WorldWideScience

Sample records for molten core relocation

  1. Thermal behavior of molten corium during TMI-2 core relocation event

    International Nuclear Information System (INIS)

    Anderson, J.L.; Sienicki, J.J.

    1988-01-01

    During the TMI-2 accident, a pool of molten corium formed in the central region of the core and was contained by solidified crusts. Failure of the crust surrounding the molten material, at approximately 224 min, resulted in a relocation of an estimated 20-25 tons of molten corium through peripheral fuel assemblies in the east side of the vessel, as well as through the core barrel assembly (CBA) at the periphery of the core. This paper presents the results of an analyses carried out to investigate the thermal interactions of molten corium with the CBA structures during the relocation event. The principal objectives of the analyses are: (a) to assess the potential for relocation to take place through the CBA versus the flow of molten core material directly downward through the core via the fuel assemblies; and (b) to understand the distribution of prior molten corium observed during vessel defueling examinations. 5 refs., 1 fig

  2. Thermophysical, hydrodynamic and mechanical aspects of molten core relocation to lower plenum

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Kune Y.; Huh, Chang Wook [Seoul National University, Seoul (Korea, Republic of)

    1997-12-31

    This paper presents the current state of knowledge on molten material relocation into the lower plenum. Consequences of movement of material to the lower head are considered with regard to the potential for reactor pressure vessel failure from both thermal hydraulic and mechanical standpoints. The models are applied to evaluating various in-vessel retention strategies for the Korean Standard Power Plant (KSNPP) reactor. The results are summarized in terms of thermal response of the reactor vessel from the very relevant severe accident management perspective. 10 refs., 1 fig., 1 tab. (Author)

  3. Thermophysical, hydrodynamic and mechanical aspects of molten core relocation to lower plenum

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Kune Y; Huh, Chang Wook [Seoul National University, Seoul (Korea, Republic of)

    1998-12-31

    This paper presents the current state of knowledge on molten material relocation into the lower plenum. Consequences of movement of material to the lower head are considered with regard to the potential for reactor pressure vessel failure from both thermal hydraulic and mechanical standpoints. The models are applied to evaluating various in-vessel retention strategies for the Korean Standard Power Plant (KSNPP) reactor. The results are summarized in terms of thermal response of the reactor vessel from the very relevant severe accident management perspective. 10 refs., 1 fig., 1 tab. (Author)

  4. Molten material relocation into the lower plenum: a status report

    International Nuclear Information System (INIS)

    1998-09-01

    This report, prepared by the task group 'Degraded Core Cooling' (DCC) for the CSNI, summarizes the experimental and theoretical knowledge of molten material relocation from a degraded core to the lower plenum of the reactor vessel under the main severe accident scenarios envisaged for both PWRs and BWRs, and boundary conditions. Consequences of movement of material to the lower head are considered with respect to the potential for reactor pressure vessel failure. The following models are reviewed: SCDAP/RELAP5, ICARE/CATHARE, ATHLET-CD/KESS, MELCOR, MAAP4, ESCADRE, etc.

  5. Molten core retention assembly

    International Nuclear Information System (INIS)

    Lampe, R.F.

    1976-01-01

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

  6. Evaluation of downmotion time interval molten materials to core catcher during core disruptive accidents postulated in LMFR

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  7. Heatup of the TMI-2 lower head during core relocation

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  8. 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

  9. The TMI-2 core relocation: Heat transfer and mechanism

    International Nuclear Information System (INIS)

    Epstein, M.; Fauske, H.K.

    1987-07-01

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

  10. Current status of investigations on molten fuel: Coolant interaction, material movement and relocation in LMFBRs in Russia

    International Nuclear Information System (INIS)

    Buksha, Yu.; Kuznetsov, I.

    1994-01-01

    The paper contains information on experimental studies and calculation codes, related to molten fuel-coolant interaction, material movement and relocation. Some calculation results for the BN-800 type reactor are presented. (author)

  11. 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

  12. Model development of UO_2-Zr dispersion plate-type fuel behavior at early phase of severe accident and molten fuel meat relocation

    International Nuclear Information System (INIS)

    Zhang Zhuohua; Yu Junchong; Peng Shinian

    2014-01-01

    According to former study on oxygen diffusion, Nb-Zr solid reaction and UO_2-Zr solid reaction, the models of oxidation, solid reaction in fuel meat and relocation of molten fuel meat are developed based on structure and material properties of UO_2-Zr dispersion plate-type fuel, The new models can supply theoretical elements for the safety analysis of the core assembled with dispersion plate-type fuel under severe accident. (authors)

  13. Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Tobita, Yoshiharu; Kamiyama, Kenji; Tagami, Hirotaka; Matsuba, Ken-ichi; Suzuki, Tohru; Isozaki, Mikio; Yamano, Hidemasa; Morita, Koji; Guo, Liancheng; Zhang, Bin

    2014-01-01

    The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior. The analysis method of molten material discharge was developed based on the computer code SIMMER-III since this code is designed to simulate the multi-phase, multi-component fluid dynamics with phase changes involved in the discharge process. Several experiments simulating the molten material discharge through duct using simulant materials were utilized as the basis of validation study of the physical models in this code. It was shown that SIMMER-III with improved physical models could simulate the molten material discharge behavior including the momentum exchange with duct wall and thermal interaction with coolant. In order to develop evaluation methodology of molten material penetration into sodium pool, a series of experiments simulating jet penetration behavior into sodium pool in SFR thermal condition were performed. These experiments revealed that the molten jet was fragmented in significantly shorter penetration length than the prediction by existing correlation for light water reactor conditions, due to the direct

  14. Molten core material holding device in a nuclear reactor

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  15. 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

  16. Thermohydraulic behaviour and heat transfer in the molten core

    International Nuclear Information System (INIS)

    Reineke, H.H.

    1977-01-01

    Increasing the application of nuclear reactors to produce electrical power extremely unprobable accidents should be investigated too. In the Federal Republic of Germany, a research program is performed for some years engaged in accidents at light water reactors in which the melting of the reactor core is presumed. A part of this program is to investigate the thermohydraulic and the heat transfer behavior in an accumulation of molten core material. The knowledge of these events is necessary to analyse the accident exactly. Further on the results of this work are of great importance to build a catcher for the molten core material. As a result of the decay heat the molten material is heated up and the density differences induce a free convection motion. In this work the thermohydraulic behavior and the distribution of the escaping heat fluxes for several accumulations of molten core material were determined. The numerical methods for solving the system of partial differential equation were used to develop computer codes, able to compute the average and local heat fluxes at the walls enclosing the molten core material and the inside increase of the temperature. The numerical computations were confirmed and verified by experimental investigations. In these investigations the molten core material was always assumed as a homogeneous fluid. In this case, the results could be reproduced by simple power laws

  17. Molten core debris-sodium interactions: M-Series experiments

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  18. Fundamental experiment on simulated molten core/concrete interaction

    International Nuclear Information System (INIS)

    Toda, S.; Katsumura, Y.

    1994-01-01

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

  19. 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.

  20. 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

  1. Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Tobita, Yoshiharu; Kamiyama, Kenji; Tagami, Hirotaka; Matsuba, Ken-ichi; Suzuki, Tohru; Isozaki, Mikio; Yamano, Hidemasa; Morita, Koji; Guo, LianCheng; Zhang, Bin

    2016-01-01

    The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt-through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior. The analysis method of molten material discharge was developed based on the computer code SIMMER-III since this code is designed to simulate the multi-phase, multi-component fluid dynamics with phase changes involved in the discharge process. Several experiments simulating the molten material discharge through duct using simulant materials were utilized as the basis of validation study of the physical models in this code. It was shown that SIMMER-III with improved physical models could simulate the molten material discharge behavior, including the momentum exchange with duct wall and thermal interaction with coolant. In order to develop an evaluation methodology of molten material penetration into sodium pool, a series of experiments simulating jet penetration behavior into sodium pool in SFR thermal condition were performed. These experiments revealed that the molten jet was fragmented in significantly shorter penetration length than the prediction by existing correlation for light water reactor conditions, due to the direct

  2. 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

  3. 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

  4. 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

  5. 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.

  6. 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

  7. 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.

  8. 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

  9. 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

  10. 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

  11. 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.

  12. 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

  13. 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)

  14. MARTINS: A foam/film flow model for molten material relocation in HWRs with U-Al-fueled multi-tube assemblies

    International Nuclear Information System (INIS)

    Kalimullah.

    1994-01-01

    Some special purpose heavy-water reactors (EM) are made of assemblies consisting of a number of coaxial aluminum-clad U-Al alloy fuel tubes and an outer Al sleeve surrounding the fuel tubes. The heavy water coolant flows in the annular gaps between the circular tubes. Analysis of severe accidents in such reactors requires a model for predicting the behavior of the fuel tubes as they melt and disrupt. This paper describes a detailed, mechanistic model for fuel tube heatup, melting, freezing, and molten material relocation, called MARTINS (Melting and Relocation of Tubes in Nuclear subassembly). The paper presents the modeling of the phenomena in MARTINS, and an application of the model to analysis of a reactivity insertion accident. Some models are being developed to compute gradual downward relocation of molten material at decay-heat power levels via candling along intact tubes, neglecting coolant vapor hydrodynamic forces on molten material. These models are inadequate for high power accident sequences involving significant hydrodynamic forces. These forces are included in MARTINS

  15. 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

  16. 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

  17. 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.

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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.

  4. 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

  5. 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.

  6. 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.)

  7. 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

  8. 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

  9. 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

  10. 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.

  11. 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

  12. 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

  13. 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.

  14. 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

  15. 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)

  16. 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

  17. 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.

  18. Technical committee meeting on material-coolant interactions and material movement and relocation in liquid metal fast reactors

    International Nuclear Information System (INIS)

    1994-01-01

    The Technical Committee Meeting on Material-Coolant Interactions and Material Movement and Relocation in Liquid Metal Fast Reactors was sponsored by the International Working Group on Fast Reactors (IWGFR), International Atomic Energy Agency (IAEA) and hosted by PNC, on behalf of the Japanese government. A broad range of technical subjects was discussed in the TCM, covering entire aspects of material motion and interactions relevant to the safety of LMFRs. Recent achievement and current status in research and development in this area were presented including European out-of-pile test of molten material movement and relocation; molten material-sodium interaction; molten fuel-coolant interaction; core disruptive accidents; sodium boiling; post accident material relocation, heat removal and relevant experiments already performed or planned

  19. Technical committee meeting on material-coolant interactions and material movement and relocation in liquid metal fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-07-01

    The Technical Committee Meeting on Material-Coolant Interactions and Material Movement and Relocation in Liquid Metal Fast Reactors was sponsored by the International Working Group on Fast Reactors (IWGFR), International Atomic Energy Agency (IAEA) and hosted by PNC, on behalf of the Japanese government. A broad range of technical subjects was discussed in the TCM, covering entire aspects of material motion and interactions relevant to the safety of LMFRs. Recent achievement and current status in research and development in this area were presented including European out-of-pile test of molten material movement and relocation; molten material-sodium interaction; molten fuel-coolant interaction; core disruptive accidents; sodium boiling; post accident material relocation, heat removal and relevant experiments already performed or planned.

  20. 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.

  1. 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

  2. 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)

  3. 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.

  4. 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)

  5. 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.)

  6. 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)

  7. 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

  8. 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

  9. 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)

  10. 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)

  11. 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)

  12. 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.

  13. 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

  14. 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.

  15. Device for relocating fuel elements and control rods in a core reactor

    International Nuclear Information System (INIS)

    Hoffmeister, B.; Schwarz, W.

    1976-01-01

    A device is described for changing the location of fuel elements and control rods in a core reactor with a guiding mast which is mounted on a movable working platform for rotation about a vertical axis and which is provided with a gripping body for grasping the fuel elements and control rods, said gripping body being suspended on a lifting mechanism and being displaceable in vertical direction within the guiding mast while being prevented from rotating relative to the guiding mast. The device furthermore comprises winding means for storing supply lines which are introduced at the top into the guiding mast and the gripper body. The device also includes power operated means for actuating the lifting mechanism. The winding means and lifting mechanism rotate with the mast and are so arranged in the mast that any water dripping therefrom runs down the inside of the mast. Supply conduits leading to the mast are connected thereby by slack loops which permit 360 0 of rotation of the mast. 11 claims, 6 drawing figures

  16. 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)

  17. 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.

  18. 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

  19. 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)

  20. 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

  1. 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

  2. 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

  3. 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.

  4. Simulation of heat and mass transfer processes in molten core debris-concrete systems. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Felde, D K

    1979-01-01

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

  5. 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.

  6. 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.

  7. 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)

  8. 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

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

    International Nuclear Information System (INIS)

    Alsmeyer, H.

    1992-11-01

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

  10. CORCON-MOD3: An integrated computer model for analysis of molten core-concrete interactions

    International Nuclear Information System (INIS)

    Bradley, D.R.; Gardner, D.R.; Brockmann, J.E.; Griffith, R.O.

    1993-10-01

    The CORCON-Mod3 computer code was developed to mechanistically model the important core-concrete interaction phenomena, including those phenomena relevant to the assessment of containment failure and radionuclide release. The code can be applied to a wide range of severe accident scenarios and reactor plants. The code represents the current state of the art for simulating core debris interactions with concrete. This document comprises the user's manual and gives a brief description of the models and the assumptions and limitations in the code. Also discussed are the input parameters and the code output. Two sample problems are also given

  11. Final results of the XR2-1 BWR metallic melt relocation experiment

    International Nuclear Information System (INIS)

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

    1997-08-01

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

  12. Molten salt reactors. Synthesis of studies realized between 1973 and 1983. Reactor core file

    International Nuclear Information System (INIS)

    1983-03-01

    The first neutronic studies were based on the MSBR project for comparisons with ORNL results. Specific effects depending on the different options are evaluated, such as cross sections in the thorium-U233 cycle, replacement of lithium by sodium, delayed neutron balance, reactivity, kinetics, residual power and recent studies concerning the core with fertile exchange zones. Fuel evolution computation taking into account chemical reprocessing is exposed. Finally different type of lattices are examined. [fr

  13. 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.).

  14. 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)

  15. 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)

  16. 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.)

  17. 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)

  18. 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)

  19. 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

  20. 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)

  1. The mechanism of translational displacements of the core of the Earth at inversion molten and solidification of substance at core-mantle boundary in opposite hemispheres

    Science.gov (United States)

    Barkin, Yu. V.

