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.

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

International Nuclear Information System (INIS)

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

1990-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1990-07-01

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

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

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

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)

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)

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-07-01

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

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

International Nuclear Information System (INIS)

1992-01-01

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

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)

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

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

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

International Nuclear Information System (INIS)

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

1982-01-01

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

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

International Nuclear Information System (INIS)

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

1982-01-01

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

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)

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

CFD approach to modeling of core-concrete interaction

International Nuclear Information System (INIS)

Vladimir V Chudanov; Anna E Aksenova; Valerii A Pervichko

2005-01-01

Full text of publication follows: A large attention is given to research behavior of concrete structures at high mechanical and thermal loadings, which those suffer at the severe accidents on Nuclear Power Plants with core melting and falling of the molten corium mass into reactor shaft. There are enough programs for analysis of heat and mass transfer processes at interaction of the molten corium with concrete. Most known among them CORCON and WECHSL, which were developed more than twenty years ago, allow considering a quasi-stationary phase decomposition of concrete and the some transition regimes. In opposing to the mentioned codes a new more generalized mathematical model and software are developed for modeling of a wide range of the heat and mass transfer processes under study of the molten core-concrete interaction. The developed mathematical model is based on the Navier-Stokes equations with variable properties with taking into account of a density jump under melting of concrete together with a heat transfer equation. The offered numerical technique is based on modern algorithms with small scheme diffusion, whose discrete approximations are constructed with use of finite-volume methods and the fully staggered grids. The developed software corresponds to modern level of development of computers and takes into account all phenomenology, used by mentioned codes, and allows to simulate the such phenomena and processes as: multidimensional heat transfer in concrete for modeling of transients for an intermediate thermal flux to concrete; direct erosion of concrete at a quasi-stationary regime of interaction with molten fuel masses; heat and mass transfer in corium and convective intermixing in a melt of corium with taking into account of its stratification on two layers of the metal and oxide components and heat transfer by radiation in a cavity of the reactor shaft; change physical properties of corium at concrete decomposition and release in corium of its

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

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

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

Radionuclide release and aerosol generation during core debris interactions with concrete

International Nuclear Information System (INIS)

Powers, D.A.

1986-01-01

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

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.

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)

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

International Nuclear Information System (INIS)

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

1992-01-01

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

A study on the modeling of molten corium-concrete interaction

International Nuclear Information System (INIS)

Park, Soo Yong

1994-02-01

The phenomenon known as molten corium concrete interaction (MCCI) has been recognized as important aspects of severe reactor accidents. The potential hazard of a MCCI is the threat to the integrity of the containment building due to the possibility of a basemat melt through, containment overpressurization by noncondensible gases, or oxidation of combustible gases. Over the past several years, a large experimental and analytical effort has been under taken in corium-concrete interaction phenomena by several organization. The purpose of this paper is to investigate the previous analytical results and computer programs, and finally to establish a new stand alone model which can predict the corium-concrete interaction. A model to predict the behavior of molten corium-concrete interaction in the reactor cavity during vessel ruptured accidents is established. Gas film model, gas bubble model, slag model and periodic contact model are employed as a major heat transfer model between corium and concrete. Solidified debris crust is considered at the boundary of molten corium. Upon the experimental observations, no layer stratification is assumed due to the strong dispersion of the metallic melt in the oxidic phase. With the assumption of temperature profile within the corium pool and crust, the temperature distribution of concrete is found by explicit solution of heat conduction equation. The sideward heat transfer rate can be obtained by considering multiplication factor to the downward heat transfer rate. The multiplication factor is treated as a user input because of its large uncertainty. Comparisons are made with two large scale experiments, SURC-2 and BETA V3.3. There is a reasonable agreement in the corium temperature, erosion depth and gas generation between the experimental data and the predicted results with periodic contact model given the uncertainties in the input data or the measurement. The gas bubble model has the highest heat transfer coefficient, and the

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

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

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

1991-12-31

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

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

International Nuclear Information System (INIS)

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

1990-01-01

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

Heat transfer in reactor cavity during core-concrete interaction

International Nuclear Information System (INIS)

Adroguer, B.; Cenerino, G.

1989-08-01

In the unlikely event of a severe accident in a nuclear power plant, the core may melt through the vessel and slump into the concrete reactor cavity. The hot mixture of the core material called corium interacts thermally with the concrete basemat. The WECHSL code, developed at K.f.K. Karlsruhe in Germany is used at the Protection and Nuclear Safety Institute (I.P.S.N.) of CEA to compute this molten corium concrete interaction (MCCI). Some uncertainties remain in the partition of heat from the corium between the basemat and the upper surrounding structures in the cavity where the thermal conditions are not computer. The CALTHER code, under development to perform a more mechanistic evaluation of the upward heat flux has been linked to WECHSL-MOD2 code. This new version enables the modelling of the feedback effects from the conditions in the cavity to the MCCI and the computation of the fraction of upward flux directly added to the cavity atmosphere. The present status is given in the paper. Preliminary calculations of the reactor case for silicate and limestone common sand (L.C.S.) concretes are presented. Significant effects are found on concrete erosion, gases release and temperature of the upper part of corium, particularly for L.C.S. concrete

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

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.)

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

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

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

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

International Nuclear Information System (INIS)

Farmer, M.T.; Basu, S.

2006-01-01

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

Viscosities of corium-concrete mixtures

International Nuclear Information System (INIS)

Seiler, J.M.; Ganzhorn, J.

1997-01-01

Severe accidents on nuclear reactors involve many situations such as pools of molten core material, melt spreading, melt/concrete interactions, etc. The word 'corium' designates mixtures of materials issued from the molten core at high temperature; these mixtures involve mainly: UO2, ZrO2, Zr and, in small amounts, Ni, Cr, Ag, In, Cd. These materials, when flowing out of the reactor vessel, may interact with the concrete of the reactor building thus introducing decomposition products of concrete into the original mixture. These decomposition products are mainly: SiO 2 , FeO, MgO, CaO and Al 2 O 3 in different amounts depending on the nature of the concrete being considered. Siliceous concrete is rich in SiO 2 , limestone concrete contains both SiO 2 and CaO. Liquidus temperatures of such mixtures are generally obove 2300 K whereas solidus temperatures are ∝1400 K. (orig.)

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

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

International Nuclear Information System (INIS)

Huntelaar, M.E.

1996-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Huntelaar, M.E.

1996-06-19

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

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)

Improvements in modelling (by ESCADRE mod1.0) radiative heat losses through gas and aerosols generated by molten corium-concrete interactions

International Nuclear Information System (INIS)

Passalacqua, R.

