, the modification of core configuration is achieved by different core zoning and various fuel channel diameters, contributing to flat flux distribution. Specifically, the core is divided into two regions and the effects of different core zoning on the both neutronics and thermal-hydraulic behavior of moderated molten salt reactor are investigated. We conclude that the flat flux distribution cannot always guarantee better performance in thermal-hydraulic perspective and can decreases the graphite lifetime significantly

SCDAP/RELAP5 lower core plate model

International Nuclear Information System (INIS)

Coryell, E.W.; Griffin, F.P.

1999-01-01

The SCDAP/RELAP5 computer code is a best-estimate analysis tool for performing nuclear reactor severe accident simulations. This report describes the justification, theory, implementation, and testing of a new modeling capability which will refine the analysis of the movement of molten material from the core region to the vessel lower head. As molten material moves from the core region through the core support structures it may encounter conditions which will cause it to freeze in the region of the lower core plate, delaying its arrival to the vessel head. The timing of this arrival is significant to reactor safety, because during the time span for material relocation to the lower head, the core may be experiencing steam-limited oxidation. The time at which hot material arrives in a coolant-filled lower vessel head, thereby significantly increasing the steam flow rate through the core region, becomes significant to the progression and timing of a severe accident. This report is a revision of a report INEEL/EXT-00707, entitled ''Preliminary Design Report for SCDAP/RELAP5 Lower Core Plate Model''

How to arrest a core meltdown accident (doing nothing)

International Nuclear Information System (INIS)

Baron, Jorge H.

2000-01-01

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

Development of a Chemical Equilibrium Model for a Molten Core-Concrete Interaction Analysis Module

Energy Technology Data Exchange (ETDEWEB)

Seo, Jae Uk; Lee, Dae Young; Park, Chang Hwan [FNC Technology Co., Yongin (Korea, Republic of)

2016-10-15

This molten core could interact with the reactor cavity region which consists of concrete. In this process, components of molten core react with components of concrete through a lot of chemical reactions. As a result, many kinds of gas species are generated and those move up forming rising bubbles into the reactor containment atmosphere. These rising bubbles are the carrier of the many kinds of the aerosols coming from the MCCI (Molten Core Concrete Interaction) layers. To evaluate the amount of the aerosols released from the MCCI layers, the amount of the gas species generated from those layers should be calculated. The chemical equilibrium state originally implies the final state of the multiple chemical reactions; therefore, investigating the equilibrium composition of molten core can be applicable to predict the gas generation status. The most common way for finding the chemical equilibrium state is a minimization of total Gibbs free energy of the system. In this paper, the method to make good guess of initial state is suggested and chemical reaction results are compared with results of CSSI report No 164. Total mass of system and the number of atoms of each element are conserved. The tendency of calculation results is similar with results presented in CSNI Report except a few species. These differences may be caused by absence of Gibbs energy data of the species such as Fe{sub 2}SiO{sub 4}, CaFe{sub 2}O{sub 4}, U(OH){sub 3}, UO(OH), UO{sub 2}(OH), U{sub 3}O{sub 7}, La, Ce.

In-core melt progression for the MAAP 4 codes

International Nuclear Information System (INIS)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Influence of Concrete Properties on Molten Core-Concrete Interaction: A Simulation Study

Directory of Open Access Journals (Sweden)

Jin-yang Jiang

2016-01-01

Full Text Available In a severe nuclear power plant accident, the molten core can be released into the reactor pit and interact with sacrificial concrete. In this paper, a simulation study is presented that aims to address the influence of sacrificial concrete properties on molten core-concrete interaction (MCCI. In particular, based on the MELCOR Code, the ferrosiliceous concrete used in European Pressurized Water Reactor (EPR is taken into account with respect to the different ablation enthalpy and Fe2O3 and H2O contents. Results indicate that the concrete ablation rate as well as the hydrogen generation rate depends much on the concrete ablation enthalpy and Fe2O3 and H2O contents. In practice, the ablation enthalpy of sacrificial concrete is the higher the better, while the Fe2O3 and H2O content of sacrificial concrete is the lower the better.

Radiation heat transfer in a pressurized water reactor lower head filled with molten corium

International Nuclear Information System (INIS)

Šadek, Siniša; Grgić, Davor; Debrecin, Nenad

2013-01-01

Highlights: ► We develop radiation heat exchange model for a reactor pressure vessel lower head. ► Model is used during a late in-vessel phase of severe accidents. ► View factors are calculated automatically for a time-dependent enclosure. ► Model is included in the RELAP5/SCDAPSIM computer code. ► Inclusion of heat radiation causes faster heat-up rate of RPV lower head structures. - Abstract: Following a core melt, molten material may slump to the lower head of a reactor pressure vessel (RPV). In that case, some structures like lower parts of fuel elements and a core support plate would remain intact. Since the melt is at high temperature and there are no obstacles between the melt and the supporting plate, the plate is exposed to an intense radiation heating. The radiation heat exchange model of the lower head was developed and applied to a finite element code COUPLE which is a part of the detailed mechanistic code RELAP5/SCDAPSIM. The radiation enclosure consisted of three surfaces: the upper surface of the relocated material, the inner surface of the RPV wall above the relocated material and the lower surface of the core support plate. View factors were calculated for the enclosure geometry that is changing in time because of intermittent accumulation of molten material. The enclosure surfaces were covered by mesh of polygonal areas and view factors were calculated, for each pair of the element areas, by solving the definite integrals using the algorithms for adaptive integrations by means of Gaussian quadrature. Algebraic equations for radiosity and irradiation vectors were solved by LU decomposition and the radiation model was explicitly coupled with the heat conduction model. The results show that there is a possibility of the core support plate failure after being heated up due to radiation heat exchange with the melt.

Molten Core - Concrete interactions in nuclear accidents. Theory and design of an experimental facility

International Nuclear Information System (INIS)

Sevon, T.

2005-11-01

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

Assessment of mass fraction and melting temperature for the application of limestone concrete and siliceous concrete to nuclear reactor basemat considering molten core-concrete interaction

Energy Technology Data Exchange (ETDEWEB)

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

2016-04-15

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

Modeling of molten core-concrete interactions and fission-product release

International Nuclear Information System (INIS)

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

1991-09-01

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

Large longitude libration of Mercury reveals a molten core.

Science.gov (United States)

Margot, J L; Peale, S J; Jurgens, R F; Slade, M A; Holin, I V

2007-05-04

Observations of radar speckle patterns tied to the rotation of Mercury establish that the planet occupies a Cassini state with obliquity of 2.11 +/- 0.1 arc minutes. The measurements show that the planet exhibits librations in longitude that are forced at the 88-day orbital period, as predicted by theory. The large amplitude of the oscillations, 35.8 +/- 2 arc seconds, together with the Mariner 10 determination of the gravitational harmonic coefficient C22, indicates that the mantle of Mercury is decoupled from a core that is at least partially molten.

Transient analyses for a molten salt fast reactor with optimized core geometry

Energy Technology Data Exchange (ETDEWEB)

Li, R., E-mail: rui.li@kit.edu [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Wang, S.; Rineiski, A.; Zhang, D. [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Merle-Lucotte, E. [Laboratoire de Physique Subatomique et de Cosmologie – IN2P3 – CNRS/Grenoble INP/UJF, 53, rue des Martyrs, 38026 Grenoble (France)

2015-10-15

Highlights: • MSFR core is analyzed by fully coupling neutronics and thermal-hydraulics codes. • We investigated four types of transients intensively with the optimized core geometry. • It demonstrates MSFR has a high safety potential. - Abstract: Molten salt reactors (MSRs) have encountered a marked resurgence of interest over the past decades, highlighted by their inclusion as one of the six candidate reactors of the Generation IV advanced nuclear power systems. The present work is carried out in the framework of the European FP-7 project EVOL (Evaluation and Viability Of Liquid fuel fast reactor system). One of the project tasks is to report on safety analyses: calculations of reactor transients using various numerical codes for the molten salt fast reactor (MSFR) under different boundary conditions, assumptions, and for different selected scenarios. Based on the original reference core geometry, an optimized geometry was proposed by Rouch et al. (2014. Ann. Nucl. Energy 64, 449) on thermal-hydraulic design aspects to avoid a recirculation zone near the blanket which accumulates heat and very high temperature exceeding the salt boiling point. Using both fully neutronics thermal-hydraulic coupled codes (SIMMER and COUPLE), we also re-confirm the efforts step by step toward a core geometry without the recirculation zone in particular as concerns the modifications of the core geometrical shape. Different transients namely Unprotected Loss of Heat Sink (ULOHS), Unprotected Loss of Flow (ULOF), Unprotected Transient Over Power (UTOP), Fuel Salt Over Cooling (FSOC) are intensively investigated and discussed with the optimized core geometry. It is demonstrated that due to inherent negative feedbacks, an MSFR plant has a high safety potential.

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

International Nuclear Information System (INIS)

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

2002-03-01

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

Penetration of molten core materials into basaltic and limestone concrete

International Nuclear Information System (INIS)

Sutherland, H.J.

1978-01-01

In conjunction with the small-scale, melt-concrete interaction tests being conducted at Sandia Laboratories, an acoustic technique has been used to monitor the penetration of molten core materials into basaltic and limestone concrete. Real time plots of the position of the melt/concrete interface have been obtained, and they illustrate that the initial penetration rate of the melt may be of the order of 80 mm/min. Phenomena deduced by the technique include a non-wetted melt/concrete interface

How to arrest a core meltdown accident (doing nothing); Como detener un accidente con fusion de nucleo (sin hacer nada)

Energy Technology Data Exchange (ETDEWEB)

Baron, Jorge H [Autoridad Regulatoria Nuclear, Buenos Aires (Argentina)

2000-07-01

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

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

Coupled study of the Molten Salt Fast Reactor core physics and its associated reprocessing unit

International Nuclear Information System (INIS)

Doligez, X.; Heuer, D.; Merle-Lucotte, E.; Allibert, M.; Ghetta, V.

2014-01-01

Highlights: • The limit on the reprocessing is due to the redox potential control. • Alkali and Earth-alkaline elements do not have to be extracted. • Criticality risks have to be studied in the reprocessing unit. • The neutronics properties are not sensitive to chemical data. • The reprocessing chemistry, from a pure numerical point of view, is an issue. - Abstract: Molten Salt Reactors (MSRs) are liquid-fuel reactors, in which the fuel is also the coolant and flows through the core. A particular configuration presented in this paper called the Molten Salt Fast Reactor consists in a Molten Salt Reactor with no moderator inside the core and a salt composition that leads to a fast neutron spectrum. Previous studies showed that this concept (previously called Thorium Molten Salt Reactor – Nonmoderated) has very promising characteristics. The liquid fuel implies a special reprocessing. Each day a small amount of the fuel salt is extracted from the core for on-site reprocessing. To study such a reactor, the materials evolution within the core has to be coupled to the reprocessing unit, since the latter cleans the salt quasi continuously and feeds the reactor. This paper details the issues associated to the numerical coupling of the core and the reprocessing. It presents how the chemistry is introduced inside the classical Bateman equation (evolution of nuclei within a neutron flux) in order to carry a numerical coupled study. To achieve this goal, the chemistry has to be modeled numerically and integrated to the equations of evolution. This paper presents how is it possible to describe the whole concept (reactor + reprocessing unit) by a system of equations that can be numerically solved. Our program is a connection between MCNP and a homemade evolution code called REM. Thanks to this tool; constraints on the fuel reprocessing were identified. Limits are specified to preserve the good neutronics properties of the MSFR. In this paper, we show that the limit

Physical properties of core-concrete systems: Al{sub 2}O{sub 3}-ZrO{sub 2} molten materials measured by aerodynamic levitation

Energy Technology Data Exchange (ETDEWEB)

Ohishi, Yuji, E-mail: ohishi@see.eng.osaka-u.ac.jp [Graduate School of Engineering, Osaka University (Japan); Kargl, F. [Institute of Materials Physics in Space, German Aerospace Center (Germany); Nakamori, F.; Muta, Hiroaki; Kurosaki, Ken [Graduate School of Engineering, Osaka University (Japan); Yamanaka, Shinsuke [Graduate School of Engineering, Osaka University (Japan); Research Institute of Nuclear Engineering, University of Fukui (Japan)

2017-04-15

During a molten core–concrete interaction, molten oxides consisting of molten core materials (UO{sub 2} and ZrO{sub 2}) and concrete (Al{sub 2}O{sub 3}, SiO{sub 2}, CaO) are formed. Reliable data on the physical properties of the molten oxides will allow us to accurately predict the progression of a nuclear reactor core meltdown accident. In this study, the viscosities and densities of molten (ZrO{sub 2}){sub x}(Al{sub 2}O{sub 3}){sub 1-x} (x = 0.356 and 0.172) were measured using an aerodynamic levitation technique. The densities of two small samples were estimated from their masses and their volumes (calculated from recorded images of the molten samples). The droplets were forced to oscillate using speakers, and their viscosities were evaluated from the damping behaviors of their oscillations. The results showed that the viscosity of molten (ZrO{sub 2}){sub x}(Al{sub 2}O{sub 3}){sub 1-x} compared to that of pure molten Al{sub 2}O{sub 3} is 25% lower for x = 0.172, while it is unexpectedly 20% higher for x = 0.356. - Highlights: •The physical properties of molten (ZrO{sub 2}){sub x}(Al{sub 2}O{sub 3}){sub 1-x} (x = 0.356 and 0.172) have been evaluated. •The measurement was conducted using an aerodynamic levitation technique. •The density and viscosity were measured.

Molten LWR core material interactions with water and with concrete

International Nuclear Information System (INIS)

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

1977-01-01

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

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.

Thermal interactions of a molten core debris pool with surrounding structural materials

International Nuclear Information System (INIS)

Baker, L. Jr.; Cheung, F.B.; Farhadieh, R.; Stein, R.P.; Gabor, J.D.; Bingle, J.D.

1979-01-01

Analytical and experimental results on individual aspects of the overall problem of the interaction of a large mass of LMFBR core debris with concrete or other materials are reviewed. Results of recent heat transfer experiments with molten UO 2 have indicated the importance of internal thermal radiation and methods to take account of this are developed. Effects of gas release and density difference are considered. The GROWS-2 Code is used to illustrate the effects of various assumptions

Status of degraded core issues. Synthesis paper prepared by G. Bandini in collaboration with the NEA task group on degraded core cooling

International Nuclear Information System (INIS)

2001-02-01

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

An analysis of molten-corium-induced failure of drain pipes in BWR Mark 2 containments

International Nuclear Information System (INIS)

Taleyarkhan, R.P.; Podowski, M.Z.

1991-01-01

This study has focused on mechanistic simulation and analysis of potential failure modes for inpedestal drywell drain pipes in the Limerick boiling water reactor (BWR) Mark 2 containment. Physical phenomena related to surface tension breakdown, heatup, melting, ablation, crust formation and failure, and core material relocation into drain pipes with simultaneous melting of pipe walls were modeled and analyzed. The results of analysis have been used to assess the possibility of drain pipe failure and the resultant loss of pressure-suppression capability. Estimates have been made for the timing and amount of molten corium released to the wetwell. The study has revealed that significantly different melt progression sequences can result depending upon the failure characteristics of the frozen metallic crust which forms over the drain cover during the initial stages of debris pour. Another important result is that it can take several days for the molten fuel to ablate the frozen metallic debris layer -- if the frozen layer has cooled below 1100 K before fuel attack. 10 refs., 3 figs., 4 tabs

Exploratory study of molten core material/concrete interactions, July 1975--March 1977

International Nuclear Information System (INIS)

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

1978-02-01

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

Fragmentation and quench behavior of corium melt streams in water

International Nuclear Information System (INIS)

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

1994-02-01

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

Analysis of the thermal hydraulics and core degradation behavior in the PHEBUS-FPT1 test train with impact/SAMPSON code

International Nuclear Information System (INIS)

Terada, Masafumi; Ikeda, Takashi; Nakahara, Katsuhiko; Shirakawa, Noriyuki; Horie, Hideki; Katsuragi, Kazuyuki; Yamagishi, Makoto; Ito, Takahiro

2003-01-01

As one of the verification studies of SAMPSON code, PHEBUS-FPT1, which is authorized as the International Standard Problem-46, was analyzed about the in-core phenomena with four modules, the molten core relocation analysis (MCRA) module, the fuel rod heat up analysis (FRHA) module, the fission product release analysis (FPRA) module, and the analysis control module (ACM) of SAMPSON. This paper describes the analysis of thermal hydraulics and core degradation behavior in the test train. Two-dimensional version of MCRA models the whole structure of the test train in the cylindrical system, including the fuel bundle and the shroud. FRHA models eighteen irradiated fuel rods, two fresh fuel rods, and one control rod in the center of the bundle. FRHA evaluates the transient behavior of fuel rods and releases failed fuel components to MCRA. MCRA evaluates the fluid dynamics of steam and debris considering the thermal and fluid mechanical interaction between them, and at the same time the thermal interaction between gas/debris and shroud material. By the phase change model of MCRA, molten debris forms debris pool and a part of them possibly freezes on fuel rods or shroud surface, then forms crust. This combination of modules of SAMPSON was proved to be capable for modeling the PHEBUS-FPT1 in-core phenomena sufficiently. The analysis has shown sufficient agreement with test results regarding to steam flow rates at the outlet, reproducing its reduction due to hydrogen generation, steam and shroud temperature, and debris relocation behavior. (author)

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

International Nuclear Information System (INIS)

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

1989-01-01

The interaction of molten core materials with water has been investigated for the pour stream mixing mode. This interaction plays a crucial role during the later stages of in-vessel core melt progression inside a light water reactor such as during the TMI-2 accident. The key issues which arise during the molten core relocation include: (1) the thermal attack and possible damage to the RPV lower head from the impinging molten fuel stream and/or the debris bed, (2) the molten fuel relocation pathways including the effects of redistribution due to core support structure and the reactor lower internals, (3) the quench rate of the molten fuel through the water in the lower plasma, (4) the steam generation and hydrogen generation during the interaction, (5) the transient pressurization of the primary system, and (6) the possibility of a steam explosion. In order to understand these issues, a series of six experiments (designated CCM-1 through -6) was performed in which molten corium passed through a deep pool of water in a long, slender pour stream mode. Results discussed include the transient temperatures and pressures, the rate and magnitude of steam/hydrogen generation, and the posttest debris characteristics. 9 refs., 29 figs

Fuel relocation modeling in the SAS4A accident analysis code system

International Nuclear Information System (INIS)

Tentner, A.M.; Miles, K.J.; Kalimullah; Hill, D.J.

1986-01-01

The SAS4A code system has been designed for the analysis of the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failure of the subassembly walls. During such postulated accident scenarios as the Loss-of-Flow (LOF) and Transient-Overpower (TOP) events, the relocation of the fuel plays a key role in determining the sequence of events and the amount of energy produced before neutronic shutdown. This paper discusses the general strategy used in modelong the various phenomena which lead to fuel relocation and presents the key fuel relocation models used in SAS4A. The implications of these models for the whole-core accident analysis as well as recent results of fuel relocation are emphasized. 12 refs

Termination of light-water reactor core-melt accidents with a chemical core catcher: the core-melt source reduction system (COMSORS)

International Nuclear Information System (INIS)

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

1996-09-01

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

Experimental and numerical thermal-hydraulics investigation of a molten salt reactor concept core

Energy Technology Data Exchange (ETDEWEB)

Yamaji, Bogdan; Aszodi, Attila [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques

2017-09-15

In the paper measurement results of experimental modelling of a molten salt fast reactor concept will be presented and compared with three-dimensional computational fluid dynamics (CFD) simulation results. Purpose of this article is twofold, on one hand to introduce a geometry modification in order to avoid the disadvantages of the original geometry and discuss new measurement results. On the other hand to present an analysis in order to suggest a method of proper numerical modelling of the problem based on the comparison of calculation results and measurement data for the new, modified geometry. The investigated concept has a homogeneous cylindrical core without any internal structures. Previous measurements on the scaled and segmented plexiglas model of the concept core and simulation results have shown that this core geometry could be optimized for better thermal-hydraulics characteristics. In case of the original geometry strong undesired flow separation could develop, that could negatively affect the characteristics of the core from neutronics point of view as well. An internal flow distributor plate was designed and installed with the purpose of optimizing the flow field in the core by enhancing its uniformity. Particle image velocimetry (PIV) measurement results of the modified experimental model will be presented and compared to numerical simulation results with the purpose of CFD model validation.

Comparison of the behaviour of two core designs for ASTRID in case of severe accidents

Energy Technology Data Exchange (ETDEWEB)

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

2016-02-15

material to inject in the pool of molten fuel in order to compensate this potential are calculated. These amounts give an order of magnitude of the strength of the mitigation actions required for each type of core in order to limit the mechanical energy release due to secondary power excursions. The calculation results have shown that in case of UTOPs (ramp ranging from several $/s to 60 $/s), the mechanical energy released by the expansion of a fuel vapor bubble is about one order of magnitude lower for the CFV core than for the SFRv2 core. Moreover, in case of core melting due to an UTOP, both cores require approximately the same amount of absorber material to be inserted in order to reach a sub-critical state. For the same degraded state a zero reactivity could be reached with a lower amount of fuel extracted from the CFV core than from the SFRv2 core. Regarding ULOFs, the assessment of the reactivity that could be induced by molten cladding relocation has shown a largely better behaviour for the CFV core. More precisely, for this latter core design, a severe power excursion induced by cladding relocation is very unlikely. Finally, the times of boiling onset, of flow redistribution in the core and of material melting, indicate that mitigation actions (absorber insertion and fuel extraction) should start earlier for the SFRv2 core than for the CFV one. This result is a good point for the CFV core because it allows later mitigation actions which are more likely to succeed because time is required to initiate the fuel relocation through in-core dedicated ducts. The results presented in this paper have contributed to retain the CFV core for the studies currently carried-out for the ASTRID design.

Liquid-liquid reductive extraction in molten fluoride salt/liquid aluminium as a core of process for the An/Ln group separation

International Nuclear Information System (INIS)

Conocar, O.

2007-06-01

This report concerns a pyrochemical process based on liquid-liquid extraction in a molten fluoride/liquid aluminium system as a core process for actinide (An)/lanthanide (Ln) group separation, studied at CEA. The basic and demonstrative experiments have established the feasibility of the An/Ln group separation in the molten fluoride/liquid aluminium system (U, Pu, Np, Am, Cm traces from Nd, Ce, Eu, Sm, Eu, La - An/Ln separation factors over 1000 - An recovery yield over 98 % in one batch). The main experimental efforts must now be targeted on the recovery of actinides from the Al matrix. A thermodynamic and bibliographical survey has been done. It shows that back-extraction in a molten chloride melt could be a promising technique for this purpose

Liquid-liquid reductive extraction in molten fluoride salt/liquid aluminium as a core of process for the An/Ln group separation

Energy Technology Data Exchange (ETDEWEB)

Conocar, O

2007-06-15

This report concerns a pyrochemical process based on liquid-liquid extraction in a molten fluoride/liquid aluminium system as a core process for actinide (An)/lanthanide (Ln) group separation, studied at CEA. The basic and demonstrative experiments have established the feasibility of the An/Ln group separation in the molten fluoride/liquid aluminium system (U, Pu, Np, Am, Cm traces from Nd, Ce, Eu, Sm, Eu, La - An/Ln separation factors over 1000 - An recovery yield over 98 % in one batch). The main experimental efforts must now be targeted on the recovery of actinides from the Al matrix. A thermodynamic and bibliographical survey has been done. It shows that back-extraction in a molten chloride melt could be a promising technique for this purpose.

New finite element-based modeling of reactor core support plate failure

Energy Technology Data Exchange (ETDEWEB)

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

2017-12-15

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

Molten salt fueled reactors with a fast salt draining

International Nuclear Information System (INIS)

Ventre, Edmond; Blum, J.M.

1976-01-01

This invention relates to a molten salt nuclear reactor which comprises a new arrangement for shutting it down in complete safety. This nuclear reactor has a molten salt primary circuit comprising, in particular, the core of this reactor. It includes a leak tight vessel the capacity of which is appreciably greater than that of the molten salt volume of the circuit and placed so that the level of the molten salt, when all the molten salt of the circuit is contained in this vessel, is less than that of the base of the core. There are facilities for establishing and maintaining an inert gas pressure in the vessel above the molten salt, for releasing the compressed gas and for connecting the vessel to the primary circuit entering this vessel at a lower level than that of the molten salt and enabling molten salt to enter or leave the vessel according to the pressure of the inert gas. The particular advantage of this reactor is that it can be shut down safely since the draining of the primary circuit no longer results from a 'positive action' but from the suppression of an arrangement essential for the operation of the reactor consisting of the build-up of the said inert gas pressure in the said vessel [fr

Heat transfer on liquid-liquid interface of molten-metal and water

International Nuclear Information System (INIS)

Tanaka, T.; Saito, Yasushi; Mishima, Kaichiro

2001-01-01

Molten-core pool had been formed in the lower-head of TMI-2 pressure vessel at the severe accident. The lower head, however, didn't receive any damage by reactor core cooling. Heat transfer at outside of the lower head and boiling heat transfer at liquid-liquid interface of molten-metal and water, however, are important for initial cooling process of the molten-core pool. The heat transfer experiments for the liquid-liquid interface of molten-metal and water are carried out over the range of natural convection to film boiling region. Phenomenon on the heat transfer experiments are visualized by using of high speed video camera. Wood's metal and U-alloy 78 are used as molten-metal. The test section of the experiments consists of a copper block with heater, wood's metal, and water. Three thermocouple probes are used for temperature measurement of water side and the molten-metal side. Stability of the liquid-liquid interface is depended on the wetness of container wall for molten metal and the temperature distribution of the interface. Entrainment phenomena of molten-metal occurs by a fluctuation of the interface after boiling on the container wall surface. The boiling curves obtained from the liquid-liquid interface experiments are agree with the nucleate boiling and the film boiling correlations of solid-liquid system. (Suetake, M.)

Analysis of an out-of-pile experiment for materials redistribution under core disruptive accident condition of fast breeder reactors

International Nuclear Information System (INIS)

Sawada, Tetsuo; Ninokata, Hisashi; Shimizu, Akinao

1995-01-01

Calculation of one of the SIMBATH experiments was performed using the SIMMER-II code. The experiments were intended to simulate the fuel pin disintegration, the molten materials relocation and following materials redistribution that could occur during core disruptive accidents assumed in fast breeder reactors. The calculation by SIMMER-II showed that the incorporated step-wise fuel pin disintegration model and the modified particle jamming model were capable of reproducing the course of materials relocation within the identified ranges of the parameters which governed the blockages formation, i.e. the characteristic radius of solid particles jamming and/or sieving out in the flow and the effective particle viscosity. In particular the final materials redistribution calculated by SIMMER-II very well reproduced the experiment. This fact made it possible to interpret theoretically the mechanisms of flow blockages formation and related materials redistribution. (author)

TMI-2 core damage: a summary of present knowledge

International Nuclear Information System (INIS)

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

1983-01-01

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

Fuel Rod Melt Progression Simulation Using Low-Temperature Melting Metal Alloy

International Nuclear Information System (INIS)

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

2002-01-01

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

Simulation of the thermalhydraulic behavior of a molten core within a structure, with the three dimensions three components TOLBIAC code

Energy Technology Data Exchange (ETDEWEB)

Spindler, B.; Moreau, G.M.; Pigny S. [Centre d`Etudes Nucleaires de Grenoble (France)

1995-09-01

The TOLBIAC code is devoted to the simulation of the behavior of a molten core within a structure (pressure vessel of core catcher), taking into account the relative position of the core components, the wall ablation and the crust formation. The code is briefly described: 3D model, physical properties and constitutive laws. wall ablation and crust model. Two results are presented: the simulation of the COPO experiment (natural convection with water in a 1/2 scale elliptic pressure vessel), and the simulation of the behavior of a corium in a PWR pressure vessel, with ablation and crust formation.

Modelling of the Molten Core Concrete Interaction (MCCI)

International Nuclear Information System (INIS)

Guillaume, M.

2008-01-01

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

Fuel relocation modeling in the SAS4A accident analysis code system

International Nuclear Information System (INIS)

Tentner, A.M.; Miles, K.J.

1985-01-01

SAS4A is a new code system which has been designed for analyzing the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failure of the subassembly walls. During such postulated accident scenarios as the Loss-of-Flow (LOF) and Transient-Overpower (TOP) events, the relocation of the fuel plays a key role in determining the sequence of events and the amount of energy produced before neutronic shutdown. This paper discusses the general strategy used in modeling the various phenomena which lead to fuel relocation and presents the key fuel relocation models used in SAS4A. The implications of these models for the whole-core accident analysis as well as recent results of fuel motion experiment analyses are also presented

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)

Break-up and quench behavior of molten material in coolant

International Nuclear Information System (INIS)

Abe, Y.; Kizu, T.; Arai, T.; Nariai, H.; Chitose, K.; Koyama, K.

2003-01-01

In a Core Disruptive Accident (CDA) of a Fast Breeder Reactor, the Post Accident Heat Removal(PAHR) is crucial for the accident mitigation. The molten core material should be solidified in the sodium coolant in the reactor vessel. The material, being fragmented while solidification and forming debris bed, will be cooled in the coolant. In the experiment, molten material jet is injected into water to experimentally obtain fragments and the visualized information of the fragmentation and boiling phenomena during PAHR in CDA. The distributed particle behavior of the molten material jet is observed with high-speed video camera. The experimental results are compared with the existing theories. Consequently, the marginal wavelength on the surface of a water jet is close to the value estimated based on the Rayleigh-Taylor instability. Moreover, the fragmented droplet diameter obtained from the interaction of molten material and water is close to the value estimated based on the Kelvin-Helmholtz instability. Once the particle diameter of the fragmented molten material could be known from a hydrodynamic model, it becomes possible to estimate the mass of the molten particle with some appropriate heat transfer model

Flowing and freezing of molten core materials during unprotected loss of flow accidents in sodium cooled fast reactors

International Nuclear Information System (INIS)

Maschek, W.; Royl, P.

1988-09-01

Flowing and freezing of mobile core materials change the fissile material distribution and core-inventory under hypothetical accident conditions and determine the path to permanent shutdown of the neutronic events and the energetic potentials. The report classifies the bondary conditions for such flowing and freezing processes by going through the different situations under which these processes can occur in the scenario of the unprotected loss of flow (ULOF) accident. The classification is based on ULOF-accident simulations for a homogeneous reactor core concept of a 300 MWe LMFBR (e. g. SNR-300), but many boundary conditions are also characteristic for other core designs. A review of the relevant experiments is then made to correlate the available experimental information with these classified boundary conditions and to look at the resulting flowing and freezing processes. Boundary conditions that have been experimentally shown to be important are assigned high priorities. The data are specifically valued in relation to these boundary conditions of high priorities. The review includes the major experimental programs with published results. The discussion shows that the results from most clean condition tests for melt relocations are valuable for a better understanding of basic phenomena and analytical model development, but are not directly applicable to real accident conditions. The database for relevant boundary conditions from the ULOF scenario is limited and largely included in integral sequence tests from which quantitative information for modelling is difficult to obtain. Needs for additional investigations are identified. The suggestions are mainly restricted to investigations of the early phase of fuel removal. They are given with reference to candidate facilities and include relocations in the subassemblies and in the inter-subassembly gaps. Particular emphasis is put on the leading edge properties and possible driving forces to which more attention

Molten salts and nuclear energy production

International Nuclear Information System (INIS)

Le Brun, Christian

2007-01-01

Molten salts (fluorides or chlorides) were considered near the beginning of research into nuclear energy production. This was initially due to their advantageous physical and chemical properties: good heat transfer capacity, radiation insensitivity, high boiling point, wide range solubility for actinides. In addition it was realised that molten salts could be used in numerous situations: high temperature heat transfer, core coolants with solid fuels, liquid fuel in a molten salt reactor, solvents for spent nuclear solid fuel in the case of pyro-reprocessing and coolant and tritium production in the case of fusion. Molten salt reactors, one of the six innovative concepts chosen by the Generation IV international forum, are particularly interesting for use as either waste incinerators or thorium cycle systems. As the neutron balance in the thorium cycle is very tight, the possibility to perform online extraction of some fission product poisons from the salt is very attractive. In this article the most important questions that must be addressed to demonstrate the feasibility of molten salt reactor will be reviewed

Modelling of molten fuel/concrete interactions

International Nuclear Information System (INIS)

Muir, J.F.; Benjamin, A.S.

1980-01-01

A computer program modelling the interaction between molten core materials and structural concrete (CORCON) is being developed to provide quantitative estimates of fuel-melt accident consequences suitable for risk assessment of light water reactors. The principal features of CORCON are reviewed. Models developed for the principal interaction phenomena, inter-component heat transfer, concrete erosion, and melt/gas chemical reactions, are described. Alternative models for the controlling phenomenon, heat transfer from the molten pool to the surrounding concrete, are presented. These models, formulated in conjunction with the development of CORCON, are characterized by the presence or absence of either a gas film or viscous layer of molten concrete at the melt/concrete interface. Predictions of heat transfer based on these models compare favorably with available experimental data

Developmental assessment of the SCDAP/RELAP5 code

International Nuclear Information System (INIS)

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

1997-01-01

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

Molten salt reactors: reactor cores

International Nuclear Information System (INIS)

1983-01-01

In this critical analysis of the MSBR I project are examined the problems concerning the reactor core. Advantages of breeding depend essentially upon solutions to technological problems like continuous reprocessing or graphite behavior under neutron irradiation. Graphite deformation, moderator unloading, control rods and core instrumentation require more studies. Neutronics of the core, influence of core geometry and salt composition, fuel evolution, and thermohydraulics are reviewed [fr

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

International Nuclear Information System (INIS)

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

1992-09-01

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

Parametric study of natural circulation flow in molten salt fuel in molten salt reactor

Energy Technology Data Exchange (ETDEWEB)

Pauzi, Anas Muhamad, E-mail: Anas@uniten.edu.my [Centre of Nuclear Energy, Universiti Tenaga Nasional (UNITEN), Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia); Cioncolini, Andrea; Iacovides, Hector [School of Mechanical, Aerospace, and Civil Engineering (MACE), University of Manchester, Oxford Road, M13 9PL Manchester (United Kingdom)

2015-04-29

The Molten Salt Reactor (MSR) is one of the most promising system proposed by Generation IV Forum (GIF) for future nuclear reactor systems. Advantages of the MSR are significantly larger compared to other reactor system, and is mainly achieved from its liquid nature of fuel and coolant. Further improvement to this system, which is a natural circulating molten fuel salt inside its tube in the reactor core is proposed, to achieve advantages of reducing and simplifying the MSR design proposed by GIF. Thermal hydraulic analysis on the proposed system was completed using a commercial computation fluid dynamics (CFD) software called FLUENT by ANSYS Inc. An understanding on theory behind this unique natural circulation flow inside the tube caused by fission heat generated in molten fuel salt and tube cooling was briefly introduced. Currently, no commercial CFD software could perfectly simulate natural circulation flow, hence, modeling this flow problem in FLUENT is introduced and analyzed to obtain best simulation results. Results obtained demonstrate the existence of periodical transient nature of flow problem, hence improvements in tube design is proposed based on the analysis on temperature and velocity profile. Results show that the proposed system could operate at up to 750MW core power, given that turbulence are enhanced throughout flow region, and precise molten fuel salt physical properties could be defined. At the request of the authors and the Proceedings Editor the name of the co-author Andrea Cioncolini was corrected from Andrea Coincolini. The same name correction was made in the Acknowledgement section on page 030004-10 and in reference number 4. The updated article was published on 11 May 2015.

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

Probability safety assessment of LOOP accident to molten salt reactor

International Nuclear Information System (INIS)

Mei Mudan; Shao Shiwei; Yu Zhizhen; Chen Kun; Zuo Jiaxu

2013-01-01

Background: Loss of offsite power (LOOP) is a possible accident to any type of reactor, and this accident can reflect the main idea of reactor safety design. Therefore, it is very important to conduct a study on probabilistic safety assessment (PSA) of the molten salt reactor that is under LOOP circumstance. Purpose: The aim is to calculate the release frequency of molten salt radioactive material to the core caused by LOOP, and find out the biggest contributor to causing the radioactive release frequency. Methods: We carried out the PSA analysis of the LOOP using the PSA process risk spectrum, and assumed that the primary circuit had no valve and equipment reliability data based on the existing mature power plant equipment reliability data. Results: Through the PSA analysis, we got the accident sequences of the release of radioactive material to the core caused by LOOP and its frequency. The results show that the release frequency of molten salt radioactive material to the core caused by LOOP is about 2×10 -11 /(reactor ·year), which is far below that of the AP1000 LOOP. In addition, through the quantitative analysis, we obtained the point estimation and interval estimation of uncertainty analysis, and found that the biggest contributor to cause the release frequency of radioactive material to the core is the reactor cavity cooling function failure. Conclusion: This study provides effective help for the design and improvement of the following molten salt reactor system. (authors)

COMSORS: A light water reactor chemical core catcher

International Nuclear Information System (INIS)

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

1997-01-01

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

Indian programme on molten salt cooled nuclear reactors

International Nuclear Information System (INIS)

DuIera, I.V.; Vijayan, P.K.; Sinha, R.K.

2013-01-01

Bhabha Atomic Research Centre (BARC) is developing a 600 MWth pebble bed high temperature reactor, cooled by natural circulation of molten fluoride salts and is capable of supplying process heat at 1000 ℃ to facilitate hydrogen production by splitting water. BARC has also initiated studies for a reactor concept in which salts of molten fluoride fuel and coolant in fluid form, flows through the reactor core of graphite moderator, resulting in nuclear fission within the molten salt. For thorium fuel cycle, this concept is very attractive, since the fuel can be re-processed on-line, enabling it to be an efficient neutron breeder. (author)

Apparatus and Method for Increasing the Diameter of Metal Alloy Wires Within a Molten Metal Pool

Science.gov (United States)

Hartman, Alan D.; Argetsinger, Edward R.; Hansen, Jeffrey S.; Paige, Jack I.; King, Paul E.; Turner, Paul C.

2002-01-29

In a dip forming process the core material to be coated is introduced directly into a source block of coating material eliminating the need for a bushing entrance component. The process containment vessel or crucible is heated so that only a portion of the coating material becomes molten, leaving a solid portion of material as the entrance port of, and seal around, the core material. The crucible can contain molten and solid metals and is especially useful when coating core material with reactive metals. The source block of coating material has been machined to include a close tolerance hole of a size and shape to closely fit the core material. The core material moves first through the solid portion of the source block of coating material where the close tolerance hole has been machined, then through a solid/molten interface, and finally through the molten phase where the diameter of the core material is increased. The crucible may or may not require water-cooling depending upon the type of material used in crucible construction. The system may operate under vacuum, partial vacuum, atmospheric pressure, or positive pressure depending upon the type of source material being used.

Molten salt reactors. Synthesis of studies realized between 1973 and 1983. General synthesis

International Nuclear Information System (INIS)

Hery, M.; Lecocq, A.

1983-03-01

After a brief recall of the MSBR project, French studies on molten salt reactors are summed up. Theoretical and experimental studies for a graphite moderated 1000 MWe reactor using molten Li, Be, Th and U fluorides cooled by salt-lead direct contact are given. These studies concern the core, molten salt chemistry, graphite, metals (molybdenum, alloy TZM), corrosion, reactor components [fr

Chemistry and technology of Molten Salt Reactors - history and perspectives

International Nuclear Information System (INIS)

Uhlir, Jan

2007-01-01

Molten Salt Reactors represent one of promising future nuclear reactor concept included also in the Generation IV reactors family. This reactor type is distinguished by an extraordinarily close connection between the reactor physics and chemical technology, which is given by the specific features of the chemical form of fuel, representing by molten fluoride salt and circulating through the reactor core and also by the requirements of continuous 'on-line' reprocessing of the spent fuel. The history of Molten Salt Reactors reaches the period of fifties and sixties, when the first experimental Molten Salt Reactors were constructed and tested in ORNL (US). Several molten salt techniques dedicated to fresh molten salt fuel processing and spent fuel reprocessing were studied and developed in those days. Today, after nearly thirty years of discontinuance, a renewed interest in the Molten Salt Reactor technology is observed. Current experimental R and D activities in the area of Molten Salt Reactor technology are realized by a relatively small number of research institutions mainly in the EU, Russia and USA. The main effort is directed primarily to the development of separation processes suitable for the molten salt fuel processing and reprocessing technology. The techniques under development are molten salt/liquid metal extraction processes, electrochemical separation processes from the molten salt media, fused salt volatilization techniques and gas extraction from the molten salt medium

Performance Evaluation of Resilience using Service Relocation for GMPLS Networks

DEFF Research Database (Denmark)

Wessing, Henrik; Herrmann, Sven; Ruepp, Sarah Renée

2015-01-01

Cloud computing today represents an ever increasing part of the data exchanged over the Internet. This influences the core network, where the special properties of cloud based services can be leveraged to increase the network efficiency. This paper explores the concept of relocation for the backup...... path when providing resiliency to cloud services in the network. The control plane of a GMPLSenabled optical network is simulated, thus evaluating the proposed Path Computation Element architecture for service relocation. The results show that for increased offered traffic in the network, the overall...

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

Energy Technology Data Exchange (ETDEWEB)

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

1992-09-01

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

Conformational selection in the molten globule state of the nuclear coactivator binding domain of CBP

DEFF Research Database (Denmark)

Kjærgaard, Magnus; Teilum, Kaare; Poulsen, Flemming M

2010-01-01

Native molten globules are the most folded kind of intrinsically disordered proteins. Little is known about the mechanism by which native molten globules bind to their cognate ligands to form fully folded complexes. The nuclear coactivator binding domain (NCBD) of CREB binding protein is particul......Native molten globules are the most folded kind of intrinsically disordered proteins. Little is known about the mechanism by which native molten globules bind to their cognate ligands to form fully folded complexes. The nuclear coactivator binding domain (NCBD) of CREB binding protein....... Biophysical studies show that despite the molten globule nature of the domain, it contains a small cooperatively folded core. By NMR spectroscopy, we have demonstrated that the folded core of NCBD has a well ordered conformer with specific side chain packing. This conformer resembles the structure of the NCBD...

Evaluation of a cavity flooding strategy for the prevention of reactor vessel failure in a severe accident

Energy Technology Data Exchange (ETDEWEB)

Park, Rae Joon; Je, Moo Sung; Park, Chang Kyoo [Korea Atomic Energy Research Institute, TaeJon (Korea, Republic of)

1994-10-01

As a part of the evaluation of accident management strategies for severe accident prevention or mitigation in a station blackout scenario for YGN 3 and 4, an external vessel cooling strategy for the prevention of reactor vessel failure has been estimated using the MAAP4 computer code. The sensitivity studies have been performed such as actuating timings and the number of spray pumps used. To explore external vessel cooling strategies, containment spray pumps were actuated by varying time spanning core uncovery, core melting and relocation of molten core material. It was shown that flooding of the reactor cavity using the containment spray system may prevent reactor vessel failure but may not prevent the failure of the relocation of molten core material during the station blackout sequence of YGN 3 and 4. Reactor vessel failure can be prevented by external vessel cooling using condensed water from the operation of two containment spray pumps at the time of core melting and using water from the operation of one containment spray pumps at the time of core melting and using water from the operation of one containment spray pump at the time of core uncovery. (Author) 46 refs., 26 figs., 5 tabs.