    2009-04-01

    thermal energy. The directed mechanical influences of the bottom shell on top (of the core on the mantle) on geological intervals of time will result in enormous additional variations of the tension state of the top shell, also ordered in space and time (besides in various time scales). This influence will be transferred to all natural processes which will have similar properties of cyclicity and orderliness. Thermodynamic stimulation of layer D " by the relative displacements of the core and mantle will result in formation of ascending mantle streams - plumes. Relative oscillations of the top shells of the Earth with boundaries on depths of 670 km, 430 km and oth. will generate the fluid formations (lenses and chambers) from a magmatic materials and fusions. Ascending fluid streams in the top mantle on system of breaks and cracks move in the top layers and on a surface of the Earth. So magmatic and volcanic activity of the Earth is realized. Over this activity again "supervise" mutually - displaced and deformed shells of the Earth. The last, in turn, are in strict "submission" at the Moon and the Sun and «are sensitively listen» to the slightest changes of their orbital motions." (Barkin, 2002, pp. 45, 46). "The powerful impacts repeating cyclically, on zones of a congestion of fluid masses (astenosphere lenses, magmatic chambers etc.) result in their growth and expansion, and at significant subsequent impacts to a effects of wedging of the top layers of lithosphere and the crust, i.e. to formation of new or to stimulation and expansion of old cracks and lineaments. Subsequent or more powerful impacts (influences) of the bottom shell on a direction of wedging will result in transport of molten mantle substances from the bottom layers in top, including outpourings of magmas and other fluids on a surface of a planet (the Earth). " (Barkin, 2002, with. 47). The mechanism of formation of plums and hot spots. "The most significant displacements of the centers of mass of

  2. 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

  3. 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

  4. 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)

  5. Irradiation of UO2 specimens with molten cores in a pressurized water loop. Test X-2-x

    International Nuclear Information System (INIS)

    Bain, A.S.

    1961-08-01

    Two Zircaloy-2 clad specimens containing stoichiometric UO 2 pellets were irradiated in a pressurized water loop for 379 hours at heat ratings sufficient to cause central melting of the UO 2 . There was no appearance of localized overheating or accelerated corrosion of the sheath, but the diametral increases were considerably larger than those observed in loop specimens irradiated at lower heat ratings. The length increases, however, were approximately the same as those measured for specimens at lower ratings. There was a clearly visible demarcation between UO 2 that had been molten and that which had not. The value of ∫ 500 o C Tm kdθ = 74 ± W/cm was essentially the same as that obtained from the short-duration tests in the Hydraulic Rabbit, indicating there is no marked decrease in thermal conductivity of the UO 2 fuel in irradiations up to 379 hours. (author)

  6. Theoretical and experimental methods to determine the properties of molten core components and reaction products. Pt. 2

    International Nuclear Information System (INIS)

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

    1975-10-01

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

  7. Core-concrete interactions using molten UO2 with zirconium on a basaltic basemat: The SURC-2 experiment

    International Nuclear Information System (INIS)

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

    1992-08-01

    An inductively heated experiment, SURC-2, using prototypic U0 2 -ZrO 2 materials was executed as part of the Integral Core-Concrete Interactions Experiments Program. The purpose of this experimental program was to measure and assess the variety of source terms produced during core debris/concrete interactions. These source terms include thermal energy released to both the reactor basemat and the containment environment, as well as flammable gas, condensable vapor and toxic or radioactive aerosols generated during the course of a severe reactor accident. The SURC-2 experiment eroded a total of 35 cm of basaltic concrete during 160 minutes of sustained interaction using 203.9 kg of prototypic U0 2 -ZrO 2 core debris material that included 18 kg of Zr metal and 3.4 kg of fission product simulants. The meltpool temperature ranged from 2400--1900 degrees C during the first 50 minutes of the test followed by steady temperatures of 1750--1800 degrees C during the middle portion of the test and increased temperatures of 1800--1900 degrees C during the final 50 minutes of testing. The total erosion during the first 50 minutes was 15 cm with an additional 7 cm during the middle part of the test and 13 cm of ablation during the final 50 minutes. Comprehensive gas flowrates, gas compositions, and aerosol release rates were also measured during the SURC-2 test. When combined with the SURC-1 results, SURC-2 forms a complete data base for prototypic U0 2 -ZrO 2 core debris interactions with concrete

  8. Core-concrete interactions using molten UO sub 2 with zirconium on a basaltic basemat: The SURC-2 experiment

    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))

    1992-08-01

    An inductively heated experiment, SURC-2, using prototypic U0{sub 2}-ZrO{sub 2} materials was executed as part of the Integral Core-Concrete Interactions Experiments Program. The purpose of this experimental program was to measure and assess the variety of source terms produced during core debris/concrete interactions. These source terms include thermal energy released to both the reactor basemat and the containment environment, as well as flammable gas, condensable vapor and toxic or radioactive aerosols generated during the course of a severe reactor accident. The SURC-2 experiment eroded a total of 35 cm of basaltic concrete during 160 minutes of sustained interaction using 203.9 kg of prototypic U0{sub 2}-ZrO{sub 2} core debris material that included 18 kg of Zr metal and 3.4 kg of fission product simulants. The meltpool temperature ranged from 2400--1900{degrees}C during the first 50 minutes of the test followed by steady temperatures of 1750--1800{degrees}C during the middle portion of the test and increased temperatures of 1800--1900{degrees}C during the final 50 minutes of testing. The total erosion during the first 50 minutes was 15 cm with an additional 7 cm during the middle part of the test and 13 cm of ablation during the final 50 minutes. Comprehensive gas flowrates, gas compositions, and aerosol release rates were also measured during the SURC-2 test. When combined with the SURC-1 results, SURC-2 forms a complete data base for prototypic U0{sub 2}-ZrO{sub 2} core debris interactions with concrete.

  9. 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.

  10. 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.

  11. 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.

  12. Aggregation-primed molten globule conformers of the p53 core domain provide potential tools for studying p53C aggregation in cancer.

    Science.gov (United States)

    Pedrote, Murilo M; de Oliveira, Guilherme A P; Felix, Adriani L; Mota, Michelle F; Marques, Mayra de A; Soares, Iaci N; Iqbal, Anwar; Norberto, Douglas R; Gomes, Andre M O; Gratton, Enrico; Cino, Elio A; Silva, Jerson L

    2018-05-31

    The functionality of the tumor suppressor p53 is altered in more than 50% of human cancers, and many individuals with cancer exhibit amyloid-like buildups of aggregated p53. An understanding of what triggers the pathogenic amyloid conversion of p53 is required for the further development of cancer therapies. Here, perturbation of the p53 core domain (p53C) with sub-denaturing concentrations of guanidine hydrochloride and high hydrostatic pressure revealed native-like molten globule (MG) states, a subset of which were highly prone to amyloidogenic aggregation. We found that MG conformers of p53C, likely representing population-weighted averages of multiple states, have different volumetric properties, as determined by pressure perturbation and size-exclusion chromatography. We also found that they bind the fluorescent dye 4,4'-dianilino-1,1'-binaphthyl-5,5'-disulfonic acid (bis-ANS) and have a native-like tertiary structure that occludes the single Trp residue in p53. Fluorescence experiments revealed conformational changes of the single Trp and Tyr residues before p53 unfolding and the presence of MG conformers, some of which were highly prone to aggregation. P53C exhibited marginal unfolding cooperativity, which could be modulated from unfolding to aggregation pathways with chemical or physical forces. We conclude that trapping amyloid precursor states in solution is a promising approach for understanding p53 aggregation in cancer. Our findings support the use of single-Trp fluorescence as a probe for evaluating p53 stability, effects of mutations, and the efficacy of therapeutics designed to stabilize p53. Published under license by The American Society for Biochemistry and Molecular Biology, Inc.

  13. 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

  14. 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.

  15. 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)

  16. 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.…

  17. 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.)

  18. 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

  19. 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''

  20. 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.)

  1. 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...

  2. 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)

  3. 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...

  4. 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

  5. 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

  6. 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

  7. A solar-thermal energy harvesting scheme: enhanced heat capacity of molten HITEC salt mixed with Sn/SiO(x) core-shell nanoparticles.

    Science.gov (United States)

    Lai, Chih-Chung; Chang, Wen-Chih; Hu, Wen-Liang; Wang, Zhiming M; Lu, Ming-Chang; Chueh, Yu-Lun

    2014-05-07

    We demonstrated enhanced solar-thermal storage by releasing the latent heat of Sn/SiO(x) core-shell nanoparticles (NPs) embedded in a eutectic salt. The microstructures and chemical compositions of Sn/SiO(x) core-shell NPs were characterized. In situ heating XRD provides dynamic crystalline information about the Sn/SiO(x) core-shell NPs during cyclic heating processes. The latent heat of ∼29 J g(-1) for Sn/SiO(x) core-shell NPs was measured, and 30% enhanced heat capacity was achieved from 1.57 to 2.03 J g(-1) K(-1) for the HITEC solar salt without and with, respectively, a mixture of 5% Sn/SiO(x) core-shell NPs. In addition, an endurance cycle test was performed to prove a stable operation in practical applications. The approach provides a method to enhance energy storage in solar-thermal power plants.

  8. 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

  9. 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.)

  10. 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)

  11. 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.

  12. 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...

  13. 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

  14. 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)

  15. 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.

  16. 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

  17. Proposition of a core model for the thorium molten salt reactor (TMSR) minimizing the graphite moderator quantity in core; Proposition d'un modele de coeur pour le RSF thorium minimisant la quantite de moderateur graphite en coeur

    Energy Technology Data Exchange (ETDEWEB)

    Nuttin, A

    2004-07-01

    This work deals with the problem of fast damage of graphite in the core of TMSR. The approach consists to minimize the quantity of graphite used in the core (by an increase of the voluminal power) and then to extract and to reprocess. (O.M.)

  18. 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

  19. 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)

  20. 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

  1. 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

  2. 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)

  3. 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)

  4. 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

  5. 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

  6. Proposal of a core model for the thorium molten salt reactor minimizing the quantity of graphite moderator in the core; Proposition d'un modele de coeur pour le RSF thorium minimisant la quantite de moderateur graphite en coeur

    Energy Technology Data Exchange (ETDEWEB)

    Nuttin, A

    2004-06-01

    In the present day TMSR design, the average power in the salt is about 200 W/cm{sup 3}, i.e. two times the one of MSBR. The average neutron flux in the core has doubled and the lifetime of graphite is two times lower. There is two approaches to solve this worrying problem: reducing the volume power to 50 W/cm{sup 3} or minimizing the amount of graphite used in the core. A solution should be to increase the volume power in order to reduce the core dimensions and thus the amount of graphite. By acting both on the total power ('economical' minimum of 1000 MWth) and on the average volume power ('physical' maximum of 500 W/cm{sup 3}) it is possible to reduce the core to a single channel or a single cylindrical ring and to concentrate graphite in a place easily accessible for its extraction and reprocessing. (J.S.)

  7. 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)

  8. 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

  9. 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

  10. 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

  11. 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.)

  12. Molten salt reactors: chemistry

    International Nuclear Information System (INIS)

    1983-01-01

    This work is a critical analysis of the 1000 MW MSBR project. Behavior of rare gases in the primary coolant circuit, their extraction from helium. Coating of graphite by molybdenum, chemistry of protactinium and niobium produced in the molten salt, continuous reprocessing of the fuel salt and use of stainless steel instead of hastelloy are reviewed [fr

  13. 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).

  14. 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

  15. 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)

  16. Molten salt breeder reactor

    International Nuclear Information System (INIS)

    Furukawa, Kazuo; Tsukada, Kineo; Nakahara, Yasuaki; Oomichi, Toshihiko; Oono, Hideo.

    1982-01-01

    Purpose: To simplify the structure, as well as improve the technical reliability and safety by the elimination of a proton beam entering window. Constitution: The nuclear reactor container main body is made of Hastelloy N and provided at the inner surface with two layers of graphite shields except for openings. An aperture was formed in the upper surface of the container, through which protons accelerated by a linear accelerator are directly entered to the liquid surface of molten salts such as 7LiF-BeF 2 -ThF 4 , 7LiF-NaF-ThF 4 , 7LiF-Rb-UF 4 , NaF-KF-UF 4 and the like. The heated molten salts are introduced by way of a pipeway into a heat exchanger where the heat is transferred to coolant salts and electric generation is conducted by way of heated steams. (Furukawa, Y.)

  17. 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)

  18. 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...

  19. 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.)

  20. 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.

  1. Research on the fundamental process of thermal-hydraulic behaviors in severe accident. Heat transfer on the liquid-liquid interface between molten core pool and coolant. JAERI's nuclear research promotion program, H10-027-6. Contract research

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Saito, Yasushi

    2002-03-01

    Heat transfer experiments under steady and transient conditions were performed using molten Wood's metal and distilled water to study heat transfer on the liquid-liquid interface between molten fuel pool and coolant under severe accident conditions. In the steady state experiment, boiling curve was measured over the range from natural convection region to film boiling region. The boiling behavior was observed using a high-speed video camera. In the transient experiment, distilled water was poured onto the hot molten metal surface, and the boiling curve was obtained in the cooling process. Comparing the measured boiling curve with existing correlations and experimental data for solid-liquid and liquid-liquid systems, the following conclusions were drawn: (a) When the interface surge is negligible and oxide layer is formed on the interface, the boiling curve at the liquid-liquid surface could be approximately reproduced by the heat transfer correlations for nucleate boiling and film boiling regions and the critical heat flux correlation for a liquid-solid system. (b) When no oxide layer is formed on the interface, the boiling curve at the liquid-liquid surface moved towards higher wall superheat than that at the liquid-solid surface, as Novakovic et al. observed in their experiment using mercury. (c) Transient heat transfer coefficient for film boiling at the liquid-liquid surface was about 100% higher than that predicted by the heat transfer correlation for a solid-liquid system. (author)

  2. 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...

  3. Molten carbonate fuel cell

    Science.gov (United States)

    Kaun, T.D.; Smith, J.L.

    1986-07-08

    A molten electrolyte fuel cell is disclosed with an array of stacked cells and cell enclosures isolating each cell except for access to gas manifolds for the supply of fuel or oxidant gas or the removal of waste gas. The cell enclosures collectively provide an enclosure for the array and effectively avoid the problems of electrolyte migration and the previous need for compression of stack components. The fuel cell further includes an inner housing about and in cooperation with the array enclosure to provide a manifold system with isolated chambers for the supply and removal of gases. An external insulated housing about the inner housing provides thermal isolation to the cell components.

  4. 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)

  5. 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)

  6. 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

  7. 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)

  8. 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

  9. 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.

  10. 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

  11. 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)

  12. 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.)

  13. 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

  14. 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

  15. 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.

  16. 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)

  17. 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.)

  18. 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

  19. 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.

  20. 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...

  1. 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.

  2. 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. Emphasize is put essentially on the fuel salt of the primary circuit inside which fission reactions occur. The reasons why the (LiF-BeF 2 -ThF 4 -UF 4 ) salt was chosen for the M.S.B.R. concept are examined; the physical, physicochemical and chemical properties of this salt are discussed with its interactions with the structural materials and its evolution in time. An important part of this volume is devoted to the continuous reprocessing of the active salt, the project designers having deemed advisable to take advantage at best from the availability of a continuous purification, in a thermal breeding. The problem of tritium formation and distribution inside the reactor is also envisaged and the fundamentals of the chemistry of the secondary coolant salt are given. The solutions proposed are: the hydrogen scavenging of the primary circuit, a reduction in metal permeability by an oxyde layer deposition on the side in contact with the vapor, and tritium absorption through an isotope exchange with the hydroxifluoroborate [fr

  3. 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)

  4. 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)

  5. 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)

  6. 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

  7. 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

  8. 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)

  9. 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

  10. 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.