1996-01-01

Aerosols generated during the molten core-concrete interaction (MCCI) influence the reactor cavity thermal hydraulics: the cloud of aerosols, located inside the reactor cavity, restrains the upward-directed heat exchange consequently the cool-down of the high-temperature molten corium for a considerable period of time. IPSN is developing a computer code system for source predictions in severe accident scenarios. This code system is named ESCADRE. WECHSL/CALTHER is internal module dealing with MCCI (it is also a stand-alone code): it models the heat transfers involving the superior volume of the cavity. When modelling the upward-directed power distribution by WECHSL/CALTHER, a faster concrete basemat penetration takes place due to the low heat losses of the closed MCCI cavity enclosure. The model, here presented, is going to be validated with data from the AEROSTAT experiment. This experiment, planned at CEA Cadarache, will evaluate the influence of aerosols on the global power distribution in the reactor cavity. Radiative heat losses are important especially for cavity configurations such as those of new plant designs (equipped with a core-catcher) where the upward power losses are promoted by the corium spreading in a flat cavity

Investigation of molten corium-concrete interaction phenomena and aerosol release

International Nuclear Information System (INIS)

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

1987-01-01

The Electric Power Research Institute is sponsoring a program of laboratory investigations at Argonne National Laboratory to study the interaction between molten core materials and reactor concrete basemats during postulated severe reactor accidents, with particular emphasis on measurements of the magnitude and chemical species present in the aerosol releases. The approach in this program is to sustain internal heat generation in reactor-material corium using direct electrical heating and to develop test operating and diagnostics capabilities with a series of small- and intermediate-scale scoping tests followed by fully instrumented large-scale testing. Real reactor materials (UO 2 , ZrO 2 , oxides of stainless steel, plus metallics) are used, with small amounts of La 2 O 3 , BaO, and SrO added to simulate nonvolatile fission products. In intermediate-scale scoping tests completed to date, corium inventories of up to 29 kg have been heated with power inputs in excess of 1 kW/kg melt. The measured concrete ablation rates have ranged from 0.9 to 3.9 mm/minute. Aerosol samples have been examined using a scanning electron microscope and show submicron particles, 2-6 micrometer spheres, and agglomerates that range from a few micrometers to string 13 micrometers in length

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

Candidate molten salt investigation for an accelerator driven subcritical core

International Nuclear Information System (INIS)

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

2013-01-01

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

Candidate molten salt investigation for an accelerator driven subcritical core

Energy Technology Data Exchange (ETDEWEB)

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

2013-09-15

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

Impact on breeding rate of different Molten Salt reactor core structures

International Nuclear Information System (INIS)

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

2013-01-01

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

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

KAPOOL experiments to simulate molten corium - sacrificial concrete interaction

International Nuclear Information System (INIS)

Eppinger, B.; Fieg, G.; Tromm, W.

2001-01-01

In future Light Water Reactors special devices (core catchers) might be required to prevent containment failure by basement erosion after reactor pressure vessel melt-through during a core meltdown accident. In the planned European Pressurized Reactor (EPR) the core melt is retained in the reactor cavity for ∼ 1 h to pick up late melts after the failure of the reactor pressure vessel. The reactor cavity is protected by a layer of sacrificial concrete and closed by a melt gate at the bottom towards the spreading compartment. After erosion of the sacrificial concrete and melt-through of the gate the core melt should be distributed homogeneously into the spreading compartment. There the melt is cooled by flooding with water. The knowledge of the sacrificial concrete erosion phase in the reactor cavity is essential for the severe accident assessment. Several KAPOOL experiments have been performed to investigate the erosion of two possible compositions of sacrificial concretes using alumina-iron thermite melts as a simulant for the core melt. Erosion rates as a function of the melt temperature and the inhomogeneity of the melt front are presented in this paper. (authors)

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

International Nuclear Information System (INIS)

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

1980-01-01

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

Candidate molten salt investigation for an accelerator driven subcritical core

Science.gov (United States)

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

2013-09-01

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

Apparatus for controlling molten core debris

International Nuclear Information System (INIS)

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

1972-01-01

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

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.

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)

Calculations of core concrete interaction using MELCOR 1.8.5

Energy Technology Data Exchange (ETDEWEB)

Kim, Hwan Yeol; Song, Jin Ho; Kim, Hee Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2005-07-01

OECD/MCCI project is scheduled for 4 years from 2002. 1 to 2005. 12 to perform a series of tests through which the data for cooling the molten core spread out at the reactor cavity and for the long-term CCI (Core Concrete Interaction) are secured. This paper deals with the transient calculations of the 2-D CCI tests performed under the OECD/MCCI project by using a well-known severe accident analysis code, MELCOR 1.8.5. The CCI test was performed at the rectangular geometry with one ablative bottom wall and two ablative and two non-ablative side walls. Since the MELCOR 1.8.5 can only accommodate a cylindrical geometry, an appropriate scaling methodology was applied to adjust the geometrical difference between the CCI test and the MELCOR calculations. The default heat transfer models contained in the CORCON-Mod3 module of MELCOR 1.8.5 were used for the base case calculation. The key parameters of the CCI phenomena such as the melt temperature, concrete ablation, cavity shape, gas generation, heat transfer rate, etc. were calculated and compared with the test results. In addition, sensitivity studies with the change of the inputs and character variables of MELCOR were also included.

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

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.)

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)

Conditions for oxygen-deficient combustion during accidents with severe core concrete thermal attack

International Nuclear Information System (INIS)

Luangdilok, W.; Elicson, G.T.; Berger, W.E. Jr.

1993-01-01

This paper addresses the interactions between MCCI (molten core-concrete interactions)-induced offgas releases, mostly the combustible gases, natural circulation between the cavity and the lower containment based on recent research developments in the area of mixed convection flow (Epstein, et al., 1989; Epstein, 1988; Epstein, 1992) between compartments, and their effects on combustion in PWR containments during prolonged severe accidents. Specifically, large dry PWR containments undergoing severe core-concrete attack during station blackouts where the containment atmosphere is expected to be inerted are objects of this analysis. The purpose of this paper, given the conditions that oxygen can be brought to the cavity, is to demonstrate that consumption of most oxygen present in the containment can be achieved in a reasonable time scale assuming that combustion is not subject to flammability limits due to the high cavity temperatures. The conditions for cavity combustion depend on several factors including good gas flowpaths between the cavity and other containment regions, and combustion processes within the cavity with the hot debris acting as the ignition source

OECD MCCI 2-D Core Concrete Interaction (CCI) tests : CCI-2 test data report-thermalhydraulic results, Rev. 0 October 15, 2004.