Transient Analyses for a Molten Salt Transmutation Reactor Using the Extended SIMMER-III Code

International Nuclear Information System (INIS)

Wang, Shisheng; Rineiski, Andrei; Maschek, Werner; Ignatiev, Victor

2006-01-01

Recent developments extending the capabilities of the SIMMER-III code for the dealing with transient and accidents in Molten Salt Reactors (MSRs) are presented. These extensions refer to the movable precursor modeling within the space-time dependent neutronics framework of SIMMER-III, to the molten salt flow modeling, and to new equations of state for various salts. An important new SIMMER-III feature is that the space-time distribution of the various precursor families with different decay constants can be computed and took into account in neutron/reactivity balance calculations and, if necessary, visualized. The system is coded and tested for a molten salt transmuter. This new feature is also of interest in core disruptive accidents of fast reactors when the core melts and the molten fuel is redistributed. (authors)

Basic visualization experiments on eutectic reaction of boron carbide and stainless steel under sodium-cooled fast reactor conditions

International Nuclear Information System (INIS)

Yamano, Hidemasa; Suzuki, Tohru; Kamiyama, Kenji; Kudo, Isamu

2016-01-01

This paper describes basic visualization experiments on eutectic reaction and relocation of boron carbide (B 4 C) and stainless steel (SS) under a high temperature condition exceeding 1500degC as well as the importance of such behaviors in molten core during a core disruptive accident in a Generation-IV sodium-cooled fast reactor (750 MWe class) designed in Japan. At first, a reactivity history was calculated using an exact perturbation calculation tool taking into account expected behaviors. This calculation indicated the importance of a relocation behavior of the B 4 C-SS eutectic because its behavior has a large uncertainty in the reactivity history. To clarify this behavior, basic experiments were carried out by visualizing the reaction of a B 4 C pellet contacted with molten SS in a high temperature-heating furnace. The experiments have shown the eutectic reaction visualization as well as freezing and relocation of the B 4 C-SS eutectic in upper part of the solidified test piece due to the density separation. (author)

Study on the quench behavior of molten fuel material jet into coolant

International Nuclear Information System (INIS)

Abe, Yutaka; Kizu, Tetsuya; Arai, Takahiro; Nariai, Hideki; Chitose, Keiko; Koyama, Kazuya

2004-01-01

In a core disruptive accident (CDA) of a Fast Breeder Reactor, the post accident heat removal (PAHR) is crucial for the accident mitigation. The molten core material should be solidified in the sodium coolant in the reactor vessel. In the present experiment, molten material jet is injected into water to experimentally obtain fragments and the visualized information of the fragmentation. The distributed particle behavior of the molten material jet is observed with high-speed video camera. The distributions of the fragmented droplet diameter from the molten material jet are evaluated by correcting the solidified particles. The experimental results of the mean fragmented droplet diameter are compared with the existing theories. Consequently, the fragmented droplet diameter is close to the value estimated based on the Kelvin-Helmholtz instability. Once the particle diameter of the fragmented molten material could be known from a hydrodynamic model, it becomes possible to estimate the mass ratio of the molten particle to the total injected mass by combining an appropriate heat transfer model. The heat transfer model used in the present study is composed of the fragmentation model based on the Kelvin-Helmholtz instability. The mass ratio of the molten fragment to total mass of the melted mixed oxide fuel in sodium coolant estimated in the present study is very small. The result means that most of the molten mixed oxide fuel material injected into the sodium coolant can be cooled down under the solidified temperature, that is so called quenched, if the amount of the coolant is sufficient. (author)

Nuclear core catchers

International Nuclear Information System (INIS)

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

1976-01-01

A receptacle is described for taking the molten fragments of a nuclear reactor during a reactor core fusion accident. The receptacle is placed under the reactor. It includes at least one receptacle for the reactor core fragments, with a dome shaped part to distribute the molten fragments and at least one outside layer of alumina bricks around the dome. The characteristic of this receptacle is that the outer layer of bricks contains neutron poison rods which pass through the bricks and protrude in relation to them [fr

Natural convection heat transfer in SIGMA experiment

International Nuclear Information System (INIS)

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

2004-01-01

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

2D model for melt progression through rods and debris

International Nuclear Information System (INIS)

Fichot, F.

2001-01-01

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Study of the pyrochemical treatment-recycling process of the Molten Salt Reactor fuel

International Nuclear Information System (INIS)

Boussier, H.; Heuer, D.

2010-01-01

The Separation Processes Studies Laboratory (Commissariat a l'energie Atomique) has made a preliminary assessment of the reprocessing system associated with Molten Salt Fast Reactor (MSFR). The scheme studied in this paper is based on the principle of reductive extraction and metal transfer that constituted the core process designed for the Molten Salt Breeder Reactor (MSBR), although the flow diagram has been adapted to the current needs of the Molten Salt Reactor Fast (MSFR).

Fast Thorium Molten Salt Reactors Started with Plutonium

International Nuclear Information System (INIS)

Merle-Lucotte, E.; Heuer, D.; Le Brun, C.; Brissot, R.; Liatard, E.; Meplan, O.; Nuttin, A.; Mathieu, L.

2006-01-01

One of the pending questions concerning Molten Salt Reactors based on the 232 Th/ 233 U fuel cycle is the supply of the fissile matter, and as a consequence the deployment possibilities of a fleet of Molten Salt Reactors, since 233 U does not exist on earth and is not yet produced in the current operating reactors. A solution may consist in producing 233 U in special devices containing Thorium, in Pressurized Water or Fast Neutrons Reactors. Two alternatives to produce 233 U are examined here: directly in standard Molten Salt Reactors started with Plutonium as fissile matter and then operated in the Th/ 233 U cycle; or in dedicated Molten Salt Reactors started and fed with Plutonium as fissile matter and Thorium as fertile matter. The idea is to design a critical reactor able to burn the Plutonium and the minor actinides presently produced in PWRs, and consequently to convert this Plutonium into 233 U. A particular reactor configuration is used, called 'unique channel' configuration in which there is no moderator in the core, leading to a quasi fast neutron spectrum, allowing Plutonium to be used as fissile matter. The conversion capacities of such Molten Salt Reactors are excellent. For Molten Salt Reactors only started with Plutonium, the assets of the Thorium fuel cycle turn out to be quickly recovered and the reactor's characteristics turn out to be equivalent to Molten Salt Reactors operated with 233 U only. Using a combination of Molten Salt Reactors started or operated with Plutonium and of Molten Salt Reactors started with 233 U, the deployment capabilities of these reactors fully satisfy the condition of sustainability. (authors)

Molten salt reactor type

International Nuclear Information System (INIS)

1977-01-01

This document is one of the three parts of a first volume devoted to the compilations of American data on the molten salt reactor concept. This part describes the MSBR core (data presented are from ORNL 4541). The principal characteristics of the core are presented in tables together with plane and elevation drawings, stress being put upon the reflector, and loading and unloading. Neutronic, and thermal and hydraulic characteristics (core and reflectors) are more detailed. The reasons why a graphite with a tight graphite layer has been chosen are briefly exposed. The physical properties of the standard graphite (irradiation behavior) have been determined for an isotropic graphite with fine granulometry; its dimensional variations largely ressemble that of Gilsonite. The mechanical stresses computed (Wigner effect) do not implicate in any way the graphite stack [fr

Assessment of steam explosion impact on KNGR plant

Energy Technology Data Exchange (ETDEWEB)

Hwang, Moon Kyu; Park, Soo Yong; Park, Ik Kyu

1999-03-01

In present day light water reactors, if complete and prolonged failure of normal and emergency coolant flow occurs, fission product decay heat could cause melting of the reactor fuel. If the molten fuel mass accumulates it may relocate into reactor lower plenum and if the lower head fails it may eventually be brought into the reactor cavity. In such course of core melt relocation, the opportunity for fuel-coolant interactions (FCI) arises as the core melt relocates into water pool in reactor vessel as well as in reactor cavity and also, as a consequence of implementing accident management strategies involving water addition to a degraded or molten core. This report presents the methodologies and their results for assessment of steam explosion impact on KNGR plant integrity. Both in-vessel and ex-vessel phenomena are addressed. For in-vessel steam explosion, TRACER-II code is used for assessment of pressure load, while bounding calculations are applied for ex-vessel analysis. Analysis shows that the integrity of reactor pressure vessel lower head is preserved during the in-vessel event and the probability that the containment integrity is challenged is very low, even when ex-vessel steam explosion is allowed due to reactor vessel failure. (Author). 15 refs., 2 tabs., 4 figs.

Assessment of steam explosion impact on KNGR plant

International Nuclear Information System (INIS)

Hwang, Moon Kyu; Park, Soo Yong; Park, Ik Kyu

1999-03-01

In present day light water reactors, if complete and prolonged failure of normal and emergency coolant flow occurs, fission product decay heat could cause melting of the reactor fuel. If the molten fuel mass accumulates it may relocate into reactor lower plenum and if the lower head fails it may eventually be brought into the reactor cavity. In such course of core melt relocation, the opportunity for fuel-coolant interactions (FCI) arises as the core melt relocates into water pool in reactor vessel as well as in reactor cavity and also, as a consequence of implementing accident management strategies involving water addition to a degraded or molten core. This report presents the methodologies and their results for assessment of steam explosion impact on KNGR plant integrity. Both in-vessel and ex-vessel phenomena are addressed. For in-vessel steam explosion, TRACER-II code is used for assessment of pressure load, while bounding calculations are applied for ex-vessel analysis. Analysis shows that the integrity of reactor pressure vessel lower head is preserved during the in-vessel event and the probability that the containment integrity is challenged is very low, even when ex-vessel steam explosion is allowed due to reactor vessel failure. (Author). 15 refs., 2 tabs., 4 figs

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Evaluation of upward heat flux in ex-vessel molten core heat transfer using MELCOR

International Nuclear Information System (INIS)

Park, S.Y.; Park, J.H.; Kim, S.D.; Kim, D.H.; Kim, H.D.

2000-01-01

The purpose of this study is to share experiences of MELCOR application to resolve the molten corium-concrete interaction (MCCI) issue in the Korea Next Generation Reactor (KNGR). In the evaluation of concrete erosion, the heat transfer modeling from the molten corium internal to the corium pool surface is very important and uncertain. MELCOR employs Kutateladze or Greene's bubble-enhanced heat transfer model for the internal heat transfer. The phenomenological uncertainty is so large that the model provides several model parameters in addition to the phenomenological model for user flexibility. However, the model parameters do not work on Kutateladze correlation at the top of the molten layer. From our experience, a code modification is suggested to match the upward heat flux with the experimental results. In this analysis, minor modification was carried out to calculate heat flux from the top molten layer to corium surface, and efforts were made to find out the best value of the model parameter based on upward heat flux of MACE test M1B. Discussion also includes its application to KNGR. (author)

Concept of the demonstration molten salt unit for the transuranium elements transmutations

International Nuclear Information System (INIS)

Alekseev, P.; Dudnikov, A.; Prusakov, V.; Subbotin, S.; Zakirov, R.; Lelek, V.; Peka, I.

1999-01-01

Fluorine reprocessing is discussed of spent fuel and of fluoride molten salt reactor in critical and subcritical modes for plutonium and minor actinides burning. International collaboration for creation of such system is proposed. Additional neutron source in the core will have positive influence on the transmutation processes in the reactor. Demonstration critical molten salt reactor of small power capacity will permit to decide the most part of problems inherent to large critical reactors and subcritical drivers. It could be expected that fluoride molten salt transmuter can work without accelerator as a critical reactor. (author)

Experimental study on thermal interaction between a high-temperature molten jet and plates

International Nuclear Information System (INIS)

Sato, K.; Saito, M.; Furutani, A.; Isozaki, M.; Imahori, S.; Konishi, K.

1994-01-01

This paper summarizes the recent simulant experiments to study molten corium-structure interactions under postulated core disruptive accident (CDA) conditions in liquid-metal fast breeder reactors (LMFMRs). These experiments were conducted in the MELT-II facility generating high-temperature molten simulants by an induction heating technique. From a series of molten jet-structure interaction experiments, the effects of the solidified crust layer and molten layer on the erosion behavior were identified, and analytical models were developed to assess the structure erosion rate with and without crust formation. Especially, we revealed the inherent mitigation mechanism that when the molten oxide jet with high melting point falls down onto the structure plate, solidified crust of the oxide can significantly reduce the erosion rate. (author)

Neutron shielding studies on an advanced molten salt fast reactor design

International Nuclear Information System (INIS)

Merk, Bruno; Konheiser, Jörg

2014-01-01

Highlights: • Material damage due to irradiation has already been discovered at the MSRE. • Neutronic analysis of MSFR with curved blanket wall geometry. • Neutron fluence limit at the wall of the outer vessel can be kept for 80 years. • Shielded MSFR core will be of same dimension than a SFR core. - Abstract: The molten salt reactor technology has gained some new interest. In contrast to the historic molten salt reactors, the current projects are based on designing a molten salt fast reactor. Thus the shielding becomes significantly more challenging than in historic concepts. One very interesting and innovative result of the most recent EURATOM project on molten salt reactors – EVOL – is the fluid flow optimized design of the inner reactor vessel using curved blanket walls. The developed structure leads to a very uniform flow distribution. The design avoids all internal structures. Based on this new geometry a model for neutron physics calculation is presented. The major steps are: the modeling of the curved geometry in the unstructured mesh neutron transport code HELIOS and the determination of the real neutron flux and power distribution for this new geometry. The developed model is then used for the determination of the neutron fluence distribution in the inner and outer wall of the system. Based on these results an optimized shielding strategy is developed for the molten salt fast reactor to keep the fluence in the safety related outer vessel below expected limit values. A lifetime of 80 years can be assured, but the size of the core/blanket system will be comparable to a sodium cooled fast reactor. The HELIOS results are verified against Monte-Carlo calculations with very satisfactory agreement for a deep penetration problem

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

Study on mechanical interaction between molten alloy and water

International Nuclear Information System (INIS)

Nishimura, Satoshi; Ueda, Nobuyuki; Nishi, Yoshihisa; Furuya, Masahiro; Kinoshita, Izumi

1999-01-01

Simulant experiments using low melting point molten alloy and water have been conducted to observe both fragmentation behavior of molten jet and boiling phenomena of water, and to measure both particle size and shape of fragmented solidified jet, focusing on post-pin-failure molten fuel-coolant interaction (FCl) which was important to evaluate the sequence of the initiating phase for metallic fueled FBR. In addition, characteristics of coolant boiling phenomena on FCIs have been investigated, focusing on the boiling heat transfer in the direct contact heat transfer mode. As a results, it is concluded that the fragmentation of poured molten alloy jet is affected by a degree of boiling of water and is classified into three modes by thermal conditions of both the instantaneous contact interface temperature of two liquids and subcooling of water. In the case of forced convection boiling in direct contact mode, it is found that the heat transfer performance is enhanced by increase of the heat transfer area, due to oscillation of the surface and fragmentation of molten alloy. As a results of preliminary investigation of FCI behavior for metallic fuel core based on these results, it is expected that the ejected molten fuel is fragmented into almost spherical particles due to the developed boiling of sodium. (author)

Fragmentation of molten metal drop with instantaneous contact temperature below the boiling point of Na

International Nuclear Information System (INIS)

Inukai, S.; Sugiyama, K.; Nishimura, S.; Kinoshita, I.

2001-01-01

The consequence of the core disruptive accidents in metallic-fueled Na-cooled reactors is strongly affected by the feedback reactivity originating in the boiling of Na and the dispersion of molten fuel due to fuel-coolant interactions. The design of the core configuration to promote the dispersion of molten fuel is therefore very important for social acceptance. It has been recognized in this context that metallic fuel has a potentiality to make liquefied fuel with fuel pin tube even in the temperature range below the boiling point of Na. If the liquefied fuel solidified without fuel-coolant interactions in the core region, this event leads the core condition to a pessimistic scenario of re-criticality. As a basic study related to this problem, the present experimental study investigates the possibility of fragmentation of metal drop with instantaneous contact temperature below the boiling point of Na (883 C). The molten Al drop, which has a melting point of 660 C above the operational temperature range of core, was selected as a simulant of liquefied fuel in the present study. Al particles of 5 g or 0.56 g were heated up to the initial temperature ranging from 850 C to 1113 C in a crucible by using an electric heater. The molten Al drop was dropped into a sodium pool adjusted the temperature from 280 C to 499 C. The Al drop at initial temperature sufficiently higher that the boiling point of Na was observed to fragment into pieces under the condition of instantaneous contact temperature below the boiling point of Na. It is confirmed that the fragmentation is caused due to the thermal interactions between the molten Al and the Na entrapped into the drop. (author)

Fragmentation of molten metal drop with instantaneous contact temperature below the boiling point of Na

Energy Technology Data Exchange (ETDEWEB)

Inukai, S.; Sugiyama, K. [Hokkaido Univ., Dept. of Nuclear Engineering, Sapporo (Japan); Nishimura, S.; Kinoshita, I. [Central Research Institute of Electric Power Industry, Tokyo (Japan)

2001-07-01

The consequence of the core disruptive accidents in metallic-fueled Na-cooled reactors is strongly affected by the feedback reactivity originating in the boiling of Na and the dispersion of molten fuel due to fuel-coolant interactions. The design of the core configuration to promote the dispersion of molten fuel is therefore very important for social acceptance. It has been recognized in this context that metallic fuel has a potentiality to make liquefied fuel with fuel pin tube even in the temperature range below the boiling point of Na. If the liquefied fuel solidified without fuel-coolant interactions in the core region, this event leads the core condition to a pessimistic scenario of re-criticality. As a basic study related to this problem, the present experimental study investigates the possibility of fragmentation of metal drop with instantaneous contact temperature below the boiling point of Na (883 C). The molten Al drop, which has a melting point of 660 C above the operational temperature range of core, was selected as a simulant of liquefied fuel in the present study. Al particles of 5 g or 0.56 g were heated up to the initial temperature ranging from 850 C to 1113 C in a crucible by using an electric heater. The molten Al drop was dropped into a sodium pool adjusted the temperature from 280 C to 499 C. The Al drop at initial temperature sufficiently higher that the boiling point of Na was observed to fragment into pieces under the condition of instantaneous contact temperature below the boiling point of Na. It is confirmed that the fragmentation is caused due to the thermal interactions between the molten Al and the Na entrapped into the drop. (author)

Preliminary analysis on in-core fuel management optimization of molten salt pebble-bed reactor

International Nuclear Information System (INIS)

Xia Bing; Jing Xingqing; Xu Xiaolin; Lv Yingzhong

2013-01-01

The Nuclear Hot Spring (NHS) is a molten salt pebble-bed reactor featured by full power natural circulation. The unique horizontal coolant flow of the NHS demands the fuel recycling schemes based on radial zoning refueling and the corresponding method of fuel management optimization. The local searching algorithm (LSA) and the simulated annealing algorithm (SAA), the stochastic optimization methods widely used in the refueling optimization problems in LWRs, were applied to the analysis of refueling optimization of the NHS. The analysis results indicate that, compared with the LSA, the SAA can survive the traps of local optimized solutions and reach the global optimized solution, and the quality of optimization of the SAA is independent of the choice of the initial solution. The optimization result gives excellent effects on the in-core power flattening and the suppression of fuel center temperature. For the one-dimensional zoning refueling schemes of the NHS, the SAA is an appropriate optimization method. (authors)

Experimental investigation of natural convection heat transfer in volumetrically heated spherical segments. Final report

International Nuclear Information System (INIS)

Asfia, F.; Dhir, V.

1998-03-01

One strategy for preventing the failure of lower head of a nuclear reactor vessel is to flood the concrete cavity with subcooled water in accidents in which relocation of core material into the vessel lower head occurs. After the core material relocates into the vessel, a crust of solid material forms on the inner wall of the vessel, however, most of the pool remains molten and natural convection exists in the pool. At present, uncertainty exists with respect to natural convection heat transfer coefficients between the pool of molten core material and the reactor vessel wall. In the present work, experiments were conducted to examine natural convection heat transfer in internally heated partially filled spherical pools with external cooling. In the experiments, Freon-113 contained in a Pyrex bell jar was used as a test liquid. The pool was bounded with a spherical segment at the bottom, and was heated with magnetrons taken from a conventional microwave oven. The vessel was cooled from the outside with natural convection of water or with nucleate boiling of liquid nitrogen

MCCI study for Pressurized Heavy Water Reactor under hypothetical accident condition

International Nuclear Information System (INIS)

Verma, Vishnu; Mukhopadhyay, Deb; Chatterjee, B.; Singh, R.K.; Vaze, K.K.

2011-01-01

In case of severe core damage accident in Pressurized Heavy Water Reactor (PHWR), large amount of molten corium is expected to come out into the calandria vault due to failure of calandria vessel. Molten corium at high temperature is sufficient to decompose and ablate concrete. Such attack could fail CV by basement penetration. Since containment is ultimate barrier for activity release. The Molten Core Concrete Interaction (MCCI) of the resulting pool of debris with the concrete has been identified as an important part of the accident sequence. MCCI Analysis has been carried out for PHWR for a hypothetical accident condition where total core material is considered to be relocated in calandria vault. Concrete ablation rate in vertical and radial direction is evaluated for rectangular geometry using MEDICIS module of ASTEC Code. Amount of gases released during MCCI is also evaluated. (author)

Workshop on large molten pool heat transfer summary and conclusions

International Nuclear Information System (INIS)

1994-01-01

The CSNI Workshop on Large Molten Heat Transfer held at Grenoble (France) in March 1994 was organised by CSNI's Principal Working Group on the Confinement of Accidental Radioactive Releases (PWG4) with the cooperation of the Principal Working Group on Coolant System Behaviour (FWG2) and in collaboration with the Grenoble Nuclear Research Centre of the French Commissariat a l'Energie Atomique (CEA). Conclusions and recommendations are given for each of the five sessions of the workshops: Feasibility of in-vessel core debris cooling through external cooling of the vessel; Experiments on molten pool heat transfer; Calculational efforts on molten pool convection; Heat transfer to the surrounding water - experimental techniques; Future experiments and ex-vessel studies (open forum discussion)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-09-15

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

Molten salt related extensions of the SIMMER-III code and its application for a burner reactor

International Nuclear Information System (INIS)

Wang Shisheng; Rineiski, Andrei; Maschek, Werner

2006-01-01

Molten salt reactors (MSRs) can be used as effective burners of plutonium (Pu) and minor actinides (MAs) from light water reactor (LWR) spent fuel. In this paper a study was made to examine the thermal hydraulic behaviour of the conceptual design of the molten salt advanced reactor transmuter (MOSART) [Ignatiev, V., Feynberg, O., Myasnikov, A., Zakirov, R., 2003a. Neutronic properties and possible fuel cycle of a molten salt transmuter. Proceedings of the 2003 ANS/ENS International Winter Meeting (GLOBAL 2003), Hyatt Regency, New Orleans, LA, USA 16-20 November 2003]. The molten salt fuel is a ternary NaF-LiF-BeF 2 system fuelled with ca. 1 mol% typical compositions of transuranium-trifluorides (PuF 3 , etc.) from light water reactor spent fuel. The MOSART reactor core does not contain graphite structure elements to guide the flow, so the neutron spectrum is rather hard in order to improve the burning performance. Without those structure elements in the core, the molten salt in core flows freely and the flow pattern could be potentially complicated and may affect significantly the fuel temperature distribution in the core. Therefore, some optimizations of the salt flow pattern may be needed. Here, the main attention has been paid to the fluid dynamic simulations of the MOSART core with the code SIMMER-III [Kondo, Sa., Morita, K., Tobita, Y., Shirakawa, K., 1992. SIMMER-III: an advanced computer program for LMFBR severe accident analysis. Proceedings of the ANP' 92, Tokyo, Japan; Kondo, Sa., Tobita, Y., Morita, K., Brear, D.J., Kamiyama, K., Yamano, H., Fujita, S., Maschek, W., Fischer, E.A., Kiefhaber, E., Buckel, G., Hesselschwerdt, E., Flad, M., Costa, P., Pigny, S., 1999. Current status and validation of the SIMMER-III LMFR safety analysis code. Proceedings of the ICONE-7, Tokyo, Japan], which was originally developed for the safety assessment of sodium-cooled fast reactors and recently extended by the authors for the thermo-hydraulic and neutronic models so as

Research on the fundamental process of thermal-hydraulic behaviors in severe accident. Breakup and cooling of molten material jets. JAERI's nuclear research promotion program, H10-027-2. Contract research

International Nuclear Information System (INIS)

Sugiyama, Ken-ichiro; Iguchi, Kentarou

2002-03-01

Core melt accidents could lead to the pouring of molten core materials into a body of water accumulating in the reactor lower head in the form of jets with a few centimeters up to a few tens of centimeters. If molten core jets penetrate the body of water without breakup. A poor coolability of the molten core bed would occur, which means the difficulty of maintaining the molten core bed in the reactor vessel. Hence, the breakup mechanism of molten core jets had to be well understood for the evaluation of the coolability of molten core bed. The objective of the present experimental study is to confirm that, even in molten material jets, the breakup of jet originating in the coolant entrained within a molten material jet due to 'the organized motion' between the coolant and the jet, which has been recognized in the field of fluid mechanics, is caused. The first series of experiment was conducted to observe this type of breakup by using molten tin jets up to 25 mm in diameter. Molten tin jet was expected to easily cause this kind of breakup of jet because of a low kinematic viscosity, which means a easy transformation of jet due to the organized motion for the coolant to entrain. The second series of experiment was conducted by using molten copper jet of 25 mm in diameter, of which kinematic viscosity is about same as that of molten UO 2 . The breakup of jet due to the entrainment of the coolant was observed up to high ambient Weber numbers, which cover the atomization regime. The mechanism of the breakup observed in the present study is able to reasonably explain the apparent difference between the breakup lengths of 150 kg-scale corium jets and the breakup lengths of about 8 kg-scale lead-bismuth alloy jets. The breakup by the mechanism reported here also assures a high coolability of molten jets because of an efficient entrainment of coolant within the jet. (author)

Molten Fuel Mass Assessment for Channel Flow Blockage Event in CANDU6

International Nuclear Information System (INIS)

Lee, Kwang Ho; Kim, Yong Bae; Choi, Hoon; Park, Dong Hwan

2011-01-01

In CANDU6, a fuel channel flow blockage causes a sudden reduction of flow through the blocked channel. Depending on the severity of the blockage, the reduced flow through the channel can result in severe heat up of the fuel, hence possibly leading to pressure tube and calandria tube failure. If the calandria tube does not fail the fuel and sheath would continue to heat up, and ultimately melting could occur. Eventually, molten material runs down onto the pressure tube. Even a thin layer of molten material in contact with the pressure tube causes the pressure tube and calandreia tube to heat up rapidly. The thermal transient is so rapid that failure temperatures are reached quickly. After channel failure, the contents of the channel, consisting of superheated coolant, fission products and possibly overheated of molten fuel, are rapidly discharged into the moderator. Fuel discharged into the moderator is quenched and cooled. The rapid discharge of hot fuel and coolant into the calandria causes the moderator pressure and temperature to increase, which may cause damage to some in-core components. Thus, the assessment results of molten fuel mass are inputs to the in-core damage analysis. In this paper, the analysis methodology and results of molten fuel mass assessment for the channel flow blockage event are presented

25 CFR 700.93 - Relocation plan.

Science.gov (United States)

2010-04-01

... 25 Indians 2 2010-04-01 2010-04-01 false Relocation plan. 700.93 Section 700.93 Indians THE OFFICE OF NAVAJO AND HOPI INDIAN RELOCATION COMMISSION OPERATIONS AND RELOCATION PROCEDURES General Policies and Instructions Definitions § 700.93 Relocation plan. The relocation plan shall be the plan prepared...

Cold crucible technique for interaction test of molten corium with structure

International Nuclear Information System (INIS)

Ha, Kwang Soon; An, Sang Mo; Min, Beong Tae; Kim, Hwan Yeol

2012-01-01

During a severe accident, the molten corium might interact with several structures in a nuclear power plant such as core peripheral structures, lower plenum, lower head vessel, and external structures of a reactor vessel. The interaction of the molten corium with the structure depends on the molten corium composition, temperature, structural materials, and environmental conditions such as pressure and humidity. For example, the interaction of a metallic molten corium containing metal uranium (U) and zirconium (Zr) with the oxidized steel structure (Fe 2O3 ) is affected by not only thermal ablation but oxidation reduction reaction because the oxidation quotients of the U and Zr are higher than that of Fe. KAERI set up an experimental facility and technique using a cold crucible melting method to verify the interaction mechanism between the metallic molten corium and structural materials. This technique includes the generation of the metallic melt, melt delivery, measurement of the interaction process, and post analyses after the test

Dynamic behaviour and neutron noise in molten salt reactors with circulating perturbations

Energy Technology Data Exchange (ETDEWEB)

Pazsit, I.; Dykin, V. [Chalmers Univ. of Tech., Nuclear Engineering, Goteborg (Sweden)

2014-07-01

This paper concerns the calculation of the neutron noise induced in Molten Salt Reactors (MSR) by the random fluctuations in space and time of the molten fuel cross sections which travel together with the fuel and pass the core region. The effect of such fluctuations was already discussed in several publications. The novelty of the present paper is that it takes into account that in addition to the delayed neutron precursors, also the cross section perturbations themselves, whose passing through the core induces the in-core neutron noise, return to the core inlet via the external loop from the core exit. The corresponding theory is developed, and some quantitative investigations are made of the characteristics of the noise, which can be attributed to the recirculation of the perturbation to the core. It is shown that the effect of the returning of the perturbations, even though it is also associated with a temporal decay, has a much stronger effect on the neutron noise spectra than that of the recirculation of the delayed neutron precursors. (author)

Dynamic behaviour and neutron noise in molten salt reactors with circulating perturbations

International Nuclear Information System (INIS)

Pazsit, I.; Dykin, V.

2014-01-01

This paper concerns the calculation of the neutron noise induced in Molten Salt Reactors (MSR) by the random fluctuations in space and time of the molten fuel cross sections which travel together with the fuel and pass the core region. The effect of such fluctuations was already discussed in several publications. The novelty of the present paper is that it takes into account that in addition to the delayed neutron precursors, also the cross section perturbations themselves, whose passing through the core induces the in-core neutron noise, return to the core inlet via the external loop from the core exit. The corresponding theory is developed, and some quantitative investigations are made of the characteristics of the noise, which can be attributed to the recirculation of the perturbation to the core. It is shown that the effect of the returning of the perturbations, even though it is also associated with a temporal decay, has a much stronger effect on the neutron noise spectra than that of the recirculation of the delayed neutron precursors. (author)

Thermal conditions and functional requirements for molten fuel containment

International Nuclear Information System (INIS)

Kang, C.S.; Torri, A.

1980-05-01

This paper discusses the configuration and functional requirements for the molten fuel containment system (MFCS) in the GCFR demonstration plant design. Meltdown conditions following a loss of shutdown cooling (LOSC) accident were studied to define the core debris volume for a realistic meltdown case. Materials and thicknesses of the molten fuel container were defined. Stainless steel was chosen as the sacrificial material and magnesium oxide was chosen as the crucible material. Thermal conditions for an expected quasi-steady state were analyzed. Highlights of the functional requirements which directly affect the MFCS design are discussed

Physical properties of molten core materials: Zr-Ni and Zr-Cr alloys measured by electrostatic levitation

Energy Technology Data Exchange (ETDEWEB)

Ohishi, Yuji, E-mail: ohishi@see.eng.osaka-u.ac.jp [Graduate School of Engineering, Osaka University (Japan); Kondo, Toshiki [Graduate School of Engineering, Osaka University (Japan); Ishikawa, Takehiko [Japan Aerospace Exploration Agency (Japan); SOKEN-DAI (Graduate University for Advanced Studies) (Japan); Okada, Junpei T. [Institute for Materials Research, Tohoku University (Japan); Watanabe, Yuki [Advanced Engineering Services Co. Ltd. (Japan); Muta, Hiroaki; Kurosaki, Ken [Graduate School of Engineering, Osaka University (Japan); Yamanaka, Shinsuke [Graduate School of Engineering, Osaka University (Japan); Research Institute of Nuclear Engineering, University of Fukui (Japan)

2017-03-15

It is important to understand the behaviors of molten core materials to investigate the progression of a core meltdown accident. In the early stages of bundle degradation, low-melting-temperature liquid phases are expected to form via the eutectic reaction between Zircaloy and stainless steel. The main component of Zircaloy is Zr and those of stainless steel are Fe, Ni, and Cr. Our group has previously reported physical property data such as viscosity, density, and surface tension for Zr-Fe liquid alloys using an electrostatic levitation technique. In this study, we report the viscosity, density, and surface tension of Zr-Ni and Zr-Cr liquid alloys (Zr{sub 1-x}Ni{sub x} (x = 0.12 and 0.24) and Zr{sub 0.77}Cr{sub 0.23}) using the electrostatic levitation technique. - Highlights: • The physical properties of Zr-Ni and Zr-Cr liquid alloys have been measured Zr{sub 1-x}Ni{sub x} (x = 0.12 and 0.24) and Zr{sub 77}Cr{sub 23}. • The measurement was conducted using the electrostatic levitation technique. • The density, viscosity, and surface tension of each liquid alloy were measured.

Physical properties of molten core materials: Zr-Ni and Zr-Cr alloys measured by electrostatic levitation

International Nuclear Information System (INIS)

Ohishi, Yuji; Kondo, Toshiki; Ishikawa, Takehiko; Okada, Junpei T.; Watanabe, Yuki; Muta, Hiroaki; Kurosaki, Ken; Yamanaka, Shinsuke

2017-01-01

It is important to understand the behaviors of molten core materials to investigate the progression of a core meltdown accident. In the early stages of bundle degradation, low-melting-temperature liquid phases are expected to form via the eutectic reaction between Zircaloy and stainless steel. The main component of Zircaloy is Zr and those of stainless steel are Fe, Ni, and Cr. Our group has previously reported physical property data such as viscosity, density, and surface tension for Zr-Fe liquid alloys using an electrostatic levitation technique. In this study, we report the viscosity, density, and surface tension of Zr-Ni and Zr-Cr liquid alloys (Zr 1-x Ni x (x = 0.12 and 0.24) and Zr 0.77 Cr 0.23 ) using the electrostatic levitation technique. - Highlights: • The physical properties of Zr-Ni and Zr-Cr liquid alloys have been measured Zr 1-x Ni x (x = 0.12 and 0.24) and Zr 77 Cr 23 . • The measurement was conducted using the electrostatic levitation technique. • The density, viscosity, and surface tension of each liquid alloy were measured.

Core degradation and fission product release

International Nuclear Information System (INIS)

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

1992-01-01

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

Prospects of subcritical molten salt reactor for minor actinides incineration in closed fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Alekseev, Pavel N.; Balanin, Andrey L.; Dudnikov, Anatoly A.; Fomichenko, Petr A.; Nevinitsa, Vladimir A.; Frolov, Aleksey A.; Lubina, Anna S.; Sedov, Aleksey A.; Subbotin, Aleksey S.; Blandinsky, Viktor Yu. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

2015-09-15

A subcritical molten salt reactor is proposed for minor actinides (separated from spent fuel VVER-1000 light water reactor) incineration and for {sup 233}U conversion from {sup 232}Th. Here the subcritical molten salt reactor with fuel composition of heavy nuclide fluorides in molten LiF - NaF - KF salt and with external neutron source, based on 1 GeV proton accelerator and molten salt cooled tungsten target is considered. The paper presents the results of parametrical analysis of equilibrium nuclide composition of molten salt reactor with minor actinides feed in dependence of core dimensions, average neutron flux and external neutron source intensity. Reactor design is defined; requirements to external neutron source are posed; heavy nuclides equilibrium and fuel cycle main parameters are calculated.

Assessment of Mass Fraction and Melting Temperature for the Application of Limestone Concrete and Siliceous Concrete to Nuclear Reactor Basemat Considering Molten Core–Concrete Interaction

Directory of Open Access Journals (Sweden)

Hojae Lee

2016-04-01

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

Unaligned instruction relocation

Energy Technology Data Exchange (ETDEWEB)

Bertolli, Carlo; O' Brien, John K.; Sallenave, Olivier H.; Sura, Zehra N.

2017-10-17

In one embodiment, a computer-implemented method includes receiving source code to be compiled into an executable file for an unaligned instruction set architecture (ISA). Aligned assembled code is generated, by a computer processor. The aligned assembled code complies with an aligned ISA and includes aligned processor code for a processor and aligned accelerator code for an accelerator. A first linking pass is performed on the aligned assembled code, including relocating a first relocation target in the aligned accelerator code that refers to a first object outside the aligned accelerator code. Unaligned assembled code is generated in accordance with the unaligned ISA and includes unaligned accelerator code for the accelerator and unaligned processor code for the processor. A second linking pass is performed on the unaligned assembled code, including relocating a second relocation target outside the unaligned accelerator code that refers to an object in the unaligned accelerator code.

Unaligned instruction relocation

Science.gov (United States)

Bertolli, Carlo; O'Brien, John K.; Sallenave, Olivier H.; Sura, Zehra N.

2018-01-23

In one embodiment, a computer-implemented method includes receiving source code to be compiled into an executable file for an unaligned instruction set architecture (ISA). Aligned assembled code is generated, by a computer processor. The aligned assembled code complies with an aligned ISA and includes aligned processor code for a processor and aligned accelerator code for an accelerator. A first linking pass is performed on the aligned assembled code, including relocating a first relocation target in the aligned accelerator code that refers to a first object outside the aligned accelerator code. Unaligned assembled code is generated in accordance with the unaligned ISA and includes unaligned accelerator code for the accelerator and unaligned processor code for the processor. A second linking pass is performed on the unaligned assembled code, including relocating a second relocation target outside the unaligned accelerator code that refers to an object in the unaligned accelerator code.

Behaviour of molten reactor fuels under accident conditions

International Nuclear Information System (INIS)

Xavier Swamikannu, A.; Mathews, C.K.

1980-01-01

The behaviour of molten reactor fuels under accident conditions has received considerable importance in recent times. The chemical processes that occur in the molten state among the fuel, the clad components and the concrete of the containment building under the conditions of a core melt down accident in oxide fuelled reactors have been reviewed with the purpose of identifying areas of developmental work required to be performed to assess and minimize the consequences of such an accident. This includes the computation and estimation of vapour pressure of various gaseous species over the fuel, the clad and the coolant, providing of sacrificial materials in the concrete in order to protect the containment building in order to prevent release of radioactive gases into the atmosphere and understanding the distribution and chemical state of fission products in the molten fuel in order to provide for the effective removal of their decay heats. (auth.)

Molten salt reactors and possible scenarios for future nuclear power deployment

International Nuclear Information System (INIS)

Merle-Lucotte, E.; Mathieu, L.; Heuer, D.; Loiseaux, J. M.; Billebaud, A.; Brissot, R.; David, S.; Garzenne, C.; Laulan, O.; Le Brun, C.; Lecarpentier, D.; Liatard, E.; Meplan, O.; Michel-Sendis, F.; Nuttin, A.; Perdu, F.

2004-01-01

An important fraction of the nature energy demand may be satisfied by nuclear power. In this context, the possibilities of worldwide nuclear deployment are studied. We are convinced that the Molten Salt Reactors may play a central role in this deployment. The Molten Salt Reactor needs to be coupled to a reprocessing unit in order to extract the Fission Products which poison the core. The efficiency of this reprocessing has a crucial influence on reactor behavior especially for the breeding ratio. The Molten Salt Breeder Reactor project was based on an intensive reprocessing for high breeding purposes. A new concept of Thorium Molten Salt Reactor is presented here. Including this new concept in the worldwide nuclear deployment, to satisfy these power needs, we consider three typical scenarios, based on three reactor types: Pressurized Water Reactor, Fast Neutron Reactor and Thorium Molten Salt Reactor. The aim of this paper is to demonstrate, in a first hand that a Thorium Molten Salt Reactor can be realistic, with correct temperature coefficients and at least iso-breeder with slow reprocessing and new geometry; on the other hand that such Molten Salt Reactors enable a successful nuclear deployment, while minimizing fuel and waste management problems. (authors)

Research on the fundamental process of thermal-hydraulic behaviors in severe accident. Breakup and cooling of molten material jets. JAERI's nuclear research promotion program, H10-027-2. Contract research

Energy Technology Data Exchange (ETDEWEB)

Sugiyama, Ken-ichiro; Iguchi, Kentarou [Hokkaido Univ., Graduate School of Engineering, Sapporo, Hokkaido (Japan)

2002-03-01

Core melt accidents could lead to the pouring of molten core materials into a body of water accumulating in the reactor lower head in the form of jets with a few centimeters up to a few tens of centimeters. If molten core jets penetrate the body of water without breakup. A poor coolability of the molten core bed would occur, which means the difficulty of maintaining the molten core bed in the reactor vessel. Hence, the breakup mechanism of molten core jets had to be well understood for the evaluation of the coolability of molten core bed. The objective of the present experimental study is to confirm that, even in molten material jets, the breakup of jet originating in the coolant entrained within a molten material jet due to 'the organized motion' between the coolant and the jet, which has been recognized in the field of fluid mechanics, is caused. The first series of experiment was conducted to observe this type of breakup by using molten tin jets up to 25 mm in diameter. Molten tin jet was expected to easily cause this kind of breakup of jet because of a low kinematic viscosity, which means a easy transformation of jet due to the organized motion for the coolant to entrain. The second series of experiment was conducted by using molten copper jet of 25 mm in diameter, of which kinematic viscosity is about same as that of molten UO{sub 2}. The breakup of jet due to the entrainment of the coolant was observed up to high ambient Weber numbers, which cover the atomization regime. The mechanism of the breakup observed in the present study is able to reasonably explain the apparent difference between the breakup lengths of 150 kg-scale corium jets and the breakup lengths of about 8 kg-scale lead-bismuth alloy jets. The breakup by the mechanism reported here also assures a high coolability of molten jets because of an efficient entrainment of coolant within the jet. (author)

Experimental investigations of long-term interactions of molten UO2 with MgO and concrete at Argonne National Laboratory

International Nuclear Information System (INIS)

Stein, R.P.; Farhadieh, R.; Pedersen, D.R.; Gunther, W.H.; Purviance, R.T.

1982-01-01

Experimental work at Argonne is being performed to investigate the long-term molten-core-debris retention capability of the ex-vessel cavity following a postulated meltdown accident. The eventual objective of the work is to determine if normal structural material (concrete) or a specifically selected sacrificial material (MgO) located in the ex-vessel cavity region can effectively contain molten core debris. The materials under investigation at ANL are various types of concrete (limestone, basalt and magnetite) and commercially-available MgO brick. Results are presented of the status of real material experimental investigation at ANL into (1) molten UO 2 pool heat transfer, (2) long-term molten UO 2 penetration into concrete and (3) long-term molten UO 2 penetration into refractory substrates. The decay heating in the fuel has been simulated by direct electrical heating permitting the study of the long-term interaction

Experimental investigations of long-term interactions of molten UO2 with MgO and concrete at Argonne National Laboratory

International Nuclear Information System (INIS)

Stein, R.P.; Farhadieh, R.; Pedersen, D.R.; Gunther, W.H.; Purviance, R.T.