  11. 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.)

  12. 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...

  13. 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...

  14. Job Relocation is High in Chemical Industry.

    Science.gov (United States)

    Chemical and Engineering News, 1979

    1979-01-01

    The chances of an employee being relocated are higher in the chemical and plastics industries than in U.S. business as a whole. But the benefits provided by chemical and plastics companies to employees shifted to other locations are generally better than average. (Author/BB)

  15. 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

  16. 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…

  17. 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

  18. 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)

  19. Catalysis in Molten Ionic Media

    DEFF Research Database (Denmark)

    Boghosian, Soghomon; Fehrmann, Rasmus

    2013-01-01

    This chapter deals with catalysis in molten salts and ionic liquids, which are introduced and reviewed briefly, while an in-depth review of the oxidation catalyst used for the manufacturing of sulfuric acid and cleaning of flue gas from electrical power plants is the main topic of the chapter...

  20. thermic oil and molten salt

    African Journals Online (AJOL)

    Boukelia T.E, Mecibah M.S and Laouafi A

    1 mai 2016 ... [27] Zavoico, AB. Solar Power Tower Design Basis Document. Tech. rep, Sandia National. Laboratories, SAND2001-2100, 2001. How to cite this article: Boukelia T.E, Mecibah M.S and Laouafi A. Performance simulation of parabolic trough solar collector using two fluids (thermic oil and molten salt).

  1. 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

  2. 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)

  3. 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

  4. 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

  5. Thermodynamic data bases and calculation code adapted to the modelling of molten core concrete interaction (M.C.C.I.) phenomena, developed jointly by Thermodata and the ''Institut de Protection et de Surete Nucleaire'' (France)

    International Nuclear Information System (INIS)

    Cenerino, G.

    1992-01-01

    An oxide data base containing the main five oxides Al 2 O 3 , CaO, SiO 2 , UO 2 and ZrO 2 of a corium obtained if the reactor core melts through the vessel and slumps into the concrete reactor cavity is developed using the GEMINI2 code. This oxide quinary system study takes into account physical realistic thermodynamical modeling of all the possible equilibrium species of the system. Two applications are presented: the determination of liquidus and solidus temperatures of some selected mixtures of the quinary system (core: UO 2 -ZrO 2 and concrete: Al 2 O 3 -CaO-SiO 2 ), a better modeling of the fission products release by vaporization from the corium. (A.B.). 5 refs., 2 figs

  6. 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.

  7. 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...

  8. 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.

  9. 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)

  10. Crisis Relocation Workshops for Transportation Industry Representatives

    Science.gov (United States)

    1979-12-01

    executive, two National Guard members, one Air Force transporta- _a ,tion representative, two Red Cross representatives, one school bus z coordinators...manaaers, local transitCD operator, and miiltary busingA authorityA Local Governa-ient 2 Fire chief, assistantI Air Force 3 Liaison Support D CPA :z...to attend the workshop. Majur trans- portation problems anticipated during crisis relocation include: 1. Transportation of carless residents; 2. The

  11. 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.)

  12. 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

  13. 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

  14. 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.

  15. 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.

  16. 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

  17. 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

  18. 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

  19. 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)

  20. 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...

  1. Records About Japanese Americans Relocated During World War II

    Data.gov (United States)

    National Archives and Records Administration — This series contains personal descriptive data about Japanese Americans evacuated from the states of Washington, Oregon, and California to ten relocation centers...

  2. 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.

  3. 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

  4. 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

  5. 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

  6. Niobium electrodeposition from molten fluorides

    International Nuclear Information System (INIS)

    Sartori, A.F.

    1987-01-01

    Niobium electrodeposition from molten alkali fluorides has been studied aiming the application of this technic to the processes of electrorefining and galvanotechnic of this metal. The effects of current density, temperature, niobium concentration in the bath, electrolysis time, substrate nature, ratio between anodic and cathodic areas, electrodes separation and the purity of anodes were investigated in relation to the cathodic current efficiency, electrorefining, electroplating and properties of the deposit and the electrolytic solution. The work also gives the results of the conctruction and operation of a pilot plant for refractory metals electrodeposition and shows the electrorefining and electroplating compared to those obtained at the laboratory scale. (author) [pt

  7. 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

  8. 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

  9. 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

  10. 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

  11. Food Relocalization for Environmental Sustainability in Cumbria

    Directory of Open Access Journals (Sweden)

    Les Levidow

    2011-04-01

    Full Text Available In the past decade, many European farmers have adopted less-intensive production methods replacing external inputs with local resources and farmers’ skills. Some have developed closer relations with consumers, also known as short food-supply chains or agro-food relocalization. Through both these means, farmers can gain more of the value that they have added to food production, as well as greater incentives for more sustainable methods and/or quality products, thus linking environmental and economic sustainability. These systemic changes encounter difficulties indicating two generic needs—for state support measures, and for larger intermediaries to expand local markets. The UK rural county of Cumbria provides a case study for exploring those two needs. Cumbria farmers have developed greater proximity to consumers, as a means to gain their support for organic, territorially branded and/or simply ‘local’ food. This opportunity has been an incentive for practices which reduce transport distances, energy costs and other inputs. Regional authorities have provided various support measures for more closely linking producers with each other and with consumers, together developing a Cumbrian food culture. Going beyond the capacity of individual producers, farmer-led intermediaries have maintained distinctive product identities in larger markets including supermarket chains. Although Cumbria’s agro-food relocalization initiatives remain marginal, they counteract the 1990s trend towards delocalization, while also indicating potential for expansion elsewhere.

  12. 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

  13. 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.)

  14. 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.).

  15. 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.)

  16. 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....

  17. Molten salt reactor related research in Switzerland

    International Nuclear Information System (INIS)

    Krepel, Jiri; Hombourger, Boris; Fiorina, Carlo

    2015-01-01

    Switzerland represented by the Paul Scherrer Institute (PSI) is a member of the Generation IV International Forum (GIF). In the past, the research at PSI focused mainly on HTR, SFR, and GFR. Currently, a research program was established also for Molten Salt Reactors (MSR). Safety is the key point and main interest of the MSR research at the Nuclear Energy and Safety (NES) department of PSI. However, it cannot be evaluated without knowing the system design, fuel chemistry, salt thermal-hydraulics features, safety and fuel cycle approach, and the relevant material and chemical limits. Accordingly, sufficient knowledge should be acquired in the other individual fields before the safety can be evaluated. The MSR research at NES may be divided into four working packages (WP): WP1: MSR core design and fuel cycle, WP2: MSR fuel behavior at nominal and accidental conditions, WP3: MSR thermal-hydraulics and decay heat removal system, WP4: MSR safety, fuel stream, and relevant limits. The WPs are proposed so that there are research topics which can be independently studied within each of them. The work plan of the four WPs is based on several ongoing or past national and international projects relevant to MSR, where NES/PSI participates. At the current stage, the program focuses on several specific and design independent studies. The safety is the key point and main long-term interest of the MSR research at NES. (author)

  18. Molten fuel behaviour during slow overpower transients

    International Nuclear Information System (INIS)

    Guerin, Y.; Boidron, M.

    1985-01-01

    In large commercial reactors as Super-Phenix, if we take into account all the uncertainties on the pins and on the core, it is no longer possible to guarantee the absence of fuel melting during incidental events such as slow overpower transients. We have then to explain what happens in the pins when fuel melting occurs and to demonstrate that a limited amount of molten fuel generates no risk of clad failure. For that purpose, we may use the results of a great number of experiments (about 40) that have been performed at C.E.A., most of them in thermal reactor, but some experiments have also been performed in Rapsodie, especially during the last run of this reactor. In a great part of these experiments, fuel melting occurred at beginning of life, but we have also some results at different burnups up to 5 at %. It is not the aim of this paper to describe all these experiments and the results of their post irradiation examination, but to summarize the main conclusions that have been set out of them and that have enabled us to determine the main characteristics of fuel element behaviour when fuel melting occurs

  19. 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)

  20. 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

  1. Simulation of Molten Salt Reactor dynamics

    International Nuclear Information System (INIS)

    Krepel, J.; Rohde, U.; Grundmann, U.

    2005-01-01

    Dynamics of the Molten Salt Reactor - one of the 'Generation IV' concepts - was studied in this paper. The graphite-moderated channel type MSR was selected for the numerical simulation of the reactor with liquid fuel. The MSR dynamics is very specific because of two physical peculiarities of the liquid fueled reactor: the delayed neutrons precursors are drifted by the fuel flow and the fission energy is immediately released directly into the coolant. Presently, there are not many accessible numerical codes appropriate for the MSR simulation, therefore the DYN3D-MSR code was developed based on the FZR in-house code DYN3D. It allows calculating of full 3D transient neutronics in combination with parallel channel type thermal-hydraulics. By means of DYN3D-MSR, several transients typical for the liquid fuel system were analyzed. Those transients were initiated by reactivity insertion, by overcooling of fuel at the core inlet, by the fuel pump start-up or coast-down, or by the blockage of selected fuel channels. In these considered transients, the response of the MSR is characterized by the immediate change of the fuel temperature with changing power and fast negative temperature feedback to the power. The response through the graphite temperature is slower. Furthermore, for big MSR cores fueled with U233 the graphite feedback coefficient can be positive. In this case the addition of erbium to the graphite can ensure the inherent safety features. The DYN3D-MSR code has been shown to be an effective tool for MSR dynamics studies. (author)

  2. 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)

  3. 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)

  4. 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...

  5. 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)

  6. 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...

  7. 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...

  8. 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…

  9. 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

  10. 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

  11. 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

  12. 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.)

  13. 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).

  14. 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.

  15. 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)

  16. 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.)

  17. 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

  18. Thermal conductivity of molten metals

    Energy Technology Data Exchange (ETDEWEB)

    Peralta-Martinez, Maria Vita

    2000-02-01

    A new instrument for the measurement of the thermal conductivity of molten metals has been designed, built and commissioned. The apparatus is based on the transient hot-wire technique and it is intended for operation over a wide range of temperatures, from ambient up to 1200 K, with an accuracy approaching 2%. In its present form the instrument operates up to 750 K. The construction of the apparatus involved four different stages, first, the design and construction of the sensor and second, the construction of an electronic system for the measurement and storage of data. The third stage was the design and instrumentation of the high temperature furnace for the melting and temperature control of the sample, and finally, an algorithm was developed for the extraction of the thermal conductivity from the raw measurement data. The sensor consists of a cylindrical platinum-wire symmetrically sandwiched between two rectangular plane sheets of alumina. The rectangular sensor is immersed in the molten metal of interest and a voltage step is applied to the ends of the platinum wire to induce heat dissipation and a consequent temperature rise which, is in part, determined by the thermal conductivity of the molten metal. The process is described by a set of partial differential equations and appropriate boundary conditions rather than an approximate analytical solution. An electronic bridge configuration was designed and constructed to perform the measurement of the resistance change of the platinum wire in the time range 20 {mu}s to 1 s. The resistance change is converted to temperature change by a suitable calibration. From these temperature measurements as a function of time the thermal conductivity of the molten metals has been deduced using the Finite Element Method for the solution of the working equations. This work has achieved its objective of improving the accuracy of the measurement of the thermal conductivity of molten metals from {+-}20% to {+-}2%. Measurements

  19. 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.)

  20. 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

  1. 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

  2. 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.

  3. 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)

  4. 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

  5. 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.

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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)

  11. 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)

  12. 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)

  13. 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.

  14. 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

  15. Choice and necessity: relocations in the Arctic and South Pacific

    Directory of Open Access Journals (Sweden)

    Robin Bronen

    2014-02-01

    Full Text Available Relocation – whereby livelihoods, housing and public infrastructure are reconstructed in another location – may be the best adaptation response for communities whose current location becomes uninhabitable or is vulnerable to future climate-induced threats.

  16. Apparatus for making molten silicon

    Science.gov (United States)

    Levin, Harry (Inventor)

    1988-01-01

    A reactor apparatus (10) adapted for continuously producing molten, solar grade purity elemental silicon by thermal reaction of a suitable precursor gas, such as silane (SiH.sub.4), is disclosed. The reactor apparatus (10) includes an elongated reactor body (32) having graphite or carbon walls which are heated to a temperature exceeding the melting temperature of silicon. The precursor gas enters the reactor body (32) through an efficiently cooled inlet tube assembly (22) and a relatively thin carbon or graphite septum (44). The septum (44), being in contact on one side with the cooled inlet (22) and the heated interior of the reactor (32) on the other side, provides a sharp temperature gradient for the precursor gas entering the reactor (32) and renders the operation of the inlet tube assembly (22) substantially free of clogging. The precursor gas flows in the reactor (32) in a substantially smooth, substantially axial manner. Liquid silicon formed in the initial stages of the thermal reaction reacts with the graphite or carbon walls to provide a silicon carbide coating on the walls. The silicon carbide coated reactor is highly adapted for prolonged use for production of highly pure solar grade silicon. Liquid silicon (20) produced in the reactor apparatus (10) may be used directly in a Czochralski or other crystal shaping equipment.

  17. 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.

  18. 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....

  19. 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...

  20. 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

  1. 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.

  2. 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.

  3. 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

  4. 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

  5. Molten Salt Breeder Reactor Analysis Based on Unit Cell Model

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

    Contemporary computer codes like the MCNP6 or SCALE are only good for solving a fixed solid fuel reactor. However, due to the molten-salt fuel, MSR analysis needs some functions such as online reprocessing and refueling, and circulating fuel. J. J. Power of Oak Ridge National Laboratory (ORNL) suggested in 2013 a method for simulating the Molten Salt Breeder Reactor (MSBR) with SCALE, which does not support continuous material processing. In order to simulate MSR characteristics, the method proposes dividing a depletion time into short time intervals and batchwise reprocessing and refueling at each step. We are applying this method by using the MCNP6 and PYTHON and NEWT-TRITON-PYTHON and PYTHON code systems to MSBR. This paper contains various parameters to analyze the MSBR unit cell model such as the multiplication factor, breeding ratio, change of amount of fuel, amount of fuel feeding, and neutron flux distribution. The result of MCNP6 and NEWT module in SCALE show some difference in depletion analysis, but it still seems that they can be used to analyze MSBR. Using these two computer code system, it is possible to analyze various parameters for the MSBR unit cells such as the multiplication factor, breeding ratio, amount of material, total feeding, and neutron flux distribution. Furthermore, the two code systems will be able to be used for analyzing other MSR model or whole core models of MSR.

  6. 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.