Energy Technology Data Exchange (ETDEWEB)

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

2011-05-23

The Melt Attack and Coolability Experiments (MACE) program addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two program objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of satisfying these objectives, the Management Board (MB) approved the conduct of two long-term 2-D Core-Concrete Interaction (CCI) experiments designed to provide information in several areas, including: (i) lateral vs. axial power split during dry core-concrete interaction, (ii) integral debris coolability data following late phase flooding, and (iii) data regarding the nature and extent of the cooling transient following breach of the crust formed at the melt-water interface. This data report provides thermal hydraulic test results from the CCI-2 experiment, which was conducted on August 24, 2004. Test specifications for CCI-2 are provided in Table 1-1. This experiment investigated the interaction of a fully oxidized 400 kg

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Molten corium concrete interaction: investigation of heat transfer in two-phase flow

International Nuclear Information System (INIS)

Amizic, Milan

2014-01-01

In the context of severe accident research for the second and the third generation of nuclear power plants, there are still open issues concerning some aspects of the concrete cavity ablation during the molten corium - concrete interaction (MCCI). The determination of heat transfer along the interfacial region between the molten corium pool and the ablating basemat concrete is crucial for the assessment of concrete ablation progression and eventually the basemat melt through. For the purpose of experimental investigation of thermal hydraulics inside a liquid pool agitated by gas bubbles, the CLARA project has been launched. The CLARA experiments are performed using simulant materials and they reveal the influence of superficial gas velocity, liquid viscosity and pool geometry on the heat transfer coefficient between the internally heated liquid pool and vertical and horizontal pool walls maintained at uniform temperature. The first test campaign has been conducted with the small pool configuration (50 cm * 25 cm * 25 cm). The tests have been performed with liquids covering a wide range of dynamic viscosity from approximately 1 mPa s to 10000 mPa s and the superficial gas velocity is varied up to 8 cm/s. This thesis comprises a brief description of MCCI phenomenology, literature reviews on the existing heat transfer correlations for two phase flow and the void fraction, a description of CLARA setup, experimental results and their interpretation. The experimental results are compared with existing models and some new models for the assessment of heat transfer coefficient in two-phase flow. (author) [fr

Importance of core/concrete interactions for German risk investigations and experimental verification

International Nuclear Information System (INIS)

Rohde, J.; Hicken, E.F.; Friederichs, H.G.; Schroedl, E.

1987-01-01

The relevance of Molten Core Concrete Interactions (MCCI) for risk oriented investigations of German LWR-plants is evaluated. The problems of MCCI have been intensely investigated since the mid seventies in connection with the German Risk Study, Phase A and B on PWR plants of German design. Many examinations of both theoretical and experimental nature have led to the development of computer codes like WECHSL. The basis for verification is the internationally well accepted BETA experiment. Code WECHSL and the knowledge gained from the BETA experiment have been applied for the final investigations in German Risk Study, Phase B. Knowledge gained will be illustrated and its importance MCCI for German LWR-concepts will be shown

OECD MCCI project 2-D Core Concrete Interaction (CCI) tests : CCI-3 test data report-thermalhydraulic results. Rev. 0 October 15, 2005.

Energy Technology Data Exchange (ETDEWEB)

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

2011-05-23

The Melt Attack and Coolability Experiments (MACE) program addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two program objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of satisfying these objectives, the Management Board (MB) approved the conduct of a third long-term 2-D Core-Concrete Interaction (CCI) experiment designed to provide information in several areas, including: (i) lateral vs. axial power split during dry core-concrete interaction, (ii) integral debris coolability data following late phase flooding, and (iii) data regarding the nature and extent of the cooling transient following breach of the crust formed at the melt-water interface. This data report provides thermal hydraulic test results from the CCI-3 experiment, which was conducted on September 22, 2005. Test specifications for CCI-3 are provided in Table 1-1. This experiment investigated the interaction of a fully oxidized 375

OECD MMCI 2-D Core Concrete Interaction (CCI) tests : CCCI-1 test data report-thermalhydraulic results. Rev 0 January 31, 2004.

Energy Technology Data Exchange (ETDEWEB)

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

2011-05-23

The Melt Attack and Coolability Experiments (MACE) program addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten coreconcrete interactions under both wet and dry cavity conditions. Achievement of these two program objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of satisfying these objectives, the Management Board (MB) approved the conduct of two long-term 2-D Core-Concrete Interaction (CCI) experiments designed to provide information in several areas, including: (i) lateral vs. axial power split during dry core-concrete interaction, (ii) integral debris coolability data following late phase flooding, and (iii) data regarding the nature and extent of the cooling transient following breach of the crust formed at the melt-water interface. This data report provides thermal hydraulic test results from the CCI-1 experiment, which was conducted on December 19, 2003. Test specifications for CCI-1 are provided in Table 1-1. This experiment investigated the interaction of a fully oxidized 400 kg

Thermodynamic evaluation of the solidification phase of molten core–concrete under estimated Fukushima Daiichi nuclear power plant accident conditions

Energy Technology Data Exchange (ETDEWEB)

Kitagaki, Toru, E-mail: kitagaki.toru@jaea.go.jp; Yano, Kimihiko; Ogino, Hideki; Washiya, Tadahiro

2017-04-01

The solidification phases of molten core–concrete under the estimated molten core–concrete interaction (MCCI) conditions in the Fukushima Daiichi Nuclear Power Plant Unit 1 were predicted using the thermodynamic equilibrium calculation tool, FactSage 6.2, and the NUCLEA database in order to contribute toward the 1F decommissioning work and to understand the accident progression via the analytical results for the 1F MCCI products. We showed that most of the U and Zr in the molten core–concrete forms (U,Zr)O{sub 2} and (Zr,U)SiO{sub 4}, and the formation of other phases with these elements is limited. However, the formation of (Zr,U)SiO{sub 4} requires a relatively long time because it involves a change in the crystal structure from fcc-(U,Zr)O{sub 2} to tet-(U,Zr)O{sub 2}, followed by the formation of (Zr,U)SiO{sub 4} by reaction with SiO{sub 2}. Therefore, the formation of (Zr,U)SiO{sub 4} is limited under quenching conditions. Other common phases are the oxide phases, CaAl{sub 2}Si{sub 2}O{sub 8}, SiO{sub 2}, and CaSiO{sub 3}, and the metallic phases of the Fe–Si and Fe–Ni alloys. The solidification phenomenon of the crust under quenching conditions and that of the molten pool under thermodynamic equilibrium conditions in the 1F MCCI progression are discussed.