1982-01-01

Experimental work at Argonne is being performed to investigate the long-term molten core debris retention capability of the ex-vessel cavity following a postulated meltdown accident. The eventual objective of the work is to determine if normal structural material (concrete) or a specifically selected sacrificial material (MgO) located in the ex-vessel cavity region can effectively contain molten core debris. The materials under investigation at ANL are various types of concrete (limestone, basalt and magnetite) and commercially-available MgO brick. Results are presented of the status of real material experimental investigation at ANL into 1) molten UO 2 pool heat transfer, 2) long-term molten UO 2 penetration into concrete and 3) long-term molten UO 2 penetration into refractory substrates. The decay heating in the fuel has been simulated by direct electrical heating permitting the study of the long-term interaction

European Experiments on 2-D Molten Core Concrete Interaction: Hecla and Vulcano

International Nuclear Information System (INIS)

Journeau, Ch.; Bonnet, J. M.; Boccaccio, E.; Piluso, P.; Monerris, J.; Breton, M.; Fritz, G.; Sevon, Tuomo; Pankakoski Pekka, H.; Holmstrom, St.; Virta, Jouko

2010-01-01

This paper presents results from two ongoing European experimental programs on molten core concrete interactions: HECLA at VTT and VULCANO at the Commissariat a l'Energie Atomique. In the HECLA experiments, metallic melt is poured into a cylindrical concrete crucible. The focus is on the initial, pouring phase of the interaction. Therefore, decay heat simulation is not required. The HECLA-2 experiment involved 50 kg of stainless steel at 1700 C and siliceous concrete. The final ablation depths were 25-30 mm in the basemat and similar to 15 mm in the side wall. The VULCANO VB experiments have been devoted to the study of the interaction of 28 to 45 kg of oxidic corium with silica-rich or limestone-rich concretes. These tests focus on long-term ablation and require the use of induction heating to simulate the decay heat fluxes. Anisotropic ablation between the horizontal and vertical direction has been observed with silica-rich concrete, confirming the CCI tests. A new series of experiments VULCANO VBS has been launched in which there are both oxide and metallic phases in the melt. In these tests, magnetic screening is used so that the induction power is provided almost only to the upper oxidic layer after stratification. (authors)

European Experiments on 2-D Molten Core Concrete Interaction: Hecla and Vulcano

Energy Technology Data Exchange (ETDEWEB)

Journeau, Ch.; Bonnet, J. M.; Boccaccio, E.; Piluso, P.; Monerris, J.; Breton, M.; Fritz, G. [CEA Cadarache, Dept Technol Nucl, Serv Technol Reacteurs Ind, Lab Essais Maitrise Accid Graves, F-13108 St Paul Les Durance (France); Sevon, Tuomo; Pankakoski Pekka, H.; Holmstrom, St.; Virta, Jouko [VTT Tech Res Ctr Finland, FI-02044 Espoo (Finland)

2010-07-01

This paper presents results from two ongoing European experimental programs on molten core concrete interactions: HECLA at VTT and VULCANO at the Commissariat a l'Energie Atomique. In the HECLA experiments, metallic melt is poured into a cylindrical concrete crucible. The focus is on the initial, pouring phase of the interaction. Therefore, decay heat simulation is not required. The HECLA-2 experiment involved 50 kg of stainless steel at 1700 C and siliceous concrete. The final ablation depths were 25-30 mm in the basemat and similar to 15 mm in the side wall. The VULCANO VB experiments have been devoted to the study of the interaction of 28 to 45 kg of oxidic corium with silica-rich or limestone-rich concretes. These tests focus on long-term ablation and require the use of induction heating to simulate the decay heat fluxes. Anisotropic ablation between the horizontal and vertical direction has been observed with silica-rich concrete, confirming the CCI tests. A new series of experiments VULCANO VBS has been launched in which there are both oxide and metallic phases in the melt. In these tests, magnetic screening is used so that the induction power is provided almost only to the upper oxidic layer after stratification. (authors)

Analysis of molten fuel behavior in coolant channel during severe accidents in KALIMER

International Nuclear Information System (INIS)

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

2004-11-01

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

Parametric studies on the fuel salt composition in thermal molten salt breeder reactors

International Nuclear Information System (INIS)

Nagy, K.; Kloosterman, J.L.; Lathouwers, D.; Van der Hagen, T.H.J.J.

2008-01-01

In this paper the salt composition and the fuel cycle of a graphite moderated molten salt self-breeder reactor operating on the thorium cycle is investigated. A breeder molten salt reactor is always coupled to a fuel processing plant which removes the fission products and actinides from the core. The efficiency of the removal process(es) has a large influence on the breeding capacity of the reactor. The aim is to investigate the effect on the breeding ratio of several parameters such as the composition of the molten salt, moderation ratio, power density and chemical processing. Several fuel processing strategies are studied. (authors)

Experimental results of core-concrete interactions using molten steel with zirconium

International Nuclear Information System (INIS)

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

1990-07-01

Four inductively sustained experiments, QT-D, QT-E, SURC-3, and SURC-3A, were performed in order to investigate the additional effects of zirconium metal oxidation on core debris-concrete interactions using molten stainless steel as the core debris simulant. The QT-D experiment ablated 18 cm of concrete axially during 50 minutes of interaction on limestone-common sand concrete using a 10 kg charge of 304 stainless steel to which 2 kg of zirconium metal was added subsequent to the onset of erosion. The QT-E experiment ablated 10 cm of limestone-common sand concrete axially and 10 cm radially during 35 minutes of sustained interaction using 50 kg of stainless steel and 10 kg of zirconium. The SURC-3 experiment had a 45 kg charge of stainless steel to which 1.1 kg of zirconium was subsequently added. SURC-3 axially eroded 33 cm of limestone concrete during two hours of interaction. The fourth experiment, SURC-3A, eroded 25 cm of limestone concrete axially and 9 cm radially during 90 minutes of sustained interaction. It utilized 40 kg of stainless steel and 2.2 kg of added zirconium as the charge material. All four experiments showed in a large increase in erosion rate, gas production, and aerosol release following the addition of Zr metal to the melt. In the SURC-3 and SURC-3A tests the measured erosion rates increased from 14 cm/hr to 27 cm/hr, gas release increased from 50 slpm to 100 slpm, and aerosol release increased from .02 q/sec to .04 q/sec. The effluent gas was composed of 80% CO, 10% CO 2 , and 2% H 2 before Zr addition and 92% CO, 4% CO 2 , 4% H 2 during the Zr interactions which lasted 10--20 minutes. Addition measurements indicated that the melt pool temperature ranged from 1600 degree C--1800 degree and that the aerosols produced were comprised primarily of Te and Fe oxides. 21 refs., 120 figs., 51 tabs

Experimental results of core-concrete interactions using molten steel with zirconium

Energy Technology Data Exchange (ETDEWEB)

Copus, E.R.; Blose, R.E.; Brockmann, J.E.; Gomez, R.D.; Lucero, D.A. (Sandia National Labs., Albuquerque, NM (USA))

1990-07-01

Four inductively sustained experiments, QT-D, QT-E, SURC-3, and SURC-3A, were performed in order to investigate the additional effects of zirconium metal oxidation on core debris-concrete interactions using molten stainless steel as the core debris simulant. The QT-D experiment ablated 18 cm of concrete axially during 50 minutes of interaction on limestone-common sand concrete using a 10 kg charge of 304 stainless steel to which 2 kg of zirconium metal was added subsequent to the onset of erosion. The QT-E experiment ablated 10 cm of limestone-common sand concrete axially and 10 cm radially during 35 minutes of sustained interaction using 50 kg of stainless steel and 10 kg of zirconium. The SURC-3 experiment had a 45 kg charge of stainless steel to which 1.1 kg of zirconium was subsequently added. SURC-3 axially eroded 33 cm of limestone concrete during two hours of interaction. The fourth experiment, SURC-3A, eroded 25 cm of limestone concrete axially and 9 cm radially during 90 minutes of sustained interaction. It utilized 40 kg of stainless steel and 2.2 kg of added zirconium as the charge material. All four experiments showed in a large increase in erosion rate, gas production, and aerosol release following the addition of Zr metal to the melt. In the SURC-3 and SURC-3A tests the measured erosion rates increased from 14 cm/hr to 27 cm/hr, gas release increased from 50 slpm to 100 slpm, and aerosol release increased from .02 q/sec to .04 q/sec. The effluent gas was composed of 80% CO, 10% CO{sub 2}, and 2% H{sub 2} before Zr addition and 92% CO, 4% CO{sub 2}, 4% H{sub 2} during the Zr interactions which lasted 10--20 minutes. Addition measurements indicated that the melt pool temperature ranged from 1600{degree}C--1800{degree} and that the aerosols produced were comprised primarily of Te and Fe oxides. 21 refs., 120 figs., 51 tabs.

Assessment of Two-Phase Flow Heat Transfer Correlations for Molten Core-Concrete Interaction Study

International Nuclear Information System (INIS)

Tourniaire, B.; Varo, O.

2006-01-01

The prediction of heat transfer between corium pool and concrete basemat is of particular significance in the framework of the study of PWR's severe accident. Heat transfer directly governs the ablation velocity of concrete in case of molten core-concrete interaction (MCCI) and, consequently, the time delay when the reactor cavity may fail. From a restricted hydrodynamic point of view, this issue is related to heat transfer between a heated bubbling pool and a porous wall with gas injection. Several experimental studies have been performed with simulant materials and many correlations have been provided to address this issue. The main purpose of this paper is to assess these correlations from comparisons against the available experimental data. After a review of these data, the different correlations are presented. Attention focuses here on the correlations generally used in MCCI study: Kutateladze-Malenkov, Konsetov and BALI correlations. The Deckwer's correlation is also included in this review. The comparisons between the results of these correlations and the experimental data are then discussed. (authors)

EPRI [Electric Power Research Institute]/ANL investigations of MCCI [molten core-concrete interactions] phenomena and aerosol release

International Nuclear Information System (INIS)

Spencer, B.W.; Gunther, W.H.; Armstrong, D.R.; Thompson, D.H.; Chasanov, M.G.; Sehgal, B.R.

1986-01-01

A program of laboratory investigations has been undertaken at Argonne National Laboratory, under sponsorship of the Electric Power Research Institute, in which the interaction between molten core materials and concrete is studied, with particular emphasis on measurements of the magnitude and chemical species present in the aerosol releases. The experiment technique used in these investigations is direct electrical heating in which a high electric current is passed through the core debris to sustain the high-temperature melt condition for potentially long periods of time. In the scoping experiments completed to date, this technique has been successfully used for corium masses of 5 and 20 kg, generating an internal heating rate of 1 kw/kg and achieving melt temperatures of 2000C. Experiments have been performed both with a concrete base and also with a cooled base with the addition of H 2 /CO sparging gas to represent chemical processes in a stratified layer. An aerosol and gas sampling system is being used to collect aerosol samples. Test results are now becoming available including masses of aerosols, x-ray diffraction, and scanning electron microscope analyses

Protection of nuclear graphite toward fluoride molten salt by glassy carbon deposit

International Nuclear Information System (INIS)

Bernardet, V.; Gomes, S.; Delpeux, S.; Dubois, M.; Guerin, K.; Avignant, D.; Renaudin, G.; Duclaux, L.

2009-01-01

Molten salt reactor represents one of the promising future Generation IV nuclear reactors families where the fuel, a liquid molten fluoride salt, is circulating through the graphite reactor core. The interactions between nuclear graphite and fluoride molten salt and also the graphite surface protection were investigated in this paper by powder X-ray diffraction, micro-Raman spectroscopy and scanning electron microscopy coupled with X-ray microanalysis. Nuclear graphite discs were covered by two kinds of protection deposit: a glassy carbon coating and a double coating of pyrolitic carbon/glassy carbon. Different behaviours have been highlighted according to the presence and the nature of the coated protection film. Intercalation of molten salt between the graphite layers did not occur. Nevertheless the molten salt adhered more or less to the surface of the graphite disc, filled more or less the graphite surface porosity and perturbed more or less the graphite stacking order at the disc surface. The behaviour of unprotected graphite was far to be satisfactory after two days of immersion of graphite in molten salt at 500 deg. C. The best protection of the graphite disc surface, with the maximum of inertness towards molten salt, has been obtained with the double coating of pyrolitic carbon/glassy carbon

Status of the French research in the field of molten salt nuclear reactors

International Nuclear Information System (INIS)

Hery, M.; Israel, M.; Fauger, P.; Lecocq, A.

1977-01-01

The research program of the CEA in the field of molten salt nuclear reactors has been concerned with MSBR type reactors (Molten Salt Breeder Reactor). The papers written after having performed the theoretical analysis are entitled: core, circuits, chemistry and economy; they include some criticisms and suggestions. The experimental studies consisted in: graphite studies, chemical studies of the salt, metallic materials, the salt loop and the lead loop [fr

Calculations of the Possible Consequences of Molten Fuel Sodium Interactions in Subassembly and Whole Core Geometries

International Nuclear Information System (INIS)

Coddington, P.; Fishlock, T.P.; Jakeman, D.

1976-01-01

In making assessments of fast reactor safety a number of accident sequences can be postulated in which molten fuel contacts sodium in a number of possible modes. In the absence of an understanding of the way in which reactor materials interact for these contact modes it is necessary to make assessments over a range of plausible conditions and assumptions. This enables those areas where an interaction might cause a new stage in the escalation of the accident to be identified and at the same time to establish what characteristics of the interaction may be important. Whether in real situations interaction of molten reactor materials can have such characteristics can then be considered from both a theoretical and experimental viewpoint. It is suggested that although high efficiency vapour explosions involving large amounts of fuel in which there is rapid and coherent fragmentation are a main source of concern in many accident sequences, interactions with other characteristics may also be important. Two areas which have been identified are: (i) the interactions of low efficiency which need only involve small fractions of the fuel or possibly could include molten clad but which can accelerate sodium and fuel sufficiently to give rise to large reactivity changes. The recent incident at a steel plant in the U.K. in which 100 tons of molten steel was ejected to a height of 10 m from a torpedo ladle when water accidentally poured into it is a particularly striking illustration of such movement; and (ii) interactions giving rise to a much slower and less coherent heat transfer which may require some degree of fragmentation but not the extensive fragmentation by the specific mechanisms associated with vapour explosions but which nevertheless on the reactor scale could lead to high slug impacts on the containment. Accident codes are being constructed in the U.K. to investigate a series of hypothetical incidents. Modules are required for these codes which enable the consequences

Fluid-mechanic/thermal interaction of a molten material and a decomposing solid

International Nuclear Information System (INIS)

Larson, D.W.; Lee, D.O.

1976-12-01

Bench-scale experiments of a molten material in contact with a decomposing solid were conducted to gain insight into the expected interaction of a hot, molten reactor core with a concrete base. The results indicate that either of two regimes can occur: violent agitation and splattering of the melt or a very quiescent settling of the melt when placed in contact with the solid. The two regimes appear to be governed by the interface temperature condition. A conduction heat transfer model predicts the critical interface temperature with reasonable accuracy. In addition, a film thermal resistance model correlates well with the data in predicting the time for a solid skin to form on the molten material

Calculation of β-effective of a molten salt reactor

International Nuclear Information System (INIS)

Hirakawa, N.; Sakaba, H.

1987-01-01

A method to calculate the β eff of a molten salt reactor was developed taking the effect of the flow of the molten salt into account. The method was applied to the 1000MW MSR design made by ORNL. The change in β eff due to the change in the residence time outside of the core of the fuel salt and to the change in the flow velocity when the total amount of the fuel salt is kept constant were investigated. It was found that β eff was reduced to 47.9% of the value when the fuel salt is at rest for the present design. (author)

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Heat removal capability of core-catcher with inclined cooling channels

International Nuclear Information System (INIS)

Suzuki, Y.; Tahara, M.; Kurita, T.; Hamazaki, R.; Morooka, S.

2009-01-01

A core-catcher is one of the mitigation systems that provide functions of molten corium cooling and stabilization during a severe accident. Toshiba has been developing a compact core-catcher to be placed at the lower drywell floor in the containment vessel for the next generation BWR as well as near term ABWR. This paper presents the evaluation of heat removal capability of the core-catcher with inclined cooling channels, our verification status and plan. The heat removal capability of the core-catcher is analyzed by using the newly developed two-phase flow analysis code which incorporates drift flux parameters for inclined channels and the CHF correlation obtained from SULTAN tests. Effects of geometrical parameters such as the inclination and the gap size of the cooling channel on the heat removal capability are also evaluated. These results show that the core-catcher has sufficient capability to cool the molten corium during a severe accident. Based on the analysis, it has been shown that the core-catcher has an efficient capability of heat removal to cool the molten corium. (author)

Current european experiments on 2d molten core concrete interaction: HECLA and VULCANO

International Nuclear Information System (INIS)

Journeau, C.; Bonnet, J. M.; Boccaccio, E.; Piluso, P.; Sevon, T.; Pankakoski, P. H.; Holmstroem, S.; Virta, J.

2008-01-01

This paper presents results from two ongoing European experimental programs on molten core concrete interactions: HECLA at VTT and VULCANO at CEA. In the HECLA experiments, metallic melt is poured into a cylindrical concrete crucible. The focus is on the initial, pouring phase of the interaction. Therefore, decay heat simulation is not required. The HECLA-2 experiment involved 50 kg of stainless steel at 1700 deg. C and siliceous concrete. The final ablation depths were 25-30 mm in the basemat and about 15 mm in the sidewall. The VULCANO VB experiments have been devoted to the study of the interaction of 28 to 45 kg of oxidic corium with silica-rich or limestone-rich concretes. These tests are focusing on long-term ablation and require the use of induction heating to simulate the decay heat fluxes. Anisotropic ablation between the horizontal and vertical direction has been observed with silica-rich concrete, confirming the CCI tests. A new series of experiments VULCANO VBS has been launched in which there are both oxide and metallic phases in the melt. In these tests, magnetic screening is used so that the induction power is provided almost only to the upper oxidic layer after stratification. (authors)

Event sequence quantification for a loss of shutdown cooling accident in the GCFR

International Nuclear Information System (INIS)

Frank, M.; Reilly, J.

1979-10-01

A summary is presented of the core-wide sequence of events of a postulated total loss of forced and natural convection decay heat removal in a shutdown Gas-Cooled Fast Reactor (GCFR). It outlines the analytical methods and results for the progression of the accident sequence. This hypothetical accident proceeds in the distinct phases of cladding melting, assembly wall melting and molten steel relocation into the interassembly spacing, and fuel relocation. It identifies the key phenomena of the event sequence and the concerns and mechanisms of both recriticality and recriticality prevention

Improvement of core degradation model in ISAAC

International Nuclear Information System (INIS)

Kim, Dong Ha; Kim, See Darl; Park, Soo Yong

2004-02-01

If water inventory in the fuel channels depletes and fuel rods are exposed to steam after uncover in the pressure tube, the decay heat generated from fuel rods is transferred to the pressure tube and to the calandria tube by radiation, and finally to the moderator in the calandria tank by conduction. During this process, the cladding will be heated first and ballooned when the fuel gap internal pressure exceeds the primary system pressure. The pressure tube will be also ballooned and will touch the calandria tube, increasing heat transfer rate to the moderator. Although these situation is not desirable, the fuel channel is expected to maintain its integrity as long as the calandria tube is submerged in the moderator, because the decay heat could be removed to the moderator through radiation and conduction. Therefore, loss of coolant and moderator inside and outside the channel may cause severe core damage including horizontal fuel channel sagging and finally loss of channel integrity. The sagged channels contact with the channels located below and lose their heat transfer area to the moderator. As the accident goes further, the disintegrated fuel channels will be heated up and relocated onto the bottom of the calandria tank. If the temperature of these relocated materials is high enough to attack the calandria tank, the calandria tank would fail and molten material would contact with the calandria vault water. Steam explosion and/or rapid steam generation from this interaction may threaten containment integrity. Though a detailed model is required to simulate the severe accident at CANDU plants, complexity of phenomena itself and inner structures as well as lack of experimental data forces to choose a simple but reasonable model as the first step. ISAAC 1.0 was developed to model the basic physicochemical phenomena during the severe accident progression. At present, ISAAC 2.0 is being developed for accident management guide development and strategy evaluation. In

Steam explosions of molten iron oxide drops: easier initiation at small pressurizations

International Nuclear Information System (INIS)

Nelson, L.S.; Duda, P.M.

1982-01-01

Steam explosions caused by hot molten materials contacting liquid water following a possible light water nuclear reactor core overheat have been investigated by releasing single drops of a core melt simulant, molten iron oxide, into liquid water. Small steam explosions were triggered shortly afterwards by applying a pressure pulse to the water. The threshold peak pulse level above which an explosion always occurs was studied at ambient pressures between 0.083 and 1.12 MPa. It was found that the threshold decreased to a minimum in the range 0.2 - 0.8 MPa and then increased again. The effect of easier initiation as ambient pressure increases may have an important role in the triggering and propagation of a large scale steam explosion through a coarsely premixed dispersion of melt in water. (U.K.)

Laboratory studies of the meltfront propagation in a borax core-catcher

International Nuclear Information System (INIS)

Dalle Donne, M.; Werle, H.

1980-08-01

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

Molten salt electrorefining method

International Nuclear Information System (INIS)

Tanaka, Hiroshi; Nakamura, Hitoshi; Shoji, Yuichi; Matsumaru, Ken-ichi.

1994-01-01

A molten cadmium phase (lower side) and a molten salt phase (upper side) are filled in an electrolytic bath. A basket incorporating spent nuclear fuels is inserted/disposed in the molten cadmium phase. A rotatable solid cathode is inserted/disposed in the molten salt phase. The spent fuels, for example, natural uranium, incorporated in the basket is dissolved in the molten cadmium phase. In this case, the uranium concentration in the molten salt phase is determined as from 0.5 to 20wt%. Then, electrolysis is conducted while setting a stirring power for stirring at least the molten salt phase of from 2.5 x 10 2 to 1 x 10 4 based on a reynolds number. Crystalline nuclei of uranium are precipitated uniformly on the surface of the solid cathode, and they grow into fine dendrites. With such procedures, since short-circuit between the cathode precipitates and the molten cadmium phase (anode) is scarcely caused, to improve the recovering rate of uranium. (I.N.)

The Analysis of Surrounding Structure Effect on the Core Degradation Progress with COMPASS Code

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

Simulation tool of the on-line reprocessing unit of a molten salt reactor

International Nuclear Information System (INIS)

Simon, Nicole; Conocar, Olivier; Boussier, Hubert; Gastaldi, Olivier; Penit, Thomas; Walle, Eric; Huguet, Anne

2006-01-01

Molten salt reactor (MSR) is an interesting technology selected in the frame of the Generation IV forum. In the MSR, actinides are diluted in a molten salt which is both the fuel and the coolant. The ability of such a reactor is the reducing of the long-lived wastes due to high burn-up and the on-site simplified reprocessing directly connected with the core which preserve the salt properties necessary for its correct operation. Here is defined a flexible computer reprocessing code which can use data from neutronic calculations and can be coupled to a neutronic code. The code allow the description the whole behaviour of MSR, including, a coupled manner, both the design of the core and the optimised reprocessing scheme effects. (authors)

Corporate relocation in venture capital finance

NARCIS (Netherlands)

Cumming, D.; Flemming, G.; Schwienbacher, A.

2009-01-01

This paper introduces an analysis of international relocation decisions of venture capital (VC)-backed companies. Relocations to the United States are motivated by economic conditions as well as an improvement in the laws of the country in which the entrepreneurial firm is based. Relocations to the

Molten salt reactor concept

International Nuclear Information System (INIS)

Sood, D.D.

1980-01-01

Molten salt reactor is an advanced breeder concept which is suited for the utilization of thorium for nuclear power production. This reactor is based on the use of solutions of uranium or plutonium fluorides in LiF-BeF 2 -ThF 4 as fuel. Unlike the conventional reactors, no external coolant is used in the reactor core and the fuel salt itself is circulated through heat exchangers to transfer the fission produced heat to a secondary salt (NaF-NaBF 4 ) for steam generation. A part of the fuel stream is continuously processed to isolate 233 Pa, so that it can decay to fissile 233 U without getting converted to 234 Pa, and for the removal of neutron absorbing fission products. This on-line processing scheme makes this reactor concept to achieve a breeding ratio of 1.07 which is the highest for any thermal breeder reactor. Experimental studies at the Bhabha Atomic Research Centre, Bombay, have established the use of plutonium as fuel for this reactor. This molten salt reactor concept is described and the work conducted at the Bhabha Atomic Research Centre is summarised. (auth.)

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

International Nuclear Information System (INIS)

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

1994-11-01

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

Resourcefulness, positive cognitions, relocation controllability and relocation adjustment among older people: a cross-sectional study of cultural differences.

Science.gov (United States)

Bekhet, Abir K; Zauszniewski, Jaclene A

2013-09-01

The population of older people in both the United States and Egypt is expected to double by the year 2030. With ageing, chronic illnesses increase and many older people need to relocate to retirement communities. Research has shown that positive cognitions and resourcefulness are positively correlated with adaptive functioning and better adjustment. The purpose of this study was to examine and compare relocation controllability, positive cognitions, resourcefulness and relocation adjustment between American and Egyptian older people living in retirement communities. The purpose of this cultural comparison is to gain insight into influencing factors in each culture that might lead to interventions to help relocated older adults in both cultures adjust to their new surroundings. A cross-sectional, descriptive design was used to compare relocation controllability, positive cognitions, resourcefulness and relocation adjustment of a convenience sample of American older people (n = 104) and a convenience sample of Egyptian older people (n = 94). The study was a secondary analysis of two studies of older people residing in six retirement communities in Northeast Ohio and in five retirement communities in Alexandria, Egypt. Examination of mean scores and standard deviations on the measure of positive cognitions using independent sample t-tests indicated that on average, the American older people reported more positive cognitions (t (131.16) = 11.29, P difference between Egyptians and Americans in resourcefulness (t (174.16) = -0.97, P > 0.05). The results provide direction for the development of positive cognition interventions and engaging older people in the decision-making process to help them to adjust to relocation. Implications for practice.  Positive thinking and resourcefulness training interventions can be used by nurses to help relocated older people to adjust to the stress of relocation to retirement communities. These interventions can be used on primary

A Scoping Analysis Of The Impact Of SiC Cladding On Late-Phase Accident Progression Involving Core–Concrete Interaction

Energy Technology Data Exchange (ETDEWEB)

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

2015-11-01

The overall objective of the current work is to carry out a scoping analysis to determine the impact of ATF on late phase accident progression; in particular, the molten core-concrete interaction portion of the sequence that occurs after the core debris fails the reactor vessel and relocates into containment. This additional study augments previous work by including kinetic effects that govern chemical reaction rates during core-concrete interaction. The specific ATF considered as part of this study is SiC-clad UO₂.

Accelerator molten-salt breeder reactor

International Nuclear Information System (INIS)

Furukawa, Kazuo; Kuroi, Hideo; Kato, Yoshio; Oomichi, Toshihiko.

1979-01-01

Purpose: To obtain fission products and to transmute transuranium elements and other radioactive wastes by the use of Accelerator Molten-Salt Breeder Reactor. Constitution: Beams from an accelerator pipe at one end of a target vessel is injected through a window into target molten salts filled inside of the target vessel. The target molten salts are subjected to pump recycling or spontaneous convection while forcively cooled by blanket molten salts in an outer vessel. Then, energy is recovered from the blanket molten salts or the target molten salts at high temperatures through electric power generation or the like. Those salts containing such as thorium 232 and uranium 238 are used as the blanket molten salts so that fission products may be produced by neutrons generated in the target molten salts. PbCl 2 -PbF 2 and LiF-BeF 2 -ThF 4 can be used as the target molten salts and as the blanket molten salts respectively. (Seki, T.)

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

Proceedings of the Workshop on in-vessel core debris retention and coolability

International Nuclear Information System (INIS)

1999-01-01

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

Fragmentation of molten core material by sodium

International Nuclear Information System (INIS)

Chu, T.Y.

1982-01-01

A series of scoping experiments was performed to study the fragmentation of prototypic high temperature melts in sodium. The quantity of melt involved was at least one order of magnitude larger than previous experiments. Two modes of contact were used: melt streaming into sodium and sodium into melt. The average bulk fragment size distribution was found to be in the range of previous data and the average size distribution was found to be insensitive to mode of contact. SEM studies showed that the metal component typically fragmented in the molten phase while the oxide component fragmented in the solid phase. For UO 2 -ZrO 2 /stainless steel melts no sigificant spatial separation of the metal and oxide was observed. The fragment size distribution was stratified vertically in the debris bed in all cases. While the bulk fragment size showed generally consistent trends, the individual experiments were sufficiently different to cause different degrees of stratification in the debris bed. For the highly stratified beds the permeability can decrease by as much as a factor of 20 from the bottom to the top of the bed

CORCON: a computer program for modelling molten fuel/concrete interactions

International Nuclear Information System (INIS)

Muir, J.F.

1980-01-01

A computer program modelling the interaction between molten core materials and structural concrete is being developed to provide a capability for making quantitative estimates of reactor fuel-melt accidents. The principal phenomenological models, inter-component heat transfer, concrete erosion, and melt/gas chemical reactions, are described. A code test comparison calculation is discussed

Improvement of molten core-concrete interaction model of the debris spreading analysis model in the SAMPSON code - 15193

International Nuclear Information System (INIS)

Hidaka, M.; Fujii, T.; Sakai, T.

2015-01-01

A debris spreading analysis (DSA) module has been developed and improved. The module is used in the severe accident analysis code SAMPSON and it has models for 3-dimensional natural convection with simultaneous spreading, melting and solidification. The existing analysis method of the quasi-3D boundary transportation to simulate downward concrete erosion for evaluation of molten-core concrete interaction (MCCI) was improved to full-3D to solve, for instance, debris lateral erosion under concrete floors at the bottom of the sump pit. In the advanced MCCI model, buffer cells were defined in order to solve numerical problems in case of trammel formation. Mass, momentum, and the advection term of energy between the debris melt cells and the buffer cells are solved. On the other hand, only the heat transfer and thermal conduction are solved between the debris melt cells and the structure cells, and the crust cells and the structure cells. As a preliminary analysis, a validation calculation was performed for erosion that occurred in the core-concrete interaction (CCI-2) test in the OECD/MCCI program. Comparison between the calculation and the CCI-2 test results showed the analysis has the ability to simulate debris lateral erosion under concrete floors. (authors)

Results of fission product release from intermediate-scale MCCI [molten core-concrete interaction] tests

International Nuclear Information System (INIS)

Spencer, B.W.; Thompson, D.H.; Fink, J.K.; Gunther, W.H.; Sehgal, B.R.

1988-01-01

A program of reactor-material molten core-concrete interaction (MCCI) tests and related analyses are under way at Argonne National Laboratory under sponsorship of the Electric Power Research Institute (EPRI). The particular objective of these tests is to provide data pertaining to the release of nonvolatile fission products such as La, Ba, and Sr, plus other aerosol materials, from the coupled thermal-hydraulic and chemical processes of the MCCI. The first stages of the program involving small and intermediate-scale tests have been completed. Three small-scale tests (/approximately/5 kg corium) and nine intermediate-scale tests (/approximately/30 kg corium) were performed between September 1985 and September 1987. Real reactor materials were used in these tests. Sustained internal heat generation at nominally 1 kW per kg of melt was provided by direct electrical heating of the corium mixture. MCCI tests were performed with both fully and partially oxidized corium mixtures that contained a variety of nonradioactive materials such as La 2 O 3 , BaO, and SrO to represent fission products. Both limestone/common sand and basaltic concrete basemats were used. The system was instrumented for characterization of the thermal hydraulic, chemical, gas release, and aerosol release processes

Molten salt breeder reactor

International Nuclear Information System (INIS)

1977-01-01

MSBR Study Group formed in October 1974 has studied molten salt breeder reactor and its various aspects. Usage of a molten salt fuel, extremely interesting as reactor chemistry, is a great feature to MSBR; there is no need for separate fuel making, reprocessing, waste storage facilities. The group studied the following, and these results are presented: molten salt technology, molten salt fuel chemistry and reprocessing, reactor characteristics, economy, reactor structural materials, etc. (Mori, K.)

Stadium Relocation in Professional Football

DEFF Research Database (Denmark)

Junghagen, Sven; Lillo, Gonzalo Leal

The aim of this paper is to conceptually illustrate the tension between commercialisation and club authenticity, which is a potential consequence of stadium relocation. There is a commercial pressure for relocating a football club to new and more modern facilities, but also a pressure from...

BWR core melt progression phenomena: Experimental analyses

International Nuclear Information System (INIS)

Ott, L.J.

1992-01-01

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

Analysis of the thermal response of a BWR Mark-I containment shell to direct contact by molten core materials

International Nuclear Information System (INIS)

Kress, T.S.; Cleveland, J.C.

1988-01-01

This study was undertaken to evaluate the thermal response of a BWR Mark-I containment shell in the event of an accident severe enough for molten core materials to fall into the cavity beneath the rector vessel and eventually come into direct contact with the shell. An existing ORNL three-dimensional transient heat transport computer code, HEATING-6, was used for a specific 2-D case (and variations) for which representative melt/shell boundary conditions required as input were available from other studies. In addition to the use of HEATING-6, a simplified analytical steady-state correlation was developed and given the name BWR Liner Analysis Program (BWRLAP). BWRLAP was ''benchmarked'' by comparison with HEATING-6 and was then used to make a number of parametric calculations to investigate the sensitivities of the results to the inputs. 5 refs., 11 figs., 2 tabs

75 FR 14442 - Federal Travel Regulation (FTR); Relocation Allowances-Relocation Income Tax Allowance (RITA) Tables

Science.gov (United States)

2010-03-25

... GENERAL SERVICES ADMINISTRATION [GSA Bulletin FTR 10-04] Federal Travel Regulation (FTR); Relocation Allowances-- Relocation Income Tax Allowance (RITA) Tables AGENCY: Office of Governmentwide Policy... (73 FR 35952) specifying that GSA would no longer publish the RITA tables found in 41 CFR Part 301-17...

The jet impingement phase of molten core-concrete interactions

International Nuclear Information System (INIS)

Sienicki, J.J.; Spencer, B.W.

1986-01-01

Scoping calculations have been carried out demonstrating that a significant and abrupt reduction in the corium temperature may be realized when molten corium drains as a jet from a localized breach in the RPV lower head to impinge upon the concrete basemat. The temperature decrease may range from a value of ∼170 K (∼140 K) for limestone (basaltic) aggregate concrete to a value approaching the initial corium superheat depending upon whether the forced convection impingement heat flux is assumed to be controlled by either thermal conduction across a slag film layer or the temperature boundary condition represented by a corium crust. The magnitude of the temperature reduction remains significant as the initial corium temperature, impinging corium mass, and initial localized breach size are varied over their range of potential values

TMI-2 core debris analysis

International Nuclear Information System (INIS)

Cook, B.A.; Carlson, E.R.

1985-01-01

One of the ongoing examination tasks for the damaged TMI-2 reactor is analysis of samples of debris obtained from the debris bed presently at the top of the core. This paper summarizes the results reported in the TMI-2 Core Debris Grab Sample Examination and Analysis Report, which will be available early in 1986. The sampling and analysis procedures are presented, and information is provided on the key results as they relate to the present core condition, peak temperatures during the transient, temperature history, chemical interactions, and core relocation. The results are then summarized

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

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

Multistage unfolding of an SH3 domain: an initial urea-filled dry molten globule precedes a wet molten globule with non-native structure.

Science.gov (United States)

Dasgupta, Amrita; Udgaonkar, Jayant B; Das, Payel

2014-06-19

The unfolding of the SH3 domain of the PI3 kinase in aqueous urea has been studied using a synergistic experiment-simulation approach. The experimental observation of a transient wet molten globule intermediate, IU, with an unusual non-native burial of the sole Trp residue, W53, provides the benchmark for the unfolding simulations performed (eight in total, each at least 0.5 μs long). The simulations reveal that the partially unfolded IU ensemble is preceded by an early native-like molten globule intermediate ensemble I*. In the very initial stage of unfolding, dry globule conformations with the protein core filled with urea instead of water are transiently observed within the I* ensemble. Water penetration into the urea-filled core of dry globule conformations is frequently accompanied by very transient burial of W53. Later during gradual unfolding, W53 is seen to again become transiently buried in the IU ensemble for a much longer time. In the structurally heterogeneous IU ensemble, conformational flexibility of the C-terminal β-strands enables W53 burial by the formation of non-native, tertiary contacts with hydrophobic residues, which could serve to protect the protein from aggregation during unfolding.

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

Materials problems related to the core catcher of sodium cooled reactors

International Nuclear Information System (INIS)

Goetzmann, O.

1975-05-01

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

Preliminary safety analysis of molten salt breeder reactor

International Nuclear Information System (INIS)

Cheng Maosong; Dai Zhimin

2013-01-01

Background: The molten salt reactor is one of the six advanced reactor concepts identified by the Generation IV International Forum as a candidate for cooperative development, which is characterized by remarkable advantages in inherent safety, fuel cycle, miniaturization, effective utilization of nuclear resources and proliferation resistance. ORNL finished the conceptual design of Molten Salt Breeder Reactor (MSBR) based on the design, building and operation of Molten Salt Reactor Experiment (MSRE). Purpose: We attempt to implement the preliminary safety analysis of MSBR in order to provide a reference for the design and optimization of MSBR in the future. Methods: According to the conceptual design of MSBR, a model of safety analysis using point kinetics coupled with the simplified heat transfer mechanism is presented. The model is applied to simulate the transient phenomena of MSBR initiated by an abnormal step reactivity addition and an abnormal ramp reactivity addition at full-power equilibrium condition. Results: The thermal power in the core increases rapidly at the beginning and is accompanied by a rise of the fuel and graphite temperatures after 100, 300, 500 and 600 pcm reactivity addition. The maximum outlet temperature of the fuel in the core is at 1250℃ in 500 pcm reactivity addition, but up to 1350℃ in 600 pcm reactivity addition. The maximum of the power and the temperature are delayed and lower in the ramp reactivity addition rather than in the step reactivity addition. Conclusions: Based on the results, when the reactivity inserted is less than 500 pcm in maximum at full power equilibrium condition, the structural material in Hastelloy-N is not melted and can keep integrity without external control action. And it is necessary to try to avoid inserting a reactivity at short time. (authors)

Transient core characteristics of small molten salt reactor coupling problem between heat transfer/flow and nuclear fission reaction

International Nuclear Information System (INIS)

Yamamoto, Takahisa; Mitachi, Koshi

2004-01-01

This paper performed the transient core analysis of a small Molten Salt Reactor (MSR). The emphasis is that the numerical model employed in this paper takes into account the interaction among fuel salt flow, nuclear reaction and heat transfer. The model consists of two group diffusion equations for fast and thermal neutron fluexs, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The results of transient analysis are that (1) fission reaction (heat generation) rate significantly increases soon after step reactivity insertion, e.g., the peak of fission reaction rate achieves about 2.7 times larger than the rated power 350 MW when the reactivity of 0.15% Δk/k 0 is inserted to the rated state, and (2) the self-control performance of the small MSR effectively works under the step reactivity insertion of 0.56% Δk/k 0 , putting the fission reaction rate back on the rated state. (author)

76 FR 41555 - Tupelo, Mississippi Railroad Relocation Project

Science.gov (United States)

2011-07-14

... Relocation Project AGENCY: Federal Railroad Administration (FRA), U.S. Department of Transportation (DOT... Tupelo, Mississippi Railroad Relocation Project (Project). SUMMARY: The Federal Railroad Administration announces the availability of the Tupelo, Mississippi Railroad Relocation Project Draft Environmental Impact...

Studies on the molten salt reactor. Code development and neutronics analysis of MSRE-type design

International Nuclear Information System (INIS)

Zhuang Kun; Cao Liangzhi; Zheng Youqi; Wu Hongchun

2015-01-01

The molten salt reactor is characterized by its use of the fluid-fuel, which serves both as a fuel and as a coolant simultaneously. The position of delayed neutron precursors continuously changes both in the core and in the external loop due to the fuel circulation, and the fission products are extracted by an online fuel reprocessing unit, which all lead to the modeling methods for the conventional reactors using solid fuel not applicable. This study establishes suitable calculation models for the neutronics analysis of the molten salt reactor and develops a new code named MOREL based on the three-dimensional diffusion steady and transient calculations. Some numerical tests are chosen to verify the code and the numerical results indicate that MOREL can be used for the analysis of the molten salt reactor. After verification, it is applied to analyze the characteristics of a typical molten salt reactor, including the steady characteristics, the influence of fuel circulation on the kinetic behaviors. Besides, the influence of online fuel reprocessing simulation is also examined. The results show that inherent safety is the character of the molten salt reactor from the aspect of reactivity feedback and the fuel circulation has great influence on the kinetic characteristics of molten salt reactor. (author)

41 CFR 302-2.5 - May I relocate from a location other than the location specified in my relocation travel...

Science.gov (United States)

2010-07-01

... 41 Public Contracts and Property Management 4 2010-07-01 2010-07-01 false May I relocate from a... Public Contracts and Property Management Federal Travel Regulation System RELOCATION ALLOWANCES INTRODUCTION 2-EMPLOYEES ELIGIBILITY REQUIREMENTS General Rules § 302-2.5 May I relocate from a location other...

Detection and removal of molten salts from molten aluminum alloys

Energy Technology Data Exchange (ETDEWEB)

K. Butcher; D. Smith; C. L. Lin; L. Aubrey

1999-08-02

Molten salts are one source of inclusions and defects in aluminum ingots and cast shapes. A selective adsorption media was used to remove these inclusions and a device for detection of molten salts was tested. This set of experiments is described and the results are presented and analyzed.

Studies of thermal hydraulics and heat transfer in cascade subcritical molten salt reactor

International Nuclear Information System (INIS)

Aysen, E.M.; Sedov, A.A.; Subbotin, A.S.

2005-01-01

Full text of publication follows: Cascade Subcritical Molten Salt Reactor (CSMSR) consists of three main parts: accelerator-driven proton-bombarded target, central and peripheral zones. External neutrons generated in the result of interaction of protons with the target nuclei are multiplied then in the central zone and leak farther into the peripheral reactor zone, where an efficient burning of Minor Actinides dissolved in a molten salt fluoride composition is produced. The bunch of target and two zones is designed so that preset subcriticality of reactor would not be less than 1% of k eff . A characteristic feature of the reactor is a high density of neutron flux (2.10 15 n/cm 2 s) in the central zone and target and very high volumetric power rate (2000 - 6000 W/cm 3 ) in all the parts of CSMSR. To provide a workability of the core structures under condition of so big level of power rate it is necessary to impose strict limitations on the temperatures and temperature gradients developed in the coolants and constructions. In this reason it has been arranged a calculational-designing study to reveal the problems of heat transfer in the coolant and core structures and to find more appropriate variant of the core and target design, which is a compromise of contradictory requirements: provision of high neutron flux and coolability of the core structures. In this paper the results of studies of thermal hydraulics and heat transfer in the core zones and proton-beam target are presented. Different variants of the target and central zone design as well as application of different kind of coolants in them are discussed and the main problems of heat removal in their structures are analyzed. Multidimensional fields of velocity and temperature got in thermal hydraulics calculations for free flow of fuelled molten salt in cylindrical-cave peripheral CSMSR zone without structures inside are demonstrated. The role of turbulent exchange of momentum and heat for free flow in the

Molten material-containing vessel

International Nuclear Information System (INIS)

Akagawa, Katsuhiko

1998-01-01

The molten material-containing vessel of the present invention comprises a vessel main body having an entrance opened at the upper end, a lid for closing the entrance, an outer tube having an upper end disposed at the lower surface of the lid, extended downwardly and having an closed lower end and an inner tube disposed coaxially with the outer tube. When a molten material is charged from the entrance to the inside of the vessel main body of the molten material-containing vessel and the entrance is closed by the lid, the outer tube and the inner tube are buried in the molten material in the vessel main body, accordingly, a fluid having its temperature elevated by absorption of the heat of the molten material rises along the inner circumferential surface of the outer tube, abuts against the lower surface of the lid and cooled by exchanging heat with the lid and forms a circulating flow. Since the heat in the molten material is continuously absorbed by the fluid, transferred to the lid and released from the lid to the atmospheric air, heat releasing efficiency can be improved compared with conventional cases. (N.H.)