  7. Molten Salt Breeder Reactor Analysis Based on Unit Cell Model

    International Nuclear Information System (INIS)

    Jeong, Yongjin; Choi, Sooyoung; Lee, Deokjung

    2014-01-01

    Contemporary computer codes like the MCNP6 or SCALE are only good for solving a fixed solid fuel reactor. However, due to the molten-salt fuel, MSR analysis needs some functions such as online reprocessing and refueling, and circulating fuel. J. J. Power of Oak Ridge National Laboratory (ORNL) suggested in 2013 a method for simulating the Molten Salt Breeder Reactor (MSBR) with SCALE, which does not support continuous material processing. In order to simulate MSR characteristics, the method proposes dividing a depletion time into short time intervals and batchwise reprocessing and refueling at each step. We are applying this method by using the MCNP6 and PYTHON and NEWT-TRITON-PYTHON and PYTHON code systems to MSBR. This paper contains various parameters to analyze the MSBR unit cell model such as the multiplication factor, breeding ratio, change of amount of fuel, amount of fuel feeding, and neutron flux distribution. The result of MCNP6 and NEWT module in SCALE show some difference in depletion analysis, but it still seems that they can be used to analyze MSBR. Using these two computer code system, it is possible to analyze various parameters for the MSBR unit cells such as the multiplication factor, breeding ratio, amount of material, total feeding, and neutron flux distribution. Furthermore, the two code systems will be able to be used for analyzing other MSR model or whole core models of MSR

  8. 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

  9. 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.

  10. 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)

  11. 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.)

  12. 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

  13. 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)

  14. 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)

  15. 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...

  16. 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...

  17. 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)

  18. 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)

  19. Molten salt burner fuel behaviour and treatment

    International Nuclear Information System (INIS)

    Ignatiev, V.V.; Zakirov, R.Y.; Grebenkine, K.F.

    2001-01-01

    The objective of this paper is to discuss the feasibility of molten salt reactor technology for treatment of Pu, minor actinides and fission products, when the reactor and fission product clean-up unit are planned as an integral system. This contribution summarises the available R and D which led to selection of the fuel compositions for the molten salt reactor of the TRU burner type (MSB). Special characteristics of behaviour of TRUs and fission products during power operation of MSB concepts are presented. The present paper briefly reviews the processing developments underlying the prior molten salt reactor programmes and relates them to the separation requirements of the MSB concept, including the permissible range of processing cycle times and removal times. Status and development needs in the thermodynamic properties of fluorides, fission product clean-up methods and container materials compatibility with the working fluids for the fission product clean-up unit are discussed. (authors)

  20. 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)

  1. 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)

  2. 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

  3. 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.

  4. Experimental studies of actinides in molten salts

    International Nuclear Information System (INIS)

    Reavis, J.G.

    1985-06-01

    This review stresses techniques used in studies of molten salts containing multigram amounts of actinides exhibiting intense alpha activity but little or no penetrating gamma radiation. The preponderance of studies have used halides because oxygen-containing actinide compounds (other than oxides) are generally unstable at high temperatures. Topics discussed here include special enclosures, materials problems, preparation and purification of actinide elements and compounds, and measurements of various properties of the molten volts. Property measurements discussed are phase relationships, vapor pressure, density, viscosity, absorption spectra, electromotive force, and conductance. 188 refs., 17 figs., 6 tabs

  5. Experimental studies of actinides in molten salts

    Energy Technology Data Exchange (ETDEWEB)

    Reavis, J.G.

    1985-06-01

    This review stresses techniques used in studies of molten salts containing multigram amounts of actinides exhibiting intense alpha activity but little or no penetrating gamma radiation. The preponderance of studies have used halides because oxygen-containing actinide compounds (other than oxides) are generally unstable at high temperatures. Topics discussed here include special enclosures, materials problems, preparation and purification of actinide elements and compounds, and measurements of various properties of the molten volts. Property measurements discussed are phase relationships, vapor pressure, density, viscosity, absorption spectra, electromotive force, and conductance. 188 refs., 17 figs., 6 tabs.

  6. 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.

  7. 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

  8. 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.)

  9. 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

  10. 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)

  11. 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)

  12. 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.

  13. 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)

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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.

  19. 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.

  20. 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.

  1. 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

  2. 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

  3. 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

  4. Recent electroanalytical studies in molten fluorides

    International Nuclear Information System (INIS)

    Manning, D.L.; Mamantov, G.

    1976-01-01

    This paper summarizes the voltametric and chronopotentiometric studies of Bi, Fe, Te, oxide and U(IV)/U(III) ratio determinations in molten LiF--BeF 2 --ThF 4 (72-16-12 mole percent) and LiF--BeF 2 --ZrF 4 (65.6-29.4-5.0 mole percent). 54 references, 11 figures

  5. 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.)

  6. Galvanic high energy cells with molten electrolytes

    Energy Technology Data Exchange (ETDEWEB)

    Borger, W.; Kappus, W.; Kunze, D.; Laig-Hoerstebrock, H.; Panesar, H.; Sterr, G.

    1981-01-01

    To develop a galvanic cell with molten salt electrolyte for electric vehicle propulsion and load leveling as well as to fabricate ten prototype cells with a capacity of at least 150 Ah (5 hour rate) and an energy density of 80 Wh/kg was the objective of this project.

  7. 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)

  8. 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.)

  9. 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

  10. 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

  11. 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%.

  12. 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

  13. 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.

  14. Static fuel molten salt reactors - simpler, cheaper and safer

    International Nuclear Information System (INIS)

    Scott, Ian

    2015-01-01

    The many conceptual designs for Molten Salt Reactors (MSR's) today are all evolutions from the prototype MSR that went critical at Oak Ridge 50 years ago. Critically, they are based on pumping the molten fuel salt from a reaction chamber where the fuel achieves critical mass through a heat exchanger where the resulting heat is transferred to another working fluid. This basic concept was not the first idea that the Oak Ridge scientists considered. Their initial preference was to put the molten salt fuel into tubes, just like solid fuel pellets in their cladding, and circulate a coolant past the tubes. They concluded however that the low thermal conductivity of the salt meant that the tubes could be no wider than 2mm which would be entirely impractical. In this analysis they ignored the contribution of convection to heat transfer in fluids, probably because they were designing an aircraft engine where varying g forces would make convection unreliable. Moltex Energy has re-examined this decision using the modern tools of computational fluid dynamics to simulate convective flow in the molten salt and discovered that in fact tubes of similar diameter to those used for solid fuels are entirely practical. Power densities of 250kW/litre of fuel salt are readily attainable providing a higher overall power density than a PWR reactor. This discovery permits MSR's to be built without any of the complex pumping, passively safe drain systems, on line degassing, filtration and chemical processing needed in pumped MSR's. Their design is very simple and they have many intrinsic safety factors including low pressure operation, chemically unreactive fluids and strongly negative fuel thermal and coolant voiding reactivity coefficients. Most importantly, the highly radioactive fission products are retained in non-volatile form within the fuel tubes in the reactor core. Radioactive fuel salt never leaves the reactor vessel except in an immobile frozen form during

  15. 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

  16. 48 CFR 31.205-35 - Relocation costs.

    Science.gov (United States)

    2010-10-01

    ... employees retain ownership of a vacated home in the old location and rent at the new location. The rented... work location (for a period of 12 months or more) of an existing employee or upon recruitment of a new employee. The following types of relocation costs are allowable as noted, subject to the limitations in...

  17. Mechanism analysis on nuclear technical relocation of the major countries

    International Nuclear Information System (INIS)

    1985-12-01

    This report is about cooperation method of nuclear energy among the USA, France and Canada. So it deals with building a cooperative system on nuclear technical relocation, researching a foreign policy on nuclear energy of the USA and cooperation method and cooperation trend of Japan, Canada, Russia, China and West Germany.

  18. Gender differences in the family-relatedness of relocation decisions

    NARCIS (Netherlands)

    Ullrich, J.; Pluut, Helen; Büttgen, M.

    2015-01-01

    Using a dyadic study design, the present study draws on research into the family-relatedness of work decisions to examine the impact of the spouse's characteristics on an employee's willingness to engage in job-related relocation. With a sample of 1234 employees, the results show that spousal

  19. 5 CFR 575.209 - Payment of relocation incentives.

    Science.gov (United States)

    2010-01-01

    ... pay periods equals 546 days, and 546 days divided by 365 days equals 1.50 years. (c)(1) An authorized... pay a relocation incentive— (1) As an initial lump-sum payment at the commencement of the service... employee in a service period may not exceed 25 percent of the annual rate of basic pay of the employee at...

  20. 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

  1. Evolution and dynamics of Earth from a molten initial stage

    Science.gov (United States)

    Louro Lourenço, D. J.; Tackley, P.

    2016-12-01

    It is now well established that most of the terrestrial planets underwent a magma ocean stage during their accretion. On Earth, it is probable that at the end of accretion, giant impacts like the hypothesised Moon-forming impact, together with other sources of heat, melted a substantial part of the mantle. The thermal and chemical evolution of the resulting magma ocean most certainly had dramatic consequences on the history of the planet. Considerable research has been done on magma oceans using simple 1-D models (e.g.: Abe, PEPI 1997; Solomatov, Treat. Geophys. 2007; Elkins-Tanton EPSL 2008). However, some aspects of the dynamics may not be adequately addressed in 1-D and require the use of 2-D or 3-D models. Moreover, new developments in mineral physics that indicate that melt can be denser than solid at high pressures (e.g.: de Koker et al., EPSL 2013) can have very important impacts on the classical views of the solidification of magma oceans (Labrosse et al., Nature 2007; Labrosse et al., The Early Earth 2015). The goal of our study is to understand and characterize the influence of melting on the long-term thermo-chemical evolution of rocky planet interiors, starting from an initial molten state (magma ocean). Our approach is to model viscous creep of the solid mantle, while parameterizing processes that involve melt as previously done in 1-D models, including melt-solid separation at all melt fractions, the use of an effective diffusivity to parameterize turbulent mixing, coupling to a parameterized core heat balance and a radiative surface boundary condition. These enhancements have been made to the numerical code StagYY (Tackley, PEPI 2008). We present results for the evolution of an Earth-like planet from a molten initial state to present day, while testing the effect of uncertainties in parameters such as melt-solid density differences, surface heat loss and efficiency of turbulent mixing. Our results show rapid cooling and crystallization until the

  2. 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

  3. 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

  4. 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.

  5. Transformation and fragmentation behavior of molten aluminum in sodium pool

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  6. 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.

  7. 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...

  8. 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...

  9. 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…

  10. 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

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

    Science.gov (United States)

    Govers, K.; Verwerft, M.

    2016-09-01

    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.

  12. 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.

  13. 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)

  14. 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

  15. Mixing of zeolite powders and molten salt

    International Nuclear Information System (INIS)

    Pereira, C.; Zyryanov, V.N.; Lewis, M.A.; Ackerman, J.P.

    1996-01-01

    Transuranics and fission products in a molten salt can be incorporated into zeolite A by an ion exchange process and by a batch mixing or blending process. The zeolite is then mixed with glass and consolidated into a monolithic waste form for geologic disposal. Both processes require mixing of zeolite powders with molten salt at elevated temperatures (>700 K). Complete occlusion of salt and a uniform distribution of chloride and fission products are desired for incorporation of the powders into the final waste form. The relative effectiveness of the blending process was studied over a series of temperature, time, and composition profiles. The major criteria for determining the effectiveness of the mixing operations were the level and uniformity of residual free salt in the mixtures. High operating temperatures (>775 K) improved salt occlusion. Reducing the chloride levels in the mixture to below 80% of the full salt capacity of the zeolite significantly reduced the free salt level in the final product

  16. Molten-salt reactor information system

    International Nuclear Information System (INIS)

    Haubenreich, P.N.; Cardwell, D.W.; Engel, J.R.

    1975-06-01

    The Molten-Salt Reactor Information System (MSRIS) is a computer-based file of abstracts of documents dealing with the technology of molten-salt reactors. The file is stored in the IBM-360 system at ORNL, and may be searched through the use of established interactive computer programs from remote terminals connected to the computer via telephone lines. The system currently contains 373 entries and is subject to updating and expansion as additional information is developed. The nature and general content of the data file, a general approach for obtaining information from it, and the manner in which material is added to the file are described. Appendixes provide the list of keywords currently in use, the subject categories under which information is filed, and simplified procedures for searching the file from remote terminals. (U.S.)

  17. Molten salt reactors. The AMSTER concept

    International Nuclear Information System (INIS)

    Vergnes, J.; Garzenne, C.; Lecarpentier, D.; Mouney, H.

    2001-01-01

    This article presents the concept of actinide molten salt transmuter (AMSTER). This reactor is graphite-moderated and is dedicated to the burning of actinides. The main difference with a molten salt reactor is that its liquid fuel undergoes an on-line partial reprocessing in which fission products are extracted and heavy nuclei are reintroduced into the fuel. In order to maintain the reactivity regular injections of 235 U-salt are made. In classical reactors, fuel burn-up is limited by the swelling of the cladding and the radiation fuel pellets resistance, in AMSTER there is no limitation to the irradiation time of the fuel, so all the actinides can be burnt or transmuted. (A.C.)

  18. Molten salt combustion of radioactive wastes

    International Nuclear Information System (INIS)

    Grantham, L.F.; McKenzie, D.E.; Richards, W.L.; Oldenkamp, R.D.

    1976-01-01

    The Atomics International Molten Salt Combustion Process reduces the weight and volume of combustible β-γ contaminated transuranic waste by utilizing air in a molten salt medium to combust organic materials, to trap particulates, and to react chemically with any acidic gases produced during combustion. Typically, incomplete combustion products such as hydrocarbons and carbon monoxide are below detection limits (i.e., 3 ) is directly related to the sodium chloride vapor pressure of the melt; >80% of the particulate is sodium chloride. Essentially all metal oxides (combustion ash) are retained in the melt, e.g., >99.9% of the plutonium, >99.6% of the europium, and >99.9% of the ruthenium are retained in the melt. Both bench-scale radioactive and pilot scale (50 kg/hr) nonradioactive combustion tests have been completed with essentially the same results. Design of three combustors for industrial applications are underway

  19. Analysis of a molten salt reactor benchmark

    International Nuclear Information System (INIS)

    Ghosh, Biplab; Bajpai, Anil; Degweker, S.B.

    2013-01-01

    This paper discusses results of our studies of an IAEA molten salt reactor (MSR) benchmark. The benchmark, proposed by Japan, involves burnup calculations of a single lattice cell of a MSR for burning plutonium and other minor actinides. We have analyzed this cell with in-house developed burnup codes BURNTRAN and McBURN. This paper also presents a comparison of the results of our codes and those obtained by the proposers of the benchmark. (author)

  20. Molten salt battery having inorganic paper separator

    Science.gov (United States)

    Walker, Jr., Robert D.

    1977-01-01

    A high temperature secondary battery comprises an anode containing lithium, a cathode containing a chalcogen or chalcogenide, a molten salt electrolyte containing lithium ions, and a separator comprising a porous sheet comprising a homogenous mixture of 2-20 wt.% chrysotile asbestos fibers and the remainder inorganic material non-reactive with the battery components. The non-reactive material is present as fibers, powder, or a fiber-powder mixture.

  1. Electrochemical studies in molten sodium fluoroborate

    International Nuclear Information System (INIS)

    Brigaudeau, M.; Wagner, J.F.