Sensitivity analysis using DECOMP and METOXA subroutines of the MAAP code in modelling core concrete interaction phenomena and post test calculations for ACE-MCCI experiment L-5

International Nuclear Information System (INIS)

Passalacqua, R.A.

1991-01-01

A parametric analysis approach was chosen in order to study core-concrete interaction phenomena. The analysis was performed using a stand-alone version of the MAAP-DECOMP model (DOE version). This analysis covered only those parameters known to have the largest effect on thermohydraulics and fission product aerosol release. Even though the main purpose of the effort was model validation, it eventually resulted in a better understanding of the core-concrete interaction physics and to a more correct interpretation of the ACE-MCCI L5 experimental data. Unusual low heat transfer fluxes from the debris pool to the cavity (corium surrounding volume) were modeled in order to have a good benchmark with the experimental data. Therefore, higher debris pool temperatures were predicted. In case of water flooding, as a consequence of the critical heat flux through the upper crust and the increase of the crust thickness, resulting high debris pool temperatures cause an increase in the concrete ablation rate in the short term. DECOMP model predicts a quick increase of the crust thickness and as a result, causes the quenching of the molten mass. However, especially for fast transient, phenomena of crust bridge formation can occur. Thus, the upward directed heat flux is minimized and the concrete erosion rate remains conspicuous also in the long term. The model validation is based, in these calculations, on post-test predictions using the MCCI L5 test data: these data are derived from results of the 'Molten Core Concrete Interaction' (MCCI) experiments, which in turn are part of the larger Advanced Containment Experiment (ACE) program. Other calculations were also performed for the new proposed MACE (Melt Debris Attack and Coolability) experiments simulating the water flooding of the cavity. Those calculations are preliminarily compared with the recent MACE scoping test results. (author) 4 tabs., 59 figs., 5 refs

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

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

The interaction of a core melt with concrete

International Nuclear Information System (INIS)

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

1977-01-01

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

Interation between a superheated uranium dioxide jet and cold concrete

International Nuclear Information System (INIS)

Howe, L.D.; Denham, M.K.; Turland, B.D.; Dop, L.M.G.; Humphreys, R.J.

1992-01-01

A scoping experiment has been carried out at the Winfrith Technology Centre using its Molten Fuel Test Facilities to examine the initial interaction between a fuel melt and concrete. A molten fuel simulant consisting of 81% UO 2 and 19% Mo with a large superheat (T≅3600 K) was poured onto a basaltic concrete target. Thermocouple data indicate that there was an initial high rate of ablation. The test demonstrated that in the case of such high superheats, a vigorous interaction between the jet and the target takes place, with much of the impinging material ejected within the first few seconds. There was a depression eroded into the target by the jet. The experiment has subsequently been modeled at Culham Laboratory using a version of the CORCON MCCI (molten core-concrete interaction) computer code. The calculations were able to produce a representation of this effect. The results of the experiment and the calculation have been compared with jetting correlations, and reasonable agreement has been found. We conclude by advising caution when applying the results of this isolated test to more prototypic interactions. (orig.)

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

Differences between silica and limestone concretes that may affect their interaction with corium

International Nuclear Information System (INIS)

Journeau, C.; Haquet, J. F.; Piluso, P.; Bonnet, J. M.

2008-01-01

Recent Molten Core Concrete Interaction tests performed at Argonne National Laboratory and at CEA Cadarache have shown that, whereas the ablation of limestone-rich concretes is almost isotropic, the ablation of silica-rich concretes is much faster towards the sides than towards the bottom of the cavity. The following differences exists between limestone-rich and silica-rich concretes: limestone concretes liberate about twice as much gas, at a given ablation rate than siliceous concretes (more than 50% more at constant heat flux) and this can affect pool hydraulics and crust stability: limestone concrete has a higher liquidus temperature than siliceous concrete and molten limestone concrete has a larger diffusion coefficient and can more easily dissolve a corium crust than siliceous melt; limestone aggregates are destroyed by de-carbonation at around 1000 K while silica aggregates melt only above 2000 K, so that floating silica aggregates can form cold spots increasing corium solidification near the interface; de-carbonation of limestone leads to a significant shrinkage of concrete melt volume compared to the cold solid that hampers the mechanical stability of overlying crusts; the chemical composition of molten mortar (sand + cement) and concrete (sand + gravel + cement) is close for limestone-rich concretes while it is different for siliceous concretes, so that the melt composition may vary significantly in case of non-simultaneous melting of the siliceous concrete constituents; molten silicates have a large viscosity, so that transport properties are different for the two types of concretes. The small range of plant concrete compositions that have been considered for MCCI experiments has not yet been found sufficient to determine which of the above-mentioned differences is paramount to explain the observed difference in ablation patterns. Separate Effect Tests using specially-designed 'artificial concretes' and prototypic corium would provide the necessary

A comparative analysis of molten corium-concrete interaction models employed in MELCOR and MAAP codes

International Nuclear Information System (INIS)

Park, Soo Yong; Song, Y. M.; Kim, D. H.; Kim, H. D.

1999-03-01

The purpose of this report are to identify the modelling differences by review phenomenological models related to MCCI, and to investigate modelling uncertainty by performing sensitivity analysis, and finally to identify models to be improved in MELCOR. As the results, the most important uncertain parameter in the MCCI area is the debris stratification/mixing, and heat transfer between molten corium and overlying water pool. MAAP has a very simple and flexible corium-water heat transfer model, which seems to be needed in MELCOR for evaluation of real plants as long as large phenomenological uncertainty still exists. During the corium-concrete interaction, there is a temperature distribution inside basemat concrete. This would affect the amount or timing of gas generation. While MAAP calculates the temperature distribution through nodalization methodology, MELCOR calculates concrete response based on one-dimensional steady-state ablation, with no consideration given to conduction into the concrete or to decomposition in advanced of the ablation front. The code may be inaccurate for analysis of combustible gas generation during MCCI. Thus there is a necessity to improve the concrete decomposition model in MELCOR. (Author). 12 refs., 5 tabs., 42 figs

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-11-01

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

Two-dimensional interaction of oxidic corium with concretes: The VULCANO VB test series

Energy Technology Data Exchange (ETDEWEB)