Results of and prospects for studies on molten salt nuclear reactors

International Nuclear Information System (INIS)

Hery, M.; Lecocq, A.

1983-04-01

This paper reviews the various studies performed in France by the EDF and CEA teams in the field of molten salt nuclear reactors. These studies include graphite moderating systems, feasibility of a 625 MWth core, lead cooling, structural materials, salts tritium diffusion and corrosion. The experience gained allows eventual development prospects of this system to appraised [fr

Molten fluoride fuel salt chemistry

International Nuclear Information System (INIS)

Toth, L.M.; Del Cul, G.D.; Dai, S.; Metcalf, D.H.

1995-01-01

The chemistry of molten fluorides is traced from their development as fuels in the Molten Salt Reactor Experiment with important factors in their selection being discussed. Key chemical characteristics such as solubility, redox behavior, and chemical activity are explained as they relate to the behavior of molten fluoride fuel systems. Development requirements for fitting the current state of the chemistry to modern nuclear fuel system are described. It is concluded that while much is known about molten fluoride behavior which can be used effectively to reduce the amount of development required for future systems, some significant molten salt chemical questions must still be addressed. copyright American Institute of Physics 1995

Fuel relocation mechanism based on microstructures of debris

International Nuclear Information System (INIS)

Strain, R.V.; Neimark, L.A.; Sanecki, J.E.

1988-05-01

Argonne National Laboratory (ANL) has performed a number of examinations to determine the microstructure and micro-chemistry of samples of debris from the TMI-2 reactor. These examinations have been a small part of the overall effort to gain an understanding of the TMI-2 accident. As a result of these overall efforts, a general scenario of the response of the core components has been established. In this paper we will describe the microstructure and micro-chemistry of debris from the lower plenum of the reactor and relate these data to a segment of the general scenario dealing with the relocation of this material. The primary tools used at ANL for the examination of material from the TMI-2 core were optical microscopy, scanning electron microscopy and Energy Dispersive X-Ray Spectroscopy, and Scanning Auger Spectroscopy. In some cases these techniques were augmented by the use of gamma spectroscopy, autoradiography, and X-ray diffraction analysis

Symbiotic molten-salt systems coupled with accelerator molten-salt breeder (AMSB) or inertial-confined fusion hybrid molten-salt breeder (IHMSB) and their comparison

International Nuclear Information System (INIS)

Furukawa, K.

1984-01-01

Two types of breeder systems are proposed. One is the combined system of Accelerator Molten-Salt Breeder (AMSB) and Molten-Salt Converter Reactor (MSCR), and the other is the combined system of Inertial-confined Fusion Hybrid Molten-Salt Breeder (IHMSB) and modified MSCR. Both apply the molten-fluorides and have technically deep relations. AMSB would be much simpler and have already high technical feasibility. This will become economical the Th breeder system having a doubling time shorter than ten years and distributing any size of power stations MSCR. (orig.) [de

Planner and nuclear crisis relocation

International Nuclear Information System (INIS)

Platt, R.H.

1984-01-01

Criticism of the Federal Emergency Management Agency (FEMA) plans to relocate two-thirds of the American people one week before a nuclear attack has been so intense that planners have renamed the crisis relocation plan to Integrated Emergency Management. A careful study of plans for the town of Greenfield, Massachusetts show that they are not only unworkable, but would seriously weaken the country even if no attack occurred. Recent warnings of a global climatic catastrophe and nuclear winter in the wake of a nuclear exchange underscore the need to denounce relocation policies as dangerous. Planners should take the lead in emphasizing the long-term consequences and informing the public that there is no civil defense against nuclear war. 7 references

Transient thermal characteristics of a core channel in a molten salt reactor

International Nuclear Information System (INIS)

Sakashita, H.; Ishiguro, R.; Sugiyama, K.

1987-01-01

The present paper deals with the thermal characteristics of Molten Salt Reactor (MSR). Analyses of the fundamental behavior of internal heat generating fluid and graphite contiguous to the fluid are performed. As a result, it is known that the transient thermal characteristics of MSR differ fundamentally from those of a solid-fuel reactor, and the simplified method of thermal analysis which is commonly used for solid-fuel reactors gives optimistic predictions than the actual phenomena. (author)

Improvement to molten salt reactors

International Nuclear Information System (INIS)

Bienvenu, Claude.

1975-01-01

The invention proposes a molten salt nuclear reactor whose core includes a mass of at least one fissile element salt to which can be added other salts to lower the melting temperature of the mass. This mass also contains a substance with a low neutron capture section that does not give rise to a chemical reaction or to an azeotropic mixture with these salts and having an atmospheric boiling point under that of the mass in operation. Means are provided for collecting this substance in the vapour state and returning it as a liquid to the mass. The kind of substance chosen will depend on that of the molten salts (fissile element salts and, where required, salts to lower the melting temperature). In actual practice, the substance chosen will have an atmospheric pressure boiling point of between 600 and 1300 0 C and a melting point sufficiently below 600 0 C to prevent solidification and clogging in the return line of the substance from the exchanger. Among the materials which can be considered for use, mention is made of magnesium, rubidium, cesium and potassium but metal cesium is not employed in the case of many fissile salts, such as fluorides, which it would reduced to the planned working temperatures [fr

Influence of corium oxidation on fission product release from molten pool

International Nuclear Information System (INIS)

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

2009-01-01

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

PWR degraded core analysis

International Nuclear Information System (INIS)

Gittus, J.H.

1982-04-01

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

Industrial relocation and energy consumption: Evidence from China

International Nuclear Information System (INIS)

Zhao Xiaoli; Yin Haitao

2011-01-01

With economic development and the change of industrial structure, industrial relocation is an inevitable trend. In the process of industrial relocation, environmental externality and social cost could occur due to market failure and government failure. Little attention has been paid to this issue. In this paper, we address it with a theoretical analysis and an empirical investigation on the relationship between China's industrial relocation in the early 1990s and energy consumption which is the primary source of CO 2 emission, an environmental externality that causes increasing concerns. The macro-policy analysis suggests that there would be a positive link between China's industrial relocation in the early 1990s and energy saving (and environmental externalities reduction). Using fixed-effect regression model and simulation method, we provide an empirical support to this argument. In order to further reduce environmental externalities and social cost in the process of industrial relocation, we provide policy suggestions as follows: First, strengthen the evaluation of environmental benefits/costs; Second, pay more attention to the coordinated social-economic development; Third, avoid long-lived investment in high-carbon infrastructure in areas with industries moved in; Fourth, address employment issue in the areas with industries moved out. - Research highlights: → Little attention has been paid to the linkage between industrial relocation and environmental externality. → Our macro-policy analysis suggests that there would be a positive link between China's industrial relocation in the early 1990s and energy saving (and environmental externalities reduction). → Using fixed-effect regression model and simulation method, we find a positive link between China's industrial relocation in the early 1990s and energy saving. → Policy suggestions to further reduce environmental externalities and social cost in the process of industrial relocation are discussed.

Partial structures in molten AgBr

Energy Technology Data Exchange (ETDEWEB)

Ueno, Hiroki [Department of Condensed Matter Chemistry and Physics, Faculty of Sciences, Kyushu University, 4-2-1 Ropponmatsu, Chuo-ku, Fukuoka 810-8560 (Japan)], E-mail: ueno@gemini.rc.kyushu-u.ac.jp; Tahara, Shuta [Faculty of Pharmacy, Niigata University of Pharmacy and Applied Life Science, Higashijima, Akiha-ku, Niigata 956-8603 (Japan); Kawakita, Yukinobu [Department of Physics, Faculty of Sciences, Kyushu University, 4-2-1 Ropponmatsu, Chuo-ku, Fukuoka 810-8560 (Japan); Kohara, Shinji [Research and Utilization Division, Japan Synchrotron Radiation Research Institute (JASRI, SPring-8), 1-1-1 Koto, Sayo-cho, Sayo-gun, Hyogo 679-5198 (Japan); Takeda, Shin' ichi [Department of Physics, Faculty of Sciences, Kyushu University, 4-2-1 Ropponmatsu, Chuo-ku, Fukuoka 810-8560 (Japan)

2009-02-21

The structure of molten AgBr has been studied by means of neutron and X-ray diffractions with the aid of structural modeling. It is confirmed that the Ag-Ag correlation has a small but well-defined first peak in the partial pair distribution function whose tail penetrates into the Ag-Br nearest neighbor distribution. This feature on the Ag-Ag correlation is intermediate between that of molten AgCl (non-superionic melt) and that of molten AgI (superionic melt). The analysis of Br-Ag-Br bond angle reveals that molten AgBr preserves a rocksalt type local ordering in the solid phase, suggesting that molten AgBr is clarified as non-superionic melt like molten AgCl.

The fuel string relocation effect - why the Bruce reactors were derated

Energy Technology Data Exchange (ETDEWEB)

Gold, M; Farooqui, M Z; Adebiyi, A S; Chu, R Y; Le, N T; Oliva, A F [Ontario Hydro, Toronto, ON (Canada); Balog, G; Qu, T; DeBuda, P G [Ontario Hydro, Tiverton, ON (Canada). Bruce Nuclear Generating Station-A

1996-12-31

In the CANDU Safety Analysis process, a series of design basis accidents are chosen and analyzed to confirm safety system effectiveness. Of all the postulated accidents, the Large Break Loss of Coolant Accident (LBLOCA) - a postulated break in the Heat Transport System piping near a component that services a large number of fuel channels - sets the most demanding requirements on the speed and reactivity depth of the shutdown system devices - shutoff rods and liquid poison injection. While the event is extremely improbable, it is reanalyzed periodically and its consequences examined to ensure continued shutdown system effectiveness. In March 1993, an additional effect was identified: if the break occurred in the piping on the inlet side of the core, this would cause sudden movement of the fuel bundles (so-called fuel string relocation) in a large number of channels. In Ontario Hydro`s Bruce NGS A, Bruce NGS B and Darlington reactors, each channel is fuelled against the flow. In this situation, the relocation of the fuel string results in a sudden positive reactivity increase. This reactivity increase is in addition to the reactivity due to the core coolant voiding. The combined reactivity effect could lead to power pulses much higher than those that would arise due to coolant voiding alone. To maintain safety margins in the event of such a postulated accident, the eight Bruce NGS A and Bruce NGS B units were initially derated to 60 percent power within 2 days of the identification and confirmation of this effect. This paper: describes the fuel string relocation phenomenon in detail; explains why the consequences differ at the various Ontario Hydro reactors; outlines the actions taken with respect to each of the Ontario Hydro reactors in the months following March 1993; describes the design solutions implemented to mitigate the problem and return the Bruce reactors to higher powers. 6 refs., 1 tab., 6 figs.

Molten Salt Breeder Reactor Analysis Methods

Energy Technology Data Exchange (ETDEWEB)

Park, Jinsu; Jeong, Yongjin; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2015-05-15

Utilizing the uranium-thorium fuel cycle shows considerable potential for the possibility of MSR. The concept of MSBR should be revised because of molten salt reactor's advantage such as outstanding neutron economy, possibility of continuous online reprocessing and refueling, a high level of inherent safety, and economic benefit by keeping off the fuel fabrication process. For the development of MSR research, this paper provides the MSBR single-cell, two-cell and whole core model for computer code input, and several calculation results including depletion calculation of each models. The calculations are carried out by using MCNP6, a Monte Carlo computer code, which has CINDER90 for depletion calculation using ENDF-VII nuclear data. From the calculation results of various reactor design parameters, the temperature coefficients are all negative at the initial state and MTC becomes positive at the equilibrium state. From the results of core rod worth, the graphite control rod alone cannot makes the core subcritical at initial state. But the equilibrium state, the core can be made subcritical state only by graphite control rods. Through the comparison of the results of each models, the two-cell method can represent the MSBR core model more accurately with a little more computational resources than the single-cell method. Many of the thermal spectrum MSR have adopted a multi-region single-fluid strategy.

Recriticality calculation with GENFLO code for the BWR core after steam explosion in the lower head

Energy Technology Data Exchange (ETDEWEB)

Miettinen, J. [VTT Processes (Finland)

2002-12-01

Recriticality of the partially degraded BWR core has been studied by assuming a severe accident phase during which the fuel rods are still intact but the control rods have experienced extensive damage. Previous NKS and EU projects have studied the same case assuming reflooding by the ECCS system In the present study it was assumed that coolant enters the core due to melt-coolant interaction in the lower plenum. In the first case specified the relocation and fragmentation of the molten control rod metal causes the level swell in the core but no steam explosion. In the second case a steam explosion in the lower head was assumed. I n the first case a prompt recriticality peak can occur, but after the peak no semistable power generation remains. In the second case the consequence of the slug entrance into the core is so violent that the fuel disintegration and melting during the first power peak may occur. After the large power peak water is rapidly pushed back from the core and no semistable power generation maintains. The fuel disintegration studies have been based on a coarse assumption that the acceptable local energy addition into the fresh fuel may be 170 cal/g, but with increasing burn-up it can be as low as 60-70 cal/g. In the level swell variations the maximum energy addition was between these limits, but in most of the steam explosion variations much above these limits. Additional variation of the assumptions related to the neutronics demonstrated that for the converged analysis result some interactions would be useful with respect to the boundary conditions and neutronic options.

Combined system of accelerator molten-salt breeder (AMSB) apd molten-salt converter reactor (MSCR)

International Nuclear Information System (INIS)

Furukawa, K.; Kato, Y.; Ohmichi, T.; Ohno, H.

1983-01-01

A design and research program is discUssed of the development of accelerator molten-salt breeder (AMSB) consisting of a proton accelerator and a molten fluoride target. The target simultaneously serves as a blanket for fissionable material prodUction. An addition of some amoUnt of fissile nuclides to a melt expands the AMSB potentialities as the fissionable material production increases and the energy generation also grows up to the level of self-provision. Besides the blanket salts may be used as nuclear fuel for molten-salt converter reactor (MSCR). The combined AM SB+MSCR system has better parameters as compared to other breeder reactors, molten-salt breeder reactors (MSBR) included

Assessment of the Neutronic and Fuel Cycle Performance of the Transatomic Power Molten Salt Reactor Design

Energy Technology Data Exchange (ETDEWEB)

Robertson, Sean [Transatomic Power Corp., Cambridge, MA (United States); Dewan, Leslie [Transatomic Power Corp., Cambridge, MA (United States); Massie, Mark [Transatomic Power Corp., Cambridge, MA (United States); Davidson, Eva E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

This report presents results from a collaboration between Transatomic Power Corporation (TAP) and Oak Ridge National Laboratory (ORNL) to provide neutronic and fuel cycle analysis of the TAP core design through the Department of Energy Gateway for Accelerated Innovation in Nuclear (GAIN) Nuclear Energy Voucher program. The TAP concept is a molten salt reactor using configurable zirconium hydride moderator rod assemblies to shift the neutron spectrum in the core from mostly epithermal at beginning of life to thermal at end of life. Additional developments in the ChemTriton modeling and simulation tool provide the critical moderator-to-fuel ratio searches and time-dependent parameters necessary to simulate the continuously changing physics in this complex system. The implementation of continuous-energy Monte Carlo transport and depletion tools in ChemTriton provide for full-core three-dimensional modeling and simulation. Results from simulations with these tools show agreement with TAP-calculated performance metrics for core lifetime, discharge burnup, and salt volume fraction, verifying the viability of reducing actinide waste production with this concept. Additional analyses of mass feed rates and enrichments, isotopic removals, tritium generation, core power distribution, core vessel helium generation, moderator rod heat deposition, and reactivity coeffcients provide additional information to make informed design decisions. This work demonstrates capabilities of ORNL modeling and simulation tools for neutronic and fuel cycle analysis of molten salt reactor concepts.

Fundamentals of molten-salt thermal technology

International Nuclear Information System (INIS)

1980-08-01

This book has been published by the Society of Molten-Salt Thermal Technology to publish a part of the achievement of its members. This book is composed of seven chapters. The chapter 1 is Introduction. The chapter 2 explains the physical properties of molten salts, such as thermal behavior, surface tension, viscosity, electrical conductivity and others. The chapter 3 presents the compatibility with construction materials. Corrosion in molten salts, the electrochemical behavior of fluoride ions on carbon electrodes in fluoride melts, the behaviors of hastelloy N and metals in melts are items of this chapter. The equipments and instruments for molten salts are described in chapter 4. The heat transfer in molten salts is discussed in chapter 5. The chapter 6 explains the application of molten salt technology. The molten salt technology can be applied not only to thermal engineering and energy engineering but also to chemical and nuclear engineerings, and the technical fundamentals, current development status, technical problems and the perspective for the future are outlined. The chapter 7 is the summary of this book. The commercialization of molten salt power reactors is discussed at the end of this book. (Kato, T.)

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

International Nuclear Information System (INIS)

Journeau, C.; Haquet, J.F.

2005-01-01

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

Computational simulation of natural convection of a molten core in lower head of a PWR pressure vessel

International Nuclear Information System (INIS)

Vieira, Camila Braga; Romero, Gabriel Alves; Jian Su

2010-01-01

Computational simulation of natural convection in a molten core during a hypothetical severe accident in the lower head of a typical PWR pressure vessel was performed for two-dimensional semi-circular geometry with isothermal walls. Transient turbulent natural convection heat transfer of a fluid with uniformly distributed volumetric heat generation rate was simulated by using a commercial computational fluid dynamics software ANSYS CFX 12.0. The Boussinesq model was used for the buoyancy effect generated by the internal heat source in the flow field. The two-equation k-ω based SST (Shear Stress Transport) turbulence model was used to mould the turbulent stresses in the Reynolds-Average Navier-Stokes equations (RANS). Two Prandtl numbers, 6:13 and 7:0, were considered. Five Rayleigh numbers were simulated for each Prandtl number used (109, 1010, 1011, 1012, and 1013). The average Nusselt numbers on the bottom surface of the semicircular cavity were in excellent agreement with Mayinger et al. (1976) correlation and only at Ra = 109 the average Nusselt number on the top flat surface was in agreement with Mayinger et al. (1976) and Kulacki and Emara (1975) correlations. (author)

Calculations of the possible consequences of molten fuel sodium interactions in subassembly and whole core geometries

International Nuclear Information System (INIS)

Coddington, P.; Fishlock, T.P.; Jakeman, D.

1976-01-01

The possible consequences of molten fuel sodium interactions are calculated using various modelling assumptions and key parameters. And the significance of the choice of assumptions and parameters are discussed. As for subassembly geometry, the results of one-dimensional code EXPEL are compared with the solutions of the one-dimensional Lagrangian equations of a compressible fluid (TOPAL was used). The adequacy of acoustic approximation used in EXPEL is discussed here. The effects of heat transfer time constant on the behaviour of peak pressure are also analyzed by parametric surveys. Other items investigated are the length and position of the interacting zone, the existence of a non-condensable gas volume, and the vapour condensation on cold clad. As for whole core geometry, a simple dynamical model of arc expanding spherical interacting zone immersed in a semi-infinite sea of cold liquid was used (SHORE code). Within the interacting zone a simple heat transfer model (including a heat transfer time and a fragmentation time) was adopted. Vapour blanketing was considered in a number of ways. Representative results of the calculations are given in a table. Containment studies were also performed for ''ducted'' design and ''open pool'' design. The development of new codes in the U.K. for these analysis are also briefly described. (Aoki, K.)

Tools and applications for core design and shielding in fast reactors

International Nuclear Information System (INIS)

Rachamin, Reuven

2013-01-01

Outline: • Modeling of SFR cores using the Serpent-DYN3D code sequence; • Core shielding assessment for the design of FASTEF-MYRRHA; • Neutron shielding studies on an advanced Molten Salt Fast Reactor (MSFR) design

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

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

Molten fuel-coolant interactions resulting from power transients in aluminium plate/water moderated reactors

International Nuclear Information System (INIS)

Storr, G.J.

1989-08-01

The behaviour of two reactors SL1 and SPERT D12, which underwent fast nuclear power transients prior to core destruction by a molten fuel-coolant interaction (MFCI) has been analysed and the results compared with measured data. The calculated spatial melt distribution and the mechanical work done during the events leads to high (∼ 250 kJ/kg) conversion efficiencies for this type of interaction when compared with molten drop experiments. A simple model for the steam explosion, using static thermodynamic properties of high temperature and pressure steam is used to calculate the dynamics of the reactors following the MFCI. 26 refs., 5 figs., 5 tabs

Gases in molten salts

CERN Document Server

Tomkins, RPT

1991-01-01

This volume contains tabulated collections and critical evaluations of original data for the solubility of gases in molten salts, gathered from chemical literature through to the end of 1989. Within the volume, material is arranged according to the individual gas. The gases include hydrogen halides, inert gases, oxygen, nitrogen, hydrogen, carbon dioxide, water vapor and halogens. The molten salts consist of single salts, binary mixtures and multicomponent systems. Included also, is a special section on the solubility of gases in molten silicate systems, focussing on slags and fluxes.

Evaluation of freshwater mussel relocation as a conservation and management strategy

Science.gov (United States)

Cope, W. Gregory; Waller, Diane L.

1995-01-01

The relocation of unionacean mussels is commonly used as a conservation and management tool in large rivers and streams. Relocation has been used to recolonize areas where mussel populations have been eliminated by prior pollution events, to remove mussels from construction zones and to re-establish populations of endangered species. More recently, relocation has been used to protect native freshwater mussels from colonization by the exotic zebra mussel Dreissena polymorpha. We conducted a literature review of mussel relocations and evaluated their relative success as a conservation and management strategy. We found that 43% of all relocations were conducted because of construction projects that were forced to comply with the Endangered Species Act 1973 and that only 16% were monitored for five or more consecutive years. Most (43%) relocation projects were conducted from July to September, presumably a period when reproductive stress is relatively low for most species and the metabolic rate is sufficient for reburrowing in the substrate. The mortality of relocated mussels was unreported in 27% of projects; reported mortality varied widely among projects and species and was difficult to assess. The mean mortality of relocated mussels was 49% based on an average recovery rate of 43%. There is little guidance on the methods for relocation or for monitoring the subsequent long-term status of relocated mussels. Based on this evaluation, research is needed to develop criteria for selecting a suitable relocation site and to establish appropriate methods and guidelines for conducting relocation projects.

Flow Boiling on a Downward-Facing Inclined Plane Wall of Core Catcher

International Nuclear Information System (INIS)

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

2013-01-01

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

Molten salt reactors

International Nuclear Information System (INIS)

Bouchter, J.C.; Dufour, P.; Guidez, J.; Simon, N.; Renault, C.

2014-01-01

Molten salt reactors are one of the 6 concepts retained for the 4. generation of nuclear reactors. The principle of this reactor is very innovative: the nuclear fuel is dissolved in the coolant which allows the online reprocessing of the fuel and the online recovery of the fission products. A small prototype: the Molten Salt Reactor Experiment (MSRE - 8 MWt) was operating a few years in the sixties in the USA. The passage towards a fast reactor by the suppression of the graphite moderator leads to the concept of Molten Salt Fast Reactor (MSFR) which is presently studied through different European projects such as MOST, ALISIA and EVOL. Worldwide the main topics of research are: the adequate materials resisting to the high level of corrosiveness of the molten salts, fuel salt reprocessing, the 3-side coupling between neutron transport, thermohydraulics and thermo-chemistry, the management of the changing chemical composition of the salt, the enrichment of lithium with Li 7 in the case of the use of lithium fluoride salt and the use of MSFR using U 233 fuel (thorium cycle). The last part of the article presents a preliminary safety analysis of the MSFR. (A.C.)

Summary of treat experiments on oxide core-disruptive accidents

International Nuclear Information System (INIS)

Dickerman, C.E.; Rothman, A.B.; Klickman, A.E.; Spencer, B.W.; DeVolpi, A.

1979-02-01

A program of transient in-reactor experiments is being conducted by Argonne National Laboratory in the Transient Reactor Test (TREAT) facility to guide and support analyses of hypothetical core-disruptive accidents (HCDA) in liquid-metal fast breeder reactors (LMFBR). Test results provide data needed to establish the response of LMFBR cores to hypothetical accidents producing fuel failure, coolant boiling, and the movement of coolant, molten fuel, and molten cladding. These data include margins to fuel failure, the modes of failure and movements, and evidence for identification of the mechanisms which determine the failure and movements. A key element in the program is the fast-neutron hodoscope, which detects fuel movement as a function of time during experiments

Behavior of concrete in contact with molten corium in the case of a hypothetical core melt accident

International Nuclear Information System (INIS)

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

1979-01-01

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

TMI-2 analysis using SCDAP/RELAP5/MOD3.1

International Nuclear Information System (INIS)

Hohorst, J.K.; Polkinghorne, S.T.; Siefken, L.J.; Allison, C.M.; Dobbe, C.A.

1994-11-01

SCDAP/RELAP5/MOD3.1, an integrated thermal hydraulic analysis code developed primarily to simulate severe accidents in nuclear power plants, was used to predict the progression of core damage during the TMI-2 accident. The version of the code used for the TMI-2 analysis described in this paper includes models to predict core heatup, core geometry changes, and the relocation of molten core debris to the lower plenum of the reactor vessel. This paper describes the TMI-2 input model, initial conditions, boundary conditions, and the results from the best-estimate simulation of Phases 1 to 4 of the TMI-2 accident as well as the results from several sensitivity calculations

Transformation and fragmentation behavior of molten metal drop in sodium pool

International Nuclear Information System (INIS)

Nishimura, Satoshi; Kinoshita, Izumi; Zhang, Zhi-gang; Sugiyama, Ken-ichiro

2006-01-01

In order to clarify the fragmentation mechanism of a metallic alloy (U-Pu-Zr) fuel on liquid phase formed by metallurgical reactions (liquefaction temperature =650degC), which is important in evaluating the sequence of core disruptive accidents for metallic fuel fast reactors, a series of experiments was carried out using molten aluminum (m.p.=660degC) and sodium mainly under the condition that the boiling of sodium does not occur. When the instantaneous contact interface temperature (T i ) between molten aluminum drop and sodium is lower than the boiling point of sodium (T c,bp ), the molten aluminum drop can be fragmented and the mass median diameter (D m ) of aluminum fragments becomes small with increasing T i . When T i is roughly equivalent to or higher than T c,bp , the fragmentation of aluminum drop is promoted by thermal interaction caused by the boiling of sodium on the surface of the drop. Furthermore, even under the condition that the boiling of sodium does not occur and the solid crust is formed on the surface of the drop, it is found from an analytical evaluation that the thermal fragmentation of molten aluminum drop with solid crust is caused by the transient pressurization within the melt confined by the crust. These results indicate the possibility that the metallic alloy fuel on liquid phase formed by the metallurgical reactions can be fragmented without occurring the boiling of sodium on the surface of the melt. (author)

Molten fuel studies at Winfrith

International Nuclear Information System (INIS)

Edwards, A.J.; Knowles, J.B.; Tattersall, R.B.

1988-01-01

This report describes the experimental facilities available for molten fuel studies at Winfrith. These include a large facility capable of testing components at full LMFBR subassembly scale and also a high pressure facility for experiments at pressures up to 25 MPa, covering the whole range of temperatures and pressures of interest for the PWR. If the hypothetical accident conditions initiating the release of molten fuel do not produce an explosive transfer of thermal energy on contact of molten fuel with the reactor coolant, then an intermediate rate of heat transfer over several hundred milliseconds may occur. Theoretical work is described which is being carried out to predict the resulting pressurisation and the degree of mechanical loading on the reactor structure. Finally the current programme of molten fuel studies and recent progress are reviewed, and future plans, which are chiefly focussed on the study of thermal interactions between molten fuel and sodium coolant for the LMFBR are outlined. (author)

Sampling device for radioactive molten salt

International Nuclear Information System (INIS)

Shindo, Masato

1998-01-01

The present invention provides a device for accurately sampling molten salts to which various kinds of metals in a molten salt storage tank are mixed for analyzing them during a spent fuel dry type reprocessing. Namely, the device comprises a sampling tube having an opened lower end to be inserted into the radioactive molten salts stored in a tank and keeps reduced pressure from the upper end, and a pressure reducing pipeline having one end connected to the sampling tube and other end connected to an evacuating pump. In this device, the top end of the sampling tube is inserted to a position for sampling the radioactive molten salts (molten salts). The pressure inside the evacuating pipeline connected to the upper portion of the sampling tube is reduced for a while. In this case, the inside of the pressure reducing pipeline is previously evacuated by the evacuating pump so as to keep a predetermined pressure. Since the pressure in the sampling tube is lowered, molten salts are inserted into the sampling tube, the sampling tube is withdrawn, and the molten salts flown in the sampling tube are analyzed. (I.S.)

Thermal performances of molten salt steam generator

International Nuclear Information System (INIS)

Yuan, Yibo; He, Canming; Lu, Jianfeng; Ding, Jing

2016-01-01

Highlights: • Thermal performances of molten salt steam generator were experimentally studied. • Overall heat transfer coefficient reached maximum with optimal molten salt flow rate. • Energy efficiency first rose and then decreased with salt flow rate and temperature. • Optimal molten salt flow rate and temperature existed for good thermal performance. • High inlet water temperature benefited steam generating rate and energy efficiency. - Abstract: Molten salt steam generator is the key technology for thermal energy conversion from high temperature molten salt to steam, and it is used in solar thermal power station and molten salt reactor. A shell and tube type molten salt steam generator was set up, and its thermal performance and heat transfer mechanism were studied. As a coupling heat transfer process, molten salt steam generation is mainly affected by molten salt convective heat transfer and boiling heat transfer, while its energy efficiency is also affected by the heat loss. As molten salt temperature increased, the energy efficiency first rose with the increase of heat flow absorbed by water/steam, and then slightly decreased for large heat loss as the absorbed heat flow still rising. At very high molten salt temperature, the absorbed heat flow decreased as boiling heat transfer coefficient dropping, and then the energy efficiency quickly dropped. As the inlet water temperature increased, the boiling region in the steam generator remarkably expanded, and then the steam generation rate and energy efficiency both rose with the overall heat transfer coefficient increasing. As the molten salt flow rate increased, the wall temperature rose and the boiling heat transfer coefficient first increased and then decreased according to the boiling curve, so the overall heat transfer coefficient first increased and then decreased, and then the steam generation rate and energy efficiency of steam generator both had maxima.

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

Molten salt reactor type

International Nuclear Information System (INIS)

1977-01-01

This document is one of the three parts of a first volume devoted to the compilations of American data on the molten salt reactor concept. This part 'CIRCUITS' regroups under a condensed form - in French and using international units - the essential information contained in both basic documents of the American project for a molten-salt breeder power plant. This part is only dealing with things relating to the CEA-EDF workshop 'CIRCUITS'. It is not concerned with information on: the reactor and the moderator replacement, the primary and secondary salts, and the fuel salt reprocessing, that are dealt with in parts 'CORE' and 'CHEMISTRY' respectively. The possible evolutions in the data - and solutions - taken by the American designers for their successive projects (1970 to 1972) are shown. The MSBR power plant comprises three successive heat transfer circuits. The primary circuit (Hastelloy N), radioactive and polluted, containing the fuel salt, includes the reactor, pumps and exchangers. The secondary circuit (pipings made of modified Hastelloy N) contaminated in the exchanger, ensures the separation between the fuel and the fluid operating the turbo-alternator. The water-steam circuit feeds the turbine with steam. This steam is produced in the steam generator flowed by the secondary fluid. Some subsidiary circuits (discharge and storage of the primary and secondary salts, ventilation of the primary circuit ...) complete the three principal circuits which are briefly described. All circuits are enclosed inside the controlled-atmosphere building of the nuclear boiler. This building also ensures the biological protection and the mechanical protection against outer aggressions [fr

47 CFR 0.186 - Emergency Relocation Board.

Science.gov (United States)

2010-10-01

..., Defense and Emergency Preparedness Functions § 0.186 Emergency Relocation Board. (a) As specified in the... functions of the Commission. Relocation may be required to accommodate a variety of emergency scenarios... require the evacuation of FCC headquarters with little or no advance notice, but for only a short duration...

44 CFR 63.6 - Reimbursable relocation costs.

Science.gov (United States)

2010-10-01

... 44 Emergency Management and Assistance 1 2010-10-01 2010-10-01 false Reimbursable relocation costs. 63.6 Section 63.6 Emergency Management and Assistance FEDERAL EMERGENCY MANAGEMENT AGENCY, DEPARTMENT... OF SECTION 1306(c) OF THE NATIONAL FLOOD INSURANCE ACT OF 1968 General § 63.6 Reimbursable relocation...

The Four-Factor Taxonomy of Relocation Outcomes

Science.gov (United States)

Matthiesen, Jane Kirsten; Tissington, Patrick

2008-01-01

Relocation, an intraorganizational geographical transfer, can be used for human resource development (HRD) because of the positive developmental effects it can induce. It is, thus, important for HRD professionals to understand the implications of relocation to ensure it is used appropriately and effectively as an HRD technique. Research on…

An Interconnected Network of Core-Forming Melts Produced by Shear Deformation

Science.gov (United States)

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

2000-01-01

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

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

International Nuclear Information System (INIS)

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

1998-01-01

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

Simulation of Molten Core-Concrete Interaction in oxide/metal stratified configuration with the TOLBIAC-ICB code

International Nuclear Information System (INIS)

Tourniaire, B.; Spindler, B.

2005-01-01

The frame of this work is the validation of the TOLBIAC-ICB code which is devoted to the simulation of Molten Core-Concrete Interaction (MCCI) for reactor safety analysis. Attention focuses here on the validation of TOLBIAC-ICB in configurations expected to be representative of the long term phase of MCCI i.e. during an interaction between an oxide/metal stratified corium melt and a concrete structure. Up to now the BETA tests performed at the Forschungszentrum Karlsruhe (FzK) are the only tests available to study such kind of interaction. The BETA tests are first described and the operating conditions are reminded. The TOLBIAC-ICB code is then briefly described, with emphasis on the models used for stratified configurations. The results of the simulations are discussed. A sensitivity study is also performed with the power generated in the oxide layer instead of the metal layer as in the test. This last calculation shows that the large axial ablation observed in the tests is probably due to the peculiar configuration of the test with input power in the bottom metal layer. Since in the reactor case the residual power would be mainly concentrated in the upper oxide layer, the conclusions of the BETA tests for the reactor applications, in term of axial ablation, must be derived with caution. (author)

Thermochemical properties of some alkaline-earth silicates and zirconates. Fission product behaviour during molten core-concrete interactions

International Nuclear Information System (INIS)

Huntelaar, M.E.

1996-01-01

This thesis aims to make a contribution to a better understanding of the chemical processes occurring during an ex-vessel MCCI accident with a western-type of nuclear reactor. Chosen is for a detailed thermochemical study of the silicates and zirconates of barium and strontium. In Chapter one a short introduction in the history of (research in) nuclear safety is given, followed by the state-of-the-art of molten core-concrete interactions in Chapter two. In both Chapters the role of chemical thermodynamics on this particular subject is dealt with. The experimental work on the silicates and zirconates of barium and strontium performed for this thesis, is described in the Chapters three, four, five, six, and parts of eight. In Chapter three the basis for all thermochemical measurements, the sample preparation is given. Because the sample preparation effects the accuracy of the thermodynamic measurements, a great deal of effort is spent in optimizing the synthesis of the silicates which resulted in the TEOS-method widely employed here. In the next Chapters the different thermochemical techniques used, are described: The low-temperature heat capacity measurements and the enthalpy increment measurements in Chapter four, the enthalpy-of-solution measurements in Chapter five, and measurements to determine the crystal structures in Chapter six. (orig.)

Thermochemical properties of some alkaline-earth silicates and zirconates. Fission product behaviour during molten core-concrete interactions

Energy Technology Data Exchange (ETDEWEB)

Huntelaar, M.E.

1996-06-19

This thesis aims to make a contribution to a better understanding of the chemical processes occurring during an ex-vessel MCCI accident with a western-type of nuclear reactor. Chosen is for a detailed thermochemical study of the silicates and zirconates of barium and strontium. In Chapter one a short introduction in the history of (research in) nuclear safety is given, followed by the state-of-the-art of molten core-concrete interactions in Chapter two. In both Chapters the role of chemical thermodynamics on this particular subject is dealt with. The experimental work on the silicates and zirconates of barium and strontium performed for this thesis, is described in the Chapters three, four, five, six, and parts of eight. In Chapter three the basis for all thermochemical measurements, the sample preparation is given. Because the sample preparation effects the accuracy of the thermodynamic measurements, a great deal of effort is spent in optimizing the synthesis of the silicates which resulted in the TEOS-method widely employed here. In the next Chapters the different thermochemical techniques used, are described: The low-temperature heat capacity measurements and the enthalpy increment measurements in Chapter four, the enthalpy-of-solution measurements in Chapter five, and measurements to determine the crystal structures in Chapter six. (orig.).

Development of Relocation Strategies Regarding Eruption Hazard in Mount Merapi

Directory of Open Access Journals (Sweden)

Rosdiana Puji Lestari

2015-09-01

Full Text Available The research was conducted to answer a questions: 1 how the relocation of the current strategy was carried out by the government 2 the problems that arise in the relocation which has yet to be resolved, 3 how the good concept of relocation so that these problems can be minimized. The research methodology used in this study combined quantitative and qualitative methods. Qualitative methods are used to explain the empirical facts that exist in each of these conditions in the relocation settlements. SWOT analysis is used to provide an overview of the strengths, weaknesses, opportunities and threats in the relocation of victims of Merapi. Information, facts and issues that arise in the study will be described, explained and analyzed primarily for the impact of residential relocation on social and economic life of the community. Relocation concept of the economic aspects of community empowerment efforts should be made related to community economic recovery after a change in livelihood. In addition, the potential of community mapping needs to be done by the government to be optimized and developed. From the aspect of government settlements seek to transfer ownership of land in the new settlement which may be the property of the citizens so that citizens obtain legal certainty while occupying the new settlement. The social aspects of the development potential of relocating residents and making it as a tourist village, residents need to support that relocation is no longer a newcomer, they have the same rights and obligations with citizens of another.

Experimental investigation of the MSFR molten salt reactor concept

Energy Technology Data Exchange (ETDEWEB)

Yamaji, Bogdan; Aszodi, Attila [Budapest University of Technology and Economics (Hungary). Inst. of Nuclear Techniques

2014-11-15

In the paper experimental modelling and investigation of the MSFR concept will be presented. MSFR is a homogeneous, single region liquid fuelled fast reactor concept. In case of molten salt reactors the core neutron flux and fission distribution is determined by the flow field through distribution and transport of fissile material and delayed neutron precursors. Since the MSFR core is a single region homogeneous volume without internal structures, it is a difficult task to ensure stable flow field, which is strongly coupled to the volumetric heat generation. These considerations suggest that experimental modelling would greatly help to understand the flow phenomena in such geometry. A scaled and segmented experimental mock-up of MSFR was designed and built in order to carry out particle image velocimetry measurements. Basic flow behaviour inside the core region can be investigated and the measurement data can also provide resource for the validation of computational fluid dynamics models. Measurement results of steady state conditions will be presented and discussed.

Evolution of fast reactor core spectra in changing a heavy liquid metal coolant by molten PB-208

Energy Technology Data Exchange (ETDEWEB)

Blokhin, D. A.; Mitenkova, E. F. [Nuclear Safety Inst., Russian Academy of Sciences, B. Tulskaya 52, Moscow, 115119 (Russian Federation); Khorasanov, G. L.; Zemskov, E. A.; Blokhin, A. I. [State Scientific Center, Russian Federation, Inst. of Physics and Power Engineering, Bondarenko Square 1, Obninsk, 249033 (Russian Federation)

2012-07-01

In the paper neutron spectra of fast reactor cooled with lead-bismuth or lead-208 are given. It is shown that in changing the coolant from lead-bismuth to lead-208 the core neutron spectra of the fast reactor FR RBEC-M are hardening in whole by several percents when a little share of low energy neutrons (5 eV - 50 keV) is slightly increasing. The shift of spectra to higher energies permits to enhance the fuel fission while the increased share of low energy neutrons provides more effective conversion of uranium-238 into plutonium due to peculiarity of {sup 238}U neutron capture cross section. Good neutron and physical features of molten {sup 208}Pb permit to assume it as perspective coolant for fast reactors and accelerator driven systems. The one-group cross sections of neutron radiation capture, {sigma}(n,g), by {sup 208}Pb, {sup 238}U, {sup 99}Tc, mix of lead and bismuth, {sup nat}Pb-Bi, averaged over neutron spectra of the fast reactor RBEC-M are given. It is shown that one-group cross sections of neutron capture by material of the liquid metal coolant consisted from lead enriched with the stable lead isotope, {sup 208}Pb, are by 4-7 times smaller {sigma}(n,g) for the coolant {sup nat}Pb-Bi. The economy of neutrons in the core cooled with {sup 208}Pb can be used for reducing reactor's initial fuel load, increasing fuel breeding and transmutation of long lived fission products, for example {sup 99}Tc. Good neutron and physical features of lead enriched with {sup 208}Pb permit to consider it as a perspective low neutron absorbing coolant for fast reactors and accelerator driven systems. (authors)

Accelerator molten-salt breeding and thorium fuel cycle

International Nuclear Information System (INIS)

Furukawa, Kazuo; Nakahara, Yasuaki; Kato, Yoshio; Ohno, Hideo; Mitachi, Kohshi.

1990-01-01

The recent efforts at the development of fission energy utilization have not been successful in establishing fully rational technology. A new philosophy should be established on the basis of the following three principles: (1) thorium utilization, (2) molten-salt fuel concept, and (3) separation of fissile-breeding and power-generating functions. Such philosophy is called 'Thorium Molten-Salt Nuclear Energy Synergetics [THORIMS-NES]'. The present report first addresses the establishment of 233 U breeding fuel cycle, focusing on major features of the Breeding and Chemical Processing Centers and a small molten-salt power station (called FUJI-II). The development of fissile producing breeders is discussed in relation to accelerator molten-salt breeder (AMSB), impact fusion molten-salt breeder, and inertial-confined fusion hybrid molten-salt breeder. Features of the accelerator molten-salt breeder are described, focusing on technical problems with accelerator breeders (or spallators), design principle of the accelerator molten-salt breeder, selection of molten salt compositions, and nuclear- and reactor-chemical aspects of AMSB. Discussion is also made of further research and development efforts required in the future for AMSB. (N.K.)

Thermal hydraulic study of a corium molten pool

International Nuclear Information System (INIS)

Pigny, S.; Grand, D.; Seiler, J.M.; Durin, M.