    1979-01-01

    Physical properties of sodium fluoroborate are recalled and first results obtained during experimental study of molten NaBF 4 are exposed. The system Cu/CuF is used as an indicator of fluoride ion activity and dissociation constant of the solvent is determined by adding NaF to NaBF 4 saturated with BF 3 at a pressure of 1 atm and found equal to 2.7x10 -3 [fr

  2. Corrosion of technical ceramics by molten aluminium

    NARCIS (Netherlands)

    Schwabe, U.; Wolff, L.R.; Loo, van F.J.J.; Ziegler, G.

    1992-01-01

    The corrosion of 8 types of ceramics, i.e., 1 grade of hot isostatically pressed reaction-bonded Si3N4 (HIPRBSN), 3 grades of hot pressed Si3N4 (HPSN), and 4 grades of RBSN, and 2 types of SiC (HIPSiC and Si-impregnated SiC (SiSiC)) in molten Al (pure Al and AlZnMgCu1.5) was studied. The HIPRBSN and

  3. 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

  4. Thorium molten-salt nuclear energy synergetics

    International Nuclear Information System (INIS)

    Furukawa, Kazuo

    1989-01-01

    One of the most practical and rational approaches for establishing the idealistic Thorium resource utilization program has been presented, which might be effective to solve the principal energy problems, concerning safety, proliferation and terrorism, resource, power size and fuel cycle economy, for the next century. The first step will be the development of Small Molten-Salt Reactors as a flexible power station, which is suitable for early commercialization of Th reactors not necessarily competing with proven Large Solid-Fuel Reactors. Therefore, the more detailed design works and practical R and D planning should be performed under the international cooperations soon, soundly depending on the basic technology established by ORNL already. R and D cost would be surprisingly low. This reactor(MSR) seems to be idealistic not only in power-size, siting, safety, safeguard and economy, but also as an effective partner of Molten-Salt Fissile Breeders(MSB) in order to establish the simplest and economical Thorium molten-salt breeding fuel cycle named THORIMS-NES in all over the world including the developing countries and isolated areas. This would be one of the most practical replies to the Lilienthal's appeal of 'A NEW START' in Nuclear Energy. (author)

  5. 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.

  6. Thermal interaction of molten copper with water

    International Nuclear Information System (INIS)

    Zyszkowski, W.

    1975-01-01

    Experimental work was performed to study the thermal interaction between molten copper particles (in the range of temperature from the copper melting point to about 1800 0 C) and water from about 15-80 0 C. The transient temperatures of the copper particles and water before and during their thermal interaction were measured. The history of the phenomena was filmed by means of a high speed FASTAX camera (to 8000 f/s). Classification of the observed phenomena and description of the heat-transfer modes were derived. One among the phenomena was the thermal explosion. The necessary conditions for the thermal explosion are discussed and their physical interpretation is given. According to the hypothesis proposed, the thermal explosion occurs when the molten metal has the temperature of its solidification and the heat transfer on its surface is sufficiently intensive. The 'sharp-change' of the crystalline structure during the solidification of the molten metal is the cause of the explosion fragmentation. (author)

  7. 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).

  8. 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)

  9. 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

  10. 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

  11. 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

  12. Different controls on sedimentary organic carbon in the Bohai Sea: River mouth relocation, turbidity and eutrophication

    Science.gov (United States)

    Xu, Yunping; Zhou, Shangzhe; Hu, Limin; Wang, Yinghui; Xiao, Wenjie

    2018-04-01

    The extractable lipids and bulk organic geochemical parameters in three sediment cores (M-1, M-3 and M-7) from southern, central and northern Bohai Sea were analyzed in order to reconstruct environmental changes since 1900. The C/N ratio and multiple biomarkers (e.g., C27 + C29 + C31n-alkanes, C24 + C26 + C28n-alkanols, branched versus isoprenoid tetraether index) suggest more terrigenous organic carbon (OC) inputs in southern Bohai Sea. The abrupt changes of biomarker indicators in core M-1 are generally synchronous with the Yellow River mouth relocation events (e.g., 1964, 1976 and 1996), suggesting the distance to the river mouth being an important factor for sedimentary OC dispersal in the southern Bohai Sea. However, in cores M-3 and M-7, terrigenous biomarkers (i.e., BIT) show a long-term declining trend, consistent with a continuous reduction of the Yellow River sediment load, whereas marine biomarkers such as cholesterol, brassicasterol and dinosterol dramatically increased post-1980, apparently related to human-induced eutrophication in the Bohai Sea. Our study suggests different controlling factors on sedimentary OC distribution in the southern (high turbidity) and other parts (less turbidity) of the Bohai Sea, which should be considered for interpretation of paleoenvironments and biogeochemical processes in the river dominated margins that are hotspots of the global carbon cycling.

  13. 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

  14. 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

  15. 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.)

  16. 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.

  17. 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)

  18. Transferring Knowledge in the Relocation of Manufacturing Units

    DEFF Research Database (Denmark)

    Yang, Cheng; Madsen, Erik; Liangsiri, Jirapha

    2010-01-01

    ABSTRACT Purpose--The global spread of production makes companies relocating their manufacturing units frequently. Not only equipments, systems and facilities, need to be moved, but the transfer of operational knowledge and experience seem to be a major challenge. However, discussions on knowledge...... transfer are derived predominantly from a cognitive perspective and normally focus on the organizational level. Thus, from the perspective of operations management, this paper aims to explore how to transfer production know-how on the shop floor level when manufacturing units are relocated and the paper...... indicate which means can be used to support this intra-firm transfer process. Design/methodology/approach--Four cases are identified from two action research projects on the basis of observations over one-and-half years and semi-structured interviews with more than 59 people. Other methods, such as surveys...

  19. 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.

  20. Temperature Discontinuity Caused by Relocation of Meteorological Stations in Taiwan

    Directory of Open Access Journals (Sweden)

    Chih-wen Hung

    2009-01-01

    Full Text Available With global warming upon us, it has be come increasingly important to identify the extent of this warming trend and in doing so be able to rank mean temperature changes in particular seasons and years. This requires a need for homogeneous climate data, which do not reflect individual anomalies in instruments, station locations or local environments (urbanization. Ac curate homogeneous long-term meteorological data helps show how temperature variations have truly occurred in the climate. Many possible factors contribute to artificial abrupt changes or sharp discontinuities in long time series data, such as the impact of station relocation, changes in observational schedules and instrumentation. Homogeneity adjustments of in situ climate data are very important processes for preparing observational data to be used in further analysis and research. Users require a well-documented history of stations to make appropriate homogeneity adjustments because precise historical back ground records of stations can provide researchers with knowledge of when artificial discontinuity has occurred and its causes. With out such de tailed historical data for each meteorological station, abrupt changes are difficult to interpret. Unfortunately, no homogeneity adjustments for temperature records have been con ducted previously in Tai wan, and present available sources of the history of Taiwan's meteorological stations exhibit in consistencies. In this study, information pertaining to station history, especially relocation records, is pro vided. This information is essential for anal y sis of continuous time series data for temperature and climate warming studies. Temperature data from several stations is given in this study to show how artificial discontinuity occurs due to station relocation. Al though there is no homogeneous adjusted climate data provided in this preliminary work, the summarizing of information regarding station relocations should be of assistance

  1. 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.

  2. Heat transfer between relocated materials and the RPV lower head

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  3. Heat transfer between relocated materials and the RPV lower head

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  4. 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)

  5. 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.

  6. 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

  7. 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)

  8. 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

  9. 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)

  10. 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)

  11. Electrochemical studies on plutonium in molten salts

    International Nuclear Information System (INIS)

    Bourges, G.; Lambertin, D.; Rochefort, S.; Delpech, S.; Picard, G.

    2007-01-01

    Electrochemical studies on plutonium have been supporting the development of pyrochemical processes involving plutonium at CEA. The electrochemical properties of plutonium have been studied in molten salts - ternary eutectic mixture NaCl-KCl-BaCl 2 , equimolar mixture NaCl-KCl and pure CaCl 2 - and in liquid gallium at 1073 K. The formal, or apparent, standard potential of Pu(III)/Pu redox couple in eutectic mixture of NaCl-KCl-BaCl 2 at 1073 K determined by potentiometry is equal to -2.56 V (versus Cl 2 , 1 atm/Cl - reference electrode). In NaCl-KCl eutectic mixture and in pure CaCl 2 the formal standard potentials deduced from cyclic voltammetry are respectively -2.54 V and -2.51 V. These potentials led to the calculation of the activity coefficients of Pu(III) in the molten salts. Chronoamperometry on plutonium in liquid gallium using molten chlorides - CaCl 2 and equimolar NaCl/KCl - led to the determination of the activity coefficient of Pu in liquid Ga, log γ = -7.3. This new data is a key parameter to assess the thermodynamic feasibility of a process using gallium as solvent metal. By comparing gallium with other solvent metals - cadmium, bismuth, aluminum - gallium appears to be, with aluminum, more favorable for the selectivity of the separation at 1073 K of plutonium from cerium. In fact, compared with a solid tungsten electrode, none of these solvent liquid metals is a real asset for the selectivity of the separation. The role of a solvent liquid metal is mainly to trap the elements

  12. Applications of molten salts in plutonium processing

    International Nuclear Information System (INIS)

    Bowersox, D.F.; Christensen, D.C.; Williams, J.D.

    1987-01-01

    Plutonium is efficiently recovered from scrap at Los Alamos by a series of chemical reactions and separations conducted at temperatures ranging from 700 to 900 0 C. These processes usually employ a molten salt or salt eutectic as a heat sink and/or reaction medium. Salts for these operations were selected early in the development cycle. The selection criteria are being reevaluated. In this article we describe the processes now in use at Los Alamos and our studies of alternate salts and eutectics

  13. Electrorecovery of tantalum in molten fluorides

    International Nuclear Information System (INIS)

    Espinola, A.; Dutra, A.J.B.; Silva, F.T. da

    1988-01-01

    Considering the privileged situation of Brazil as a productor of tantaliferous minerals, the authors have in view the development of a technology for production of metallic tantalum via molten salts electrolysis; this has the advantage of improving the aggregate value of exportation products, additionally to tantalum oxide and tantalum concentrates. Having in view the preliminary determintion of better conditions of temperature, electrolyte composition and current density for this process, electrolysis were conducted with a solvent composed of an eutetic mixture of lithium, sodium and potassium fluoride for dipotassium fluotantalate and occasionally for tantalum oxide. Current efficiencies as high as 83% were obtained in favoured conditions. (author) [pt

  14. Safe actinide disposition in molten salt reactors

    International Nuclear Information System (INIS)

    Gat, U.

    1997-01-01

    Safe molten salt reactors (MSR) can readily accommodate the burning of all fissile actinides. Only minor compromises associated with plutonium are required. The MSRs can dispose safely of actinides and long lived isotopes to result in safer and simpler waste. Disposing of actinides in MSRs does increase the source term of a safety optimized MSR. It is concluded that the burning and transmutation of actinides in MSRs can be done in a safe manner. Development is needed for the processing to handle and separate the actinides. Calculations are needed to establish the neutron economy and the fuel management. 9 refs

  15. SAFARI-1 research reactor beryllium reflector element replacement, management and relocation

    Energy Technology Data Exchange (ETDEWEB)

    Kock, Marisa De; Vlok, Jwh; Steynberg, B J [South Africa Atomic Energy Corporation (Necsa) (South Africa)

    2012-03-15

    The beryllium (Be) reflector elements of the SAFARI-1 Research Reactor were replaced in October 2011 as part of the Ageing Management Programme of the reactor. After more than three million MWh of operation over a period of 47 years, core reloading became more difficult due to the geometric deformation of the beryllium reflector elements. During the replacement of the reflector elements, criticality and reactivity worth experiments were performed and found to compare favorably with calculated values. A Beryllium Management Programme was established at SAFARI-1 to identify and apply effective and appropriate actions and practices for managing the ageing of the new beryllium reflector elements. This will provide timely detection and mitigation of ageing mechanisms relevant to beryllium reflector elements, supporting the life extension of these elements. These actions and practices include monitoring of the tritium levels in the primary water, calculating and measuring the fluxes within the beryllium reflector positions, measuring the straightness of the elements to track geometric deformation and visually inspecting the reflector elements for crack formation. Acceptance criteria indicating the end of life of the elements were established. These criteria take into account the smallest gap that could exist between elements, sudden changes in the tritium levels and formation of cracks. All the data obtained through the Beryllium Management Programme are recorded in a database. Additional benefits gained through a Beryllium Management Programme are the availability of a complete irradiation history of the beryllium reflector elements at any point in time and the establishment of a knowledge base to assists in the understanding of the behavior of the beryllium reflector elements in an irradiation environment. Straightness baseline measurements of the new beryllium reflector elements were performed with a beryllium straightness measurement tool, designed at SAFARI-1. The

  16. SAFARI-1 research reactor beryllium reflector element replacement, management and relocation

    International Nuclear Information System (INIS)

    Kock, Marisa De; Vlok, Jwh; Steynberg, B.J.

    2012-01-01

    The beryllium (Be) reflector elements of the SAFARI-1 Research Reactor were replaced in October 2011 as part of the Ageing Management Programme of the reactor. After more than three million MWh of operation over a period of 47 years, core reloading became more difficult due to the geometric deformation of the beryllium reflector elements. During the replacement of the reflector elements, criticality and reactivity worth experiments were performed and found to compare favorably with calculated values. A Beryllium Management Programme was established at SAFARI-1 to identify and apply effective and appropriate actions and practices for managing the ageing of the new beryllium reflector elements. This will provide timely detection and mitigation of ageing mechanisms relevant to beryllium reflector elements, supporting the life extension of these elements. These actions and practices include monitoring of the tritium levels in the primary water, calculating and measuring the fluxes within the beryllium reflector positions, measuring the straightness of the elements to track geometric deformation and visually inspecting the reflector elements for crack formation. Acceptance criteria indicating the end of life of the elements were established. These criteria take into account the smallest gap that could exist between elements, sudden changes in the tritium levels and formation of cracks. All the data obtained through the Beryllium Management Programme are recorded in a database. Additional benefits gained through a Beryllium Management Programme are the availability of a complete irradiation history of the beryllium reflector elements at any point in time and the establishment of a knowledge base to assists in the understanding of the behavior of the beryllium reflector elements in an irradiation environment. Straightness baseline measurements of the new beryllium reflector elements were performed with a beryllium straightness measurement tool, designed at SAFARI-1. The

  17. Relocation of earthquakes at southwestern Indian Ocean Ridge and its tectonic significance

    Science.gov (United States)

    Luo, W.; Zhao, M.; Haridhi, H.; Lee, C. S.; Qiu, X.; Zhang, J.