Journeau, Christophe [CEA, DEN, STRI/LMA, Cadarache, F-13108 St Paul lez Durance (France)], E-mail: christophe.journeau@cea.fr; Piluso, Pascal; Haquet, Jean-Francois; Boccaccio, Eric; Saldo, Valerie; Bonnet, Jean-Michel; Malaval, Sophie; Carenini, Laure [CEA, DEN, STRI/LMA, Cadarache, F-13108 St Paul lez Durance (France); Brissonneau, Laurent [CEA, DEN, STPA/LPC, Cadarache, F-13108 St Paul lez Durance (France)

2009-10-15

Three two-dimensional Molten Core-Concrete Interaction tests have been conducted in the VULCANO facility with prototypic oxidic corium. The major finding is that for the two tests with silica-rich concrete, the ablation was anisotropic while it was isotropic for limestone-rich concrete. The cause of this behaviour is not yet well understood. Post Test Examinations have indicated that for the silica-rich concrete, the corium melt mixed specifically with mortar, while, for limestone-rich concretes, the analysed samples were in accordance with a corium-concrete mixing. The experimental results are described and compared to numerical codes. Separate Effect Tests with Artificial Concretes and prototypic corium are proposed to understand the phenomena governing the ablation geometry.

Two-dimensional interaction of oxidic corium with concretes: The VULCANO VB test series

International Nuclear Information System (INIS)

Journeau, Christophe; Piluso, Pascal; Haquet, Jean-Francois; Boccaccio, Eric; Saldo, Valerie; Bonnet, Jean-Michel; Malaval, Sophie; Carenini, Laure; Brissonneau, Laurent

2009-01-01

Three two-dimensional Molten Core-Concrete Interaction tests have been conducted in the VULCANO facility with prototypic oxidic corium. The major finding is that for the two tests with silica-rich concrete, the ablation was anisotropic while it was isotropic for limestone-rich concrete. The cause of this behaviour is not yet well understood. Post Test Examinations have indicated that for the silica-rich concrete, the corium melt mixed specifically with mortar, while, for limestone-rich concretes, the analysed samples were in accordance with a corium-concrete mixing. The experimental results are described and compared to numerical codes. Separate Effect Tests with Artificial Concretes and prototypic corium are proposed to understand the phenomena governing the ablation geometry.

Analysis of core-concrete interaction event with flooding for the Advanced Neutron Source reactor

International Nuclear Information System (INIS)

Kim, S.H.; Taleyarkhan, R.P.; Georgevich, V.; Navarro-Valenti, S.

1993-01-01

This paper discusses salient aspects of the methodology, assumptions, and modeling of various features related to estimation of source terms from an accident involving a molten core-concrete interaction event (with and without flooding) in the Advanced Neutron Source (ANS) reactor at the Oak Ridge National Laboratory. Various containment configurations are considered for this postulated severe accident. Several design features (such as rupture disks) are examined to study containment response during this severe accident. Also, thermal-hydraulic response of the containment and radionuclide transport and retention in the containment are studied. The results are described as transient variations of source terms, which are then used for studying off-site radiological consequences and health effects for the support of the Conceptual Safety Analysis Report for ANS. The results are also to be used to examine the effectiveness of subpile room flooding during this type of severe accident

Molten Corium-Concrete Interaction Behavior Analyses for Severe Accident Management in CANDU Reactor

International Nuclear Information System (INIS)

Choi, Y.; Kim, D. H.; Song, Y. M.

2014-01-01

After the last few severe accidents, the importance of accident management in nuclear power plants has increased. Many countries, including the United States (US) and Canada, have focused on understanding severe accidents in order to identify ways to further improve the safety of nuclear plants. It has been recognized that severe accident analyses of nuclear power plants will be beneficial in understanding plant-specific vulnerabilities during severe accidents. The objectives of this paper are to describe the molten corium behavior to identify a plant response with various concrete specific components. Accident analyses techniques using ISSAC can be useful tools for MCCI behavior in severe accident mitigation

Large longitude libration of Mercury reveals a molten core.

Science.gov (United States)

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

2007-05-04

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

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

BNL program in support of LWR degraded-core accident analysis

International Nuclear Information System (INIS)

Ginsberg, T.; Greene, G.A.

1982-01-01

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-05-23

Although extensive research has been conducted over the last several years in the areas of Core-Concrete Interaction (CCI) and debris coolability, two important issues warrant further investigation. The first issue concerns the effectiveness of water in terminating a CCI by flooding the interacting masses from above, thereby quenching the molten core debris and rendering it permanently coolable. This safety issue was investigated in the EPRI-sponsored Melt Attack and Coolability Experiments (MACE) program. The approach was to conduct large scale, integral-type reactor materials experiments with core melt masses ranging up to two metric tons. These experiments provided unique, and for the most part repeatable, indications of heat transfer mechanism(s) that could provide long term debris cooling. However, the results did not demonstrate definitively that a melt would always be completely quenched. This was due to the fact that the crust anchored to the test section sidewalls in every test, which led to melt/crust separation, even at the largest test section lateral span of 1.20 m. This decoupling is not expected for a typical reactor cavity, which has a span of 5-6 m. Even though the crust may mechanically bond to the reactor cavity walls, the weight of the coolant and the crust itself is expected to periodically fracture the crust and restore contact with the melt. Although crust fracturing does not ensure that coolability will be achieved, it nonetheless provides a pathway for water to recontact the underlying melt, thereby allowing other debris cooling mechanisms to proceed. A related task of the current program, which is not addressed in this particular report, is to measure crust strength to check the hypothesis that a corium crust would not be strong enough to sustain melt/crust separation in a plant accident. The second important issue concerns long-term, two-dimensional concrete ablation by a prototypic core oxide melt. As discussed by Foit the existing

Design and analysis of concrete reactor vessels: New developments, problems and trends

International Nuclear Information System (INIS)

Bazant, Z.P.

1984-01-01

This lecture reviews new developments in analysis and design of prestressed concrete reactor vessels (PCRV). After a brief assessment of the current status and experience, the advantages, disadvantages, and especially the safety features of PCRV, are discussed. Attention is then focused on the design of penetrations and openings, and on the design for high-temperature resistance - areas in which further developments are needed. Various possible designs for high-temperature exposure of concrete in a hypothetical accident are analyzed. Considered are not only PCRVs for gas-cooled reactors (GCR), but also guard vessels for liquid metal fast breeder reactors (LMFBR), for which designs mitigating the adverse effects of molten sodium, molten steel, and core melt are surveyed. Realistic analysis of the problems requires further development in the knowledge of material behavior and its mathematical modeling. Recent advances in the modeling of high-temperature response of concrete, including pore water transfer, pore pressure, creep and shrinkage are outlined. This is followed by a discussion of new developments in the analysis of cracking of concrete, where the need of switching from stress criteria to energy criteria for fracture is emphasized. The lecture concludes with a brief discussion of long-time behavior, the effect of aging, and probabilistic analysis of creep. (orig.)