1993-01-01

The thermohydraulic behaviour of a mass of molten core is investigated, in the frame of PWR severe accidents studies. The corium may be located in the vessel lower head or in an external core-catcher. It is assumed to be present in the container instantaneously. Its motion is described by one velocity field. It may be homogeneous or made of two stratified fluids. The residual power is assumed to be constant and uniform in the UO 2 phase. The radiative losses and the external water-cooling are taken into account. The thermal resistance of a peripheral crust is considered. The influence of the crust on the pool geometry may be studied. The wall behaviour is analysed by a conduction calculation. The interest of a sacrificial layer is underlined, so as the necessity of a multicomponent multiphase model to study the behaviour of a core catcher. It is also concluded that some experiments are needed for code validation about volume heated natural convection and multiphase flows. (author). 14 figs., 3 refs

Characterisation of Ceramic-Coated 316LN Stainless Steel Exposed to High-Temperature Thermite Melt and Molten Sodium

Science.gov (United States)

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

2017-11-01

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

Transformation and fragmentation behavior of molten metal drop in sodium pool

International Nuclear Information System (INIS)

Nishimura, Satoshi; Zhang Zhigang; Sugiyama, Ken-Ichiro; Kinoshita, Izumi

2007-01-01

In order to clarify the fragmentation mechanism of a metallic alloy (U-Pu-Zr) fuel on liquid phase formed by metallurgical reactions (liquefaction temperature = 650 deg. C), which is important in evaluating the sequence of core disruptive accidents for metallic fuel fast reactors, a series of experiments was carried out using molten aluminum (melting point 660 deg. C) and sodium mainly under the condition that the boiling of sodium does not occur. When the instantaneous contact interface temperature (T i ) between molten aluminum drop and sodium is lower than the boiling point of sodium (T c,bp ), the molten aluminum drop can be fragmented and the mass median diameter (D m ) of aluminum fragments becomes small with increasing T i . When T i is roughly equivalent to or higher than T c,bp , the fragmentation of aluminum drop is promoted by thermal interaction caused by the boiling of sodium on the surface of the drop. Furthermore, even under the condition that the boiling of sodium does not occur and the solid crust is formed on the surface of the drop, it is confirmed from an analytical evaluation that the thermal fragmentation of molten aluminum drop with solid crust has a potential to be caused by the transient pressurization within the melt confined by the crust. These results indicate the possibility that the metallic alloy fuel on liquid phase formed by the metallurgical reactions can be fragmented without occurring the boiling of sodium on the surface of the melt

Relocation impacts of a major release from SRTC

International Nuclear Information System (INIS)

Blanchard, A.; Thompson, E.A.; Thompson, J.M.

1999-01-01

The relocation impacts of an accidental release, scenario 1-RD-3 , are evaluated for the Savannah River Technology Center. The extent of the area potentially contaminated to a level that would result in doses exceeding the relocation protective action guide(PAG)is calculated. The maximum calculated distance downwind from the accident at which the relocation PAG is exceeded is also determined. The consequences of the particulate portion of the release are evaluated using the HOTSPOT model and an EXCEL spreadsheet. The consequences of the tritium release are evaluated using UFOTRI

Influence of corium oxidation on fission product release from molten pool

Energy Technology Data Exchange (ETDEWEB)

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

2010-05-15

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

Core catcher cooling for a gas-cooled fast breeder

International Nuclear Information System (INIS)

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

1976-01-01

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

Strategic environmental policy when the governments are threatened by relocation

International Nuclear Information System (INIS)

Greaker, Mads

2003-01-01

This paper analyzes how the threat of relocation influences environmental policy. The stringency of environmental policy is determined in a game between two governments. There is one firm in each jurisdiction, and both firms threaten to relocate their production to the other jurisdiction. Because there is asymmetric information about the cost of relocation, the governments do not know the credibility of the threat. We compare the outcome of this game with the outcome of a game in which relocation is not possible. We find that the threat of relocation can increase both the level of environmental regulation and welfare. The profit tax level proves to be the most decisive for the result; that is, the higher the profit tax level, the lower the level of environmental regulation

Compatibility of molten salt and structural materials

International Nuclear Information System (INIS)

Kawakami, Masahiro

1994-01-01

As the important factors for considering the compatibility of fuel salt and coolant salt with structural materials in molten salt reactors, there are the moisture remaining in molten salt and the fluorine potential in molten salt. In this study, as for the metals which are the main components of corrosion resistant alloys, the corrosion by the moisture remaining in molten salt and the dependence of the corrosion on fluorine potential were examined. As the molten salts, an eutectic molten salt LiF-BeF 2 was mainly used, and LiF-KF was used in combination. As the metallic materials, Cr, Ni and Cu which are the main components of corrosion resistant and heat resistant alloys, Hastelloy and Monel, were used. In the experiment, the metal pieces were immersed in the molten salt, and by sampling the molten salt, the change with time lapse of the concentration of the dissolved metals was examined. Besides, the electrochemical measurement was carried out for Cr, of which the corrosion was remarkable, and the change with time lapse of the dissolved ions was examined. The experimental setup, the experimental method, and the results of the immersion test and the electrochemical test are reported. The experiment on the corrosion of metals depending on fluorine potential is also reported. (K.I.)

Rapidly changing flows in the Earth's core

DEFF Research Database (Denmark)

Olsen, Nils; Mandea, M.

2008-01-01

A large part of the Earth's magnetic field is generated by fluid motion in the molten outer core(1). As a result of continuous satellite measurements since 1999, the core magnetic field and its recent variations can now be described with a high resolution in space and time(2). These data have...... field occurring over only a few months, indicative of fluid flow at the top of the core, can in fact be resolved. Using nine years of magnetic field data obtained by satellites as well as Earth-based observatories, we determine the temporal changes in the core magnetic field and flow in the core. We...

Executive summary. Session summaries

International Nuclear Information System (INIS)

2011-01-01

The possibility of stopping and/or delaying the progression of a core melt accident by the use of a recovered water source or by taking benefit of specific engineered systems is taken into account in a number of PSA (Probabilistic Safety Assessment) studies. The likelihood to stop the progression of a core melt-down accident by water injection is generally considered as high in the early phase of core degradation and depends on reactor specific features, nevertheless even in later sequences, e.g. during the relocation in the lower head, cooling still can be achieved but depends on reactor specific features and the accident scenario.. Ongoing, starting and planned experimental programmes address the coolability issues in the different relevant configurations, i.e. reflooding of bundles, debris beds, molten pools and Reactor Pressure Vessel external cooling. There is still a difficulty with present models to predict reliably if reflooding during the early core degradation would or not trigger a cladding oxidation runaway. Whether this is due to deficiencies in thermal-hydraulics description or problems for taking into account the oxidation of melts is a matter of discussion. The code developments are promisingly directed towards a more mechanistic approach using a porous medium modelling able to treat the different configurations of a degraded core. Secondly, the models to describe adequately the relocation of parts of the molten core to the lower head and the debris bed formation still need further development and qualification. Their validation is expected against the results of ongoing experimental programmes. The transposition of results to the reactor scale where multi-dimensional effects are expected needs to be evaluated, all the more as larger scale experiments are probably not feasible

Relocation and Livelihood Concerns of Sariska Tiger Project, Rajasthan: A Pride or Plight?

Directory of Open Access Journals (Sweden)

Muraree Lal Meena

2015-11-01

Full Text Available It is well known that Sariska Tiger Reserve is a home to the India’s national animal - the Bengal Tiger (Panthera tigris tigris. The crux of this research is to examine the role of local peoples in the conservation of Sariska Tiger project, which was declared a wildlife reserve way back in 1955 and then further raised to a status of a Tiger Reserve in 1978, and a National Park in 1982. According to the Government officials, the people around the reserve are not only responsible for degrading the reserve, it has also emerged as a safe haven for the poachers involved in illegal hunting of the tigers. One of the reasons identified to be the cause of tiger extinction is the human habitation in the core and in the peripheral areas surrounding the tiger reserves. Despite efforts being made to protect, the Sariska Tiger Reserve, located in Alwar (Rajasthan is seriously facing the problem of tiger extinction. It is estimated that there are around 12 thousand people residing inside the tiger reserve, with 11 villages in the Core Zone-1 area and about 170 villages situated along the peripheries of the reserve. Among the prime measures undertaken are diversions of traffic, relocation of the villages located inside the forest. The Tiger Task Force (2005 has recommended relocation of three key villages surrounding the central area of the Sariska Tiger Reserve. However, this relocation of villages is quite contrary to the life style of the people residing within these villages. Their social and cultural attributes, needs and aspirations have not been given due consideration and the villagers are strongly resenting this move by the government. These displaced villagers have been living in perfect harmony with their environment from time immemorial and it is hard to understand how these villagers can be held responsible for degrading their natural environment, which is their lifeline. The government has failed to take into consideration the role of the

Social Relationships, Psychosocial Adaptation, and Intrainstitutional Relocation of Elderly Men.

Science.gov (United States)

Tesch, Stephanie A.; And Others

1989-01-01

Studied intrainstitutional relocation of 40 men aged 53 to 91 and its effects on social integration, and behavioral and self-reported measures of psychosocial adaptation. Found in both longitudinal and time-lag analyses of pre- and post-relocation data, relocated men had lower scores on attitude toward aging dimension, fewer friends, and evidenced…

Compatibility of molten salt and structural materials

Energy Technology Data Exchange (ETDEWEB)

Kawakami, Masahiro [Toyohashi Univ. of Technology, Aichi (Japan)

1994-12-01

As the important factors for considering the compatibility of fuel salt and coolant salt with structural materials in molten salt reactors, there are the moisture remaining in molten salt and the fluorine potential in molten salt. In this study, as for the metals which are the main components of corrosion resistant alloys, the corrosion by the moisture remaining in molten salt and the dependence of the corrosion on fluorine potential were examined. As the molten salts, an eutectic molten salt LiF-BeF{sub 2} was mainly used, and LiF-KF was used in combination. As the metallic materials, Cr, Ni and Cu which are the main components of corrosion resistant and heat resistant alloys, Hastelloy and Monel, were used. In the experiment, the metal pieces were immersed in the molten salt, and by sampling the molten salt, the change with time lapse of the concentration of the dissolved metals was examined. Besides, the electrochemical measurement was carried out for Cr, of which the corrosion was remarkable, and the change with time lapse of the dissolved ions was examined. The experimental setup, the experimental method, and the results of the immersion test and the electrochemical test are reported. The experiment on the corrosion of metals depending on fluorine potential is also reported. (K.I.).

Numerical predictions of natural convection in a uniformly heated pool

International Nuclear Information System (INIS)

Tzanos, C.P.; Cho, D.H.

1993-01-01

In the event of a core meltdown accident, one of the accident progression paths is fuel relocation to the lower reactor plenum. In the heavy-water new production reactor (NPR-HWR) design, the reactor cavity is flooded with water. In such a design, decay heat removal to the water in the reactor cavity and thence to the containment may be adequate to keep the reactor vessel temperature below failure limits. If this is the case, the accident progression can be arrested by retaining a coolable corium configuration in the lower reactor plenum. The strategy of reactor cavity flooding to prevent reactor vessel failure from molten corium relocation to the reactor vessel lower head has also been considered for commercial pressurized water reactors

Franchise Relocations, Expansions, and Mergers in Professional Sports Leagues

OpenAIRE

Daniel, Rascher

2008-01-01

All three sections in this chapter are interrelated. Expansions and relocations, especially in the early years of a league, are often the response to upstart rival leagues. More recently, relocations have occurred because another city offers a better facility lease regardless of whether the league as a whole is better off or not. Relocations, more so than expansions, often end up in court whether as an antitrust case accusing the league of monopolistically restricting business or as an emi...

The molten salt reactor adventure

International Nuclear Information System (INIS)

MacPherson, H.G.

1985-01-01

A personal history of the development of molten salt reactors in the United States is presented. The initial goal was an aircraft propulsion reactor, and a molten fluoride-fueled Aircraft Reactor Experiment was operated at Oak Ridge National Laboratory in 1954. In 1956, the objective shifted to civilian nuclear power, and reactor concepts were developed using a circulating UF 4 -ThF 4 fuel, graphite moderator, and Hastelloy N pressure boundary. The program culminated in the successful operation of the Molten Salt Reactor Experiment in 1965 to 1969. By then the Atomic Energy Commission's goals had shifted to breeder development; the molten salt program supported on-site reprocessing development and study of various reactor arrangements that had potential to breed. Some commercial and foreign interest contributed to the program which, however, was terminated by the government in 1976. The current status of the technology and prospects for revived interest are summarized

A calculational procedure for neutronic and depletion analysis of Molten-Salt reactors based on SCALE6/TRITON

International Nuclear Information System (INIS)

Sheu, R.J.; Chang, J.S.; Liu, Y.-W. H.

2011-01-01

Molten-Salt Reactors (MSRs) represent one of the selected categories in the GEN-IV program. This type of reactor is distinguished by the use of liquid fuel circulating in and out of the core, which makes it possible for online refueling and salt processing. However, this operation characteristic also complicates the modeling and simulation of reactor core behaviour using conventional neutronic codes. The TRITON sequence in the SCALE6 code system has been designed to provide the combined capabilities of problem-dependent cross-section processing, rigorous treatment of neutron transport, and coupled with the ORIGEN-S depletion calculations. In order to accommodate the simulation of dynamic refueling and processing scheme, an in-house program REFRESH together with a run script are developed for carrying out a series of stepwise TRITON calculations, that makes the work of analyzing the neutronic properties and performance of a MSR core design easier. As a demonstration and cross check, we have applied this method to reexamine the conceptual design of Molten Salt Actinide Recycler & Transmuter (MOSART). This paper summarizes the development of the method and preliminary results of its application on MOSART. (author)

Study of the pyrochemical treatment-recycling process of the Molten Salt Reactor fuel; Estudio de sistema de un proceso de tratamiento-reciclaje piroquimico del combustible de un reactor de sales fundidas

Energy Technology Data Exchange (ETDEWEB)

Boussier, H.; Heuer, D.

2010-07-01

The Separation Processes Studies Laboratory (Commissariat a l'energie Atomique) has made a preliminary assessment of the reprocessing system associated with Molten Salt Fast Reactor (MSFR). The scheme studied in this paper is based on the principle of reductive extraction and metal transfer that constituted the core process designed for the Molten Salt Breeder Reactor (MSBR), although the flow diagram has been adapted to the current needs of the Molten Salt Fast Reactor (MSFR).

Relocation: Its Effect on Health, Functioning and Mortality.

Science.gov (United States)

Borup, Jerry H.; And Others

1980-01-01

Relocation of older patients had a positive effect on hypochondria, stamina, hygiene, and daily functioning but no effect on health status. Self-health assessments, stamina, hypochondria, and hygiene had no effect on the mortality rate of relocated patients, but daily functioning did effect the mortality rate. (Author)

Individual relocation decisions after tornadoes: a multi-level analysis.

Science.gov (United States)

Cong, Zhen; Nejat, Ali; Liang, Daan; Pei, Yaolin; Javid, Roxana J

2018-04-01

This study examines how multi-level factors affected individuals' relocation decisions after EF4 and EF5 (Enhanced Fujita Tornado Intensity Scale) tornadoes struck the United States in 2013. A telephone survey was conducted with 536 respondents, including oversampled older adults, one year after these two disaster events. Respondents' addresses were used to associate individual information with block group-level variables recorded by the American Community Survey. Logistic regression revealed that residential damage and homeownership are important predictors of relocation. There was also significant interaction between these two variables, indicating less difference between homeowners and renters at higher damage levels. Homeownership diminished the likelihood of relocation among younger respondents. Random effects logistic regression found that the percentage of homeownership and of higher income households in the community buffered the effect of damage on relocation; the percentage of older adults reduced the likelihood of this group relocating. The findings are assessed from the standpoint of age difference, policy implications, and social capital and vulnerability. © 2018 The Author(s). Disasters © Overseas Development Institute, 2018.

Experimental studies of thermal and chemical interactions between molten aluminum and nuclear dispersion fuels with water

International Nuclear Information System (INIS)

Farahani, A.A.

1997-01-01

Because of the possibility of rapid physical and chemical molten fuel-water interactions during a core melt accident in noncommercial or experimental reactors, it is important to understand the interactions that might occur if these materials were to contact water. An existing vertical 1-D shock tube facility was improved and a gas sampling device to measure the gaseous hydrogen in the upper chamber of the shock tube was designed and built to study the impact of a water column driven downward by a pressurized gas onto both molten aluminum (6061 alloy) and oxide and silicide depleted nuclear dispersion fuels in aluminum matrices. The experiments were carried out with melt temperatures initially at 750 to 1,000 C and water at room temperature and driving pressures of 0.5 and 1 MPa. Very high transient pressures, in many cases even larger than the thermodynamic critical pressure of the water (∼ 20 MPa), were generated due to the interactions between the water and the crucible and its contents. The molten aluminum always reacted chemically with the water but the reaction did not increase consistently with increasing melt temperature. An aluminum ignition occurred when water at room temperature impacted 28.48 grams of molten aluminum at 980.3 C causing transient pressures greater than 69 MPa. No signs of aluminum ignition were observed in any of the experiments with the depleted nuclear dispersion fuels, U 3 O 8 -Al and U 3 Si 2 -Al. The greater was the molten aluminum-water chemical reaction, the finer was the debris recovered for a given set of initial conditions. Larger coolant velocities (larger driving pressures) resulted in more melt fragmentation but did not result in more molten aluminum-water chemical reaction. Decreasing the water temperature also resulted in more melt fragmentation and did not suppress the molten aluminum-water chemical reaction

Structure and thermodynamics of molten salts

International Nuclear Information System (INIS)

Papatheodorou, G.N.

1983-01-01

This chapter investigates single-component molten salts and multicomponent salt mixtures. Molten salts provide an important testing ground for theories of liquids, solutions, and plasmas. Topics considered include molten salts as liquids (the pair potential, the radial distribution function, methods of characterization), single salts (structure, thermodynamic correlations), and salt mixtures (the thermodynamics of mixing; spectroscopy and structure). Neutron and X-ray scattering techniques are used to determine the structure of molten metal halide salts. The corresponding-states theory is used to obtain thermodynamic correlations on single salts. Structural information on salt mixtures is obtained by using vibrational (Raman) and electronic absorption spectroscopy. Charge-symmetrical systems and charge-unsymmetrical systems are used to examine the thermodynamics of salt mixtures

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)

Definition of breeding gain for molten salt reactors - 147

International Nuclear Information System (INIS)

Nagy, K.; Kloosterman, J.L.; Lathouwers, D.; Van der Hagen, T.H.J.J.

2010-01-01

The graphite-moderated Molten Salt Reactor (MSR) is a potential breeder reactor using the thorium fuel cycle. The MSR has unique properties due to the possibility of making changes to the salt composition during operation. Most important is the extraction of protactinium, which separates the fissile uranium production into two volumes: the reactor core and the external stockpile. The paper focuses on the definition of breeding gain in such a system. The prospects of using breeding gain expressions defined for solid fuel reactors are investigated and new definitions are given which incorporate the processes occurring in the reactor core and the external stockpile. The difference of the growth rate of the mass of fissile material and breeding gain is pointed out. The new definitions are applied to an optimization study of the graphite-salt lattice of a breeder MSR. (authors)

Results of recent reactor-material tests on dispersal of oxide fuel from a disrupted core

International Nuclear Information System (INIS)

Spencer, B.W.; Wilson, R.J.; Vetter, D.L.; Erickson, E.G.; Dewey, G.

1985-01-01

The results of experimental investigations and related analyses are reported addressing the dispersal of molten oxide fuel from a disrupted core via various available pathways for the CRBR system. These investigations included the GAPFLOW tests in which pressure-driven and gravity drainage tests were performed using dispersal pathways mocking up the intersubassembly gaps, the CAMEL C6 and C7 tests in which molten fuel entered sodium-filled control assembly ducts under prototypic thermal-hydraulic conditions, and the Lower Internals Drainage (LID) tests in which molten fuel drained downward through simulated below-core structure (orifice plate stacks) as the bottom of control assembly ducts. The results of SHOTGUN tests addressing basic freezing of molten UO 2 and UO 2 /metal mixtures flowing through circular tubes are also reported. Test results have invariably shown the existance of stable UO 2 crusts on the inside surfaces of the flow paths. Appreciable removal of fuel was indicated prior to freezing-induced immobilization. Application of heat transfer models based upon the presence of stable, insulating fuel crusts tends to overpredict the removal process

Heat transfer analysis to investigate the core catcher plate assembly in SFR

International Nuclear Information System (INIS)

Patil, Swapnil; Sharma, Anil Kumar; Velusamy, K.; Nashine, B.K.; Selvaraj, P.

2015-01-01

Severe accident scenario in Sodium Cooled Fast Reactor (SFR) is the major concern for public acceptance. After severe accident, the molten core continuously generates substantial decay heat. However, an in-vessel core catcher plate is provided to remove the decay heat passively. The numerical investigation of pool hydraulics phenomena in sodium pool of typical Indian SFR has been carried out. The debris may form a heap with different angle over the core catcher plate due to molten fuel density and interaction force. Therefore, the debris bed with different heap angle has been analyzed for steady and transient state conditions. The governing equation of fluid flow and heat transfer are solved by finite volume method based solver with the k-ε turbulent model. The time period Δ for which temperature is exceeding above safety limit with different debris heap angle have been established. (author)

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

International Nuclear Information System (INIS)

Kleinhietpass, I.D.

2006-01-01

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

Lunar core formation: New constraints from metal-silicate partitioning of siderophile elements

NARCIS (Netherlands)

Rai, N.; van Westrenen, W.

2014-01-01

Analyses of Apollo era seismograms, lunar laser ranging data and the lunar moment of inertia suggest the presence of a small, at least partially molten Fe-rich metallic core in the Moon, but the chemical composition and formation conditions of this core are not well constrained. Here, we assess

Potentiometric Sensor for Real-Time Monitoring of Multivalent Ion Concentrations in Molten Salt

International Nuclear Information System (INIS)

Zink, Peter A.; Jue, Jan-Fong; Serrano, Brenda E.; Fredrickson, Guy L.; Cowan, Ben F.; Herrmann, Steven D.; Li, Shelly X.

2010-01-01

Electrorefining of spent metallic nuclear fuel in high temperature molten salt systems is a core technology in pyroprocessing, which in turn plays a critical role in the development of advanced fuel cycle technologies. In electrorefining, spent nuclear fuel is treated electrochemically in order to effect separations between uranium, noble metals, and active metals, which include the transuranics. The accumulation of active metals in a lithium chloride-potassium chloride (LiCl-KCl) eutectic molten salt electrolyte occurs at the expense of the UCl3-oxidant concentration in the electrolyte, which must be periodically replenished. Our interests lie with the accumulation of active metals in the molten salt electrolyte. The real-time monitoring of actinide concentrations in the molten salt electrolyte is highly desirable for controlling electrochemical operations and assuring materials control and accountancy. However, real-time monitoring is not possible with current methods for sampling and chemical analysis. A new solid-state electrochemical sensor is being developed for real-time monitoring of actinide ion concentrations in a molten salt electrorefiner. The ultimate function of the sensor is to monitor plutonium concentrations during electrorefining operations, but in this work gadolinium was employed as a surrogate material for plutonium. In a parametric study, polycrystalline sodium beta double-prime alumina (Na-β(double p rime)-alumina) discs and tubes were subject to vapor-phase exchange with gadolinium ions (Gd3+) using a gadolinium chloride salt (GdCl3) as a precursor to produce gadolinium beta double-prime alumina (Gd-β(double p rime)-alumina) samples. Electrochemical impedance spectroscopy and microstructural analysis were performed on the ion-exchanged discs to determine the relationship between ion exchange and Gd3+ ion conductivity. The ion-exchanged tubes were configured as potentiometric sensors in order to monitor real-time Gd3+ ion concentrations in

Fuel pellet fracture and relocation

International Nuclear Information System (INIS)

Walton, L.A.; Husser, D.L.

1983-01-01

The model used to describe fuel pellet fracture and relocation is an important feature of a fuel performance computer code. This model becomes especially important if the computer code is principally to be used for the evaluation of pellet clad interaction. The fracture and relocation model being developed for the B and W fuel performance code FUMAC was derived from an extensive data base. Cross sections of irradiated fuel rods were photographically magnified and measured to determine the configuration of the fragments of the fractured fuel pellets. Data, representing a wide range of LWR fuel designs and as-manufactured mechanical configurations, were catalogued and systematically reduced and then correlated as a function of the likely independent variables. These correlations define the key phenomenological behavior patterns which the relocation model must duplicate and indicate which mechanistic approaches are viable explanations of this behavior. The data base covers the burnup range from approximately one to 35 GWd/mtU and linear heat rates from less than 100 to nearly 700 W/Cm. This paper presents the correlated data base and the methods used to derive and interpret it. It was determined from this data base that pellet cracking is initially both power level and burnup dependent but tends to saturate eventually with continued steady irradiation. Fuel pellet relocation was found to be much more extensive than would be deduced from thermal considerations alone. Even at very low burnups fuel fragments were found to move outward until restrained by the cladding. The results also suggest that changes in internal resistance to heat flow within the pellets due to the opening of cracks may be as important as peripheral gap changes to the thermal modeler. The transient response and thermal implications of this model are recommended as primary areas for future investigation

Uncertainty analysis and flow measurements in an experimental mock-up of a molten salt reactor concept

Energy Technology Data Exchange (ETDEWEB)

Yamaji, Bogdan; Aszodi, Attila [Budapest University of Technology and Economics (Hungary). Inst. of Nuclear Techniques

2016-09-15

In the paper measurement results from the experimental modelling of a molten salt reactor concept will be presented along with detailed uncertainty analysis of the experimental system. Non-intrusive flow measurements are carried out on the scaled and segmented mock-up of a homogeneous, single region molten salt fast reactor concept. Uncertainty assessment of the particle image velocimetry (PIV) measurement system applied with the scaled and segmented model is presented in detail. The analysis covers the error sources of the measurement system (laser, recording camera, etc.) and the specific conditions (de-warping of measurement planes) originating in the geometry of the investigated domain. Effect of sample size in the ensemble averaged PIV measurements is discussed as well. An additional two-loop-operation mode is also presented and the analysis of the measurement results confirm that without enhancement nominal and other operation conditions will lead to strong unfavourable separation in the core flow. It implies that use of internal flow distribution structures will be necessary for the optimisation of the core coolant flow. Preliminary CFD calculations are presented to help the design of a perforated plate located above the inlet region. The purpose of the perforated plate is to reduce recirculation near the cylindrical wall and enhance the uniformity of the core flow distribution.

In-Vessel Retention Modeling Capabilities of SCDAP/RELAP5-3DC

International Nuclear Information System (INIS)

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

2002-01-01

Molten core materials may relocate to the lower head of a reactor vessel in the latter stages of a severe accident. Under such circumstances, in-vessel retention (IVR) of the molten materials is a vital step in mitigating potential severe accident consequences. Whether IVR occurs depends on the interactions of a number of complex processes including heat transfer inside the accumulated molten pool, heat transfer from the molten pool to the reactor vessel (and to overlying fluids), and heat transfer from exterior vessel surfaces. SCDAP/RELAP5-3D C has been developed at the Idaho National Engineering and Environmental Laboratory to facilitate simulation of the processes affecting the potential for IVR, as well as processes involved in a wide variety of other reactor transients. In this paper, current capabilities of SCDAP/RELAP5-3D C relative to IVR modeling are described and results from typical applications are provided. In addition, anticipated developments to enhance IVR simulation with SCDAP/RELAP5-3D C are outlined. (authors)

Physical properties of molten carbonate electrolyte

Energy Technology Data Exchange (ETDEWEB)

Kojima, T.; Yanagida, M.; Tanimoto, K. [Osaka National Research Institute (Japan)] [and others

1996-12-31

Recently many kinds of compositions of molten carbonate electrolyte have been applied to molten carbonate fuel cell in order to avoid the several problems such as corrosion of separator plate and NiO cathode dissolution. Many researchers recognize that the addition of alkaline earth (Ca, Sr, and Ba) carbonate to Li{sub 2}CO{sub 3}-Na{sub 2}CO{sub 3} and Li{sub 2}CO{sub 3}-K{sub 2}CO{sub 3} eutectic electrolytes is effective to avoid these problems. On the other hand, one of the corrosion products, CrO{sub 4}{sup 2-} ion is found to dissolve into electrolyte and accumulated during the long-term MCFC operations. This would affect the performance of MCFC. There, however, are little known data of physical properties of molten carbonate containing alkaline earth carbonates and CrO{sub 4}{sup 2-}. We report the measured and accumulated data for these molten carbonate of electrical conductivity and surface tension to select favorable composition of molten carbonate electrolytes.

The Live program - Results of test L1 and joint analyses on transient molten pool thermal hydraulics

Energy Technology Data Exchange (ETDEWEB)

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

2010-07-01

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

Aluminum titanate crucible for molten uranium

International Nuclear Information System (INIS)

Asbury, J.J.

1975-01-01

An improved crucible for molten uranium is described. The crucible or crucible liner is formed of aluminum titanate which essentially eliminates contamination of uranium and uranium alloys during molten states thereof. (U.S.)

Mississippi Labor Mobility Demonstration Project--Relocating the Unemployed: Dimensions of Success.

Science.gov (United States)

Speight, John F.; And Others

The document provides an analysis of relocation stability of individuals relocated during the March, 1970-November, 1971 contract period. Data bases were 1,244 applicants with screening information and 401 individuals with follow-up interview information. Approximately one half were in new areas six months after being relocated. Reasons for…

Improvements and validation of the transient analysis code MOREL for molten salt reactors

International Nuclear Information System (INIS)

Zhuang Kun; Zheng Youqi; Cao Liangzhi; Hu Tianliang; Wu Hongchun

2017-01-01

The liquid fuel salt used in the molten salt reactors (MSRs) serves as the fuel and coolant simultaneously. On the one hand, the delayed neutron precursors circulate in the whole primary loop and part of them decay outside the core. On the other hand, the fission heat is carried off directly by the fuel flow. These two features require new analysis method with the coupling of fluid flow, heat transfer and neutronics. In this paper, the recent update of MOREL code is presented. The update includes: (1) the improved quasi-static method for the kinetics equation with convection term is developed. (2) The multi-channel thermal hydraulic model is developed based on the geometric feature of MSR. (3) The Variational Nodal Method is used to solve the neutron diffusion equation instead of the original analytic basis functions expansion nodal method. The update brings significant improvement on the efficiency of MOREL code. And, the capability of MOREL code is extended for the real core simulation with feedback. The numerical results and experiment data gained from molten salt reactor experiment (MSRE) are used to verify and validate the updated MOREL code. The results agree well with the experimental data, which prove the new development of MOREL code is correct and effective. (author)

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

International Nuclear Information System (INIS)

Rogers, J.T.

1984-06-01

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

Relocation Impacts of an H-Separations Coil and Tube Failure

International Nuclear Information System (INIS)

Blanchard, A.

1999-01-01

The relocation impacts of an accidental release, scenario 1-RD-2.35, are evaluated for H-Separations. The extent of the area potentially contaminated to a level that would result in doses exceeding the relocation protective action guide (PAG) is calculated. The maximum calculated distance downwind from the accident at which the relocation PAG is exceeded is also determined. The consequences of the release are evaluated using the HOTSPOT model and an EXCEL spreadsheet

Relocation impacts of an H-Separations coil and tube failure

International Nuclear Information System (INIS)

Blanchard, A.

2000-01-01

The relocation impacts of an accidental release, scenario 1-RD-2.35, are evaluated for H-Separations. The extent of the area potentially contaminated to a level that would result in doses exceeding the relocation protective action guide (PAG) is calculated. The maximum calculated distance downwind from the accident at which the relocation PAG is exceeded is also determined. The consequences of the release are evaluated using the HOTSPOT model and an EXCEL spreadsheet

Process for recovering tritium from molten lithium metal

Science.gov (United States)

Maroni, Victor A.

1976-01-01

Lithium tritide (LiT) is extracted from molten lithium metal that has been exposed to neutron irradiation for breeding tritium within a thermonuclear or fission reactor. The extraction is performed by intimately contacting the molten lithium metal with a molten lithium salt, for instance, lithium chloride - potassium chloride eutectic to distribute LiT between the salt and metal phases. The extracted tritium is recovered in gaseous form from the molten salt phase by a subsequent electrolytic or oxidation step.

LIFE Materails: Molten-Salt Fuels Volume 8

Energy Technology Data Exchange (ETDEWEB)

Moir, R; Brown, N; Caro, A; Farmer, J; Halsey, W; Kaufman, L; Kramer, K; Latkowski, J; Powers, J; Shaw, H; Turchi, P

2008-12-11

The goals of the Laser Inertial Fusion Fission Energy (LIFE) is to use fusion neutrons to fission materials with no enrichment and minimum processing and have greatly reduced wastes that are not of interest to making weapons. Fusion yields expected to be achieved in NIF a few times per day are called for with a high reliable shot rate of about 15 per second. We have found that the version of LIFE using TRISO fuel discussed in other volumes of this series can be modified by replacing the molten-flibe-cooled TRISO fuel zone with a molten salt in which the same actinides present in the TRISO particles are dissolved in the molten salt. Molten salts have the advantage that they are not subject to radiation damage, and hence overcome the radiation damage effects that may limit the lifetime of solid fuels such as TRISO-containing pebbles. This molten salt is pumped through the LIFE blanket, out to a heat exchanger and back into the blanket. To mitigate corrosion, steel structures in contact with the molten salt would be plated with tungsten or nickel. The salt will be processed during operation to remove certain fission products (volatile and noble and semi-noble fission products), impurities and corrosion products. In this way neutron absorbers (fission products) are removed and neutronics performance of the molten salt is somewhat better than that of the TRISO fuel case owing to the reduced parasitic absorption. In addition, the production of Pu and rare-earth elements (REE) causes these elements to build up in the salt, and leads to a requirement for a process to remove the REE during operation to insure that the solubility of a mixed (Pu,REE)F3 solid solution is not exceeded anywhere in the molten salt system. Removal of the REE will further enhance the neutronics performance. With molten salt fuels, the plant would need to be safeguarded because materials of interest for weapons are produced and could potentially be removed.

LIFE Materails: Molten-Salt Fuels Volume 8

International Nuclear Information System (INIS)

Moir, R.; Brown, N.; Caro, A.; Farmer, J.; Halsey, W.; Kaufman, L.; Kramer, K.; Latkowski, J.; Powers, J.; Shaw, H.; Turchi, P.

2008-01-01

The goals of the Laser Inertial Fusion Fission Energy (LIFE) is to use fusion neutrons to fission materials with no enrichment and minimum processing and have greatly reduced wastes that are not of interest to making weapons. Fusion yields expected to be achieved in NIF a few times per day are called for with a high reliable shot rate of about 15 per second. We have found that the version of LIFE using TRISO fuel discussed in other volumes of this series can be modified by replacing the molten-flibe-cooled TRISO fuel zone with a molten salt in which the same actinides present in the TRISO particles are dissolved in the molten salt. Molten salts have the advantage that they are not subject to radiation damage, and hence overcome the radiation damage effects that may limit the lifetime of solid fuels such as TRISO-containing pebbles. This molten salt is pumped through the LIFE blanket, out to a heat exchanger and back into the blanket. To mitigate corrosion, steel structures in contact with the molten salt would be plated with tungsten or nickel. The salt will be processed during operation to remove certain fission products (volatile and noble and semi-noble fission products), impurities and corrosion products. In this way neutron absorbers (fission products) are removed and neutronics performance of the molten salt is somewhat better than that of the TRISO fuel case owing to the reduced parasitic absorption. In addition, the production of Pu and rare-earth elements (REE) causes these elements to build up in the salt, and leads to a requirement for a process to remove the REE during operation to insure that the solubility of a mixed (Pu,REE)F3 solid solution is not exceeded anywhere in the molten salt system. Removal of the REE will further enhance the neutronics performance. With molten salt fuels, the plant would need to be safeguarded because materials of interest for weapons are produced and could potentially be removed.

Mobility of partially molten crust, heat and mass transfer, and the stabilization of continents

Science.gov (United States)

Teyssier, Christian; Whitney, Donna L.; Rey, Patrice F.

2017-04-01

The core of orogens typically consists of migmatite terrains and associated crustal-derived granite bodies (typically leucogranite) that represent former partially molten crust. Metamorphic investigations indicate that migmatites crystallize at low pressure (cordierite stability) but also contain inclusions of refractory material (mafic, aluminous) that preserve evidence of crystallization at high pressure (HP), including HP granulite and eclogite (1.0-1.5 GPa), and in some cases ultrahigh pressure (2.5-3.0 GPa) when the continental crust was subducted (i.e. Norwegian Caledonides). These observations indicate that the partially molten crust originates in the deep crust or at mantle depths, traverses the entire orogenic crust, and crystallizes at shallow depth, in some cases at the near-surface ( 2 km depth) based on low-T thermochronology. Metamorphic assemblages generally show that this nearly isothermal decompression is rapid based on disequilibrium textures (symplectites). Therefore, the mobility of partially molten crust results in one of the most significant heat and mass transfer mechanisms in orogens. Field relations also indicate that emplacement of partially molten crust is the youngest major event in orogeny, and tectonic activity essentially ceases after the partially molten crust is exhumed. This suggests that flow and emplacement of partially molten crust stabilize the orogenic crust and signal the end of orogeny. Numerical modeling (open source software Underworld; Moresi et al., 2007, PEPI 163) provides useful insight into the mechanisms of exhumation of partially molten crust. For example, extension of thickened crust with T-dependent viscosity shows that extension of the shallow crust initially drives the mobility of the lowest viscosity crust (T>700°C), which begins to flow in a channel toward the zone of extension. This convergent flow generates channel collision and the formation of a double-dome of foliation (two subdomes separated by a steep

Event course analysis of core disruptive accidents

International Nuclear Information System (INIS)

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

1995-01-01

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

Molten-salt converter reactors

International Nuclear Information System (INIS)

Perry, A.M.

1975-01-01

Molten-salt reactors appear to have substantial promise as advanced converters. Conversion ratios of 0.85 to 0.9 should be attainable with favourable fuel cycle costs, with 235 U valued at $12/g. An increase in 235 U value by a factor of two or three ($10 to $30/lb. U 3 O 8 , $75/SWU) would be expected to increase the optimum conversion ratio, but this has not been analyzed in detail. The processing necessary to recover uranium from the fuel salt has been partially demonstrated in the MSRE. The equipment for doing this would be located at the reactor, and there would be no reliance on an established recycle industry. Processing costs are expected to be quite low, and fuel cycle optimization depends primarily on inventory and burnup or replacement costs for the fuel and for the carrier salt. Significant development problems remain to be resolved for molten-salt reactors, notably the control of tritium and the elimination of intergranular cracking of Hastelloy-N in contact with tellurium. However, these problems appear to be amenable to solution. It is appropriate to consider separating the development schedule for molten-salt reactors from that for the processing technology required for breeding. The Molten-Salt Converter Reactor should be a useful reactor in its own right and would be an advance towards the achievement of true breeding in thermal reactors. (author)

A heat transfer correlation based on a surface renewal model for molten core concrete interaction study

International Nuclear Information System (INIS)

Tourniaire, B. . E-mail bruno.tourniaire@cea.fr

2006-01-01

The prediction of heat transfer between corium pool and concrete basemat is of particular significance in the framework of the study of PWR's severe accident. Heat transfer directly governs the ablation velocity of concrete in case of molten core concrete interaction (MCCI) and, consequently, the time delay when the reactor cavity may fail. From a restricted hydrodynamic point of view, this issue is related to heat transfer between a heated bubbling pool and a porous wall with gas injection. Several experimental studies have been performed with simulant materials and many correlations have been provided to address this issue. The comparisons of the results of these correlations with the measurements and their extrapolation to reactor materials show that strong discrepancies between the results of these models are obtained which probably means that some phenomena are not well taken into account. The main purpose of this paper is to present an alternative heat transfer model which was originally developed for chemical engineering applications (bubble columns) by Deckwer. A part of this work is devoted to the presentation of this model, which is based on a surface renewal assumption. Comparison of the results of this model with available experimental data in different systems are presented and discussed. These comparisons clearly show that this model can be used to deal with the particular problem of MCCI. The analyses also lead to enrich the original model by taking into account the thermal resistance of the wall: a new formulation of the Deckwer's correlation is finally proposed

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.

A method of measuring a molten metal liquid pool volume

Science.gov (United States)

Garcia, G.V.; Carlson, N.M., Donaldson, A.D.

1990-12-12

A method of measuring a molten metal liquid pool volume and in particular molten titanium liquid pools, including the steps of (a) generating an ultrasonic wave at the surface of the molten metal liquid pool, (b) shining a light on the surface of a molten metal liquid pool, (c) detecting a change in the frequency of light, (d) detecting an ultrasonic wave echo at the surface of the molten metal liquid pool, and (e) computing the volume of the molten metal liquid. 3 figs.

Data forwarding mechanism for supporting real-time services during relocations in UMTS systems

Science.gov (United States)

Cai, Wei; Liao, Xianglong; Zheng, Liang; Liu, Zehong

2004-04-01

To minimize the interruption during the handovers or relocations invoked by subscribers moving is a very critical factor to enhance the performance of the UMTS systems. We know that the 2G systems have been optimized to minimize the interruption of speech during handovers by two main technologies: one is the bi-casting for the DL traffic and the other is the fast radio resynchronization by the UE for the UL traffic. In the UMTS systems, we have also implemented lossless relocations for non real-time services with high reliability by data buffering in the source RNC and target RNC for the UE. However, the UMTS systems support four QoS classes traffic flow: conversational class, streaming class, interactive class and background class. The main distinguishing factor between these QoS classes is how delay sensitive the traffic is: Conversational and Streaming classes are mainly used to carry real-time traffic flows, like video telephony, interactive and background classes are mainly used by traditional Internet applications like WWW, E-mail and FTP. It"s essential to provide the solutions for supporting real-time services to meet the requirement for QoS in UMTS systems. Apparently, the Data buffering mechanism is not adapted to real-time services because of it"s delay may exceed the basic requirement for real-time services. Under this background, the paper discussed two data forwarding solutions for real-time services from the PS domain in the UMTS systems: packet duplication and Core Network bi-casting. The former mechanism does not require any new procedures, messages nor information elements. The later mechanism requires that the GGSN or SGSN is able to bi-cast the DL traffic to the target RNC according to the relocations involving two SGSNs or just involving one SGSN. It also implicitly shows that we need change procedures at the nodes SGSN, GGSN and RNC which are involved in the relocation procedure based on existing procedures that we have already designed if

SCDAP/RELAP5 applications to RCS natural circulation

International Nuclear Information System (INIS)

Bayless, P.D.

1988-01-01

The effects of natural circulation flows in the reactor coolant system during a TMLB' sequence were investigated. Both in-vessel circulation and hot leg countercurrent flow were modeled in the Surry nuclear power plant using the SCDAP/RELAP5 computer code. The transient was analyzed until after fuel rod relocation had begun. The delays in the onset of relocation resulting from the natural circulation flows were not significant compared to SCDAP/RELAP5 calculations without natural circulation modeled, but were large compared to the analyses presented in NUREG-1150. The most significant aspect of the natural circulations flows was the heating of ex-vessel structures. Surge line failure is likely to occur before the vessel is breached by the molten core, while steam generator tube failure is not expected

Molten salt processes in special materials preparation

International Nuclear Information System (INIS)

Krishnamurthy, N.; Suri, A.K.

2013-01-01

As a class, molten salts are the largest collection of non aqueous inorganic solvents. On account of their stability at high temperature and compatibility to a number of process requirements, molten salts are considered indispensable to realize many of the numerous benefits of high temperature technology. They play a crucial role and form the basis for numerous elegant processes for the preparation of metals and materials. Molten salt are considered versatile heat transfer media and have led to the evolution of many interesting reactor concepts in fission and possibly in fusion. They also have been the basis of thinking for few novel processes for power generation. While focusing principally on the actual utilization of molten salts for a variety of materials preparation efforts in BARC, this lecture also covers a few of the other areas of technological applications together with the scientific basis for considering the molten salts in such situations. (author)

Studies on components for a molten salt reactor

International Nuclear Information System (INIS)

Nejedly, M.; Matal, O.