    2015-12-01

    The southwest Indian Ridge (SWIR) is a typical ultra-slow spreading ridge (Dick et al., 2003) and further plate boundary where the earthquakes often occurred. Due to the lack of the seismic stations in SWIR, positioning of earthquakes and micro-earthquakes is not accurate. The Ocean Bottom Seismometers (OBS) seismic experiment was carried out for the first time in the SWIR 49 ° 39 'E from Jan. to March, 2010 (Zhao et al., 2013). These deployed OBS also recorded the earthquakes' waveforms during the experiment. Two earthquakes occurred respectively in Feb. 7 and Feb. 9, 2010 with the same magnitude of 4.4 mb. These two earthquakes were relocated using the software HYPOSAT based on the spectrum analysis and band-pass (3-5 Hz) filtering and picking up the travel-times of Pn and Sn. Results of hypocentral determinations show that there location error is decreased significantly by joined OBS's recording data. This study do not only provide the experiences for the next step deploying long-term wide-band OBSs, but also deepen understanding of the structure of SWIR and clarify the nature of plate tectonic motivation. This research was granted by the Natural Science Foundation of China (41176053, 91028002, 91428204). Keywords: southwest Indian Ridge (SWIR), relocation of earthquakes, Ocean Bottom Seismometers (OBS), HYPOSAT References:[1] Dick, H. J. B., Lin J., Schouten H. 2003. An ultraslow-spreading class of ocean ridge. Nature, 426(6965): 405-412. [2] Zhao M. H., et al. 2013. Three-dimensional seismic structure of the Dragon Flag oceanic core complex at the ultraslow spreading Southwest Indian Ridge (49°39' E). Geochemistry Geophysics Geosystems, 14(10): 4544-4563.

  18. 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)

  19. 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)

  20. 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

  1. Advancing Molten Salts and Fuels at Sandia National Laboratories

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, Salvador B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-09-26

    SNL has a combination of experimental facilities, nuclear engineering, nuclear security, severe nuclear accidents, and nuclear safeguards expertise that can enable significant progress towards molten salts and fuels for Molten Salt Reactors (MSRs). The following areas and opportunities are discussed in more detail in this white paper.

  2. On the ionic equilibrium between complexes in molten fluoroaluminates

    International Nuclear Information System (INIS)

    Akdeniz, Z.; Tankeshwar, K.; Tosi, M.P.

    1991-02-01

    We discuss theoretically (i) the effect of the alkali cation species on the ionic equilibrium between (AlF 6 ) 3- and (AlF 4 ) - complexes in molten alkali fluoroaluminates, and (ii) the possible presence of (AlF 5 ) 2 - complexes in molten cryolite, in relation to very recent Raman scattering experiments by Gilbert and Materne. (author). 7 refs, 2 tabs

  3. 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...

  4. 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

  5. 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.

  6. 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.

  7. 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)

  8. 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

  9. Fuel processing for molten-salt reactors

    International Nuclear Information System (INIS)

    Hightower, J.R. Jr.

    1976-01-01

    Research devoted to development of processes for the isolation of protactinium and for the removal of fission products from molten-salt breeder reactors is reported. During this report period, engineering development progressed on continuous fluorinators for uranium removal, the metal transfer process for rare-earth removal, the fuel reconstitution step, and molten salt--bismuth contactors to be used in reductive extraction processes. The metal transfer experiment MTE-3B was started. In this experiment all parts of the metal transfer process for rare-earth removal are demonstrated using salt flow rates which are about 1 percent of those required to process the fuel salt in a 1000-MW(e) MSBR. During this report period the salt and bismuth phases were transferred to the experimental vessels, and two runs with agitator speeds of 5 rps were made to measure the rate of transfer of neodymium from the fluoride salt to the Bi--Li stripper solution. The uranium removed from the fuel salt by fluorination must be returned to the processed salt in the fuel reconstitution step before the fuel salt is returned to the reactor. An engineering experiment to demonstrate the fuel reconstitution step is being installed. In this experiment gold-lined equipment will be used to avoid introducing products of corrosion by UF 6 and UF 5 . Alternative methods for providing the gold lining include electroplating and mechanical fabrication

  10. Organic waste processing using molten salt oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, M. G., LLNL

    1998-03-01

    Molten Salt Oxidation (MSO) is a thermal means of oxidizing (destroying) the organic constituents of mixed wastes, hazardous wastes, and energetic materials while retaining inorganic and radioactive constituents in the salt. For this reason, MSO is considered a promising alternative to incineration for the treatment of a variety of organic wastes. The U. S. Department of Energy`s Office of Environmental Management (DOE/EM) is currently funding research that will identify alternatives to incineration for the treatment of organic-based mixed wastes. (Mixed wastes are defined as waste streams which have both hazardous and radioactive properties.) One such project is Lawrence Livermore National Laboratory`s Expedited Technology Demonstration of Molten Salt Oxidation (MSO). The goal of this project is to conduct an integrated demonstration of MSO, including off-gas and spent salt treatment, and the preparation of robust solid final forms. Livermore National Laboratory (LLNL) has constructed an integrated pilot-scale MSO treatment system in which tests and demonstrations are presently being performed under carefully controlled (experimental) conditions. The system consists of a MSO process vessel with dedicated off-gas treatment, a salt recycle system, feed preparation equipment, and equipment for preparing ceramic final waste forms. In this paper we describe the integrated system and discuss its capabilities as well as preliminary process demonstration data. A primary purpose of these demonstrations is to identify the most suitable waste streams and waste types for MSO treatment.

  11. Effect of fuel string relocation on the consequences of postulated inlet header LBLOCA in KANUPP reactor

    International Nuclear Information System (INIS)

    Ahmed, I.; Chow, H.C.; Younis, M.H.

    1996-01-01

    An investigation aimed at determining the effect of fuel string relocation on reactivity excursion and power pulse following a hypothetical Large Break Loss of Coolant Accident in KANUPP reactor is reported. The assessment of reactivity insertion was performed making use of global (reactor) core analysis computer code RFSP. The reactor kinetics module CERBERUS of the RFSP code and the SOPHT (thermal-hydraulics code) were subsequently employed for the neutronic transient analysis. The effect was evaluated in context of determining the adequacy of moderator dump shutdown system. Because of the presence of the gap between the inlet shield plug and the fuel string, the fuel bundles may shift in such a manner that low-irradiated fuel is moved towards the core centre. This represents an additional reactivity increase to be accounted for in the analysis. The reactivity excursion, however, is alleviated by an earlier reactor trip. The net impact is that the energy deposited in the maximum rated fuel pencil is increased from 56% of the 960 kJ/kg fuel-centre-line melting limit to 63%. The result demonstrated the adequacy of the shutdown system against the maximum credible accident event. (author)

  12. 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.

  13. 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

  14. 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.

  15. 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

  16. 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)

  17. 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

  18. 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

  19. 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.)

  20. Settlement Relocation Modeling: Reacting to Merapi’s Eruption Incident

    Science.gov (United States)

    Pramitasari, A.; Buchori, I.

    2018-02-01

    Merapi eruption has made severe damages in Central Java Province. Klaten was one of the most affected area, specifically in Balerante Village. This research is made to comprehend GIS model on finding alternative locations for impacted settlement in hazardous zones of eruption. The principal objective of the research study is to identify and analyze physical condition, community characteristics, and local government regulation related to settlements relocation plan for impacted area of eruption. The output is location map which classified into four categories, i.e. not available, available with low accessibility, available with medium accessibility, and available with high accessibility.

  1. Relocation of radioactive residuals store: environment effects statement

    International Nuclear Information System (INIS)

    1984-11-01

    This Environment Effects Statement describes and assesses the likely environmental effects of the proposal to relocate the Health Commission's existing radioactive residuals store to a site within the established Dutson Downs waste disposal area, located 20 km south-east of Sale and 225 km east of Melbourne. The information presented demonstrates that the siting and construction of the proposed radioactive residuals store and the procedures to be adopted for the handling and storage of materials will not present an unacceptable risk to public health and safety, nor will it involve any significant adverse environmental effects

  2. Relocation and the characteristics of hospital and hostel regimes.

    Science.gov (United States)

    Booth, T; Simons, K; Booth, W

    1991-01-01

    Drawing on evidence from a research evaluation of a local community care programme, this paper explores whether relocation from a British National Health Service mental handicap hospital into local authority hostels (supervised residential facilities) brought about a qualitative change in the residential environment of movers towards less restrictive management practices and caring routines, more responsive attitudes towards their rights and needs as individuals, and greater control over their own lives. The conclusions point to the existence of a substantial measure of overlap in the fundamental characteristics of the hospital and hostel regimes.

  3. (Re)locating the gay Filipino: resistance, postcolonialism, and identity.

    Science.gov (United States)

    Manalansan, M F

    1993-01-01

    This paper attempts to critically analyze issues of postcolonial displacement, immigration, and homosexuality by examining the works of two Filipino gay immigrant writers, John Silva and Ralph Peña. Using postcolonial and critical theories, anthropological studies, and ethnographic fieldwork in New York City, this paper focuses on the role of language, memory, the body, race/ethnicity, and social class in the narrative strategies of the two writers. This paper argues that gay postcolonial writers such as these two relocate and reconfigure homosexual/gay identity in the face of new and oppressive hierarchies, identities, and practices.

  4. 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

  5. Proteomics Core

    Data.gov (United States)

    Federal Laboratory Consortium — Proteomics Core is the central resource for mass spectrometry based proteomics within the NHLBI. The Core staff help collaborators design proteomics experiments in a...

  6. 41 CFR 302-3.2 - As a new appointee or student trainee what relocation expenses may my agency pay or reimburse me...

    Science.gov (United States)

    2010-07-01

    ... student trainee what relocation expenses may my agency pay or reimburse me for incident to a permanent... Management Federal Travel Regulation System RELOCATION ALLOWANCES RELOCATION ALLOWANCES 3-RELOCATION ALLOWANCE BY SPECIFIC TYPE New Appointee § 302-3.2 As a new appointee or student trainee what relocation...

  7. 41 CFR 302-12.101 - May we enter into a contract with a relocation services company for the company to provide...

    Science.gov (United States)

    2010-07-01

    ... contract with a relocation services company for the company to provide relocation services? 302-12.101... the company to provide relocation services? Yes, you may enter into a contract with a relocation services company for the company to provide relocation services. ...

  8. Corrosion study in molten fluoride salt

    International Nuclear Information System (INIS)

    Keny, S.J.; Kumbhar, A.G.; Rangarajan, S.; Gupta, V.K.; Maheshwari, N.K.; Vijayan, P.K.

    2013-01-01

    Corrosion behaviors of two alloys viz. Inconel 625 and Inconel 617 were tested in molten fluoride salts of lithium, sodium and potassium (FLiNaK) in the temperature range of 550-750 ℃ in a nickel lined Inconel vessel. Electrochemical polarization (Tafel plot) technique was used for this purpose. For both alloys, the corrosion rate was found to increase sharply beyond 650 ℃ . At 600 ℃ , Inconel 625 showed a decreasing trend in the corrosion rate over a period of 24 hours, probably due to changes in the surface conditions. After fifteen days, re-testing of Inconel 625 in the same melt showed an increase in the corrosion rate. Inconel 625 was found to be more corrosion resistant than Inconel 617. (author)

  9. The Integral Molten Salt Reactor (IMSR)

    Energy Technology Data Exchange (ETDEWEB)

    Leblanc, D. [Terrestrial Energy, Mississauga, Ontario (Canada)

    2014-12-15

    The Integral Molten Salt Reactor is a simple burner or converter design that seeks to maximize passive and inherent safety features in order to minimize development time and achieve true cost innovation. Its integration of all primary systems into a unit sealed for the design life of the reactor will be reviewed with focus on the unique design aspects that make this a pragmatic approach. The IMSR is being developed by Terrestrial Energy in a range of power outputs with initial focus on an 80 MWth (32.5 MWe) unit primarily for remote energy needs. Similar units of modestly larger dimension and up to 600 MWth (291 MWe) are planned that remain truck transportable and able to compete in base load electricity markets worldwide. (author)

  10. The Integral Molten Salt Reactor (IMSR)

    Energy Technology Data Exchange (ETDEWEB)

    LeBlanc, D., E-mail: dleblanc@terrestrialenergy.com [Terrestrial Energy, Mississauga, Ontario (Canada)

    2014-07-01

    The Integral Molten Salt Reactor is a simple burner or converter design that seeks to maximize passive and inherent safety features in order to minimize development time and achieve true cost innovation. Its integration of all primary systems into a unit sealed for the design life of the reactor will be reviewed with focus on the unique design aspects that make this a pragmatic approach. The IMSR is being developed by Terrestrial Energy in a range of power outputs with initial focus on an 80 MWth (32.5 MWe) unit primarily for remote energy needs. Similar units of modestly larger dimension and up to 600 MWth (291 MWe) are planned that remain truck transportable and able to compete in base load electricity markets worldwide. (author)

  11. Structure of partly quenched molten copper chloride

    International Nuclear Information System (INIS)

    Pastore, G.; Tosi, M.P.

    1995-09-01

    The structural modifications induced in a model of molten CuCl by quenching the chlorine component into a microporous disordered matrix are evaluated using the hypernetted-chain closure in Ornstein-Zernike relations for the pair distribution functions in random systems. Aside from obvious changes in the behaviour of long-wavelength density fluctuations, the main effect of partial quenching is an enhanced delocalization of the Cu + ions. The model suggests that the ionic mobility in a superionic glass is enhanced relative to the melt at the same temperature and density. Only very minor quantitative differences are found in the structural functions when the replica Ornstein-Zernike relations derived by Given and Stell for a partly quenched system are simplified to those given earlier by Madden and Glandt. (author). 19 refs, 6 figs

  12. Terrestrial Energy bets on molten salt reactors

    International Nuclear Information System (INIS)

    Anon.

    2015-01-01

    Terrestrial Energy is a Canadian enterprise, founded in 2013, for marketing the integral molten salt reactor (IMSR). A first prototype (called MSRE and with an energy output of 8 MW) was designed and operated between 1965 and 1969 by the Oak Ridge National Laboratory. IMSR is a small, modular reactor with a thermal energy output of 400 MW. According to Terrestrial Energy the technology of conventional power reactors is too complicated and too expensive. On the contrary IMSR's technology appears to be simple, easy to operate and affordable. With a staff of 30 people Terrestrial Energy appears to be a start-up in the nuclear sector. A process of pre-licensing will be launched in 2016 with the Canadian nuclear safety authority. (A.C.)

  13. 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

  14. Focusing patterns of seismicity with relocation and collapsing

    Science.gov (United States)

    Li, Ka Lok; Gudmundsson, Ólafur; Tryggvason, Ari; Bödvarsson, Reynir; Brandsdóttir, Bryndís

    2016-07-01

    Seismicity is generally concentrated on faults or in fault zones of varying, sometimes complex geometry. An earthquake catalog, compiled over time, contains useful information about this geometry, which can help understanding the tectonics of a region. Interpreting the geometrical distribution of events in a catalog is often complicated by the diffuseness of the earthquake locations. Here, we explore a number of strategies to reduce this diffuseness and hence simplify the seismicity pattern of an earthquake catalog. These strategies utilize information about event locations contained in their overall catalog distribution. They apply this distribution as an a priori constraint on relocations of the events, or as an attractor for each individual event in a collapsing scheme, and thereby focus the locations. The latter strategy is not a relocation strategy in a strict sense, although event foci are moved, because the movements are not driven by data misfit. Both strategies simplify the seismicity pattern of the catalog and may help to interpret it. A synthetic example and a real-data example from an aftershock sequence in south west Iceland are presented to demonstrate application of the strategies. Entropy is used to quantify their effect.