Decontamination and concrete core sampling by teleoperated robot at Fukushima Daiichi reactor buildings

International Nuclear Information System (INIS)

Watanabe, Masaru; Onitsuka, Hironori; Shimonabe, Noriaki; Fujita, Jun; Matsumura, Takumi; Okumura, Atsushi

2015-01-01

For decommissioning of Fukushima daiichi nuclear power station, reduction of the dose equivalent rates inside the reactor buildings is an important issue. Concrete core sampling from the buildings to investigate the contamination is necessary for study about effective decontamination. However, dose rate inside the reactor buildings is very high. For example, dose rate of 1st floor on the Unit 1 is 1.2 - 1820 [mSv / h], the Unit 2 is 2.5 - 220 [mSv / h] and Unit 3 is 2.2 - 4780 [mSv / h]. So it is difficult for workers to work long hours. Therefore, a teleoperated robot, named 'MHI-MEISTeR (Mitsubishi Heavy Industries - Maintenance Equipment Integrated System of Telecontrol Robot)', has been developed to conduct operations like concrete core samples from the reactor buildings. Actually, some concrete core samples from Fukushima daiichi were taken by MHI-MEISTeR. In addition, MHI-MEISTeR is designed as a versatile robot, and so it can conduct suction / blast decontamination works as well as concrete core sampling. The above operations were performed by MHI-MEISTeR in Fukushima daiichi nuclear power station. (author)

Analysis of top flooding during molten corium concrete interaction (MCCI) with the code MEDICIS using a simplified approach for the combined effect of crust formation and boiling

International Nuclear Information System (INIS)

Spengler, C.

2012-01-01

The objective of this work is to provide adequate models in the code MEDICIS for the molten corium concrete interaction (MCCI) phase in a severe accident. Here, the multidimensional distribution of heat fluxes from the molten pool of corium to the sidewall and bottom wall concrete structures in the reactor pit and to the top surface is a persistent subject of international research activities on MCCI. In recent experi-ments with internally heated oxide melts it was observed that the erosion progress may be anisotropic - with an apparent preference of the sidewall compared to the bottom wall - or isotropic, in dependence of the type of concrete with which the cori-um interacts. The lumped parameter code MEDICIS, which is part of the severe accident codes ASTEC and COCOSYS - developed and used at IRSN/GRS respectively GRS for the latter one -, is dedicated to simulate the phenomenology during MCCI. In this work a simplified modelling in MEDICIS is tested to account for the observed ablation behaviour during MCCI, with focus on the heat transfer to the top surface under flooded conditions. This approach is assessed by calculations for selected MCCI experiments involving the top flooding of the melt. (orig.)

Materials problems related to the core catcher of sodium cooled reactors

International Nuclear Information System (INIS)

Goetzmann, O.

1975-05-01

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

KATS experiments to simulate corium spreading in the EPR core catcher concept

International Nuclear Information System (INIS)

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

2001-01-01

In future Light Water Reactors special devices (core catchers) might be required to prevent containment failure by basement erosion after reactor pressure vessel melt-through during a core meltdown accident. Quick freezing of the molten core masses is desirable to reduce release of radioactivity. Several concepts of core catcher de-vices have been proposed based on the spreading of corium melt onto flat surfaces with subsequent cooling by flooding with water. Therefore a series of experiments to investigate high temperature melt spreading on flat surfaces has been carried out using alumina-iron thermite melts as a simulant. The oxidic thermite melt is conditioned by adding other oxides to simulate a realistic corium melt as close as possible. Spreading of oxidic and metallic melts have been performed in one- and two-dimensional geometry. Substrates were chemically inert ceramic layers, dry concrete and concrete with a shallow water layer on top. (authors)

The WECHSL-Mod2 code: A computer program for the interaction of a core melt with concrete including the long term behavior

International Nuclear Information System (INIS)

Reimann, M.; Stiefel, S.

1989-06-01

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

Fission product release from core-concrete mixtures

International Nuclear Information System (INIS)

Roche, M.F.; Settle, J.; Leibowitz, L.; Johnson, C.E.; Ritzman, R.L.

1988-01-01

The objective of this research is to measure the amount of strontium, barium, and lanthanum that is vaporized from core-concrete mixtures. The measurements are being done using a transpiration method. Mixtures of limestone-aggregated concrete, urania doped with a small amount of La, Sr, Ba, and Zr oxides, and stainless steel were vaporized at 2150 K from a zirconia crucible into flowing He-6% H 2 -0.06% H 2 O (a partial molar free energy of oxygen of -420 kJ). The amounts that were vaporized was determined by weight change and by chemical analyses on condensates. The major phases present in the mixture were inferred from electron probe microanalysis (EPM). They were: (1) urania containing calcia and zirconia, (2) calcium zirconate, (3) a calcium magnesium silicate, and (4) magnesia. About 10% of the zirconia crucible was dissolved by the concrete-urania mixture during the experiment, which accounts for the presence of zirconia-containing major phases. To circumvent the problem of zirconia dissolution, we repeated the experiments using mixtures of the limestone-aggregate concrete and the doped urania in molybdenum crucibles. These studies show that thermodynamic calculations of the release of refractory fission products will yield release fractions that are a factor of sixteen too high if the effects of zirconate formation are ignored

CORCON-MOD1 modelling improvements

International Nuclear Information System (INIS)

Corradini, M.L.; Gonzales, F.G.; Vandervort, C.L.