2003-01-01

The aim is contribute to a design of selected components of molten salt reactors with fuel in the molten fluoride salt matrix. Molten salt reactors (MSRs) permit the utilization of plutonium and minor actinides as new nuclear fuel from a traditional nuclear power station with production of electric energy. Results of preliminary feasibility studies of an intermediate heat exchanger, a small power molten salt pump and a modular conception of a steam generator for a demonstration unit of the MSR (30 MW) are summarized. (author)

Investigating thermal-hydraulic characteristic of molten fluoride salt in a circular pipe using a CFD methodology

International Nuclear Information System (INIS)

Chi Chenwei; Ferng Yuhming; Pei Baushei; Liang Jenqhorng

2011-01-01

In recent years, the molten salt reactor (MSR) has attracted increasing attention and become one of the most important 'Generation IV reactor' designs. In particular, the fact that molten fluoride salts are utilized as liquid fuel and coolant constitutes the main feature of the reactor. Furthermore, since the molten fluoride salt has a high Prandtl number and contains quite different behaviors to those of ordinary water and gas, an in-depth investigation of molten fluoride salt is thus highly demanded. Hence, it is the central objective of this study to examine the thermal-hydraulic characteristics of molten salt especially for the optimal design of reactor core and its safety operation. In this study, the dependence of pressure drop, Nusselt number and entrance length on the inlet Reynolds number for a molten fluoride salt (LiF(46.5)-NaF(11.5)-KF(42)) are computed using a comprehensive computational fluid dynamics (CFD) methodology. The methodology employs the continuity equation, momentum equation, energy equation, and standard k - ε turbulence model to conduct fluid dynamics simulation. For simplicity, the geometry employed in this study is a circular tube. The simulated results indicated that the pressure drop and Nusselt number and entrance length increase as the inlet Reynolds number increases. And the computed pressure drop corresponds well to theoretical value. It is also given a new correlation of computed entrance length in this paper. In addition, two well-known Nusselt number correlations such as, Hausen, Gnielinski, are employed to make comparisons with the computed results. It is also found that the computed Nusselt numbers overestimate the Hausen ones in the high Reynolds number region. However, the computed Nusselt numbers correspond well to the Gnielinski ones in all the Reynolds numbers region. Also notice that an experimental setup is currently in progress in order to validate the present CFD simulation. (author)

Concept of the demonstration molten salt unit for the transuranium elements transmutation

International Nuclear Information System (INIS)

Alekseev, P.; Dudnikov, A.; Prusakov, V.; Subbotin, S.; Zakirov, R.; Lelek, V.; Peka, I.

1999-01-01

In this report it is considered fluorine reprocessing of spent fuel and fluoride molten salt reactor in critical and subcritical modes for plutonium and minor actinides burning. International collaboration for creation of such system is proposed. It is without any doubt that additional neutron source in the core will have positive influence on the transmutation process in the reactor. On the other side there is a lot of problems to realize it technically and to ensure stable work of the whole complex. (Authors)

Waste treatment using molten salt oxidation

International Nuclear Information System (INIS)

Navratil, J.D.; Stewart, A.E.

1996-01-01

MSO technology can be characterized as a submerged oxidation process; the basic concept is to introduce air and wastes into a bed of molten salt, oxidize the organic wastes in the molten salt, use the heat of oxidation to keep the salt molten and remove the salt for disposal or processing and recycling. The molten salt (usually sodium carbonate at 900-1000 C) provides four waste management functions: providing a heat transfer medium, catalyzing the oxidation reaction, preventing the formation of acid gases by forming stable salts, and efficiently capturing ash particles and radioactive materials by the combined effects of wetting, encapsulation and dissolution. The MSO process requires no wet scrubbing system for off-gas treatment. The process has been developed through bench-scale and pilot-scale testing, with successful destruction demonstration of a wide variety of hazardous and mixed (radioactive and hazardous wastes). (author). 24 refs, 2 tabs, 2 figs

Molten fuel-moderator interaction

International Nuclear Information System (INIS)

Lee, J.H.S.; Kynstautas, R.

1987-02-01

A critical review of the current understanding of vapor explosions was carried out. It was concluded that, on the basis of actual industrial accidents and large scale experiments, energetic high yield steam explosion cannot be regarded as an improbable event if large quantities of molten fuel and coolant are mixed together. This study also reviewed a hydrodynamic transient model proposed by Henry and Fauske Associates to assess a molten fuel-moderator interaction event. It was found that the proposed model negates a priori the possibility of a violent event, by introducing two assumptions: 1) fine fragmentation of the molten fuel, and ii) rapid heat transfer from the fine fragments to form steam. Using the Hicks and Menzies thermodynamic model, maximum work potential and pressure rise in the calandria were estimated. However, it is recommended that a more representative upper bound model based on an underwater explosion of a pressurized volume of steam be developed

Feasibility study of passive gamma spectrometry of molten core material from Fukushima Daiichi Nuclear Power Station unit 1, 2, and 3 cores for special nuclear material accountancy - low-volatile FP and special nuclear material inventory analysis and fundamental characteristics of gamma-rays from fuel debris

International Nuclear Information System (INIS)

Sagara, Hiroshi; Tomikawa, Hirofumi; Watahiki, Masaru; Kuno, Yusuke

2014-01-01

The technologies applied to the analysis of the Three Mile Island accident were examined in a feasibility study of gamma spectrometry of molten core material from the Fukushima Daiichi Nuclear Power Station unit 1, 2, and 3 cores for special nuclear material accountancy. The focus is on low-volatile fission products and heavy metal inventory analysis, and the fundamental characteristics of gamma-rays from fuel debris with respect to passive measurements. The inventory ratios of the low-volatile lanthanides, "1"5"4Eu and "1"4"4Ce, to special nuclear materials were evaluated by the entire core inventories in units 1, 2, and 3 with an estimated uncertainty of 9%-13% at the 1σ level for homogenized molten fuel material. The uncertainty is expected to be larger locally owing to the use of the irradiation cycle averaging approach. The ratios were also evaluated as a function of burnup for specific fuel debris with an estimated uncertainty of 13%-25% at the 1σ level for units 1 and 2, and most of the fuels in unit 3, although the uncertainty regarding the separated mixed oxide fuel in unit 3 would be significantly higher owing to the burnup dependence approach. Source photon spectra were also examined and cooling-time-dependent data sets were prepared. The fundamental characteristics of high-energy gamma-rays from fuel debris were investigated by a bare-sphere model transport calculation. Mass attenuation coefficients of fuel debris were evaluated to be insensitive to its possible composition in a high-energy region. The leakage photon ratio was evaluated using a variety of parameters, and a significant impact was confirmed for a certain size of fuel debris. Its correlation was summarized with respect to the leakage photopeak ratio of source "1"5"4Eu. Finally, a preliminary study using a hypothetical canister model of fuel debris based on the experience at Three Mile Island was presented, and future plans were introduced. (author)

Metal Production by Molten Salt Electrolysis

DEFF Research Database (Denmark)

Grjotheim, K.; Kvande, H.; Qingfeng, Li

Chemistry and electrochemistry of molten salts are reviewed. Technological aspects of electrolytic production of aluminium, magnesium, and other metals are comprehensively surveyed.......Chemistry and electrochemistry of molten salts are reviewed. Technological aspects of electrolytic production of aluminium, magnesium, and other metals are comprehensively surveyed....

Thorium-based Molten Salt Reactor (TMSR) project in China

International Nuclear Information System (INIS)

Dai, Zhimin; Liu, Wei

2013-01-01

Making great efforts in development of nuclear energy is one of the long-term-plan in China's energy strategies. The advantages of Thorium-based nuclear energy are: rich resource in nature, less nuclear waste, low toxicity, nuclear non-proliferation and so on. Furthermore, China is a country with abundant thorium, thus it is necessary to develop the Thorium-based Molten Salt Reactor (TMSR) in China. Shanghai Institute of Applied Physics, Chinese Academy of Sciences (SINAP) had designed and constructed the first China's light-water reactor and developed a zero-power thorium-based molten salt reactor successfully in the early 1970s. The applied research project 'thorium molten salt reactor nuclear power system' by SINAP together with several other institutes had been accepted and granted by China government in 2011. The whole project has been divided into three stages: Firstly, built a 2 MW-zero-power high temperature solid molten salt reactor in 2015 and a 2 MW-zero-power high temperature liquid molten salt reactor in 2017. Secondly, in 2020 built a 10 MW high temperature liquid molten salt reactor. Thirdly, on the base of previous work, a 100 MW high temperature molten salt reactor should be achieving in 2030. After more than one years of efforts, a high quality scientific research team has been formed, which is able to design the molten salt reactor, the molten salt loop and related key equipment, the systems of molten salt preparation, purification and the radioactive gas removal. In the past one year, the initial physical design of high temperature molten salt reactor has been completed; the nuclear chemistry and radiation chemical laboratory has been built, a high temperature salt (HTS) loop and radioactive gas removal experiment device system have been successfully developed and constructed. Further, the preliminary study on reactor used carbon-carbon composite material has been investigated. (author)

Inertia-confining thermonuclear molten salt reactors

International Nuclear Information System (INIS)

Furukawa, Kazuo; Yamanaka, Chiyoe; Nakai, Sadao; Imon, Shunji; Nakajima, Hidenori; Nakamura, Norio; Kato, Yoshio.

1984-01-01

Purpose: To increase the heat generating efficiency while improving the reactor safety and thereby maintaining the energy balance throughout the reactor. Constitution: In an inertia-confining type D-T thermonuclear reactor, the blanket is made of lithium-containing fluoride molten salts (LiF.BeF 2 , LiF.NaF.KF, LiF.KF, etc) which are cascaded downwardly in a large thickness (50 - 100 cm) along the inner wall of the thermonuclear reaction vessel, and neutrons generated by explosive compression are absorbed to lithium in the molten salts to produce tritium, Heat transportation is carried out by the molten salts. (Ikeda, J.)

Thorium Molten Salt Nuclear Energy Synergetic System (THORIMS-NES)

International Nuclear Information System (INIS)

Yoshioka, Ritsuo; Mitachi, Koshi

2013-01-01

The authors have been promoting nuclear energy technology based on thorium molten salt as Thorium Molten Salt Nuclear Energy Synergetic System (THORIMS-NES). This system is a combination of fission power reactor of Molten Salt Reactor (MSR), and Accelerator Molten Salt Breeder (AMSB) for production of fissile 233 U with connecting chemical processing facility. In this paper, concept of THORIMS-NES, advantages of thorium and molten salt recent MSR design results such as FUJI-U3 using 233 U fuel, FUJI-Pu, large sized super-FUJI, pilot plant miniFUJI, AMSB, and chemical processing facility are described. (author)

Molten salt: Corrosion problems and electrometallurgy in nuclear applications

International Nuclear Information System (INIS)

Santarini, G.

1981-01-01

A bibliographic survey is given of corrosion problems and electrometallurgical problems of molten salt in nuclear reactor applications. Due to the high potential to be achieved, their high ionic conductivity and the rapidity of reactions in a molten salt atmosphere, molten salts are interesting solvents for various electrometallurgical processes. Another important field of application is in the separation or electrolytical refining of various metals (Be, U, Pu, Th, Hf, Zr). However, these very characteristics of molten salts may also cause serious corrosion problems. Results obtained for the molten-salt reactor and the different causes of corrosion are reviewed an possible countermeasures analyzed. (orig.)

A Heuristic Procedure for the Outbound Container Relocation Problem during Export Loading Operations

Directory of Open Access Journals (Sweden)

Roberto Guerra-Olivares

2015-01-01

Full Text Available During export ship loading operations, it is often necessary to perform relocation movements with containers that interfere with access to the desired container in the ship loading sequence. This paper presents a real-time heuristic procedure for the container relocation problem employing reachstacker vehicles as container handling equipment. The proposed heuristic searches for good relocation coordinates within a set of nearby bays. The heuristic has a parameter that determines how far from the original bay a container may be relocated. The tradeoff between reducing relocation movements and limiting vehicle travel distances is examined and the performance of the heuristic is compared with a common practice in the smaller container terminals in Chile and Mexico. Finally, a mathematical model for the container relocation problem is presented.

Ceramics for Molten Materials Transfer

Science.gov (United States)

Standish, Evan; Stefanescu, Doru M.; Curreri, Peter A.

2009-01-01

The paper reviews the main issues associated with molten materials transfer and handling on the lunar surface during the operation of a hig h temperature electrowinning cell used to produce oxygen, with molten iron and silicon as byproducts. A combination of existing technolog ies and purposely designed technologies show promise for lunar exploi tation. An important limitation that requires extensive investigation is the performance of refractory currently used for the purpose of m olten metal containment and transfer in the lunar environment associa ted with electrolytic cells. The principles of a laboratory scale uni t at a scale equivalent to the production of 1 metric ton of oxygen p er year are introduced. This implies a mass of molten materials to be transferred consistent with the equivalent of 1kg regolithlhr proces sed.

Development of a safety analysis code for molten salt reactors

International Nuclear Information System (INIS)

Zhang Dalin; Qiu Suizheng; Su Guanghui

2009-01-01

The molten salt reactor (MSR) well suited to fulfill the criteria defined by the Generation IV International Forum (GIF) is presently revisited all around the world because of different attractive features of current renewed relevance. The MSRs are characterized by using the fluid-fuel, so that their technologies are fundamentally different from those used in the conventional solid-fuel reactors. In this work, in particular, the attention is focused on the safety characteristic analysis of the MSRs, in which a point kinetic model considering the flow effects of the fuel salt is established for the MSRs and calculated by developing a microcomputer code coupling with a simplified heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the molten salt actinide recycler and transmuter system (MOSART) by simulating three types of basic transient conditions including the unprotected loss of flow, unprotected overcooling accident and unprotected transient overpower. Some reasonable results are obtained for the MOSART, which show that the MOSART conceptual design is an inherently stable reactor design. The present study provides some valuable information for the research and design of the new generation MSRs.

The Experiences and Challenges in Drilling into Semi molten or Molten Intrusive in Menengai Geothermal Field

Science.gov (United States)

Mortensen, A. K.; Mibei, G. K.

2017-12-01

Drilling in Menengai has experienced various challenges related to drilling operations and the resource itself i.e. quality discharge fluids vis a vis gas content. The main reason for these challenges is related to the nature of rocks encountered at depths. Intrusives encountered within Menengai geothermal field have been group into three based on their geological characteristics i.e. S1, S2 and S3.Detailed geology and mineralogical characterization have not been done on these intrusive types. However, based on physical appearances, S1 is considered as a diorite dike, S2 is syenite while S3 is molten rock material. This paper summarizes the experiences in drilling into semi molten or molten intrusive (S3).

New rational nuclear energy system composed of accelerator molten-salt breeder (AMSB) and molten-salt power stations (MSCR)

International Nuclear Information System (INIS)

Furukawa, K.

1985-01-01

For the next century, it was predicted that some rational fission energy system breeding in significantly short doubling time less than 10 years should be developed replacing the fossil fuels. In practice, this rationality, that is, simplicity and high economy could be realized by the natural combination of: molten salt fuel concept; accelerator (spallation) breeding concept; and Thorium fuel cycle concept, in the symbiont system of Accelerator Molten-Salt breeders and Molten-Salt Power Stations. The economy of this system might significantly become better than the other breeder systems, although the prediction in Chapter 6 was too much conservative. Its more important aspect is the low cost of future R and D, which depend on the rational character of Molten-Fluoride Technology and really is verified by the basic R and D cost (only $0.13 B) in Oak Ridge N.L. It is interesting that molten-salt technology will be able to apply to chemical processing of U-Pu oxide fuels by the developing effort by USSR in near future. This fact and the demand of small power stations such as 150MWe MSCR presented here will be able to bridge between the present and the next century

WASA-BOSS. Development and application of Severe Accident Codes. Evaluation and optimization of accident management measures. Subproject F. Contributions to code validation using BWR data and to evaluation and optimization of accident management measures. Final report; WASA-BOSS. Weiterentwicklung und Anwendung von Severe Accident Codes. Bewertung und Optimierung von Stoerfallmassnahmen. Teilprojekt F. Beitraege zur Codevalidierung anhand von SWR-Daten und zur Bewertung und Optimierung von Stoerfallmassnahmen. Schlussbericht

Energy Technology Data Exchange (ETDEWEB)

Di Marcello, Valentino; Imke, Uwe; Sanchez Espinoza, Victor

2016-09-15

The exact knowledge of the transient course of events and of the dominating processes during a severe accident in a nuclear power station is a mandatory requirement to elaborate strategies and measures to minimize the radiological consequences of core melt. Two typical experiments using boiling water reactor assemblies were modelled and simulated with the severe accident simulation code ATHLET-CD. The experiments are related to the early phase of core degradation in a boiling water reactor. The results reproduce the thermal behavior and the hydrogen production due to oxidation inside the bundle until relocation of material by melting. During flooding of the overheated assembly temperatures and hydrogen oxidation are under estimated. The deviations from the experimental results can be explained by the missing model to simulate bore carbide oxidation of the control rods. On basis of a hypothetical loss of coolant accident in a typical German boiling water reactor the effectivity of flooding the partial degraded core is investigated. This measure of mitigation is efficient and prevents failure of the reactor pressure vessel if it starts before molten material is relocated into the lower plenum. Considerable amount of hydrogen is produced by oxidation of the metallic components.

WASA-BOSS. Development and application of Severe Accident Codes. Evaluation and optimization of accident management measures. Subproject F. Contributions to code validation using BWR data and to evaluation and optimization of accident management measures. Final report

International Nuclear Information System (INIS)

Di Marcello, Valentino; Imke, Uwe; Sanchez Espinoza, Victor

2016-09-01

The exact knowledge of the transient course of events and of the dominating processes during a severe accident in a nuclear power station is a mandatory requirement to elaborate strategies and measures to minimize the radiological consequences of core melt. Two typical experiments using boiling water reactor assemblies were modelled and simulated with the severe accident simulation code ATHLET-CD. The experiments are related to the early phase of core degradation in a boiling water reactor. The results reproduce the thermal behavior and the hydrogen production due to oxidation inside the bundle until relocation of material by melting. During flooding of the overheated assembly temperatures and hydrogen oxidation are under estimated. The deviations from the experimental results can be explained by the missing model to simulate bore carbide oxidation of the control rods. On basis of a hypothetical loss of coolant accident in a typical German boiling water reactor the effectivity of flooding the partial degraded core is investigated. This measure of mitigation is efficient and prevents failure of the reactor pressure vessel if it starts before molten material is relocated into the lower plenum. Considerable amount of hydrogen is produced by oxidation of the metallic components.

Transfer characteristics of a lithium chloride–potassium chloride molten salt

Directory of Open Access Journals (Sweden)

Eve Mullen

2017-12-01

Full Text Available Pyroprocessing is an alternative method of reprocessing spent fuel, usually involving the dissolving spent fuel in a molten salt media. The National Nuclear Laboratory designed, built, and commissioned a molten salt dynamics rig to investigate the transfer characteristics of molten lithium chloride–potassium chloride eutectic salt. The efficacy and flow characteristics of a high-temperature centrifugal pump and argon gas lift were obtained for pumping the molten salt at temperatures up to 500°C. The rig design proved suitable on an industrial scale and transfer methods appropriate for use in future molten salt systems. Corrosion within the rig was managed, and melting techniques were optimized to reduce stresses on the rig. The results obtained improve the understanding of molten salt transport dynamics, materials, and engineering design issues and support the industrialization of molten salts pyroprocessing.

Compatibility studies of potential molten-salt breeder reactor materials in molten fluoride salts

International Nuclear Information System (INIS)

Keiser, J.R.

1977-05-01

The molten fluoride salt compatibility studies carried out during the period 1974--76 in support of the Molten-Salt Reactor Program are summarized. Thermal-convection and forced-circulation loops were used to measure the corrosion rate of selected alloys. Results confirmed the relationship of time, initial chromium concentration, and mass loss developed by previous workers. The corrosion rates of Hastelloy N and Hastelloy N modified by the addition of 1--3 wt percent Nb were well within the acceptable range for use in an MSBR. 13 figures, 3 tables

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Heat and Fission Product Transport in a Molten U-Zr-O Pool With Crust

International Nuclear Information System (INIS)

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

2002-01-01

Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the pool. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products in the pool was estimated by product of the mass concentration and energy conversion factor of each fission product. For the calculation of heat generation rate in the pool, twenty-nine elements were chosen and classified by their chemical properties. The mass concentration of a fission product is obtained from released fraction and the tabular output of the ORIGEN 2 code. The initial core and pool inventories at each time can also be estimated using ORIGEN 2. The released fraction of each fission product is calculated based on the bubble dynamics and mass transport. Numerical analysis was performed for the TMI-2 accident. The pool is assumed to be a partially filled hemispherical geometry and the change of pool geometry during the numerical calculation was neglected. Results of the numerical calculation revealed that the peak temperature of the molten pool significantly decreased and most of the volatile fission products were released from the molten pool during the accident. (authors)

Simulant melt experiments on performance of the in-vessel core catcher

International Nuclear Information System (INIS)

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

2007-01-01

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

Electrochemistry of plutonium in molten halides

International Nuclear Information System (INIS)

McCurry, L.E.; Moy, G.M.M.; Bowersox, D.F.

1987-01-01

The electrochemistry of plutonium in molten halides is of technological importance as a method of purification of plutonium. Previous authors have reported that plutonium can be purified by electrorefining impure plutonium in various molten haldies. Work to eluciate the mechanism of the plutonium reduction in molten halides has been limited to a chronopotentiometric study in LiCl-KCl. Potentiometric studies have been carried out to determine the standard reduction potential for the plutonium (III) couple in various molten alkali metal halides. Initial cyclic voltammetric experiments were performed in molten KCL at 1100 K. A silver/silver chloride (10 mole %) in equimolar NaCl-KCl was used as a reference electrode. Working and counter electrodes were tungsten. The cell components and melt were contained in a quartz crucible. Background cyclic voltammograms of the KCl melt at the tungsten electrode showed no evidence of electroactive impurities in the melt. Plutonium was added to the melt as PuCl/sub 3/, which was prepared by chlorination of the oxide. At low concentrations of PuCl/sub 3/ in the melt (0.01-0.03 molar), no reduction wave due to the reduction of Pu(III) was observed in the voltammograms up to the potassium reduction limit of the melt. However on scan reversal after scanning into the potassium reduction limit a new oxidation wave was observed

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-09-15

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

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.

Means for Transferring Knowledge in the Relocation of Manufacturing Units

DEFF Research Database (Denmark)

Yang, Cheng; Madsen, Erik Skov; Liangsiri, Jirapha

2009-01-01

The global spread of production makes companies relocating their manufacturing units to achieve economies of scale, enjoy low-cost labor, or access to new markets. For the relocation, not only equipments, systems and facilities, need to be moved, but also operational knowledge and experience. Based...

Quotas, Relocation and Jobs

Science.gov (United States)

Blumrosen, Alfred

1975-01-01

This testimony, before a public hearing of the New York City Commission on Human Rights in May 1974, has two express focuses: one is the question of quotas, and the other is the problem of relocation of plants and facilities in the suburban areas and the consequent difficulties minorities face in retaining equal employment opportunities.…

Infectious Disease Prevalence and Factors Associated with Upper Respiratory Infection in Cats Following Relocation

Directory of Open Access Journals (Sweden)

Mehnaz Aziz

2018-06-01

Full Text Available Feline relocation is used increasingly in animal welfare to decrease shelter euthanasia rates and increase positive outcomes. Concerns about infectious disease introduction and transmission are often expressed; however, little research has been conducted on even the baseline prevalence of infectious disease following relocation. This study, which collected data on 430 cats relocated through an established program over 7 months, evaluated the prevalence of upper respiratory infection (URI, feline panleukopenia virus (FPV and dermatophytosis at one destination agency. The period prevalence was 25.8% for URI, 1.6% for FPV and 0.9% for dermatophytosis. Mixed-effects logistic regression was performed to investigate factors associated with URI. Younger age, increased time in transport, and increased length of stay at the destination agency were associated with increased URI prevalence following relocation. The findings of this study reveal that certain highly contagious and environmentally persistent infectious diseases, such as FPV and dermatophytosis, are uncommon following relocation in an established program; however, URI in relocated cats should be proactively managed. Animal welfare agencies can use this information to guide shelter and relocation operations and mitigate the impact of URI in relocated cats.

Proposals on the organization of a fuel cycle of the cascade sub-critical molten salt reactor (CSMSR)

International Nuclear Information System (INIS)

Bychkov, A.V.; Kormilitsyn, M.V.; Melnik, M.I.; Babikov, L.G.; Ponomarev, L.I.

2002-01-01

At present the approach of burning out long-lived radioactive waste (RW) in the reactor core neutron flux is the most feasible one. Currently the way of closing nuclear fuel cycle (NFC) on the basis of the nuclear chemical concept of the cascade sub-critical molten salt reactor (CSMSR) is considered as the most promising one. It is characterised by a number of advantages. CSMSR controlled by a beam of protons or electrons is the optimal reactor for closing the NFC using non-aqueous fluoride methods of fuel reprocessing. They, in comparison with aqueous methods, are characterised by a small waste quantity and are less laborious because of the absence of severe requirements to the product purity. A high productivity of high-temperature electrochemical processes allows the implementation of the fuel recycling process as part of the CSMSR total technological cycle. It can be conducted in the 'on-line' mode in the bypass molten salt circuit that brings the transportation volume of high-activity materials to a minimum. In order to reprocess the CSMSR irradiated molten salt fuel on the basis of salt composition LiF-NaF-(BeF 2 ) an option, based on the following three main operations of the melt treatment, was proposed at SSC RF RIAR: (i) On-line argon treatment of molten salt fuel for removal of gaseous fission products (FP) and also FP that form volatile fluorides and aerosols; (ii) Organisation of the fuel-active metal (probably with a fine-dispersed plutonium alloy) interaction in the on-line mode for removal of 'noble' and 'semi-noble' FP and corrosion products such as Ni, Fe, Cr (when using Pu alloy it allows to regenerate at the same time of the burned-out plutonium component); (iii) Portion-by-portion (fuel composition partially being removed from the CSMSR molten salt circuit) pyroelectrochemical reprocessing of the molten salt composition aimed at the removal of lanthanides - FP followed by a return of actinides to the CSMSR fuel cycle. This technology will allow

Environmental assessment for relocation of NREL research experiments

International Nuclear Information System (INIS)

1992-06-01

In compliance with the National Environmental Policy Act (NEPA), the Department of Energy (DOE) has prepared this environmental assessment (DOE/EA-0619) to evaluate the environmental consequences associated with the conduct of ongoing research activities of its National Renewable Energy Laboratory proposed to be relocated to leased commercial laboratory and warehouse space at 6800 Joyce Sum, in Arvada, Colorado. NREL is currently leasing space in Golden, Colorado, for conduct of the research actions discussed in the EA. The research project proposed for relocation is the Amorphous silicon Deposition Laboratory (ASDL). Additionally, it may be financially desirable to relocate the Scanning Hartman Optical Tester (SHOT) and the Whole Building Test Facility at a later date, therefore, the consequences of their operation at the proposed facility is evaluated in the EA to support such future decisions. The new location can also provide additional warehouse space required by NREL

Knowledge Transfer and Manufacturing Relocation in International Manufacturing Networks

DEFF Research Database (Denmark)

Madsen, Erik Skov

2014-01-01

This paper is built on six longitudinal case studies of knowledge transfer in manufacturing relocation. By focusing on tacit and explicit knowledge the paper introduces a model for identification of knowledge in relation to four task situations on the shop floor in a manufacturing environment...... after relocation. Finally the paper discusses how “dispatching capacity” and “absorptive capacity” can improve the process....

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-07-01

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

Molten salt engineering for thorium cycle. Electrochemical studies as examples

International Nuclear Information System (INIS)

Ito, Yasuhiko

1998-01-01

A Th-U nuclear energy system utilizing accelerator driven subcritical molten salt breeder reactor has several advantages compared to conventional U-Pu nuclear system. In order to obtain fundamental data on molten salt engineering of Th-U system, electrochemical study was conducted. As the most primitive simulated study of beam irradiation of molten salt, discharge electrolysis was investigated in molten LiCl-KCl-AgCl system. Stationary discharge was generated under atmospheric argon gas and fine Ag particles were obtained. Hydride ion (H - ) behavior in molten salts was also studied to predict the behavior of tritide ion (T - ) in molten salt fuel. Finally, hydrogen behavior in metals at high temperature was investigated by electrochemical method, which is considered to be important to confine and control tritium. (author)

Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

Energy Technology Data Exchange (ETDEWEB)

Laureau, A., E-mail: laureau.axel@gmail.com; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

2017-05-15

Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

International Nuclear Information System (INIS)

Laureau, A.; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

2017-01-01

Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

Coupled neutronics/thermal-hydraulics for analysis of molten salt reactor

International Nuclear Information System (INIS)

Guo, Zhangpeng; Zhou, Jianjun; Zhang, Dalin; Chaudri, Khurrum Saleem; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

2013-01-01

Highlights: ► A multiple-channel analysis code (MAC) is developed to be coupled with MCNP. ► 1/8 of core is simulated in MCNP and thermal-hydraulic code. ► The coupling calculation can achieve stable state after a few iterations. ► The coupling calculation results are in reasonable agreement with the analytic solutions of the ORNL. ► Parametric studies of MSR are performed to provide valuable information for future design MSR. -- Abstract: The Generation IV International Forum (GIF) selected molten salt reactor (MSR) among six advanced reactor types. It is characterized by a liquid circulating fuel that also serves as coolant. In this study, a multiple-channel analysis code (MAC) is developed and it is coupled with MCNP4c to analyze the neutronics/thermal-hydraulics behavior of molten salt reactor experiment (MSRE). The MAC calculates thermal-hydraulic parameters, such as temperature distribution, flow distribution and pressure drop. MCNP4c performs the analysis of effective multiplication factor, neutron flux and power distribution. A linkage code is developed to exchange data between MAC and MCNP to implement coupling iteration process until the power convergence is achieved. The coupling calculation can achieve converged solution after a few iterations. The results are in reasonable agreement with the analytic solutions from the ORNL. For further design analysis, parametric studies are performed to provide valuable information for new design of MSR. The effect of inlet temperature, graphite to molten salt volume ratio (G/Ms) from varying channel diameter and different power levels on the effective multiplication factor, neutron flux, graphite lifetime and temperature distribution are discussed in detail

BNL program in support of LWR degraded-core accident analysis

International Nuclear Information System (INIS)

Ginsberg, T.; Greene, G.A.

1982-01-01

Two major sources of loading on dry watr reactor containments are steam generatin from core debris water thermal interactions and molten core-concrete interactions. Experiments are in progress at BNL in support of analytical model development related to aspects of the above containment loading mechanisms. The work supports development and evaluation of the CORCON (Muir, 1981) and MARCH (Wooton, 1980) computer codes. Progress in the two programs is described in this paper. 8 figures

Feet sunk in molten aluminium: The burn and its prevention.

Science.gov (United States)

Alonso-Peña, David; Arnáiz-García, María Elena; Valero-Gasalla, Javier Luis; Arnáiz-García, Ana María; Campillo-Campaña, Ramón; Alonso-Peña, Javier; González-Santos, Jose María; Fernández-Díaz, Alaska Leonor; Arnáiz, Javier

2015-08-01

Nowadays, despite improvements in safety rules and inspections in the metal industry, foundry workers are not free from burn accidents. Injuries caused by molten metals include burns secondary to molten iron, aluminium, zinc, copper, brass, bronze, manganese, lead and steel. Molten aluminium is one of the most common causative agents of burns (60%); however, only a few publications exist concerning injuries from molten aluminium. The main mechanisms of lesion from molten aluminium include direct contact of the molten metal with the skin or through safety apparel, or when the metal splash burns through the pants and rolls downward along the leg. Herein, we report three cases of deep dermal burns after 'soaking' the foot in liquid aluminium and its evolutive features. This paper aims to show our experience in the management of burns due to molten aluminium. We describe the current management principles and the key features of injury prevention. Copyright © 2014 Elsevier Ltd and ISBI. All rights reserved.

Fragmentation of molten copper drop caused by entrapment of liquid sodium

International Nuclear Information System (INIS)

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

2001-01-01

In core meltdown accidents, it is possible to occur thermal interactions between molten fuel and coolant. Analysis of the steam explosion, which is one of the most severe phenomena in such thermal interactions, is important for the safety evaluation. The steam explosion is a phenomenon that intensive pressure waves are caused by the explosive thermal interaction between high and low temperature liquids, and is considered to be one of the phenomena that can cause a serious failure of the nuclear reactor structures. In a large-scale steam explosion, the fragmentation of hot molten material causes a rapid increase of heat transfer area, and it is achieved to transmit instantaneously a large amount of heat to coolant. Two ideas are chiefly considered as the mechanism of the fragmentation. The one is the hypothesis that hydrodynamic effect causes fragmentation of hot liquid. According to this hypothesis, the high temperature drops flake off from the surface. The other is that fragmentation is caused by the interface instability accompanied by collapse of the steam bubble formed around a hot liquid. In this research, the possibility of the internal fragmentation caused by the coolant jet is focused in. Experiments were conducted on the condition that the surface of melt drops solidify at the moment drops contact the coolant. The possibility of the fragmentation of hot liquid from its surface was eliminated in this condition. To satisfy this condition, molten copper was chosen as hot liquid, and liquid sodium was used as coolant to verify the effect of the driving force of the sodium jet. (author)

On-line reprocessing of a molten salt reactor: a simulation tool

International Nuclear Information System (INIS)

Simon, Nicole; Gastaldi, Olivier; Penit, Thomas; Cohin, Olivier; Campion, Pierre-Yves

2008-01-01

The molten salt reactor (MSR) is one of the concepts studied in the frame of GEN IV road-map. Due to the specific features of its liquid fuel, the reprocessing unit may be directly connected to the reactor. A modelling of this unit is presented. The final objective is to create a flexible computer reprocessing code which can use data from neutron calculations and can be coupled to a neutron code. Such a code allows the description of the whole behaviour of MSR, including, in a coupled manner, both the design of the core and the optimised reprocessing scheme effects. (authors)

Selective Adsorption of Sodium Aluminum Fluoride Salts from Molten Aluminum

Energy Technology Data Exchange (ETDEWEB)

Leonard S. Aubrey; Christine A. Boyle; Eddie M. Williams; David H. DeYoung; Dawid D. Smith; Feng Chi

2007-08-16

Aluminum is produced in electrolytic reduction cells where alumina feedstock is dissolved in molten cryolite (sodium aluminum fluoride) along with aluminum and calcium fluorides. The dissolved alumina is then reduced by electrolysis and the molten aluminum separates to the bottom of the cell. The reduction cell is periodically tapped to remove the molten aluminum. During the tapping process, some of the molten electrolyte (commonly referred as “bath” in the aluminum industry) is carried over with the molten aluminum and into the transfer crucible. The carryover of molten bath into the holding furnace can create significant operational problems in aluminum cast houses. Bath carryover can result in several problems. The most troublesome problem is sodium and calcium pickup in magnesium-bearing alloys. Magnesium alloying additions can result in Mg-Na and Mg-Ca exchange reactions with the molten bath, which results in the undesirable pickup of elemental sodium and calcium. This final report presents the findings of a project to evaluate removal of molten bath using a new and novel micro-porous filter media. The theory of selective adsorption or removal is based on interfacial surface energy differences of molten aluminum and bath on the micro-porous filter structure. This report describes the theory of the selective adsorption-filtration process, the development of suitable micro-porous filter media, and the operational results obtained with a micro-porous bed filtration system. The micro-porous filter media was found to very effectively remove molten sodium aluminum fluoride bath by the selective adsorption-filtration mechanism.

24 CFR 1003.209 - Prohibition on use of assistance for employment relocation activities.

Science.gov (United States)

2010-04-01

... employment relocation activities. (a) Prohibition. ICDBG funds may not be used to directly assist a business, including a business expansion, in the relocation of a plant, facility, or operation from one Identified... business to undertake one or more of these activities as a condition of the business relocating a facility...

20 CFR 627.215 - Relocation.

Science.gov (United States)

2010-04-01

... the original location. (b) For 120 days after the commencement or the expansion of commercial... original location. (c) For the purposes of this section, relocating establishment means a business entity... review should include names under which the establishment does business, including successors-in-interest...

Breakup Behavior of Molten Wood's Metal Jet in Subcooled Water

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

There are safety characteristics of the metal fueled sodium fast-cooled reactor (SFR), by identifying the possibility of early termination of severe accidents. If the molten fuel is ejected from the cladding, the ejected molten fuel can interact with the coolant in the reactor vessel. This phenomenon is called as fuel-coolant interaction (FCI). The FCI occurs at the initial phase leading to severe accidents like core disruptive accident (CDA) in the SFR. A part of the corium energy is intensively transferred to the coolant in a very short time during the FCI. The coolant vaporizes at high pressure and expands so results in steam explosion that can threat to the integrity of nuclear reactor. The intensity of steam explosion is determined by jet breakup and the fragmentation behavior. Therefore, it is necessary to understand the jet breakup between the molten fuel jet and the coolant in order to evaluate whether the steam explosion occurs or not. The liquid jet breakup has been studied in various areas, such as aerosols, spray and combustion. In early studies, small diameter jets of low density liquids were studied. The jet breakup for large density liquids has been studied in nuclear reactor field with respect to safety. The existence of vapor film layer between the melt and liquid fluid is only in case of large density breakup. This paper deals with the jet breakup experiment in non-boiling conditions in order to analyze hydraulic effect on the jet behavior. In the present study, the wood's metal was used as the jet material. It has similar properties to the metal fuel. The physical properties of molten materials and coolants are listed in Table I, respectively. It is easy to conduct the experiment due to low melting point of the wood's metal. In order to clarify the dominant factors determining jet breakup and size distribution of the debris, the experiment that the molten wood's metal was injected into the subcooled condition was conducted. The

Nonlinear core deflection in injection molding

Science.gov (United States)

Poungthong, P.; Giacomin, A. J.; Saengow, C.; Kolitawong, C.; Liao, H.-C.; Tseng, S.-C.

2018-05-01

Injection molding of thin slender parts is often complicated by core deflection. This deflection is caused by molten plastics race tracking through the slit between the core and the rigid cavity wall. The pressure of this liquid exerts a lateral force of the slender core causing the core to bend, and this bending is governed by a nonlinear fifth order ordinary differential equation for the deflection that is not directly in the position along the core. Here we subject this differential equation to 6 sets of boundary conditions, corresponding to 6 commercial core constraints. For each such set of boundary conditions, we develop an explicit approximate analytical solution, including both a linear term and a nonlinear term. By comparison with finite difference solutions, we find our new analytical solutions to be accurate. We then use these solutions to derive explicit analytical approximations for maximum deflections and for the core position of these maximum deflections. Our experiments on the base-gated free-tip boundary condition agree closely with our new explicit approximate analytical solution.

Divided parents, shared children
Conflicting approaches to relocation disputes in the USA

Directory of Open Access Journals (Sweden)

Theresa Glennon

2008-06-01

Full Text Available Custody relocation disputes pose intractable dilemmas for courts in a highly mobile society. The custodial parent, most often a woman, seeks self-determination, freedom of movement and a continued custodial relationship with the child. The non-custodial parent seeks to preserve a geographically close relationship with the child. Courts must identify the best interests of the children amidst these multiple and conflicting interests. They make decisions that may determine the course of custodial parents’ lives, affecting remarriage, employment, education, and proximity to family. A narrow doctrinal focus on children’s best interests ignores these key aspects of relocation disputes. This article examines the varied legislative and judicial approaches to relocation disputes in the US and proposed principles for resolution of these disputes. It reviews scholarship analyzing relocation disputes from a wide range of perspectives, including: conflicting social science research; competing ideologies of the post-divorce family; alternative dispute resolution; parents’ constitutional rights; domestic violence victims; and proposals to eliminate geographic presumptions and remedy the economic effect of restraints on relocation. Relocation doctrine in the US should be realigned to address these complex perspectives. States should also assist post-divorce families to support children through the common experience of relocation.

Experimental studies of oxidic molten corium-vessel steel interaction

International Nuclear Information System (INIS)

Bechta, S.V.; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Lopukh, D.B.; Petrov, Yu.B.; Petchenkov, A.Yu.; Kulagin, I.V.; Granovsky, V.S.; Kovtunova, S.V.; Martinov, V.V.; Gusarov, V.V.

2001-01-01

The experimental results of molten corium-steel specimen interaction with molten corium on the 'Rasplav-2' test facility are presented. In the experiments, cooled vessel steel specimens positioned on the molten pool bottom and uncooled ones lowered into the molten pool were tested. Interaction processes were studied for different corium compositions, melt superheating and in alternative (inert and air) overlying atmosphere. Hypotheses were put forward explaining the observed phenomena and interaction mechanisms. The studies presented in the paper were aimed at the detection of different corium-steel interaction mechanisms. Therefore certain identified phenomena are more typical of the ex-vessel localization conditions than of the in-vessel corium retention. Primarily, this can be referred to the phenomena of low-temperature molten corium-vessel steel interaction in oxidizing atmosphere

Experimental studies of oxidic molten corium-vessel steel interaction

Energy Technology Data Exchange (ETDEWEB)

Bechta, S.V. E-mail: niti-npc@sbor.net; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Lopukh, D.B.; Petrov, Yu.B.; Petchenkov, A.Yu.; Kulagin, I.V.; Granovsky, V.S.; Kovtunova, S.V.; Martinov, V.V.; Gusarov, V.V

2001-12-01

The experimental results of molten corium-steel specimen interaction with molten corium on the 'Rasplav-2' test facility are presented. In the experiments, cooled vessel steel specimens positioned on the molten pool bottom and uncooled ones lowered into the molten pool were tested. Interaction processes were studied for different corium compositions, melt superheating and in alternative (inert and air) overlying atmosphere. Hypotheses were put forward explaining the observed phenomena and interaction mechanisms. The studies presented in the paper were aimed at the detection of different corium-steel interaction mechanisms. Therefore certain identified phenomena are more typical of the ex-vessel localization conditions than of the in-vessel corium retention. Primarily, this can be referred to the phenomena of low-temperature molten corium-vessel steel interaction in oxidizing atmosphere.

Molten salt oxidation of organic hazardous waste with high salt content.

Science.gov (United States)

Lin, Chengqian; Chi, Yong; Jin, Yuqi; Jiang, Xuguang; Buekens, Alfons; Zhang, Qi; Chen, Jian

2018-02-01

Organic hazardous waste often contains some salt, owing to the widespread use of alkali salts during industrial manufacturing processes. These salts cause complications during the treatment of this type of waste. Molten salt oxidation is a flameless, robust thermal process, with inherent capability of destroying the organic constituents of wastes, while retaining the inorganic ingredients in the molten salt. In the present study, molten salt oxidation is employed for treating a typical organic hazardous waste with a high content of alkali salts. The hazardous waste derives from the production of thiotriazinone. Molten salt oxidation experiments have been conducted using a lab-scale molten salt oxidation reactor, and the emissions of CO, NO, SO 2 , HCl and dioxins are studied. Impacts are investigated from the composition of the molten salts, the types of feeding tube, the temperature of molten carbonates and the air factor. Results show that the waste can be oxidised effectively in a molten salt bath. Temperature of molten carbonates plays the most important role. With the temperature rising from 600 °C to 750 °C, the oxidation efficiency increases from 91.1% to 98.3%. Compared with the temperature, air factor has but a minor effect, as well as the composition of the molten salts and the type of feeding tube. The molten carbonates retain chlorine with an efficiency higher than 99.9% and the emissions of dioxins are below 8 pg TEQ g -1 sample. The present study shows that molten salt oxidation is a promising alternative for the disposal of organic hazardous wastes containing a high salt content.