  15. Study of Profile Changes during Mechanical Polishing using Relocation Profilometry

    Science.gov (United States)

    Kumaran, S. Chidambara; Shunmugam, M. S.

    2017-10-01

    Mechanical polishing is a finishing process practiced conventionally to enhance quality of surface. Surface finish is improved by mechanical cutting action of abrasive particles on work surface. Polishing is complex in nature and research efforts have been focused on understanding the polishing mechanism. Study of changes in profile is a useful method of understanding behavior of the polishing process. Such a study requires tracing same profile at regular process intervals, which is a tedious job. An innovative relocation technique is followed in the present work to study profile changes during mechanical polishing of austenitic stainless steel specimen. Using special locating fixture, micro-indentation mark and cross-correlation technique, the same profile is traced at certain process intervals. Comparison of different parameters of profiles shows the manner in which metal removal takes place in the polishing process. Mass removal during process estimated by the same relocation technique is checked with that obtained using weight measurement. The proposed approach can be extended to other micro/nano finishing processes and favorable process conditions can be identified.

  16. Fuel processing for molten-salt reactors

    International Nuclear Information System (INIS)

    Hightower, J.R. Jr.

    1975-01-01

    Progress is reported on the development of processes for the isolation of protactinium and for the removal of fission products from molten-salt breeder reactors. The metal transfer experiment MTE-3 (for removing rare earths from MSRE fuel salt) was completed and the equipment used in that experiment was examined. The examination showed that no serious corrosion had occurred on the internal surfaces of the vessels, but that serious air oxidation occurred on the external surfaces of the vessels. Analyses of the bismuth phases indicated that the surfaces in contact with the salts were enriched in thorium and iron. Mass transfer coefficients in the mechanically agitated nondispersing contactors were measured in the Salt/Bismuth Flow-through Facility. The measured mass transfer coefficients are about 30 to 40 percent of those predicted by the preferred literature correlation, but were not as low as those seen in some of the runs in MTE-3. Additional studies using water--mercury systems to simulate molten salt-bismuth systems indicated that the model used to interpret results from previous measurements in the water--mercury system has significant deficiencies. Autoresistance heating studies were continued to develop a means of internal heat generation for frozen-wall fluorinators. Equipment was built to test a design of a side arm for the heating electrode. Results of experiments with this equipment indicate that for proper operation the wall temperature must be held much lower than that for which the equipment was designed. Studies with an electrical analog of the equipment indicate that no regions of abnormally high current density exist in the side arm. (JGB)

  17. Fast Spectrum Molten Salt Reactor Options

    Energy Technology Data Exchange (ETDEWEB)

    Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Patton, Bruce W [ORNL; Howard, Rob L [ORNL; Harrison, Thomas J [ORNL

    2011-07-01

    During 2010, fast-spectrum molten-salt reactors (FS-MSRs) were selected as a transformational reactor concept for light-water reactor (LWR)-derived heavy actinide disposition by the Department of Energy-Nuclear Energy Advanced Reactor Concepts (ARC) program and were the subject of a preliminary scoping investigation. Much of the reactor description information presented in this report derives from the preliminary studies performed for the ARC project. This report, however, has a somewhat broader scope-providing a conceptual overview of the characteristics and design options for FS-MSRs. It does not present in-depth evaluation of any FS-MSR particular characteristic, but instead provides an overview of all of the major reactor system technologies and characteristics, including the technology developments since the end of major molten salt reactor (MSR) development efforts in the 1970s. This report first presents a historical overview of the FS-MSR technology and describes the innovative characteristics of an FS-MSR. Next, it provides an overview of possible reactor configurations. The following design features/options and performance considerations are described including: (1) reactor salt options-both chloride and fluoride salts; (2) the impact of changing the carrier salt and actinide concentration on conversion ratio; (3) the conversion ratio; (4) an overview of the fuel salt chemical processing; (5) potential power cycles and hydrogen production options; and (6) overview of the performance characteristics of FS-MSRs, including general comparative metrics with LWRs. The conceptual-level evaluation includes resource sustainability, proliferation resistance, economics, and safety. The report concludes with a description of the work necessary to begin more detailed evaluation of FS-MSRs as a realistic reactor and fuel cycle option.

  18. 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)

  19. 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.

  20. 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.

  1. Novel waste printed circuit board recycling process with molten salt

    OpenAIRE

    Riedewald, Frank; Sousa-Gallagher, Maria

    2015-01-01

    The objective of the method was to prove the concept of a novel waste PCBs recycling process which uses inert, stable molten salts as the direct heat transfer fluid and, simultaneously, uses this molten salt to separate the metal products in either liquid (solder, zinc, tin, lead, etc.) or solid (copper, gold, steel, palladium, etc.) form at the operating temperatures of 450?470??C. The PCB recovery reactor is essentially a U-shaped reactor with the molten salt providing a continuous fluid, a...

  2. Mechanical structure and problem of thorium molten salt reactor

    International Nuclear Information System (INIS)

    Kamei, Takashi

    2011-01-01

    After Fukushima Daiichi accident, there became great interest in Thorium Molten Salt Reactor (MSR) for the safety as station blackout leading to auto drainage of molten salts with freeze valve. This article described mechanical structure of MSR and problems of materials and pipes. Material corrosion problem by molten salts would be solved using modified Hastelloy N with Ti and Nb added, which should be confirmed by operation of an experimental reactor. Trends in international activities of MSR were also referred including China declaring MSR development in January 2011 to solve thorium contamination issues at rare earth production and India rich in thorium resources. (T. Tanaka)

  3. 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

  4. 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

  5. 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.

  6. 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

  7. Coverage improvement in clustered wireless sensor networks by relocating mobile nodes based on waypoints

    DEFF Research Database (Denmark)

    Mathur, Prateek; Nielsen, Rasmus Hjorth; Prasad, Neeli R.

    2013-01-01

    by clusters with relocation of mobile nodes between the clusters. Mobile nodes are guided by waypoints between source and destination clusters without use of localization services. The simulation results have been presented which state the effectiveness of the proposed approach. Mobile node relocation has...

  8. 76 FR 48863 - Federal Travel Regulation (FTR); Relocation Allowances-Standard Mileage Rate for Moving Purposes

    Science.gov (United States)

    2011-08-09

    ... Travel Regulation (FTR); Relocation Allowances--Standard Mileage Rate for Moving Purposes AGENCY: Office... agencies will reimburse an employee for using a privately owned vehicle for relocation on a worldwide basis...), Office of Travel, Transportation and Asset Management (MT), General Services Administration at (202) 208...

  9. Relocation Stress, Coping, and Sense of Control among Resettlers Resulting from China's Three Gorges Dam Project

    Science.gov (United States)

    Xi, Juan; Hwang, Sean-Shong

    2011-01-01

    The involuntary relocation of people for development purposes has become prevalent across the world in recent decades. Depression is one of the documented negative outcomes of involuntary relocation among resettlers. Viewing the affected population simply as passive victims, past studies have largely ignored the coping strategies employed by…

  10. 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.

  11. 41 CFR 302-2.8 - When must I complete all aspects my relocation?

    Science.gov (United States)

    2010-07-01

    ... 41 Public Contracts and Property Management 4 2010-07-01 2010-07-01 false When must I complete all aspects my relocation? 302-2.8 Section 302-2.8 Public Contracts and Property Management Federal Travel Regulation System RELOCATION ALLOWANCES INTRODUCTION 2-EMPLOYEES ELIGIBILITY REQUIREMENTS General Rules Time...

  12. 41 CFR 302-2.3 - What determines my entitlements and allowances for relocation?

    Science.gov (United States)

    2010-07-01

    ... 41 Public Contracts and Property Management 4 2010-07-01 2010-07-01 false What determines my entitlements and allowances for relocation? 302-2.3 Section 302-2.3 Public Contracts and Property Management Federal Travel Regulation System RELOCATION ALLOWANCES INTRODUCTION 2-EMPLOYEES ELIGIBILITY REQUIREMENTS...

  13. 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.

  14. 76 FR 32340 - Federal Travel Regulation; Temporary Duty (TDY) Travel Allowances (Taxes); Relocation Allowances...

    Science.gov (United States)

    2011-06-06

    ... reflection of the actual tax impact on the employee. Therefore, this proposed rule offers the one-year RITA... to estimate the additional income tax liability that you incur as a result of relocation benefits and... Allowances (Taxes); Relocation Allowances (Taxes) AGENCY: Office of Governmentwide Policy (OGP), General...

  15. 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...

  16. 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...

  17. 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

  18. 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.

  19. 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.

  20. 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

  1. 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.

  2. Advanced Additive Manufacturing Feedstock from Molten Regolith Electrolysis

    Data.gov (United States)

    National Aeronautics and Space Administration — Demonstrate the feasibility of Molten Regolith Electrolysis (MRE) Reactor start by initiating resistive-heating of the regolith past its melting point using...

  3. High Surface Iridium Anodes for Molten Oxide Electrolysis, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Processing of lunar regolith into oxygen for habitat and propulsion is needed to support future space missions. Direct electrochemical reduction of molten regolith...

  4. System Requirements Document for the Molten Salt Reactor Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Aigner, R.D.

    2000-04-01

    The purpose of the conversion process is to convert the {sup 233}U fluoride compounds that are being extracted from the Molten Salt Reactor Experiment (MSRE) equipment to a stable oxide for long-term storage at Bldg. 3019.

  5. 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

  6. Density and Structure Analysis of Molten Ni-W Alloys

    Institute of Scientific and Technical Information of China (English)

    Feng XIAO; Liang FANG

    2004-01-01

    Density of molten Ni and Ni-W alloys was measured in the temperature range of 1773~1873 K with a sessile drop method.The density of molten Ni and Ni-W alloys trends to decrease with increasing temperature. The density and molar volume of the alloys trend to increase with increasing W concentration in the alloys. The calculation result shows an ideal mixing of Ni-W alloys.

  7. Molten fluoride mixtures as possible fission reactor fuels

    International Nuclear Information System (INIS)

    Grimes, W.R.

    1978-01-01

    Molten mixtures of fluorides with UF 4 as a component have been used as combined fuel and primary heat transfer agent in experimental high-temperature reactors and have been proposed for use in breeders or converters of 233 U from thorium. Such use places stringent and diverse demands upon the fluid fuel. A brief review of chemical behavior of molten fluorides is given to show some of their strengths and weaknesses for such service

  8. Internal cation mobilities in molten lithium. Potassium fluoride

    International Nuclear Information System (INIS)

    Matsuura, Haruaki; Ohashi, Ryo; Chou, Pao-Hwa; Takagi, Ryuzo

    2006-01-01

    Relative differences between internal cation mobilities in molten (Li, K) F have been measured by countercurrent electromigration (Klemm method) at 1023 K. Internal mobilities of K + are larger than those of Li + in all composition on which we have measured so far. More striking feature is that the isotherms have minimum of mobilities at ca. x K =0.5. The local structural parameters would be highly related to the ionic conduction behavior in molten fluorides. (author)

  9. Measurement of emittance of metal interface in molten salt

    International Nuclear Information System (INIS)

    Araki, N.; Makino, A.; Nakamura, Y.

    1995-01-01

    A new technique for measuring the total normal emittance of a metal in a semi-transparent liquid has been proposed and this technique has been applied to measure the emittance of stainless steel (SUS304), nickel, and gold in molten potassium nitrate KNO 3 . These emittance data are indispensable to analyzing the radiative heat transfer between a metal and a semitransparent liquid, such as a molten salt

  10. Hydrogen permeation through Flinabe fluoride molten salts for blanket candidates

    Energy Technology Data Exchange (ETDEWEB)

    Nishiumi, Ryosuke, E-mail: r.nishiumi@aees.kyushu-u.ac.jp; Fukada, Satoshi; Nakamura, Akira; Katayama, Kazunari

    2016-11-01

    Highlights: • H{sub 2} diffusivity, solubility and permeability in Flinabe as T breeder are determined. • Effects in composition differences among Flibe, Fnabe and Flinabe are compared. • Changes of pressure dependence of Flinabe permeation rate are clarified. - Abstract: Fluoride molten salt Flibe (2LiF + BeF{sub 2}) is a promising candidate for the liquid blanket of a nuclear fusion reactor, because of its large advantages of tritium breeding ratio and heat-transfer fluid. Since its melting point is higher than other liquid candidates, another new fluoride molten salt Flinabe (LiF + NaF + BeF{sub 2}) is recently focused on because of its lower melting point while holding proper breeding properties. In this experiment, hydrogen permeation behavior through the three molten salts of Flibe (2LiF + BeF{sub 2}), Fnabe (NaF + BeF{sub 2}) and Flinabe are investigated in order to clarify the effects of their compositions on hydrogen transfer properties. After making up any of the three molten salts and purifying it using HF, hydrogen permeability, diffusivity and solubility of the molten salts are determined experimentally by using a system composed of tertiary cylindrical tubes. Close agreement is obtained between experimental data and analytical solutions. H{sub 2} permeability, diffusivity and solubility are correlated as a function of temperature and are compared among the three molten salts.

  11. Novel waste printed circuit board recycling process with molten salt.

    Science.gov (United States)

    Riedewald, Frank; Sousa-Gallagher, Maria

    2015-01-01

    The objective of the method was to prove the concept of a novel waste PCBs recycling process which uses inert, stable molten salts as the direct heat transfer fluid and, simultaneously, uses this molten salt to separate the metal products in either liquid (solder, zinc, tin, lead, etc.) or solid (copper, gold, steel, palladium, etc.) form at the operating temperatures of 450-470 °C. The PCB recovery reactor is essentially a U-shaped reactor with the molten salt providing a continuous fluid, allowing molten salt access from different depths for metal recovery. A laboratory scale batch reactor was constructed using 316L as suitable construction material. For safety reasons, the inert, stable LiCl-KCl molten salts were used as direct heat transfer fluid. Recovered materials were washed with hot water to remove residual salt before metal recovery assessment. The impact of this work was to show metal separation using molten salts in one single unit, by using this novel reactor methodology. •The reactor is a U-shaped reactor filled with a continuous liquid with a sloped bottom representing a novel reactor concept.•This method uses large PCB pieces instead of shredded PCBs as the reactor volume is 2.2 L.•The treated PCBs can be removed via leg B while the process is on-going.