1986-01-01

Given the unlikely occurrence of a severe accident in a light water reactor (LWR), the core may melt and slump into the reactor cavity below the reactor vessel. The interaction of the molten core with exposed concrete (a molten-core-concrete-interaction, MCCI) causes copious gas production which influences further heat transfer and concrete attack and may threaten containment integrity. In this paper the authors focus on the low-temperature phase of the MCCI where the molten pool is partially solidified, but is still capable of attacking concrete. The authors have developed some improved phenomenological models for pool freezing and molten core-coolant heat transfer and have incorporated them into the CORCON-MOD1 computer program. In the paper the authors compare the UW-CORCON/MOD1 calculations to CORCON/MOD2 and WECHSL results as well as the BETA experiments which are being conducted in Germany

Interaction and penetration of heated UO2 with limestone concrete

International Nuclear Information System (INIS)

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

1982-01-01

To safeguard the environment against radiological releases, the major question of concern in PAHR safety assessment, following an HCDA, involves confinement and dilution of the molten core-debris. Significant to the study is the directional growth of the core-debris in the concrete foundation of the reactor building or the concrete below the reactor cavity. The real material experiments were carried out in the test apparatus shown. Casts of CRBRP limestone concrete were prepared in graphite cylinders, each having an internal diameter of 8.9 cm and a depth of 30.5 cm. The 17.8-cm-deep concrete samples were allowed to cure for at least 28 days. Experiments were conducted within two months of curing time. The cavity above concrete was packed with 3 kg of pure UO 2 particles (1 to 3 mm). A uranothermic mixture was placed on the top of UO 2 powder. Heating and possible melting of UO 2 was achieved resistively after the ignition of the thermite. Total experimental time was about 60 minutes, during which time a maximum electrical power input of 1.8 watts/gr was applied to the UO 2 . Three gas samples were taken at temperatures of 100, 600, and 950 0 C, measured in the plane of the No. 2 thermocouple. Selection of three temperatures were to study the amount and the type of gases released from different phases of concrete

Response of concrete exposed to a high heat flux on one surface

International Nuclear Information System (INIS)

Muir, J.F.

1977-11-01

Experiments were performed to investigate the response of concrete to severe thermal environments such as might be encountered during the interaction of molten reactor core materials with the containment substructure following a hypothetical fuel melt accident. The dominant mechanism for erosion of both limestone and basaltic concrete appears to be melting of the cementitious material in the matrix. The erosion proceeded in a quiescent manner with negligible spallation. The erosion rate increased with heat flux, becoming as large as approximately 70 cm/hr for a net surface heat flux of roughly 190 W/cm 2 . Analyses reveal the surface temperature to be the single most significant parameter affecting the net surface heat flux, through its importance to emitted radiation; and that the greatest fraction of the net energy transmitted to the concrete goes into sensible heat

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

International Nuclear Information System (INIS)

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

1996-09-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-09-15

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

Numerical Analysis of Heat Storage and Heat Conductivity in the Concrete Hollow Core Deck Element

DEFF Research Database (Denmark)

Pomianowski, Michal Zbigniew; Heiselberg, Per; Jensen, Rasmus Lund

2011-01-01

extent these simplified models estimate the heat storage potential of precast hollow-core concrete decks correctly. This study investigates various approaches on how to model the heat transfer within the air void in the deck. Furthermore, it is analysed how different heat transfer models influence...... the overall heat transfer and heat storage in the hollow-core decks. The presented results allow comparison between detailed results from 2D-COMSOL simulations and simple 1D calculations from the whole building simulation tool such as BSim program and moreover, it is possible to validate the calculation...... method in BSim for the concrete deck element with air voids. Finally, this paper presents a comparison of the calculated heat conductivity of the hollow-core concrete deck and the measured heat conductivity for the same deck by using hot box apparatus....

Influence of gas generation on high-temperature melt/concrete interactions

International Nuclear Information System (INIS)

Powers, D.A.

1979-01-01

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

Study of technological features of tubular compressed concrete members in concreting

Directory of Open Access Journals (Sweden)

Voskobiinyk Olena

2017-01-01

Full Text Available The technological features of core concreting were analyzed as the main factor in ensuring of strength and reliability of compressed concrete-filled steel tubular (CFST members. We have conducted the analysis of existing concreting methods of CFST members. In this respect, the most dangerous types of possible technological defects of concrete core of CFST members are inhomogeneity along the height, voids, caverns, and concrete “weak spots”. The authors considered the influence of such technological factors of concreting: placeability, time, concrete mixture compaction method, concreting height on the concrete core strength of CFST members. Based on the experimental studies conducted we suggested the regression correlations for determining the concrete strength of CFST members of different length depending on the movability of concrete mixture and a time for its compaction. The authors performed the correlation analysis of technological factors of concreting on the strength of the concrete core. We carried out the comparison of data on the concrete core strength of CFST members, that were determined by non-destructive methods (sclerometer test results, ultrasonic method and direct compression strength tests. We experimentally proved that using movable mixtures with the slump of about 4 – 9 cm the overall variation coefficient of concrete core strength of CFST members along the height reaches nearly 13%. Based on the experimental studies conducted we suggested the guidelines on optimal regimes of concrete compaction during manufacturing CFST members at a construction site environment.

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

International Nuclear Information System (INIS)

Conocar, O.

2007-06-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Conocar, O

2007-06-15

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

Molten salt fueled reactors with a fast salt draining

International Nuclear Information System (INIS)

Ventre, Edmond; Blum, J.M.

1976-01-01

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Petrographic Analysis of Portland Cement Concrete Cores from Pease Air National Guard Base, New Hampshire

Science.gov (United States)

2016-11-01

Petrographic Analysis of Portland Cement Concrete Cores from Pease Air National Guard Base, New Hampshire E n g in e e r R e s e a rc h a n d...id, age of the concrete being evaluated and tests performed...4 3 Preface This study was conducted in support of the Air Force Civil Engineer Center (AFCEC) to assess concrete obtained from Pease

Core reilforced braided composite armour as a substitute to steel in concrete reinforcement

OpenAIRE

Fangueiro, Raúl; Sousa, Guilherme José Miranda de; Araújo, Mário Duarte de; Pereira, C. Gonilho; Jalali, Said

2006-01-01

This paper presents the work that is being done at the University of Minho concerning the development of brainded rods concrete reinforcement. Several samples of core reinforced braided fabrics have been produced varying the type of braided fabric (core reinforced and hybrid), the linear density of the core reinforcing yarns and the type of braiding structure (with or without ribs). The tensile properties of braided fabrics has also been analysed. Core reinforced braided composites rods were ...

Dynamic and Static Behavior of Hollow-Core FRP-Concrete-Steel and Reinforced Concrete Bridge Columns under Vehicle Collision

Directory of Open Access Journals (Sweden)

Omar I. Abdelkarim

2016-12-01

Full Text Available This paper presents the difference in behavior between hollow-core fiber reinforced polymer-concrete-steel (HC-FCS columns and conventional reinforced concrete (RC columns under vehicle collision in terms of dynamic and static forces. The HC-FCS column consisted of an outer FRP tube, an inner steel tube, and a concrete shell sandwiched between the two tubes. The steel tube was hollow inside and embedded into the concrete footing with a length of 1.5 times the tube diameter while the FRP tube stopped at the top of footing. The RC column had a solid cross-section. The study was conducted through extensive finite element impact analyses using LS-DYNA software. Nine parameters were studied including the concrete material model, unconfined concrete compressive strength, material strain rate, column height-to-diameter ratio, column diameter, column top boundary condition, axial load level, vehicle velocity, and vehicle mass. Generally, the HC-FCS columns had lower dynamic forces and higher static forces than the RC columns when changing the values of the different parameters. During vehicle collision with either the RC or the HC-FCS columns, the imposed dynamic forces and their equivalent static forces were affected mainly by the vehicle velocity and vehicle mass.