24 CFR 570.210 - Prohibition on use of assistance for employment relocation activities.

Science.gov (United States)

2010-04-01

... relocation activities. (a) Prohibition. CDBG funds may not be used to directly assist a business, including a business expansion, in the relocation of a plant, facility, or operation from one LMA to another LMA if the... business to undertake one or more of these activities as a condition of the business relocating a facility...

Dynamics of the Molten Contact Line

Science.gov (United States)

Sonin, Ain A.; Duthaler, Gregg; Liu, Michael; Torresola, Javier; Qiu, Taiqing

1999-01-01

The purpose of this program is to develop a basic understanding of how a molten material front spreads over a solid that is below its melting point, arrests, and freezes. Our hope is that the work will contribute toward a scientific knowledge base for certain new applications involving molten droplet deposition, including the "printing" of arbitrary three-dimensional objects by precise deposition of individual molten microdrops that solidify after impact. Little information is available at this time on the capillarity-driven motion and arrest of molten contact line regions. Schiaffino and Sonin investigated the arrest of the contact line of a molten microcrystalline wax spreading over a subcooled solid "target" of the same material. They found that contact line arrest takes place at an apparent liquid contact angle that depends primarily on the Stefan number S=c(T(sub f) -T(sub t)/L based on the temperature difference between the fusion point and the target temperature, and proposed that contact line arrest occurs when the liquid's dynamic contact angle approaches the angle of attack of the solidification front just behind the contact line. They also showed, however, that the conventional continuum equations and boundary conditions have no meaningful solution for this angle. The solidification front angle is determined by the heat flux just behind the contact line, and the heat flux is singular at that point. By comparing experiments with numerical computations, Schiaffino and Sonin estimated that the conventional solidification model must break down within a distance of order 0.1 - 1 microns of the contact line. The physical mechanism for this breakdown is as yet undetermined, and no first-principles theory exists for the contact angle at arrest. Schiaffino and Sonin also presented a framework for understanding how to moderate Weber number molten droplet deposition in terms of similarity laws and experimentation. The study is based on experiments with three molten

Fission product removal from molten salt using zeolite

International Nuclear Information System (INIS)

Pereira, C.; Babcock, B.D.

1996-01-01

Spent nuclear fuel (SNF) can be treated in a molten salt electrorefiner for conversion into metal and mineral waste forms for geologic disposal. The fuel is dissolved in molten chloride salt. Non-transuranic fission products in the molten salt are ion-exchanged into zeolite A, which is subsequently mixed with glass and consolidated. Zeolite was found to be effective in removing fission product cations from the molten salt. Breakthrough of cesium and the alkaline earths occurred more rapidly than was observed for the rare earths. The effluent composition as a function of time is presented, as well as results for the distribution of fission products along the length of the column. Effects of temperature and salt flow rate are also discussed

Materials interaction tests to identify base and coating materials for an enhanced in-vessel core catcher design

Energy Technology Data Exchange (ETDEWEB)

Rempe, J.L.; Knudson, D.L.; Condie, K.G.; Swank, W.D. [Idaho National Engineering and Environmental Laboratory, Idaho Falls ID (United States); Cheung, F.B. [Pennsylvania State University, Department of Mechanical and Nuclear Engineering, University Park PA (United States); Suh, K.Y. [Seoul National University, Department of Nuclear Engineering, Seoul (Korea, Republic of); Kim, S.B. [Korea Atomic Energy Research Institute, Severe Accident Research Project, Taejon (Korea, Republic of)

2004-07-01

An enhanced in-vessel core catcher is being designed and evaluated, it must ensure In-Vessel Retention of core materials that may relocate under severe accident conditions in advanced reactors. To reduce cost and simplify manufacture and installation, this new core catcher design consists of several interlocking sections that are machined to fit together when inserted into the lower head. If needed, the core catcher can be manufactured with holes to accommodate lower head penetrations. Each section of the core catcher consists of two material layers with an option to add a third layer (if deemed necessary): a base material, which has the capability to support and contain the mass of core materials that may relocate during a severe accident; an insulating oxide coating material on top of the base material, which resists interactions with high-temperature core materials; and an optional coating on the bottom side of the base material to prevent any potential oxidation of the base material during the lifetime of the reactor. Initial evaluations suggest that a thermally-sprayed oxide material is the most promising candidate insulator coating for a core catcher. Tests suggest that 2 coatings can provide adequate protection to a stainless steel core catcher: -) a 500 {mu}m thick zirconium dioxide coating over a 100-200 {mu}m Inconel 718 bond coating, and -) a 500 {mu}m thick magnesium zirconate coating.

EXPEL - a computing module for molten fuel/coolant interactions in fast reactor sub-assemblies

International Nuclear Information System (INIS)

Fishlock, T.P.

1975-10-01

This report describes a module for computing the effects of a molten fuel/coolant interaction in a fast reactor subassembly. The module is to be incorporated into the FRAX code which calculates the consequences of hypothetical whole core accidents. Details of the interaction are unknown and in consequence the model contains a large number of parameters which must be set by assumption. By variation of these parameters the interaction may be made mild or explosive. Results of a parametric survey are included. (author)

Managing water addition to a degraded core

International Nuclear Information System (INIS)

Kuan, P.; Hanson, D.J.; Odar, F.

1992-01-01

In this paper the authors present information that can be used in severe accident management by providing an improved understanding of the effects of water addition to a degraded core. This improved understanding is developed using a diagram showing a sequence of core damage states. Whenever possible, a temperature and a time after accident initiation are estimated for each damage state in the sequence diagram. This diagram can be used to anticipate the evolution of events during an accident. Possible responses of plant instruments are described to identify these damage states and the effects of water addition. The rate and amount of water addition needed (a) to remove energy from the core, (b) to stabilize the core or (c) to not adversely affect the damage progression, are estimated. Analysis of the capability to remove energy from large cohesive and particulate debris beds indicates that these beds may not be stabilized in the core region and they may partially relocate to the lower plenum of the reactor vessel

30 CFR 780.33 - Relocation or use of public roads.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 3 2010-07-01 2010-07-01 false Relocation or use of public roads. 780.33... PLAN § 780.33 Relocation or use of public roads. Each application shall describe, with appropriate maps... line of any public road, except where mine access or haul roads join that right-of-way; or (b...

Relocating a Company within the EU

DEFF Research Database (Denmark)

Werlauff, Erik

2008-01-01

This article analyses how a public or private limited company can be relocated in another EU country via establishment of a subsidiary in the second country, followed by a reverse vertical merger (the parent ceases to exist and the subsidiary is the continuing company). The consideration for the ...... for the merger is shares in the subsidiary, which can be effected by giving the parent company's shares in the subsidiary to the parent company's shareholders, but after the shares have taken a brief respite in the subsidiary as own shares......This article analyses how a public or private limited company can be relocated in another EU country via establishment of a subsidiary in the second country, followed by a reverse vertical merger (the parent ceases to exist and the subsidiary is the continuing company). The consideration...

The Future of Earthquake Relocation Tools

Science.gov (United States)

Lecocq, T.; Caudron, C.

2010-12-01

Many scientists around the world use earthquake relocation software for their research. Some use "known" software like HYPODD or COMPLOC, while others use their own algorithms and codes. Often, beginners struggle to get one tool running or to properly configure input parameters. This Poster will be witness of debates that will take place during the Meeting, for example adressing questions like "Which program for which application?" ; "Standardized In/Outs?" , "Tectonic / Volcanic / Other ?" ; "All programs inside one single Super-Package?" ; "Common/Base Bibliography for the Relocation-Beginner?" ; "Continuous or Layered Velocity Model?" etc... We will also present the scheme of a Super-Package we are working on, grouping HYPODD [Waldhauser 2001], COMPLOC [Lin&Shearer 2006], LOTOS [Koulakov 2009] ; allowing standard in/outs for the 3 programs, and thus, the comparison of their outputs.

Electrochemical ion separation in molten salts

Science.gov (United States)

Spoerke, Erik David; Ihlefeld, Jon; Waldrip, Karen; Wheeler, Jill S.; Brown-Shaklee, Harlan James; Small, Leo J.; Wheeler, David R.

2017-12-19

A purification method that uses ion-selective ceramics to electrochemically filter waste products from a molten salt. The electrochemical method uses ion-conducting ceramics that are selective for the molten salt cations desired in the final purified melt, and selective against any contaminant ions. The method can be integrated into a slightly modified version of the electrochemical framework currently used in pyroprocessing of nuclear wastes.

Molten salt reactor as asymptotic safety nuclear system

International Nuclear Information System (INIS)

Novikov, V.M.; Ignatyev, V.V.

1989-01-01

Safety is becoming the main and priority problem of the nuclear power development. An increase of the active safety measures could hardly be considered as the proper way to achieve the asymptotically high level of nuclear safety. It seem that the more realistic way to achieve such a goal is to minimize risk factors and to maximize the use of inherent and passive safety properties. The passive inherent safety features of the liquid fuel molten salt reactor (MSR) technology are making it attractive for future energy generation. The achievement of the asymptotic safety in MSR is being connected with the minimization of such risk factors as a reactivity excess, radioactivity stored, decay heat, non nuclear energy stored in core. In this paper safety peculiarities of the different MSR concepts are discussed

46 CFR 151.50-55 - Sulfur (molten).

Science.gov (United States)

2010-10-01

... BULK LIQUID HAZARDOUS MATERIAL CARGOES Special Requirements § 151.50-55 Sulfur (molten). (a.... Heat transfer media shall be steam, and alternate media will require specific approval of the... 46 Shipping 5 2010-10-01 2010-10-01 false Sulfur (molten). 151.50-55 Section 151.50-55 Shipping...

Flow effect on {sup 135}I and {sup 135}Xe evolution behavior in a molten salt reactor

Energy Technology Data Exchange (ETDEWEB)

Wu, Jianhui; Guo, Chen [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); Cai, Xiangzhou [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Yu, Chenggang; Zou, Chunyan [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); Han, Jianlong [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Chen, Jingen, E-mail: chenjg@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China)

2017-04-01

Highlights: • {sup 135}Xe and {sup 135}I evolution law in a molten salt reactor is analytically deduced. • The circulation of fuel salt through the primary loop decreases the concentration of {sup 135}I and {sup 135}Xe. • {sup 135}I and {sup 135}Xe concentration reduction is independent with the mass flow rate at normal core operating condition. • Increasing the external core volume would raise {sup 135}I and {sup 135}Xe concentration reduction caused by the flow effect. - Abstract: Molten Salt Reactor (MSR) employs fissile material dissolved in the fluoride salt as fuel which continuously circulates through the primary loop with the flow cycle time being a few tens of seconds. The nuclei evolution law is quite different from that in a solid fuel reactor. In this paper, we analytically deduce the nuclei evolution law of {sup 135}Xe and {sup 135}I which are entrained in the flowing salt, evaluate its concentration changing with the burnup time, and validate the result with the SCALE6. The circulation of fuel salt could decrease the concentration of {sup 135}Xe and {sup 135}I, and the reduction can achieve to around 40% and 50% for {sup 135}Xe and {sup 135}I respectively at a small power level (e.g., 2 MW) when the core has the same fuel salt volume as that of the outer-loop. Furthermore, it can be found that the reduction is inversely proportional to the core to outer-loop volume ratio, but uncorrelated with the mass flow rate under normal operating condition of a MSR. At low core power scale, the flow effect on {sup 135}Xe concentration reduction is apparent, but it is mitigated as the core power scale increases because of the rise of {sup 135}I concentration, which raises its decay to {sup 135}Xe and compensates the loss of {sup 135}Xe due to decay at the outer-loop. The decreased {sup 135}Xe concentration results in a core reactivity increase varying from around 150 pcm to 1000 pcm depending on the core power and core to outer-loop volume ratio.

Thorium Molten-Salt Nuclear Energy Synergetics

International Nuclear Information System (INIS)

Furukawa, Kazuo; Lecocq, A.; Kato, Yoshio; Mitachi, Kohshi.

1990-01-01

In the next century, the 'fission breeder' concept will not be practical to solve the global energy problems, including environmental and North-South problems. As a new measure, a simple rational Th molten salt breeding fuel cycle system, named 'Thorium Molten-Salt Nuclear Energy Synergetics (THORIMS-NES)', which composed of simple power stations and fissile producers, is proposed. This is effective to establish the essential improvement in issues of resources, safety, power-size flexibility, anti-nuclear proliferation and terrorism, radiowaste, economy, etc. securing the simple operation, maintenance, chemical processing, and rational breeding fuel cycle. As examples, 155 MWe fuel self-sustaining power station 'FUJI-II', 7 MWe pilot-plant 'miniFUJI-II', 1 GeV-300 mA proton Accelerator Molten-Salt Breeder 'AMSB', and their combined fuel cycle system are explained. (author)

Factors and aspects of the relocation strategies of companies

Directory of Open Access Journals (Sweden)

Serghei Margulescu

2013-05-01

Full Text Available In the last two decades there have been significant changes in the factors that determine the geographic location or relocation of R &D, production and marketing of all transnational companies. Supply chains have expanded to new areas of the globe and big traditional providers have also expanded their global presence by an increasing trend of co-localization with their main customers. Contract manufacturers have multiplied and strengthened, expanding their geographical distribution. A more obvious trend was that of the geographic dispersion of other global value chain functions such as business services and logistics support functions. Relocation to countries with cheap labor is not always a successful strategy. Therefore the decision to relocate in international geographical area, regardless of the structure formula, offshoring or outsourcing, must be based on a more diverse set of factors.

30 CFR 784.18 - Relocation or use of public roads.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 3 2010-07-01 2010-07-01 false Relocation or use of public roads. 784.18... PLAN § 784.18 Relocation or use of public roads. Each application shall describe, with appropriate maps...-of-way line of any public road, except where mine access or haul roads join that right-of-way; or (b...

Conceptual design of Indian molten salt breeder reactor

International Nuclear Information System (INIS)

Vijayan, P.K.; Basak, A.; Dulera, I.V.; Vaze, K.K.; Basu, S.; Sinha, R.K.

2014-01-01

The fuel in a molten salt breeder reactor is in the form of a continuously circulating molten salt. Fluoride based salts have been almost universally proposed. A crucial part for achieving reasonable breeding in such reactors is the need to reprocess the salt continuously, either online or in batch mode. This constitutes a major technological challenge for this type of reactors. India has recently started carrying out fundamental studies so as to arrive at a conceptual design of Indian Molten Salt Breeder Reactor (IMSBR). Presently various design options and possibilities are being studied from the point of view of reactor physics and thermal hydraulic design. In parallel fundamental studies as regards various molten salts have also been initiated. This paper would discuss conceptual design of these reactors, as well as associated issues and technologies

Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

International Nuclear Information System (INIS)

Govers, K.; Verwerft, M.

2016-01-01

The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive. - Highlights: • We performed Discrete Element Methods simulation for fuel relocation and dispersal during LOCA transients. • The approach provides a mechanistic description of these phenomena. • The approach shows the ability of the technique to reproduce experimental observations. • The packing fraction in the balloon is shown to stabilize at 50–60%.

The Problem of Faculty Relocation.

Science.gov (United States)

Tabachnick, Stephen E.

1992-01-01

A faculty move to a new campus can be traumatic, but colleges and universities can take steps to lessen the strain. Solutions to faculty relocation problems should be a standard part of any hiring package, not left to chance and individual negotiation. Some problems are inexpensive and easy to solve. (MSE)

Measurement and analyses of molten Ni-Co alloy density

Institute of Scientific and Technical Information of China (English)

XIAO Feng; K. MUKAI; FANG Liang; FU Ya; YANG Ren-hui

2006-01-01

With the advent of powerful mathematical modeling techniques for material phenomena, there is renewed interest in reliable data for the density of the Ni-based superalloys. Up to now, there has been few report on the density of molten Ni-Co alloy.In order to obtain more accurate density data for molten Ni-Co alloy, the density of molten Ni-Co alloy was measured with a modified sessile drop method, and the accommodation of different atoms in molten Ni-Co alloy was analyzed. The density of alloy is found to decrease with increasing temperature and Co concentration in the alloy. The molar volume of molten Ni-Co alloy increases with increasing Co concentration. The molar volume of Ni-Co alloy determined shows a positive deviation from the linear molar volume, and the deviation of molar volume from ideal mixing increases with increasing Co concentration over the experimental concentration range.

Wages and commuting: quasi-natural experiments' evidence from firms that relocate

DEFF Research Database (Denmark)

Mulalic, Ismir; N. Van Ommeren, Jos; Pilegaard, Ninette

We examine workers' compensating differentials for commuting distance in a quasi-natural experiment setting by examining how wages of workers belonging to the same firm respond to changes in commuting distance induced by firm relocations. This setup enables us to test for the relevance of non...... explanations. We focus on wage changes in the year after, as well as three years after, the firm relocation. We demonstrate that a one km increase in commuting distance induces a wage increase of only 0.06 percent in the year after the relocation, but a more substantial wage increase of about 0.2 percent three...

Molten salt reactors: A new beginning for an old idea

International Nuclear Information System (INIS)

LeBlanc, David

2010-01-01

Molten salt reactors have seen a marked resurgence of interest over the past decade, highlighted by their inclusion as one of six Generation IV reactor types. The most active development period however was between the mid 1950s and early 1970s at Oak Ridge National Laboratories (ORNL) and any new re-examination of this concept must bear in mind the far different priorities then in place. High breeding ratios and short doubling times were paramount and this guided the evolution of the Molten Salt Breeder Reactor (MSBR) program. As the inherent advantages of the molten salt concept have become apparent to an increasing number of researchers worldwide it is important to not simply look to continue where ORNL left off but to return to basics in order to offer the best design using updated goals and abilities. A major potential change to the traditional Single Fluid, MSBR design and a subject of this presentation is a return to the mode of operation that ORNL proposed for the majority of its MSR program. That being the Two Fluid design in which separate salts are used for fissile 233 UF 4 and fertile ThF 4 . Oak Ridge abandoned this promising route due to what was known as the 'plumbing problem'. It will be shown that a simple yet crucial modification to core geometry can solve this problem and enable the many advantages of the Two Fluid design. In addition, another very promising route laid out by ORNL was simplified Single Fluid converter reactors that could obtain far superior lifetime uranium utilization than LWR or CANDU without the need for any fuel processing beyond simple chemistry control. Updates and potential improvements to this very attractive concept will also be explored.

Preliminary model for core/concrete interactions

International Nuclear Information System (INIS)

Murfin, W.B.

1977-08-01

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

Application of lithium in molten-salt reduction processes

International Nuclear Information System (INIS)

Gourishankar, K. V.

1998-01-01

Metallothermic reductions have been extensively studied in the field of extractive metallurgy. At Argonne National Laboratory (ANL), we have developed a molten-salt based reduction process using lithium. This process was originally developed to reduce actinide oxides present in spent nuclear fuel. Preliminary thermodynamic considerations indicate that this process has the potential to be adapted for the extraction of other metals. The reduction is carried out at 650 C in a molten-salt (LiCl) medium. Lithium oxide (Li 2 O), produced during the reduction of the actinide oxides, dissolves in the molten salt. At the end of the reduction step, the lithium is regenerated from the salt by an electrowinning process. The lithium and the salt from the electrowinning are then reused for reduction of the next batch of oxide fuel. The process cycle has been successfully demonstrated on an engineering scale in a specially designed pyroprocessing facility. This paper discusses the applicability of lithium in molten-salt reduction processes with specific reference to our process. Results are presented from our work on actinide oxides to highlight the role of lithium and its effect on process variables in these molten-salt based reduction processes

Structure of molten Bi-Sb-alloys by means of neutron diffraction

International Nuclear Information System (INIS)

Lamparter, P.; Knoll, W.; Steeb, S.

1976-01-01

The structural investigations with melts can be subdivided into two groups: The first group contains molten metals and molten alloys, and one can state that the structure of molten metals and of molten alloys nowadays is rather well understood. Interference functions of molten metals may be described by a hard sphere model. This is valid also for molten alloys with statistical distribution. For the second group, namely molten non-metals and molecular melts, the interference functions as well as the pair correlation functions are very offen rather complicated and not well understood. The present study is concerned with the transition region between these two groups. It is shown that the melts of the Bi-Sb system exhibit a change from metallic to non-metallic structure. Regarding the experimental details: the experiments were done with the two-axes spectrometer D 4 at the high-flux reactor at Grenoble. The containers consisted of cylindrical quartz tubes with a wall thickness of 0.1 cm. The furnace consisted of a direct-heated vanadium tube. The wavelength of the neutrons was 0.695 A. The final result is that the structure in molten Bi-Sb-alloys consists of primitive tetrahedra with coordination number 3. There are less tetrahedra in molten Bi than in molten Sb. Also with rising temperature the number of tetrahedra decreases. It is shown how to compose the coordination numbers of two structures to get the observed coordination number. The observed values are always the mean values of the two structures. (orig./HK) [de

Molten salt extractive distillation process for zirconium-hafnium separation

International Nuclear Information System (INIS)

McLaughlin, D.F.; Stoltz, R.A.

1989-01-01

This patent describes an improvement in a process for zirconium-hafnium separation. It utilizes an extractive distillation column with a mixture of zirconium and hafnium tetrachlorides introduced into a distillation column having a top and bottom with hafnium enriched overheads taken from the top of the column and a molten salt solvent circulated through the column to provide a liquid phase, and with molten salt solvent containing zirconium chloride being taken from the bottom of the distillation column. The improvements comprising: utilizing a molten salt solvent consisting principally of lithium chloride and at least one of sodium, potassium, magnesium and calcium chlorides; stripping of the zirconium chloride taken from the bottom of the distillation column by electrochemically reducing zirconium from the molten salt solvent; and utilizing a pressurized reflux condenser on the top of the column to add the hafnium chloride enriched overheads to the molten salt solvent previously stripped of zirconium chloride

Advances in molten salt electrochemistry towards future energy systems

International Nuclear Information System (INIS)

Ito, Yasuhiko

2005-01-01

This review article describes some selected novel molten salt electrochemical processes which have been created/developed by the author and his coworkers, with emphasis on the applications towards future energy systems. After showing a perspective of the applications of molten salt electrochemistry from the viewpoints of energy and environment, several selected topics are described in detail, which include nitride fuel cycle in a nuclear field, hydrogen energy system coupled with ammonia economy, thermally regenerative fuel cell systems, novel Si production process for solar cell and novel molten salt electrochemical processes for various energy and environment related functional materials including nitrides, rare earth-transition metal alloys, fine particles obtained by plasma-induced electrolysis, and carbon film. And finally, the author stresses again, the importance and potential of molten salt electrochemistry, and encourages young students, scientists and researchers to march in a procession hand in hand towards a bright future of molten salts. (author)

Evaluation of relocation of unionid mussels into artificial ponds

Science.gov (United States)

Newton, T.J.; Monroe, E.M.; Kenyon, R.; Gutreuter, S.; Welke, K.I.; Thiel, P.A.

2001-01-01

Relocation of unionid mussels into refuges (e.g., hatchery ponds) has been suggested as a management tool to protect these animals from the threat of zebra mussel (Dreissena polymorpha) invasion. To evaluate the efficacy of relocation, we experimentally relocated 768 mussels, representing 5 species (Leptodea fragilis, Obliquaria reflexa, Fusconaia flava, Amblema plicata, and Quadrula quadrula) into an earthen pond at a National Fish Hatchery or back into the river. In both locations, mussels were placed into 1 of 4 treatments (mesh bags, corrals, and buried or suspended substrate-filled trays). Mussels were examined annually for survival, growth (shell length and wet mass), and physiological condition (glycogen concentration in foot and mantle and tissue condition index) for 36 mo in the pond or 40 mo in the river. We observed significant differences in mortality rates between locations (mortality was 4 times greater in the pond than in the river), among treatments (lowest mortality in the suspended trays), and among species (lower mortality in the amblemines than lamp-silines). Overall survival in both locations averaged 80% the 1st year; survival in the pond decreased dramatically after that. Although length and weight varied between locations and over time, these changes were small, suggesting that their utility as short-term measures of well being in long-lived unionids is questionable. Mussels relocated to the pond were in poor physiological condition relative to those in the river, but the magnitude of these differences was small compared to the inherent variability in physiological condition of reference mussels. These data suggest that relocation of unionids into artificial ponds is a high-risk conservation strategy; alternatives such as introduction of infected host fish, identification of mussel beds at greatest risk from zebra mussels, and a critical, large-scale assessment of the factors contributing to their decline should be explored.

Event course analysis of core disruptive accidents; Ereignisablaufanalyse kernzerstoerender Unfaelle

Energy Technology Data Exchange (ETDEWEB)

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

1995-08-01

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

Short-Period Surface Wave Based Seismic Event Relocation

Science.gov (United States)

White-Gaynor, A.; Cleveland, M.; Nyblade, A.; Kintner, J. A.; Homman, K.; Ammon, C. J.

2017-12-01

Accurate and precise seismic event locations are essential for a broad range of geophysical investigations. Superior location accuracy generally requires calibration with ground truth information, but superb relative location precision is often achievable independently. In explosion seismology, low-yield explosion monitoring relies on near-source observations, which results in a limited number of observations that challenges our ability to estimate any locations. Incorporating more distant observations means relying on data with lower signal-to-noise ratios. For small, shallow events, the short-period (roughly 1/2 to 8 s period) fundamental-mode and higher-mode Rayleigh waves (including Rg) are often the most stable and visible portion of the waveform at local distances. Cleveland and Ammon [2013] have shown that teleseismic surface waves are valuable observations for constructing precise, relative event relocations. We extend the teleseismic surface wave relocation method, and apply them to near-source distances using Rg observations from the Bighorn Arche Seismic Experiment (BASE) and the Earth Scope USArray Transportable Array (TA) seismic stations. Specifically, we present relocation results using short-period fundamental- and higher-mode Rayleigh waves (Rg) in a double-difference relative event relocation for 45 delay-fired mine blasts and 21 borehole chemical explosions. Our preliminary efforts are to explore the sensitivity of the short-period surface waves to local geologic structure, source depth, explosion magnitude (yield), and explosion characteristics (single-shot vs. distributed source, etc.). Our results show that Rg and the first few higher-mode Rayleigh wave observations can be used to constrain the relative locations of shallow low-yield events.

Partially molten magma ocean model

International Nuclear Information System (INIS)

Shirley, D.N.

1983-01-01

The properties of the lunar crust and upper mantle can be explained if the outer 300-400 km of the moon was initially only partially molten rather than fully molten. The top of the partially molten region contained about 20% melt and decreased to 0% at 300-400 km depth. Nuclei of anorthositic crust formed over localized bodies of magma segregated from the partial melt, then grew peripherally until they coverd the moon. Throughout most of its growth period the anorthosite crust floated on a layer of magma a few km thick. The thickness of this layer is regulated by the opposing forces of loss of material by fractional crystallization and addition of magma from the partial melt below. Concentrations of Sr, Eu, and Sm in pristine ferroan anorthosites are found to be consistent with this model, as are trends for the ferroan anorthosites and Mg-rich suites on a diagram of An in plagioclase vs. mg in mafics. Clustering of Eu, Sr, and mg values found among pristine ferroan anorthosites are predicted by this model

Molten salts processes and generic simulation

International Nuclear Information System (INIS)

Ogawa, Toru; Minato, Kazuo

2001-01-01

Development of dry separation process (pyrochemical process) using molten salts for the application of spent-nuclear fuel reprocessing requires a rather complete fundamental database as well as process simulation technique with wide applicability. The present report concerns recent progress and problems in this field taking behaviors of co-electrodeposition of UO 2 and PuO 2 in molten salts as an example, and using analytical simulation of local equilibrium combined with generic diffusion. (S. Ohno)

Modelling of heat transfer between molten core and concrete with account of phase changes in the melt

International Nuclear Information System (INIS)

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

1992-01-01

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

A Novel Molten Salt Reactor Concept to Implement the Multi-Step Time-Scheduled Transmutation Strategy

International Nuclear Information System (INIS)

Csom, Gyula; Feher, Sandor; Szieberthj, Mate

2002-01-01

Nowadays the molten salt reactor (MSR) concept seems to revive as one of the most promising systems for the realization of transmutation. In the molten salt reactors and subcritical systems the fuel and material to be transmuted circulate dissolved in some molten salt. The main advantage of this reactor type is the possibility of the continuous feed and reprocessing of the fuel. In the present paper a novel molten salt reactor concept is introduced and its transmutation capabilities are studied. The goal is the development of a transmutation technique along with a device implementing it, which yield higher transmutation efficiencies than that of the known procedures and thus results in radioactive waste whose load on the environment is reduced both in magnitude and time length. The procedure is the multi-step time-scheduled transmutation, in which transformation is done in several consecutive steps of different neutron flux and spectrum. In the new MSR concept, named 'multi-region' MSR (MRMSR), the primary circuit is made up of a few separate loops, in which salt-fuel mixtures of different compositions are circulated. The loop sections constituting the core region are only neutronically and thermally coupled. This new concept makes possible the utilization of the spatial dependence of spectrum as well as the advantageous features of liquid fuel such as the possibility of continuous chemical processing etc. In order to compare a 'conventional' MSR and a proposed MRMSR in terms of efficiency, preliminary calculational results are shown. Further calculations in order to find the optimal implementation of this new concept and to emphasize its other advantageous features are going on. (authors)

Refractory thermowell for continuous high temperature measurement of molten metal

International Nuclear Information System (INIS)

Thiesen, T.J.

1992-01-01

This patent describes a vessel for handling molten metal having an interior refractory lining, apparatus for continuous high temperature measurement of the molten metal. It comprises a thermowell; the thermowell containing a multiplicity of thermocouples; leads being coupled to a means for continuously indicating the temperature of the molten metal in the vessel

A basic study on fluoride-based molten salt electrolysis technology

Energy Technology Data Exchange (ETDEWEB)

Hwang, Il Soon [Seoul National University, Seoul (Korea); Kim, Kwang Bum [Yonsei University, Seoul (Korea); Park, Byung Gi [Seoul National University, Seoul (Korea)

2001-04-01

The objective of this project is to study on the physicochemical properties of fluoride molten salt, to develop numerical model for simulation of molten salt electrolysis, and to establish experimental technique of fluoride molten salt. Physicochemical data of fluoride molten salt are investigated and summarized. The numerical model, designated as REFIN is developed with diffusion-layer theory and electrochemical reaction kinetics. REFIN is benchmarked with published experimental data. REFIN has a capability to simulate multicomponent electrochemical system at transient conditions. Experimental device is developed to measure electrochemical properties of structural material for fluoride molten salt. Ni electrode is measured with cyclic voltammogram in the conditions of 600 .deg. C LiF-BeF{sub 2} and 700 .deg. C LiF-BeF{sub 2}. 74 refs., 23 figs., 57 tabs. (Author)

An experimental test plan for the characterization of molten salt thermochemical properties in heat transport systems

International Nuclear Information System (INIS)

Calderoni, Pattrick

2010-01-01

Molten salts are considered within the Very High Temperature Reactor program as heat transfer media because of their intrinsically favorable thermo-physical properties at temperatures starting from 300 C and extending up to 1200 C. In this context two main applications of molten salt are considered, both involving fluoride-based materials: as primary coolants for a heterogeneous fuel reactor core and as secondary heat transport medium to a helium power cycle for electricity generation or other processing plants, such as hydrogen production. The reference design concept here considered is the Advanced High Temperature Reactor (AHTR), which is a large passively safe reactor that uses solid graphite-matrix coated-particle fuel (similar to that used in gas-cooled reactors) and a molten salt primary and secondary coolant with peak temperatures between 700 and 1000 C, depending upon the application. However, the considerations included in this report apply to any high temperature system employing fluoride salts as heat transfer fluid, including intermediate heat exchangers for gas-cooled reactor concepts and homogeneous molten salt concepts, and extending also to fast reactors, accelerator-driven systems and fusion energy systems. The purpose of this report is to identify the technical issues related to the thermo-physical and thermo-chemical properties of the molten salts that would require experimental characterization in order to proceed with a credible design of heat transfer systems and their subsequent safety evaluation and licensing. In particular, the report outlines an experimental R and D test plan that would have to be incorporated as part of the design and operation of an engineering scaled facility aimed at validating molten salt heat transfer components, such as Intermediate Heat Exchangers. This report builds on a previous review of thermo-physical properties and thermo-chemical characteristics of candidate molten salt coolants that was generated as part

An experimental test plan for the characterization of molten salt thermochemical properties in heat transport systems

Energy Technology Data Exchange (ETDEWEB)

Pattrick Calderoni

2010-09-01

Molten salts are considered within the Very High Temperature Reactor program as heat transfer media because of their intrinsically favorable thermo-physical properties at temperatures starting from 300 C and extending up to 1200 C. In this context two main applications of molten salt are considered, both involving fluoride-based materials: as primary coolants for a heterogeneous fuel reactor core and as secondary heat transport medium to a helium power cycle for electricity generation or other processing plants, such as hydrogen production. The reference design concept here considered is the Advanced High Temperature Reactor (AHTR), which is a large passively safe reactor that uses solid graphite-matrix coated-particle fuel (similar to that used in gas-cooled reactors) and a molten salt primary and secondary coolant with peak temperatures between 700 and 1000 C, depending upon the application. However, the considerations included in this report apply to any high temperature system employing fluoride salts as heat transfer fluid, including intermediate heat exchangers for gas-cooled reactor concepts and homogenous molten salt concepts, and extending also to fast reactors, accelerator-driven systems and fusion energy systems. The purpose of this report is to identify the technical issues related to the thermo-physical and thermo-chemical properties of the molten salts that would require experimental characterization in order to proceed with a credible design of heat transfer systems and their subsequent safety evaluation and licensing. In particular, the report outlines an experimental R&D test plan that would have to be incorporated as part of the design and operation of an engineering scaled facility aimed at validating molten salt heat transfer components, such as Intermediate Heat Exchangers. This report builds on a previous review of thermo-physical properties and thermo-chemical characteristics of candidate molten salt coolants that was generated as part of the

Rheological behavior and constitutive equations of heterogeneous titanium-bearing molten slag

Science.gov (United States)

Jiang, Tao; Liao, De-ming; Zhou, Mi; Zhang, Qiao-yi; Yue, Hong-rui; Yang, Song-tao; Duan, Pei-ning; Xue, Xiang-xin

2015-08-01

Experimental studies on the rheological properties of a CaO-SiO2-Al2O3-MgO-TiO2-(TiC) blast furnace (BF) slag system were conducted using a high-temperature rheometer to reveal the non-Newtonian behavior of heterogeneous titanium-bearing molten slag. By measuring the relationships among the viscosity, the shear stress and the shear rate of molten slags with different TiC contents at different temperatures, the rheological constitutive equations were established along with the rheological parameters; in addition, the non-Newtonian fluid types of the molten slags were determined. The results indicated that, with increasing TiC content, the viscosity of the molten slag tended to increase. If the TiC content was less than 2wt%, the molten slag exhibited the Newtonian fluid behavior when the temperature was higher than the critical viscosity temperature of the molten slag. In contrast, the molten slag exhibited the non-Newtonian pseudoplastic fluid characteristic and the shear thinning behavior when the temperature was less than the critical viscosity temperature. However, if the TiC content exceeded 4wt%, the molten slag produced the yield stress and exhibited the Bingham and plastic pseudoplastic fluid behaviors when the temperature was higher and lower than the critical viscosity temperature, respectively. When the TiC content increased further, the yield stress of the molten slag increased and the shear thinning phenomenon became more obvious.

Dynamics and control of molten-salt breeder reactor

Energy Technology Data Exchange (ETDEWEB)

Sing, Vikram; Lish, Matthew R.; Chvala, Ondrej; Upadhyaya, Belle R. [Dept. of Nuclear Engineering, University of Tennessee, Knoxville (United States)

2017-08-15

Preliminary results of the dynamic analysis of a two-fluid molten-salt breeder reactor (MSBR) system are presented. Based on an earlier work on the preliminary dynamic model of the concept, the model presented here is nonlinear and has been revised to accurately reflect the design exemplified in ORNL-4528. A brief overview of the model followed by results from simulations performed to validate the model is presented. Simulations illustrate stable behavior of the reactor dynamics and temperature feedback effects to reactivity excursions. Stable and smooth changes at various nodal temperatures are also observed. Control strategies for molten-salt reactor operation are discussed, followed by an illustration of the open-loop load-following capability of the molten-salt breeder reactor system. It is observed that the molten-salt breeder reactor system exhibits “self-regulating” behavior, minimizing the need for external controller action for load-following maneuvers.

Boric Ester-Type Molten Salt via Dehydrocoupling Reaction

Directory of Open Access Journals (Sweden)

Noriyoshi Matsumi

2014-11-01

Full Text Available Novel boric ester-type molten salt was prepared using 1-(2-hydroxyethyl-3-methylimidazolium chloride as a key starting material. After an ion exchange reaction of 1-(2-hydroxyethyl-3-methylimidazolium chloride with lithium (bis-(trifluoromethanesulfonyl imide (LiNTf2, the resulting 1-(2-hydroxyethyl-3-methylimidazolium NTf2 was reacted with 9-borabicyclo[3.3.1]nonane (9-BBN to give the desired boric ester-type molten salt in a moderate yield. The structure of the boric ester-type molten salt was supported by 1H-, 13C-, 11B- and 19F-NMR spectra. In the presence of two different kinds of lithium salts, the matrices showed an ionic conductivity in the range of 1.1 × 10−4–1.6 × 10−5 S cm−1 at 51 °C. This was higher than other organoboron molten salts ever reported.

Dynamics and control of molten-salt breeder reactor

Directory of Open Access Journals (Sweden)

Vikram Singh

2017-08-01

Full Text Available Preliminary results of the dynamic analysis of a two-fluid molten-salt breeder reactor (MSBR system are presented. Based on an earlier work on the preliminary dynamic model of the concept, the model presented here is nonlinear and has been revised to accurately reflect the design exemplified in ORNL-4528. A brief overview of the model followed by results from simulations performed to validate the model is presented. Simulations illustrate stable behavior of the reactor dynamics and temperature feedback effects to reactivity excursions. Stable and smooth changes at various nodal temperatures are also observed. Control strategies for molten-salt reactor operation are discussed, followed by an illustration of the open-loop load-following capability of the molten-salt breeder reactor system. It is observed that the molten-salt breeder reactor system exhibits “self-regulating” behavior, minimizing the need for external controller action for load-following maneuvers.

Core catcher concepts future PWR-Plants

International Nuclear Information System (INIS)

Alsmeyer, H.; Werle, H.

1994-01-01

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

A condensed review of the core catcher in the LMR

International Nuclear Information System (INIS)

Lee, Yong Bum; Hahn, Do hee

2001-03-01

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

Molten salts processes and generic simulation

Energy Technology Data Exchange (ETDEWEB)

Ogawa, Toru; Minato, Kazuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-12-01

Development of dry separation process (pyrochemical process) using molten salts for the application of spent-nuclear fuel reprocessing requires a rather complete fundamental database as well as process simulation technique with wide applicability. The present report concerns recent progress and problems in this field taking behaviors of co-electrodeposition of UO{sub 2} and PuO{sub 2} in molten salts as an example, and using analytical simulation of local equilibrium combined with generic diffusion. (S. Ohno)

Molten salt reactors - safety options galore

International Nuclear Information System (INIS)

Gat, U.; Dodds, H.L.

1997-01-01

Safety features and attributes of molten salt reactors (MSR) are described. The unique features of fluid fuel reactors of on-line continuous processing and the ability for so-called external cooling result in simple and safe designs with low excess reactivity, low fission product inventory, and small source term. These, in turn, make a criticality accident unlikely and reduce the severity of a loss of coolant to where they are no longer severe accidents. A melt down is not an accident for a reactor that uses molten fuel. The molten salts are stable, non-reactive and efficient heat transfer media that operate at high temperatures at low pressures and are highly compatible with selected structural materials. All these features reduce the accident plethora. Freeze valves can be used for added safety. An ultimate safe reactor (U.S.R) is described with safety features that are passive, inherent and non-tamperable (PINT)

Development of viscometers for molten salts

International Nuclear Information System (INIS)

Hayashi, Hirokazu; Kato, Yoshio; Ogawa, Toru; Sato, Yuzuru.

1997-06-01

Viscometers specially designed for molten salts were made. One is a oscillating cup type and the other is a capillary type. In the case of the oscillating cup viscometer, the viscosity is determined absolutely through the period and the logarithmic decrement of oscillation with other physical parameters. The period and the logarithmic decrement are calculated from the time intervals between two photo-detectors' intercepts of the reflected laser beam. The capillary viscometer used is made of quartz and the sample is sealed under vacuum, which is placed in a transparent furnace. Efflux time is measured by direct visual observation. Cell constants are determined with distilled water as a calibrating liquid. Viscosities of molten KCl are measured with each viscometer. The differences between measured and standard values of molten KCl at several temperatures are within 5% for the oscillating cup viscometer and within 3% for the capillary viscometer. (author)

Computer simulation on molten ionic salts

International Nuclear Information System (INIS)

Kawamura, K.; Okada, I.

1978-01-01

The extensive advances in computer technology have since made it possible to apply computer simulation to the evaluation of the macroscopic and microscopic properties of molten salts. The evaluation of the potential energy in molten salts systems is complicated by the presence of long-range energy, i.e. Coulomb energy, in contrast to simple liquids where the potential energy is easily evaluated. It has been shown, however, that no difficulties are encountered when the Ewald method is applied to the evaluation of Coulomb energy. After a number of attempts had been made to approximate the pair potential, the Huggins-Mayer potential based on ionic crystals became the most often employed. Since it is thought that the only appreciable contribution to many-body potential, not included in Huggins-Mayer potential, arises from the internal electrostatic polarization of ions in molten ionic salts, computer simulation with a provision for ion polarization has been tried recently. The computations, which are employed mainly for molten alkali halides, can provide: (1) thermodynamic data such as internal energy, internal pressure and isothermal compressibility; (2) microscopic configurational data such as radial distribution functions; (3) transport data such as the diffusion coefficient and electrical conductivity; and (4) spectroscopic data such as the intensity of inelastic scattering and the stretching frequency of simple molecules. The computed results seem to agree well with the measured results. Computer simulation can also be used to test the effectiveness of a proposed pair potential and the adequacy of postulated models of molten salts, and to obtain experimentally inaccessible data. A further application of MD computation employing the pair potential based on an ionic model to BeF 2 , ZnCl 2 and SiO 2 shows the possibility of quantitative interpretation of structures and glass transformation phenomena

The MELTSPREAD Code for Modeling of Ex-Vessel Core Debris Spreading Behavior, Code Manual – Version3-beta

Energy Technology Data Exchange (ETDEWEB)

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

2017-09-01

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

Understanding the Experiences of Relocatees During Forced Relocation in Chinese Urban Restructuring

NARCIS (Netherlands)

Li, X.; van Ham, M.; Kleinhans, R.J.

Despite the massive forced relocation of residents during urban restructuring in China, there are no systematic studies on how residents undergo the process. Most studies concerning urban restructuring in China directly equate forced relocation with displacement, which has a negative connotation.

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

Clinical validity of a relocation stress scale for the families of patients transferred from intensive care units.