  12. Novel waste printed circuit board recycling process with molten salt

    Science.gov (United States)

    Riedewald, Frank; Sousa-Gallagher, Maria

    2015-01-01

    The objective of the method was to prove the concept of a novel waste PCBs recycling process which uses inert, stable molten salts as the direct heat transfer fluid and, simultaneously, uses this molten salt to separate the metal products in either liquid (solder, zinc, tin, lead, etc.) or solid (copper, gold, steel, palladium, etc.) form at the operating temperatures of 450–470 °C. The PCB recovery reactor is essentially a U-shaped reactor with the molten salt providing a continuous fluid, allowing molten salt access from different depths for metal recovery. A laboratory scale batch reactor was constructed using 316L as suitable construction material. For safety reasons, the inert, stable LiCl–KCl molten salts were used as direct heat transfer fluid. Recovered materials were washed with hot water to remove residual salt before metal recovery assessment. The impact of this work was to show metal separation using molten salts in one single unit, by using this novel reactor methodology. • The reactor is a U-shaped reactor filled with a continuous liquid with a sloped bottom representing a novel reactor concept. • This method uses large PCB pieces instead of shredded PCBs as the reactor volume is 2.2 L. • The treated PCBs can be removed via leg B while the process is on-going. PMID:26150977

  13. Selection of Models for Ingestion Pathway and Relocation Radii Determination

    International Nuclear Information System (INIS)

    Blanchard, A.

    1998-01-01

    The distance at which intermediate phase protective actions (such as food interdiction and relocation) may be needed following postulated accidents at three Savannah River Site nonreactor nuclear facilities will be determined by modeling. The criteria used to select dispersion/deposition models are presented. Several models were considered, including ARAC, MACCS, HOTSPOT, WINDS (coupled with PUFF-PLUME), and UFOTRI. Although ARAC and WINDS are expected to provide more accurate modeling of atmospheric transport following an actual release, analyses consistent with regulatory guidance for planning purposes may be accomplished with comparatively simple dispersion models such as HOTSPOT and UFOTRI. A recommendation is made to use HOTSPOT for non-tritium facilities and UFOTRI for tritium facilities

  14. Selection of Models for Ingestion Pathway and Relocation

    International Nuclear Information System (INIS)

    Blanchard, A.; Thompson, J.M.

    1998-01-01

    The area in which intermediate phase protective actions (such as food interdiction and relocation) may be needed following postulated accidents at three Savannah River Site nonreactor nuclear facilities will be determined by modeling. The criteria used to select dispersion/deposition models are presented. Several models are considered, including ARAC, MACCS, HOTSPOT, WINDS (coupled with PUFF-PLUME), and UFOTRI. Although ARAC and WINDS are expected to provide more accurate modeling of atmospheric transport following an actual release, analyses consistent with regulatory guidance for planning purposes may be accomplished with comparatively simple dispersion models such as HOTSPOT and UFOTRI. A recommendation is made to use HOTSPOT for non-tritium facilities and UFOTRI for tritium facilities. The most recent Food and Drug Administration Derived Intervention Levels (August 1998) are adopted as evaluation guidelines for ingestion pathways

  15. Selection of Models for Ingestion Pathway and Relocation

    International Nuclear Information System (INIS)

    Blanchard, A.; Thompson, J.M.

    1999-01-01

    The area in which intermediate phase protective actions (such as food interdiction and relocation) may be needed following postulated accidents at three Savannah River Site nonreactor nuclear facilities will be determined by modeling. The criteria used to select dispersion/deposition models are presented. Several models are considered, including ARAC, MACCS, HOTSPOT, WINDS (coupled with PUFF-PLUME), and UFOTRI. Although ARAC and WINDS are expected to provide more accurate modeling of atmospheric transport following an actual release, analyses consistent with regulatory guidance for planning purposes may be accomplished with comparatively simple dispersion models such as HOTSPOT and UFOTRI. A recommendation is made to use HOTSPOT for non-tritium facilities and UFOTRI for tritium facilities. The most recent Food and Drug Administration Derived Intervention Levels (August 1998) are adopted as evaluation guidelines for ingestion pathways

  16. Understanding Housing Delays and Relocations Within the Housing First Model.

    Science.gov (United States)

    Zerger, Suzanne; Pridham, Katherine Francombe; Jeyaratnam, Jeyagobi; Hwang, Stephen W; O'Campo, Patricia; Kohli, Jaipreet; Stergiopoulos, Vicky

    2016-01-01

    This study explores factors contributing to delays and relocations during the implementation of the Housing First model in Toronto, Ontario. While interruptions in housing tenure are expected en route to recovery and housing stability, consumer and service provider views on finding and keeping housing remain largely unknown. In-person interviews and focus groups were conducted with 48 study participants, including 23 case managers or housing workers and 25 consumers. The following three factors contributed to housing delays and transfers: (1) the effectiveness of communication and collaboration among consumers and service providers, (2) consumer-driven preferences and ambivalence, and (3) provider prioritization of consumer choice over immediate housing access. Two strategies--targeted communications and consumer engagement in housing searches--supported the housing process. Several factors affect the timing and stability of housing. Communication between and among providers and consumers, and a shared understanding of consumer choice, can further support choice and recovery.

  17. Future analytical provision - Relocation of Sellafield Ltd Analytical Services Laboratory

    International Nuclear Information System (INIS)

    Newell, B.

    2015-01-01

    Sellafield Ltd Analytical Services provide an essential view on the environmental, safety, process and high hazard risk reduction performances by analysis of samples. It is the largest and most complex analytical services laboratory in Europe, with 150 laboratories (55 operational) and 350 staff (including 180 analysts). Sellafield Ltd Analytical Services Main Laboratory is in need of replacement. This is due to the age of the facility and changes to work streams. This relocation is an opportunity to -) design and commission bespoke MA (Medium-Active) cells, -) modify HA (High-Active) cell design to facilitate an in-cell laboratory, -) develop non-destructive techniques, -) open light building for better worker morale. The option chosen was to move the activities to the NNL Central laboratory (NNLCL) that is based at Sellafield and is the UK's flagship nuclear research and development facility. This poster gives a time schedule

  18. 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

  19. 41 CFR 302-3.508 - What relocation expenses are not authorized for new appointees or student trainees?

    Science.gov (United States)

    2010-07-01

    ... are not authorized for new appointees or student trainees? 302-3.508 Section 302-3.508 Public Contracts and Property Management Federal Travel Regulation System RELOCATION ALLOWANCES RELOCATION... relocation expenses are not authorized for new appointees or student trainees? You must not pay any expenses...

  20. 41 CFR 302-3.507 - Once we authorize relocation expenses for new appointees or student trainees what expenses must...

    Science.gov (United States)

    2010-07-01

    ... relocation expenses for new appointees or student trainees what expenses must we pay? 302-3.507 Section 302-3.507 Public Contracts and Property Management Federal Travel Regulation System RELOCATION ALLOWANCES....507 Once we authorize relocation expenses for new appointees or student trainees what expenses must we...

  1. 33 CFR 150.515 - What are the requirements for weight testing of newly installed or relocated craft?

    Science.gov (United States)

    2010-07-01

    ... weight testing of newly installed or relocated craft? 150.515 Section 150.515 Navigation and Navigable... testing of newly installed or relocated craft? (a) The operator must perform installation weight testing... (a) of this section, when survival crafts are relocated to another deepwater port. ...

  2. Materials testing for molten carbonate fuel cells

    International Nuclear Information System (INIS)

    Di Mario, F.; Frangini, S.

    1995-01-01

    Unlike conventional generation systems fuel cells use an electrochemical reaction between a fossil fuel and an oxidant to produce electricity through a flame less combustion process. As a result, fuel cells offer interesting technical and operating advantages in terms of conversion efficiencies and environmental benefits due to very low pollutant emissions. Among the different kinds of fuel cells the molten carbonate fuel cells are currently being developed for building compact power generation plants to serve mainly in congested urban areas in virtue of their higher efficiency capabilities at either partial and full loads, good response to power peak loads, fuel flexibility, modularity and, potentially, cost-effectiveness. Starting from an analysis of the most important degradative aspects of the corrosion of the separator plate, the main purpose of this communication is to present the state of the technology in the field of corrosion control of the separator plate in order to extend the useful lifetime of the construction materials to the project goal of 40,000 hours

  3. Molten aluminum alloy fuel fragmentation experiments

    International Nuclear Information System (INIS)

    Gabor, J.D.; Purviance, R.T.; Cassulo, J.C.; Spencer, B.W.

    1992-01-01

    Experiments were conducted in which molten aluminum alloys were injected into a 1.2 m deep pool of water. The parameters varied were (i) injectant material (8001 aluminum alloy and 12.3 wt% U-87.7 wt% Al), (ii) melt superheat (O to 50 K), (iii) water temperature (313, 343 and 373 K) and (iv) size and geometry of the pour stream (5, 10 and 20 mm diameter circular and 57 mm annular). The pour stream fragmentation was dominated by surface tension with large particles (∼30 mm) being formed from varicose wave breakup of the 10-mm circular pours and from the annular flow off a 57 mm diameter tube. The fragments produced by the 5 mm circular et were smaller (∼ mm), and the 20 mm jet which underwent sinuous wave breakup produced ∼100 mm fragments. The fragments froze to form solid particles in 313 K water, and when the water was ≥343 K, the melt fragments did not freeze during their transit through 1.2 m of water

  4. Molten salt destruction process for mixed wastes

    International Nuclear Information System (INIS)

    Upadhye, R.S.; Wilder, J.G.; Karlsen, C.E.

    1993-04-01

    We are developing an advanced two-stage process for the treatment of mixed wastes, which contain both hazardous and radioactive components. The wastes, together with an oxidant gas, such as air, are injected into a bed of molten salt comprising a mixture of sodium-, potassium-, and lithium-carbonates, with a melting point of about 580 degree C. The organic constituents of the mixed waste are destroyed through the combined effect of pyrolysis and oxidation. Heteroatoms. such as chlorine, in the mixed waste form stable salts, such as sodium chloride, and are retained in the melt. The radioactive actinides in the mixed waste are also retained in the melt because of the combined action of wetting and partial dissolution. The original process, consists of a one-stage unit, operated at 900--1000 degree C. The advanced two-stage process has two stages, one for pyrolysis and one for oxidation. The pyrolysis stage is designed to operate at 700 degree C. The oxidation stage can be operated at a higher temperature, if necessary

  5. Actuation method of molten carbonate fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Yasuhiko; Kimoto, Mamoru; Murakami, Shuzo; Furukawa, Nobuhiro

    1987-10-17

    A molten carbonate fuel cell uses reformed gas of crude fuel as fuel gas, but in this gas, CO/sub 2/ is contained in addition to H/sub 2/ and CO which participate the reaction in its fuel electrode. In order to make the reaction of the cell by these gases smoothly, CO/sub 2/ in the exhaust gas from the fuel electrode must be introduced efficiently to its oxygen electrode, however since unreacted H/sub 2/ and CO are contained in the above exhaust gas, they are oxidated and burned once in a boiler and transformed into H/sub 2/O (steam) and CO/sub 2/, then CO/sub 2/ generated in the fuel electrode is added thereto, and afterwards these gases with the air are introduced into the oxygen electrode. However, since this method hinders the high power generation efficiency, in this invention, the exhaust gas from the fuel electrode which burns the reformed gas is introduced into separation chambers separated with CO/sub 2/ permselective membranes, and the mixture of CO/sub 2/ in the above exhaust gas separated with the aforementioned permeable membranes and the air is supplied to the oxygen electrode. At the same time, H/sub 2/ and CO in the above exhaust gas which were not separated with the above permeable membranes are recirculated to the above fuel electrode. (3 figs)

  6. 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

  7. 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)

  8. 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

  9. 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.

  10. 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

  11. 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.

  12. 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)

  13. 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

  14. 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

  15. 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...

  16. Environmental Assessment for the vacuum process laboratory (VPL) relocation at the Lawrence Livermore National Laboratory

    International Nuclear Information System (INIS)

    1992-04-01

    This Environmental Assessment (EA) evaluates the potential environmental impacts of relocating a vacuum process laboratory (VPL) from Building 321 to Building 2231 at Lawrence Livermore National Laboratory (LLNL). The VPL provides the latest technology in the field of vacuum deposition of coatings onto various substrates for several weapons-related and energy-related programs at LLNL. Operations within the VPL at LLNL will not be expanded nor reduced by the relocation. No significant environmental impacts are expected as a result of the relocation of the VPL

  17. An optimized symbiotic fusion and molten-salt fission reactor system

    International Nuclear Information System (INIS)

    Blinkin, V.L.; Novikov, V.M.

    A symbiotic fusion-fission reactor system which breeds nuclear fuel is discussed. In the blanket of the controlled thermonuclear reactor (CTR) uranium-233 is generated from thorium, which circulates in the form of ThF 4 mixed with molten sodium and beryllium fluorides. The molten-salt fission reactor (MSR) burns up the uranium-233 and generates tritium for the fusion reactor from lithium, which circulates in the form of LiF mixed with BeF 2 and 233 UF 4 through the MSR core. With a CTR-MSR thermal power ratio of 1:11 the system can produce electrical energy and breed fuel with a doubling time of 4-5 years. The system has the following special features: (1) Fuel reprocessing is much simpler and cheaper than for contemporary fission reactors; reprocessing consists simply in continuous removal of 233 U from the salt circulating in the CTR blanket by the fluorination method and removal of xenon from the MSR fuel salt by gas scavenging; the MSR fuel salt is periodically exchanged for fresh salt and the 233 U is then removed from it; (2) Tritium is produced in the fission reactor, which is a much simpler system than the fusion reactor; (3) The CTR blanket is almost ''clean''; no tritium is produced in it and fission fragment activity does not exceed the activity induced in the structural materials; (4) Almost all the thorium introduced into the CTR blanket can be used for producing 233 U

  18. 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.

  19. 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.

  20. Relocation of a nucletron microselectron-HDR brachytherapy system

    Energy Technology Data Exchange (ETDEWEB)

    Bartrum, T; Tran, T; Freeman, N; Morales, J [St Vincents Hospital, Darlinghurst, NSW (Australia)

    2004-12-15

    Full text: For a period of four weeks, our clinical Nucletron microSelectron high dose rate (HDR) brachytherapy system was pulled out of clinical use and relocated to a new building. During this period decommission tests, de-wiring of the treatment unit and its associated safety system (such as radiation detector, emergency off circuits and door interlocks), transportation of all equipment, re-wiring of this equipment in the new location and recommission tests were carried out. The decommission and recommission test program was designed upon consultation with the manufacturer's (Nucletron) acceptance test procedures and work carried out by others. The ACPSEM tolerances for remote afterloaders was used as a guideline. In addition to mandatory dosimetry, positional, workstation database and safety tests, two Australian Standard compliance tests were carried out. The compliance tests involved one for remote afterloaders and another for treatment room design. This testing program was designed and implemented with the aim of ensuring ongoing safe delivery of brachytherapy doses to the patient. The testing program consisted of two parts. The first involved a series of decommissioning tests that consisted of dosimetry tests such as source and check cable positional accuracy and source calibration tests. In addition to these tests an inventory of standard plans, patient records and system configuration information was catalogued. The second part involved a series of recommission tests and involved carrying out dosimetry tests on the brachytherapy system (positional accuracy and calibration tests), simulating common treatment scenarios (prostate, cervical, vaginal and bile duct) and checking standard plans; patient records and system configuration had remained unchanged. During this period, other tests were carried out. These included Nucletron acceptance and preventative maintenance tests, Australian Standards compliance testing and integrity of network transfer of