Science.gov (United States)

Oh, HyunSoo; Lee, Seul; Kim, JiSun; Lee, EunJu; Min, HyoNam; Cho, OkJa; Seo, WhaSook

2015-07-01

This study was conducted to develop a family relocation stress scale by modifying the Son's Relocation Stress Syndrome Scale, to examine its clinical validity and reliability and to confirm its suitability for measuring family relocation stress. The transfer of ICU patients to general wards is a significant anxiety-producing event for family members. However, no relocation stress scale has been developed specifically for families. A nonexperimental, correlation design was adopted. The study subjects were 95 family members of 95 ICU patients at a university hospital located in Incheon, South Korea. Face and construct validities of the devised family relocation stress scale were examined. Construct validity was examined using factor analysis and by using a nomological validity test. Reliability was also examined. Face and content validity of the scale were verified by confirming that its items adequately measured family relocation stress. Factor analysis yielded four components, and the total variance explained by these four components was 63·0%, which is acceptable. Nomological validity was well supported by significant relationships between relocation stress and degree of preparation for relocation, patient self-care ability, family burden and satisfaction with the relocation process. The devised scale was also found to have good reliability. The family relocation stress scale devised in this study was found to have good validity and reliability, and thus, is believed to offer a means of assessing family relocation stress. The findings of this study provide a reliable and valid assessment tool when nurses prepare families for patient transfer from an ICU to a ward setting, and may also provide useful information to those developing an intervention programme for family relocation stress management. © 2015 John Wiley & Sons Ltd.

Validation of the TRACE code for the system dynamic simulations of the molten salt reactor experiment and the preliminary study on the dual fluid molten salt reactor

International Nuclear Information System (INIS)

He, Xun

2016-01-01

Molten Salt Reactor (MSR), which was confirmed as one of the six Generation IV reactor types by the GIF (Generation IV International Forum in 2008), recently draws a lot of attention all around the world. Due to the application of liquid fuels the MSR can be regarded as the most special one among those six GEN-IV reactor types in a sense. A unique advantage of using liquid nuclear fuel lies in that the core melting accident can be thoroughly eliminated. Besides, a molten salt reactor can have several fuel options, for instance, the fuel can be based on "2"3"5U, "2"3"2Th-"2"3"3U, "2"3"8U-"2"3"9Pu cycle or even the spent nuclear fuel (SNF), so the reactor can be operated as a breeder or as an actinides burner both with fast, thermal or epi-thermal neutron spectrum and hence, it has excellent features of the fuel sustainability and for the non-proliferation. Furthermore, the lower operating pressure not only means a lower risk of the explosion as well as the radioactive leakage but also implies that the reactor vessel and its components can be lightweight, thus lowering the cost of equipments. So far there is no commercial MSR being operated. However, the MSR concept and its technical validation dates back to the 1960s to 1970s, when the scientists and engineers from ORNL (Oak Ridge National Laboratory) in the United States managed to build and run the world's first civilian molten salt reactor called MSRE (Molten Salt Reactor Experiment). The MSRE was an experimental liquid-fueled reactor with 10 MW thermal output using "4LiF-BeF_2-ZrF_4-UF_4 as the fuel also as the coolant itself. The MSRE is usually taken as a very important reference case for many current researches to validate their codes and simulations. Without exception it works also as a benchmark for this thesis. The current thesis actually consists of two main parts. The first part is about the validation of the current code for the old MSRE concept, while the second one is about the demonstration of a new

Validation of the TRACE code for the system dynamic simulations of the molten salt reactor experiment and the preliminary study on the dual fluid molten salt reactor

Energy Technology Data Exchange (ETDEWEB)

He, Xun

2016-06-14

Molten Salt Reactor (MSR), which was confirmed as one of the six Generation IV reactor types by the GIF (Generation IV International Forum in 2008), recently draws a lot of attention all around the world. Due to the application of liquid fuels the MSR can be regarded as the most special one among those six GEN-IV reactor types in a sense. A unique advantage of using liquid nuclear fuel lies in that the core melting accident can be thoroughly eliminated. Besides, a molten salt reactor can have several fuel options, for instance, the fuel can be based on {sup 235}U, {sup 232}Th-{sup 233}U, {sup 238}U-{sup 239}Pu cycle or even the spent nuclear fuel (SNF), so the reactor can be operated as a breeder or as an actinides burner both with fast, thermal or epi-thermal neutron spectrum and hence, it has excellent features of the fuel sustainability and for the non-proliferation. Furthermore, the lower operating pressure not only means a lower risk of the explosion as well as the radioactive leakage but also implies that the reactor vessel and its components can be lightweight, thus lowering the cost of equipments. So far there is no commercial MSR being operated. However, the MSR concept and its technical validation dates back to the 1960s to 1970s, when the scientists and engineers from ORNL (Oak Ridge National Laboratory) in the United States managed to build and run the world's first civilian molten salt reactor called MSRE (Molten Salt Reactor Experiment). The MSRE was an experimental liquid-fueled reactor with 10 MW thermal output using {sup 4}LiF-BeF{sub 2}-ZrF{sub 4}-UF{sub 4} as the fuel also as the coolant itself. The MSRE is usually taken as a very important reference case for many current researches to validate their codes and simulations. Without exception it works also as a benchmark for this thesis. The current thesis actually consists of two main parts. The first part is about the validation of the current code for the old MSRE concept, while the second

Nest enlargement in leaf-cutting ants: relocated brood and fungus trigger the excavation of new chambers.

Directory of Open Access Journals (Sweden)

Daniela Römer

Full Text Available During colony growth, leaf-cutting ants enlarge their nests by excavating tunnels and chambers housing their fungus gardens and brood. Workers are expected to excavate new nest chambers at locations across the soil profile that offer suitable environmental conditions for brood and fungus rearing. It is an open question whether new chambers are excavated in advance, or will emerge around brood or fungus initially relocated to a suitable site in a previously-excavated tunnel. In the laboratory, we investigated the mechanisms underlying the excavation of new nest chambers in the leaf-cutting ant Acromyrmex lundi. Specifically, we asked whether workers relocate brood and fungus to suitable nest locations, and to what extent the relocated items trigger the excavation of a nest chamber and influence its shape. When brood and fungus were exposed to unfavorable environmental conditions, either low temperatures or low humidity, both were relocated, but ants clearly preferred to relocate the brood first. Workers relocated fungus to places containing brood, demonstrating that subsequent fungus relocation spatially follows the brood deposition. In addition, more ants aggregated at sites containing brood. When presented with a choice between two otherwise identical digging sites, but one containing brood, ants' excavation activity was higher at this site, and the shape of the excavated cavity was more rounded and chamber-like. The presence of fungus also led to the excavation of rounder shapes, with higher excavation activity at the site that also contained brood. We argue that during colony growth, workers preferentially relocate brood to suitable locations along a tunnel, and that relocated brood spatially guides fungus relocation and leads to increased digging activity around them. We suggest that nest chambers are not excavated in advance, but emerge through a self-organized process resulting from the aggregation of workers and their density

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

Directory of Open Access Journals (Sweden)

Saas Laurent

2017-01-01

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

Nest relocation and excavation in the Florida harvester ant, Pogonomyrmex badius.

Directory of Open Access Journals (Sweden)

Walter R Tschinkel

Full Text Available The Florida harvester ant (Pogonomyrmex badius excavates deep nests in the sandy soils of the Gulf and Atlantic coastal plains. Nest relocations of over 400 colonies in a north Florida coastal plains pine forest were tracked and mapped from 2010 to 2013. Individual colonies varied from one move in two years to four times a year, averaging about one per year. Almost all moves occurred between May and November peaking in July when more than 1% of the colonies moved per day. Move directions were random, and averaged 4 m, with few moves exceeding 10 m. Distance moved was not related to colony size. Over multiple moves, paths were random walks around the original nest location. Relocation is probably intrinsic to the life history of this species, and the causes of relocation remain obscure--the architecture of old and new nests was very similar, and neither the forest canopy nor the density or size of neighbors was correlated with relocation. Monitoring entire relocations (n = 20 showed that they were usually completed in 4 to 6 days. Moves were diurnal, peaking in the mornings and afternoons dipping during mid-day, and ceasing before sundown. Workers excavated the new nest continuously during the daytime throughout the move and beyond. A minority of workers carried seeds, charcoal and brood, with seeds being by far the most common burden. The proportion of burdened workers increased throughout the move. Measured from year to year, small colonies gained size and large ones lost it. Colonies moving more than once in two years lost more size than those moving less often, suggesting that moving may bear a fitness cost. Colony relocation is a dramatic and consistent feature of the life history of the Florida harvester ant, inviting inquiry into its proximal and ultimate causes.

Reexamining the Impact of Employee Relocation Assistance on Housing Prices

OpenAIRE

Marcus T. Allen; Ronald C. Rutherford; Thomas M. Springer

1997-01-01

In this paper, we reexamine the issue of whether corporate relocation assistance programs for transferred employees significantly affect sale prices of single-family homes. We estimate a hedonic price equation that includes physical housing characteristics, location factors, occupancy status, and type of seller for a sample of 2,441 transactions. Seller types include (a) transferred employees who were given direct relocation assistance, (b) transferred employees who were not given direct relo...

Technique for estimating relocated gap width for gap conductance calculations

International Nuclear Information System (INIS)

Klink, P.H.

1978-01-01

Thermally induced fuel fragmentation and relocation has been demonstrated to influence the thermal behavior of a fuel rod in two ways. The effective fuel pellet conductivity is decreased and pellet-to-cladding heat transfer is improved. This paper presents a correlation between as-built and relocated gap width which, used with the Ross and Stoute Gap Conductance Correlation and an appropriate fuel thermal expansion model, closely predicts the measured gap conductances

Development of High Temperature Transport System for Molten Salt

International Nuclear Information System (INIS)

Lee, S. H.; Lee, H. S.; Kim, J. G.

2011-01-01

Pyroprocessing technology is one of the the most promising technologies for the advanced fuel cycle with favorable economic potential and intrinsic proliferation-resistance. The electrorefining process, one of main processes which is composed of pyroprocess to recover the useful elements from spent fuel, is under development at the Korea Atomic Energy Research Institute as a sub process of the pyrochemical treatment of spent PWR fuel. High-temperature molten salt transport technologies are required because a molten salt should be transported from the electrorefiner to electrowiner after the electrorefining process. Therefore, in pyrometallurgical processing, the development of high-temperature molten salt transport technologies is a crucial prerequisite. However, there have been a few transport studies on high-temperature molten salt. In this study, an apparatus for suction transport experiments was designed and constructed for the development of high temperature transport technology for molten salt, and the performance test of the apparatus was performed. And also, predissolution test of the salt was carried out using the reactor with furnace in experimental apparatus

Molten salt fueled nuclear facility with steam-and gas turbine cycles of heat transformation

International Nuclear Information System (INIS)

Ananich, P.I.; Bunin, E.N.; Kazazyan, V.T.; Nemtsev, V.A.; Sikorin, S.N.

2001-01-01

The molten salt fueled nuclear facilities with fuel circulating in the primary circuit have a series of the potential advantages in comparison with the traditional thermal and fast reactors with solid fuel elements. These advantages are ensured by the possibility to receive effective neutron balance in the core, minimum margin reactivity, more deep fuel burnup, unbroken correctness of the fuel physical and chemical properties and by low prices of the fuel cycle. The neutron and thermal-physical calculations of the various variants of the MSFNF with steam-water and gas turbine power circuits and their technical and economical comparison are carried out in this article. Calculations of molten salt nuclear power plant with gas turbine power circuit have been carried out using chemically reacting working body ''nitrin'' (N304 + 1%NO). The molten salt fueled reactors with the thermal power near of 2300 MW with two fuel compositions have been considered. The base variant has been taken the design of NPP with VVER NP-1000 when comparing the results of the calculations. Its economical performances are presented in prices of 1990. The results of the calculations show that it is difficult to determine the advantages of any one of the variants considered in a unique fashion. But NPP with MSR possesses large reserves in the process of optimization of cycle and energy equipment parameters that can improve its technical and economical performances sufficiently. (author)

Present status and needs of research on severe core damage

International Nuclear Information System (INIS)

1982-05-01

The needs for research on severe core damage accident have been emphasized recently, in particular, since TMI-2 accident. The Severe Core Damage Research Task Force was established by the Divisions of Reactor Safety and Reactor Safety Evaluation to evaluate individual phenomenon, to survey the present status of research and to provide the recommended research subjects on severe accidents. This report describes the accident phenomena involving some analytical results, status of research and recommended research subjects on severe core damage accidents, divided into accident sequence, fuel damage, and molten material behavior, fission product behavior, hydrogen generation and combustion, steam explosion and containment integrity. (author)

Results of out-of-pile experiments to investigate the possibilities of cooling a core melt with internal heat production

International Nuclear Information System (INIS)

Fieg, G.

1976-01-01

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

Controlling the discharge of molten material

International Nuclear Information System (INIS)

Geel, J. van; Dobbels, F.; Theunissen, W.

1980-01-01

A method and device are described for controlling the discharge of molten material from a melter or an intermediate vessel, in which a primary outflow is fed to an overflow system, the working level of which is regulated by means of pneumatic pressure on a communicating chamber pertaining to the overflow system. Molten material may be led into a primary overflow by means of a pneumatic lift. The material melted may be a glass used for disposing of radioactive liquid wastes. (author)

Moving as a gift: relocation in older adulthood.

Science.gov (United States)

Perry, Tam E

2014-12-01

While discussions of accessibility, mobility and activities of daily living frame relocation studies, in older adulthood, the paper explores the emotional motivation of gift giving as a rationale for moving. This ethnographic study investigates the processes of household disbandment and decision-making of older adults in the Midwestern United States relocating in post-Global Financial Crisis contexts. In this study, relationships are created and sustained through the process of moving, linking older adults (n=81), their kin (n=49), and professionals (n=46) in the Midwestern United States. Using Marcel Mauss' The Gift (1925/1990) as a theoretical lens, relocation in older adulthood is conceptualized as a gift in two ways: to one's partner, and one's kin. Partners may consider gift-giving in terms of the act of moving to appease and honor their partner. Kin who were not moving themselves were also recipients of the gift of moving. These gifts enchain others in relationships of reciprocity. However these gifts, like all gifts, are not without costs or danger, so this paper examines some of the challenges that emerge along with gift-giving. Copyright © 2014 Elsevier Inc. All rights reserved.

Broadband phase difference method for ultrasonic velocimetry in molten glass

International Nuclear Information System (INIS)

Kikura, Hiroshige; Ihara, Tomonori

2016-01-01

This study aims to develop ultrasonic Doppler velocimetry in molten glass. Realization of such a technique has two difficulties: ultrasonic transmission into molten salt and Doppler signal processing. Buffer rod technique was developed in our research to transmit ultrasound into high temperature molten glass. This article discusses newly developed signal processing technique named broadband phase difference method. (J.P.N.)

A review of the core catcher design in LMR

International Nuclear Information System (INIS)

Lee, Yong Bum; Hahn, Do Hee

2001-08-01

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

Experimental studies on natural circulation in molten salt loops

International Nuclear Information System (INIS)

Srivastava, A.K.; Borgohain, A.; Maheshwari, N.K.; Vijayan, P.K.

2015-01-01

Molten salts are increasingly getting attention as a coolant and storage medium in solar thermal power plants and as a liquid fuel, blanket and coolant in Molten Salt Reactors (MSR’s). Two different test facilities named Molten Salt Natural Circulation Loop (MSNCL) and Molten Active Fluoride salt Loop (MAFL) have been setup for thermal hydraulics, instrument development and material related studies relevant to MSR and solar power plants. The working medium for MSNCL is a molten nitrate salt which is a mixture of NaNO 3 and KNO 3 in 60:40 ratio and proposed as one of the coolant option for molten salt based reactor and coolant as well as storage medium for solar thermal power application. On the other hand, the working medium for MAFL is a eutectic mixture of LiF and ThF 4 and proposed as a blanket salt for Indian Molten Salt Breeder Reactor (MSBR). Steady state natural circulation experiments at different power level have been performed in the MSNCL. Transient studies for startup of natural circulation, loss of heat sink, heater trip and step change in heater power have also been carried out in the same. A 1D code LeBENC, developed in-house to simulate the natural circulation characteristics in closed loops, has been validated with the experimental data obtained from MSNCL. Further, LeBENC has been used for Pretest analysis of MAFL. This paper deals with the description of both the loops and experimental studies carried out in MSNCL. Validation of LeBENC along with the pretest analysis of MAFL using the same are also reported in this paper. (author)

Compatibility tests between molten salts and metal materials (2)

International Nuclear Information System (INIS)

Shiina, Yasuaki

2003-08-01

Latent heat storage technology using molten salts can reduce temperature fluctuations of heat transfer fluid by latent heat for middle and high temperature regions. This enables us to operate several heat utilization systems in cascade connected to High Temperature Gas Cooled Reactors (HTGRs) from high to low temperature range by setting the latent heat storage system after a heat utilization system to reduce thermal load after the heat utilization systems. This latent heat technology is expected to be used for effective use of heat such as equalization of electric load between night and daytime. In the application of the latent heat technology, compatibility between molten salts and metal materials is very important because molten salts are corrosive, and heat transfer pipes and vessels will contact with the molten salts. It will be necessary to prevail the latent heat storage technique that normal metal materials can be used for the pipes and vessels. However, a few studies have been reported of compatibility between molten salts and metals in middle and high temperature ranges. In this study, four molten salts, range of the melting temperature from 490degC to 800degC, are selected and five metals, high temperature and corrosion resistance steels of Alloy600, HastelloyB2, HastelloyC276, SUS310S and pure Nickel are selected for the test with the consideration of metal composition. Test was performed in an electric furnace by setting the molten salts and the metals in melting pots in an atmosphere of nitrogen. Results revealed excellent corrosion resistance of pure Nickel and comparatively low corrosion resistance of nickel base alloys such as Alloy600 and Hastelloys against Li 2 CO 3 . Corrosion resistance of SUS310S was about same as nickel based alloys. Therefore, if some amount of corrosion is permitted, SUS310S would be one of the candidate alloys for structure materials. These results will be used as reference data to select metals in latent heat technology

The earths innermost core

International Nuclear Information System (INIS)

Nanda, J.N.

1989-01-01

A new earth model is advanced with a solid innermost core at the centre of the Earth where elements heavier than iron, over and above what can be retained in solution in the iron core, are collected. The innermost core is separated from the solid iron-nickel core by a shell of liquid copper. The innermost core has a natural vibration measured on the earth's surface as the long period 26 seconds microseisms. The earth was formed initially as a liquid sphere with a relatively thin solid crust above the Byerly discontinuity. The trace elements that entered the innermost core amounted to only 0.925 ppm of the molten mass. Gravitational differentiation must have led to the separation of an explosive thickness of pure 235 U causing a fission explosion that could expel beyond the Roche limit a crustal scab which would form the centre piece of the moon. A reservoir of helium floats on the liquid copper. A small proportion of helium-3, a relic of the ancient fission explosion present there will spell the exciting magnetic field. The field is stable for thousands of years because of the presence of large quantity of helium-4 which accounts for most of the gaseous collisions that will not disturb the atomic spin of helium-3 atoms. This field is prone to sudden reversals after long periods of stability. (author). 14 refs

An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Aufiero, M.; Cammi, A.; Fiorina, C. [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy); Leppänen, J. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland); Luzzi, L., E-mail: lelio.luzzi@polimi.it [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy); Ricotti, M.E. [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy)

2013-10-15

In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

Experimental study on forced convection boiling heat transfer on molten alloy

International Nuclear Information System (INIS)

Nishimura, Satoshi; Ueda, Nobuyuki; Nishi, Yoshihisa; Furuya, Masahiro; Kinoshita, Izumi

1999-01-01

In order to clarify the characteristics of forced convection boiling heat transfer on molten metal, basic experiments have been carried out with subcooled water flowing on molten Wood's alloy pool surface. In these experiments, water flows horizontally in a rectangular duct. A cavity filled with Wood's alloy is present in a portion of the bottom of the duct. Wood's alloy is heated by a copper conductor at the bottom of the cavity. The experiments have been carried out with various velocities and subcoolings of water, and temperature of Wood's alloy. Boiling curves on the molten alloy surface were obtained and compared with that on a solid heat transfer surface. It is observed that the boiling curve on molten alloy is in a lower superheat region than the boiling curve on a solid surface. This indicates that the heat transfer performance of forced convection boiling on molten alloy is enhanced by increase of the heat transfer area, due to oscillation of the surface and fragmentation of molten alloy

Compatibility of AlN ceramics with molten lithium

Energy Technology Data Exchange (ETDEWEB)

Yoneoka, Toshiaki; Sakurai, Toshiharu; Sato, Toshihiko; Tanaka, Satoru [Tokyo Univ., Department of Quantum Engineering and Systems Science, Tokyo (Japan)

2002-04-01

AlN ceramics were a candidate for electrically insulating materials and facing materials against molten breeder in a nuclear fusion reactor. In the nuclear fusion reactor, interactions of various structural materials with solid and liquid breeder materials as well as coolant materials are important. Therefore, corrosion tests of AlN ceramics with molten lithium were performed. AlN specimens of six kinds, different in sintering additives and manufacturing method, were used. AlN specimens were immersed into molten lithium at 823 K. Duration for the compatibility tests was about 2.8 Ms (32 days). Specimens with sintering additive of Y{sub 2}O{sub 3} by about 5 mass% formed the network structure of oxide in the crystals of AlN. It was considered that the corrosion proceeded by reduction of the oxide network and the penetration of molten lithium through the reduced pass of this network. For specimens without sintering additive, Al{sub 2}O{sub 3} containing by about 1.3% in raw material was converted to fine oxynitride particles on grain boundary or dissolved in AlN crystals. After immersion into lithium, these specimens were found to be sound in shape but reduced in electrical resistivity. These degradation of the two types specimens were considered to be caused by the reduction of oxygen components. On the other hand, a specimen sintered using CaO as sintering additive was finally became appreciably high purity. This specimen showed good compatibility for molten lithium at least up to 823 K. It was concluded that the reduction of oxygen concentration in AlN materials was essential in order to improve the compatibility for molten lithium. (author)

78 FR 2319 - Relocation of Regulations

Science.gov (United States)

2013-01-11

...) operations; and Bank assistance for the Resolution Funding Corporation (RefCorp). This final rule relocates... Board Part 996/FHFA Part 1271--Subpart E (Authority for Bank Assistance of the Resolution Funding... National Mortgage Association and any affiliate thereof, and any Federal Home Loan Bank. Resolution Funding...

Investigation of molten salt fast reactor

International Nuclear Information System (INIS)

Kubota, Kenichi; Konomura, Mamoru

2002-01-01

On survey research for practicability strategy of fast reactor (FR) (phase 1), to extract future practicability image candidates of FR from wide options, in addition to their survey and investigation objects of not only solid fuel reactors of conventional research object but also molten salt reactor as a flowing fuel reactor, investigation on concept of molten salt FR plant was carried out. As a part of the first step of the survey research for practicability strategy, a basic concept on plant centered at nuclear reactor facility using chloride molten salt reactor capable of carrying out U-Pu cycle was examined, to perform a base construction to evaluate economical potential for a practical FBR. As a result, a result could be obtained that because of inferior fuel inventory and heat transmission to those in Na cooling reactor in present knowledge, mass of reactor vessel and intermediate heat exchanger were to widely increased to expect reduction of power generation unit price even on considering cheapness of its fuel cycle cost. Therefore, at present step further investigation on concept design of the chloride molten salt reactor plant system is too early in time, and it is at a condition where basic and elementary researches aiming at upgrading of economical efficiency such as wide reduction of fuel inventory, a measure expectable for remarkable rationalization effect of reprocessing system integrating a reactor to a processing facility, and so on. (G.K.)

Metalcasting: Filtering Molten Metal

International Nuclear Information System (INIS)

Lauren Poole; Lee Recca

1999-01-01

A more efficient method has been created to filter cast molten metal for impurities. Read about the resulting energy and money savings that can accrue to many different industries from the use of this exciting new technology

Relocation of belt conveyors along the final slope of the Belacevac surface coal mine

Energy Technology Data Exchange (ETDEWEB)

Maksimovic, N; Rosic, Z

1987-07-01

Describes how following a landslide on the northern wing of the Belacevac surface coal mine the belt conveyor lines had to be relocated in order to assure a reliable coal supply to the Kosovo A and Kosovo B thermal power stations. The relocation was achieved in three phases with new approach cuts being made, necessitating the removal of 280,000 m/sup 3/ of overburden in the first phase and 870,000 m/sup 3/ in the second phase of the reconstruction. Illustrates the relocation of the conveyor system by means of site plans and notes that the production of coal and the removal of overburden were not interrupted during the relocation exercise. 2 refs.

Molten metal feed system controlled with a traveling magnetic field

International Nuclear Information System (INIS)

Praeg, W.F.

1991-01-01

This patent describes a continuous metal casting system in which the feed of molten metal controlled by means of a linear induction motor capable of producing a magnetic traveling wave in a duct that connects a reservoir of molten metal to a caster. The linear induction motor produces a traveling magnetic wave in the duct in opposition to the pressure exerted by the head of molten metal in the reservoir

Visualization study of molten metal-water interaction by using neutron radiography

International Nuclear Information System (INIS)

Mishima, K.; Hibiki, T.; Saito, Y.

1999-01-01

The purpose of this study is to visualize the behavior of molten metal dropped into water during the premixing process by means of neutron radiography which makes use of the difference in the attenuation characteristics of materials. For this purpose, a high-sensitive, high-frame-rate imaging system using neutron radiography was constructed and was applied to visualization of the behavior of molten metal dropped into water. The test rig consisted of a furnace and a test section. The furnace could heat the molten metal up to 650 C. The test section was a rectangular tank made of aluminum alloy. The tank was filled with heavy water and molten Wood's metal was dropped into heavy water. Visualization study was carried out with use of the high-frame-rate neutron radiography to see the breakup of molten metal jet or lump dropped into heavy water pool. In the images obtained, water, steam or air bubbles, molten metal jets or droplets, cloud of small particles of molten metal after atomization could be distinguished. The debris of Wood's metal was collected after the experiment, and the relation between the break-up behavior and the size and the shape of the debris particles was investigated. (orig.)

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)

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.

Tritium loss in molten flibe systems

Energy Technology Data Exchange (ETDEWEB)

Longhurst, G.R.; Anderl, R.A. [Idaho National Eng. and Environ. Lab., Idaho Falls, ID (United States); Scott Willms, R. [Los Alamos National Lab., NM (United States)

2000-04-01

An emerging issue relative to beryllium technology in fusion involves tritium interactions with molten beryllium-bearing salts. Innovative designs for fusion reactors, both magnetic and inertially confined, feature the molten salt mixture 2LiF.BeF{sub 2}, commonly called Flibe, as a tritium breeder and coolant. Tritium is bred in the Flibe as neutrons from the plasma are absorbed by Li atoms, which then transmute to tritium and helium. Transmutation of tritium from Be also occurs. Among the issues to be resolved for such coolant systems is the potential loss of tritium from the Flibe coolant to the walls of the system, particularly through heat exchanger tubes, and from there into secondary coolants or working fluids and the environment. Effectively removing tritium from Flibe in clean-up units is also important. In quiescent or low Reynolds number flow, tritium movement through Flibe is governed by diffusion. For Flibe in turbulent flow, as in heat exchanger tubes, transport is by turbulent mixing, and the same flow conditions and structural design features that maximize heat transfer to the heat exchanger walls will enhance the transport of tritium to those same surfaces. Analyses have been performed to estimate the fractional loss of tritium through heat exchanger tubes and release rates from Flibe droplets in vacuum disengagers in molten Flibe systems. The calculations suggest unacceptably large losses of tritium through heat exchanger tubes. The gravity of the implications of these estimates calls for experimental verification to determine if tritium losses through molten Flibe heat exchangers or other Flibe systems can really be so high and whether vacuum disengagers will really work. There is also a need for better information on evolution of tritium from Flibe droplets in a vacuum. Several experiments are presently being planned to address these issues and are discussed. These include experiments to induce tritium in Flibe using spontaneous fission neutrons

Tritium loss in molten flibe systems

International Nuclear Information System (INIS)

Longhurst, G.R.; Anderl, R.A.; Scott Willms, R.

2000-01-01

An emerging issue relative to beryllium technology in fusion involves tritium interactions with molten beryllium-bearing salts. Innovative designs for fusion reactors, both magnetic and inertially confined, feature the molten salt mixture 2LiF.BeF 2 , commonly called Flibe, as a tritium breeder and coolant. Tritium is bred in the Flibe as neutrons from the plasma are absorbed by Li atoms, which then transmute to tritium and helium. Transmutation of tritium from Be also occurs. Among the issues to be resolved for such coolant systems is the potential loss of tritium from the Flibe coolant to the walls of the system, particularly through heat exchanger tubes, and from there into secondary coolants or working fluids and the environment. Effectively removing tritium from Flibe in clean-up units is also important. In quiescent or low Reynolds number flow, tritium movement through Flibe is governed by diffusion. For Flibe in turbulent flow, as in heat exchanger tubes, transport is by turbulent mixing, and the same flow conditions and structural design features that maximize heat transfer to the heat exchanger walls will enhance the transport of tritium to those same surfaces. Analyses have been performed to estimate the fractional loss of tritium through heat exchanger tubes and release rates from Flibe droplets in vacuum disengagers in molten Flibe systems. The calculations suggest unacceptably large losses of tritium through heat exchanger tubes. The gravity of the implications of these estimates calls for experimental verification to determine if tritium losses through molten Flibe heat exchangers or other Flibe systems can really be so high and whether vacuum disengagers will really work. There is also a need for better information on evolution of tritium from Flibe droplets in a vacuum. Several experiments are presently being planned to address these issues and are discussed. These include experiments to induce tritium in Flibe using spontaneous fission neutrons

Molten salt scrubbing of zirconium or hafnium tetrachloride

International Nuclear Information System (INIS)

Lee, E.D.; McLaughlin, D.F.

1990-01-01

This patent describes a continuous process for removing impurities of iron or aluminum chloride or both from vaporous zirconium or hafnium chloride or both. It comprises: introducing impure zirconium or hafnium chloride vapor or both into a middle portion of an absorbing column containing a molten salt phase, the molten salt phase absorbing the impurities of iron or aluminum chloride or both to produce chloride vapor stripped of zirconium or hafnium chloride; introducing sodium or potassium chloride or both into a top portion of the column; controlling the top portion of the column to between 300--375 degrees C.; heating a bottom portion of the column to 450--550 degrees C. To vaporize zirconium chloride or hafnium chloride or hafnium and zirconium chloride from the molten salt; withdrawing molten salt substantially free of zirconium and hafnium chloride from the bottom portion of the column; and withdrawing zirconium chloride or hafnium chloride or hafnium and zirconium chloride vapor substantially free of impurities of iron and aluminum chloride from the top of the column

The Earth's core: its composition, formation and bearing upon the origin of the Earth

International Nuclear Information System (INIS)

Ringwood, A.E.

1984-01-01

The density of the outer core is about 3% smaller than pure iron, which implies that the core contains a substantial amount of one or more low atomic mass elements. New experimental data on the solubility of FeO in molten iron are compatible with oxygen being the other element. At atmospheric pressure FeO is extensively soluble in iron at 2500 0 C, completely miscible above 2800 0 C. Also the solubility of FeO in molten iron is considerably increased at higher pressures. The density measurements can be explained if the core contains about 35% FeO; the new data show this to be possible. A model for the formation of the core based on a high FeO content in the Bulk Earth can be explained if the Earth accreted from a mixture of two components: A, a highly reduced, metal-rich devolatilized assemblage and B, a highly oxidized, volatile-rich assemblage similar to Cl chondrites. The formation of these components in the solar nebula is discussed. The large amount of FeO now inferred to be present in the Earth was mainly produced during accretion by oxidation of metallic iron from component A by water from component B. (U.K.)

Molar Volume Analysis of Molten Ni-Al-Co Alloy by Measuring the Density

Institute of Scientific and Technical Information of China (English)

XIAO Feng; FANG Liang; FU Yuechao; YANG Lingchuan

2004-01-01

The density of molten Ni-Al-Co alloys was measured in the temperature range of 1714～1873K using a modified pycnometric method, and the molar volume of molten alloys was analyzed. The density of molten Ni-Al-Co alloys was found to decrease with increasing temperature and Co concentration in alloys. The molar volume of molten Ni-Al-Co alloys increases with increasing Co concentration in alloys. The molar volume of molten Ni-Al-Co alloys shows a negative deviation from the linear molar volume.

Interaction of calcium oxide with molten alkali metal chlorides

International Nuclear Information System (INIS)

Volkovich, A.V.; Zhuravlev, V.I.; Ermakov, D.S.; Magurina, M.V.

1999-01-01

Calcium oxide solubility in molten lithium, sodium, potassium, cesium chlorides and their binary mixtures is determined in a temperature range of 973-1173 K by the method of isothermal saturation. Mechanisms of calcium oxide interaction with molten alkali metal chlorides are proposed

Performance Analysis of a Forecasting Relocation Model for One-Way Carsharing

Directory of Open Access Journals (Sweden)

Ganjar Alfian

2017-06-01

Full Text Available A carsharing service can be seen as a transport alternative between private and public transport that enables a group of people to share vehicles based at certain stations. The advanced carsharing service, one-way carsharing, enables customers to return the car to another station. However, one-way implementation generates an imbalanced distribution of cars in each station. Thus, this paper proposes forecasting relocation to solve car distribution imbalances for one-way carsharing services. A discrete event simulation model was developed to help evaluate the proposed model performance. A real case dataset was used to find the best simulation result. The results provide a clear insight into the impact of forecasting relocation on high system utilization and the reservation acceptance ratio compared to traditional relocation methods.

Thermal Characterization of Molten Salt Systems

Energy Technology Data Exchange (ETDEWEB)

Toni Y. Gutknecht; Guy L. Fredrickson

2011-09-01

The phase stability of molten salts in an electrorefiner (ER) may be adversely affected by the buildup of sodium, fission products, and transuranics in the electrolyte. Potential situations that need to be avoided are the following: (1) salt freezing due to an unexpected change in the liquidus temperature, (2) phase separation or non-homogeneity of the molten salt due to the precipitation of solids or formation of immiscible liquids, and (3) any mechanism that can result in the separation and concentration of fissile elements from the molten salt. Any of these situations would result in an off-normal condition outside the established safety basis for electrorefiner (ER) operations. The stability (and homogeneity) of the phases can potentially be monitored through the thermal characterization of the salts, which can be a function of impurity concentration. This report describes the experimental results of typical salts compositions, which consist of chlorides of potassium, lithium, strontium, samarium, praseodymium, lanthanum, barium, cerium, cesium, neodymium, sodium and gadolinium chlorides as a surrogate for both uranium and plutonium, used for the processing of used nuclear fuels.

Deep-Earth Equilibration between Molten Iron and Solid Silicates

Science.gov (United States)

Brennan, M.; Zurkowski, C. C.; Chidester, B.; Campbell, A.

2017-12-01

Elemental partitioning between iron-rich metals and silicate minerals influences the properties of Earth's deep interior, and is ultimately responsible for the nature of the core-mantle boundary. These interactions between molten iron and solid silicates were influential during planetary accretion, and persist today between the mantle and liquid outer core. Here we report the results of diamond anvil cell experiments at lower mantle conditions (40 GPa, >2500 K) aimed at examining systems containing a mixture of metals (iron or Fe-16Si alloy) and silicates (peridotite). The experiments were conducted at pressure-temperature conditions above the metallic liquidus but below the silicate solidus, and the recovered samples were analyzed by FIB/SEM with EDS to record the compositions of the coexisting phases. Each sample formed a three-phase equilibrium between bridgmanite, Fe-rich metallic melt, and an oxide. In one experiment, using pure Fe, the quenched metal contained 6 weight percent O, and the coexisting oxide was ferropericlase. The second experiment, using Fe-Si alloy, was highly reducing; its metal contained 10 wt% Si, and the coexisting mineral was stishovite. The distinct mineralogies of the two experiments derived from their different starting metals. These results imply that metallic composition is an important factor in determining the products of mixed phase iron-silicate reactions. The properties of deep-Earth interfaces such as the core-mantle boundary could be strongly affected by their metallic components.

Facile preparation of highly pure KF-ZrF4 molten salt

Science.gov (United States)

Zong, Guoqiang; Cui, Zhen-Hua; Zhang, Zhi-Bing; Zhang, Long; Xiao, Ji-Chang

2018-03-01

The preparation of highly pure KF-ZrF4 (FKZr) molten salt, a potential secondary coolant in molten salt reactors, was realized simply by heating a mixture of (NH4)2ZrF6 and KF. X-ray diffraction analysis indicated that the FKZr molten salt was mainly composed of KZrF5 and K2ZrF6. The melting point of the prepared FKZr molten salt was 420-422 °C under these conditions. The contents of all metal impurities were lower than 20 ppm, and the content of oxygen was lower than 400 ppm. This one-step protocol avoids the need for a tedious procedure to prepare ZrF4 and for an additional purification process to remove oxide impurities, and is therefore a convenient, efficient and economic preparation method for high-purity FKZr molten salt.

Improvements in modelling (by ESCADRE mod1.0) radiative heat losses through gas and aerosols generated by molten corium-concrete interactions

International Nuclear Information System (INIS)

Passalacqua, R.

1996-01-01

Aerosols generated during the molten core-concrete interaction (MCCI) influence the reactor cavity thermal hydraulics: the cloud of aerosols, located inside the reactor cavity, restrains the upward-directed heat exchange consequently the cool-down of the high-temperature molten corium for a considerable period of time. IPSN is developing a computer code system for source predictions in severe accident scenarios. This code system is named ESCADRE. WECHSL/CALTHER is internal module dealing with MCCI (it is also a stand-alone code): it models the heat transfers involving the superior volume of the cavity. When modelling the upward-directed power distribution by WECHSL/CALTHER, a faster concrete basemat penetration takes place due to the low heat losses of the closed MCCI cavity enclosure. The model, here presented, is going to be validated with data from the AEROSTAT experiment. This experiment, planned at CEA Cadarache, will evaluate the influence of aerosols on the global power distribution in the reactor cavity. Radiative heat losses are important especially for cavity configurations such as those of new plant designs (equipped with a core-catcher) where the upward power losses are promoted by the corium spreading in a flat cavity

Molten Salts for High Temperature Reactors: University of Wisconsin Molten Salt Corrosion and Flow Loop Experiments -- Issues Identified and Path Forward

Energy Technology Data Exchange (ETDEWEB)

Piyush Sabharwall; Matt Ebner; Manohar Sohal; Phil Sharpe; Thermal Hydraulics Group

2010-03-01

Considerable amount of work is going on regarding the development of high temperature liquid salts technology to meet future process needs of Next Generation Nuclear Plant. This report identifies the important characteristics and concerns of high temperature molten salts (with lesson learned at University of Wisconsin-Madison, Molten Salt Program) and provides some possible recommendation for future work

Cracking and relocation of UO2 fuel during nuclear operation

International Nuclear Information System (INIS)

Appelhans, A.D.; Dagbjartsson, S.J.

1981-01-01

Cracking and relocation of light water reactor (LWR) fuel pellets affect the axial gas flow path within nuclear reactor fuel rods and the thermal performance of the fuel. As part of the Nuclear Regulatory Commission's Water Reactor Safety Research Fuel Behavior Program, the Thermal Fuels Behavior Program of EG and G Idaho, Inc., is conducting fuel rod behavior studies in the Heavy Boiling Water Reactor in Halden, Norway. The Instrumental Fuel Assembly-430 (IFA-430) operated in that facility is a multipurpose assembly designed to provide information on fuel cracking and relocation, the long-term thermal response of LWR fuel rods subjected to various internal pressures and gas compositions, and the release of fission gases. This report presents the results of an analysis of fuel cracking and relocation phenomena as deduced from fuel rod axial gas flow and fuel temperature data from the first 6.5 GWd/tUO 2 burnup of the IFA-430

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

Advanced heat exchanger development for molten salts

Energy Technology Data Exchange (ETDEWEB)

Sabharwall, Piyush, E-mail: Piyush.Sabharwall@inl.gov [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Clark, Denis; Glazoff, Michael [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Zheng, Guiqiu; Sridharan, Kumar; Anderson, Mark [University of Wisconsin, Madison (United States)

2014-12-15

Highlights: • Hastelloy N and 242, shows corrosion resistance to molten salt at nominal operating temperatures. • Both diffusion welds and sheet material in Hastelloy N were corrosion tested in at 650, 700, and 850 °C for 200, 500, and 1000 h. • Thermal gradients and galvanic couples in the molten salts enhance corrosion rates. • Corrosion rates found were typically <10 mils per year. - Abstract: This study addresses present work concerned with advanced heat exchanger development for molten salt in nuclear and non-nuclear thermal systems. The molten salt systems discussed herein use alloys, such as Hastelloy N and 242, that show good corrosion resistance in molten salt at nominal operating temperatures up to 700 °C. These alloys were diffusion welded, and the corresponding information is presented. Test specimens were prepared for exposing diffusion welds to molten salt environments. Hastelloy N and 242 were found to be weldable by diffusion welding, with ultimate tensile strengths about 90% of base metal values. Both diffusion welds and sheet material in Hastelloy N were corrosion tested in 58 mol% KF and 42 mol% ZrF{sub 4} at 650, 700, and 850 °C for 200, 500, and 1000 h. Corrosion rates were similar between welded and nonwelded materials, typically <100 μm per year after 1000 h of corrosion tests. No catastrophic corrosion was observed in the diffusion welded regions. For materials of construction, nickel-based alloys and alloys with dense nickel coatings are effectively inert to corrosion in fluorides, but not so in chlorides. Hence, additional testing of selected alloys for resistance to intergranular corrosion is needed, as is a determination of corrosion rate as a function of the type of salt impurity and alloy composition, with respect to chromium and carbon, to better define the best conditions for corrosion resistance. Also presented is the division of the nuclear reactor and high-temperature components per American Society of Mechanical

Natural convection heat transfer in the molten metal pool

International Nuclear Information System (INIS)

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

1997-01-01

Analytical studies using the FLOW-3D computer program have been performed on natural convection heat transfer of a high density molten metal pool, in order to evaluate the coolability of the corium pool. The FLOW-3D results on the temperature distribution and the heat transfer rate in the molten metal pool region have been compared and evaluated with the experimental data. The FLOW-3D results have shown that the developed natural convection flow contributes to the solidified crust formation of the high density molten metal pool. The present FLOW-3D results, on the relationship between the Nusselt number and the Rayleigh number in the molten metal pool region, are more similar to the calculated results of Globe and Dropkin's correlation than any others. The natural convection heat transfer in the low aspect ratio case is more substantial than that in the high aspect ratio case. The FLOW-3D results, on the temperature profile and on the heat transfer rate in the molten metal pool region, are very similar to the experimental data. The heat transfer rate of the internal heat generation case is higher than that of the bottom heating case at the same heat supply condition. (author)

Postaccident heat removal: large-scale molten-fuel-sodium interaction experiments

International Nuclear Information System (INIS)

Johnson, T.R.; Pavlik, J.R.; Baker, L. Jr.

1975-02-01

Kilogram-scale interactions between molten UO 2 and sodium were performed in an unrestrained geometry to study the resulting energetics and fragmentation. The molten UO 2 was producted by the exothrmic reaction between uranium and MoO 3 powders. Under the conditions of the experiments completed to date, the short-rise-time pressure pulses created in the liquid phase had negligible work potential, and their magnitude did not increase with the amount of molten fuel. No significant gas-phase shock pressures were generated. The largest potential for mechanical work was the sodium vapor generated over a period of roughly 1 sec. About 20 percent of the heat was effective in generating vapor. The ex- perimental results show a marked tendency of molten UO 2 to form particulate after passage through only a few inches of sodium. Particle size distributions obtained under the conditions of the experiments were not significantly different from those obtained in prior small-scale tests and in TREAT tests. Also, the results indicate that the metallic component of the molten mixture formed larger particles than the oxide component. (U.S.)