The calculation of phase equilibria of oxide core-concrete systems

International Nuclear Information System (INIS)

Ball, R.G.J.; Mignanelli, M.A.; Barry, T.I.; Gisby, J.A.

1993-01-01

Thermodynamic models have been developed to describe the phase equilibria of oxide solutions appropriate for the understanding of the chemical interactions between nuclear reactor core debris and concrete. For this purpose, the Gibbs energy of the liquid phase is described by the inclusion of associate species and nonideal interactions between the components and associate species. Assessments of the thermodynamic and phase equilibrium data for the subsystems of the CaO-Al 2 O 3 -SiO 2 -UO 2 -ZrO 2 system have been used to obtain a thermodynamic description of the crystalline and liquid phases in good agreement with published data. The data for the subsystems have then been combined, using well established principles, to predict the phase relationships in the ternary and quaternary sytsems and in the overall quinary system. The results show that he overall system cannot properly be treated as a pseudo-ideal liquid and solid solution, as used in some computer codes which attempt to model the physics and chemistry of core-concrete interactions. The limitations of the current model are discussed. (orig.)

A transient analysis of decomposition and erosion of concrete exposed to a surface heat flux

International Nuclear Information System (INIS)

Kilic, A.N.

1994-01-01

A simple approximation for predicting the concrete erosion rate and depth is derived based on the heat balance integral method for conduction with the time-dependent boundary conditions. The problem is considered a four-region model including separate, moving heat sinks at the boundaries due to endothermic decomposition reactions. Polynomial temperature profiles are assumed, and the results are compared with previous experimental data and other analytical solutions. Since the technique provides an approximate temperature distribution on the average, it does not give the real temperature evaluation but provides a simple prediction of the erosion rates and the depth of defaulted concrete in terms of the parameters that are important during the physical phenomena. Because of its simplicity and reliability, the model might be useful for the larger molten core/concrete interaction codes and aerosol generation models

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

MCCI study for Pressurized Heavy Water Reactor under hypothetical accident condition

International Nuclear Information System (INIS)

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

2011-01-01

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

Thermal protection system for the concrete core support floor at Fort St. Vrain

International Nuclear Information System (INIS)

Jones, H.; Hedgecock, P.D.

1976-01-01

A unique feature of the Fort St. Vrain HTGR is its steel jacketed concrete core support floor. The construction of this floor generally resembles that of the prestressed concrete reactor vessel, but its location immediately below the core hot gas outlets generates some particularly severe thermal protection requirements. A thermal barrier is used over the entire outer surface of the floor and in the twelve hot gas ducts which convey the primary coolant through the floor to the steam generators. A cooling water system of square tubes welded to the inside of the steel jacket is used to remove that heat which does pass through the thermal barrier and to maintain the concrete at acceptable temperatures. The design approach to the floor itself and to the thermal barriers and cooling system will be described, but the main emphasis of the paper will be on the total experience gained during construction and pre-operational testing. A particular problem experienced during construction was leakage from some cooling tubes, after their embedment in concrete. The solution to that problem was to develop a method for injecting catalyzed epoxy into the leaking tube. This method, which has general usefulness for in-service repairs, will be described. (author)

HECLA experiments on interaction between metallic melt and hematite-containing concrete

Energy Technology Data Exchange (ETDEWEB)

Sevon, Tuomo, E-mail: tuomo.sevon@vtt.f [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT, Espoo (Finland); Kinnunen, Tuomo; Virta, Jouko; Holmstroem, Stefan; Kekki, Tommi; Lindholm, Ilona [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT, Espoo (Finland)

2010-10-15

In a hypothetical severe accident in a nuclear power plant, molten materials may come into contact with concrete, causing concrete ablation. In five HECLA experiments the interaction between metallic melt and concrete was investigated by pouring molten stainless steel at almost 1800 {sup o}C into cylindrical concrete crucibles. The tests were transient, i.e. no decay heat simulation was used. The main objective was to test the behavior of the FeSi concrete, containing hematite (Fe{sub 2}O{sub 3}) and siliceous aggregates. This special concrete type is used as a sacrificial layer in the Olkiluoto 3 EPR reactor pit, and very scarce experimental data is available about its behavior at high temperatures. It is concluded that no clear differences between the ablation of FeSi concrete and ordinary siliceous concrete were observed. The ablation depths were small, 25 mm at maximum. No dramatic effects, such as cracking of large pieces of concrete due to the thermal shock, took place. An important side result of the test series was gaining knowledge of the properties of the special concrete type. Chemical analyses were conducted and mechanical properties were measured.

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.

Thermohydraulics in a high-temperature gas-cooled reactor prestressed-concrete reactor vessel during unrestricted core-heatup accidents

International Nuclear Information System (INIS)

Kroeger, P.G.; Colman, J.; Araj, K.

1983-01-01

The hypothetical accident considered for siting considerations in High Temperature Gas-Cooled Reactors (HTGR) is the so called Unrestricted Core Heatup Accident (UCHA), in which all forced circulation is lost at initiation, and none of the auxillary cooling loops can be started. The result is a gradual slow core heatup, extending over days. Whether the liner cooling system (LCS) operates during this time is of crucial importance. If it does not, the resulting concrete decomposition of the prestressed concrete reactor vessel (PCRV) will ultimately cause containment building (CB) failure after about 6 to 10 days. The primary objective of the work described here was to establish for such accident conditions the core temperatures and approximate fuel failure rates, to check for potential thermal barrier failures, and to follow the PCRV concrete temperatures, as well as PCRV gas releases from concrete decomposition. The work was done for the General Atomic Corporation Base Line Zero reactor of 2240 MW(t). Most results apply at least qualitatively also to other large HTGR steam cycle designs

Melting experiment on concrete waste using a hollow type plasma torch mounted on furnace

International Nuclear Information System (INIS)

Moon, Y. P.; Kim, T. W.; Kim, H. S.; Shin, S. U.; Lee, M. C.

2000-01-01

A furnace coupled with a hollow type plasma torch was manufactured and installed in order to develop a volume reduction technology for non-combustible radioactive waste using plasma. A melting test with 10kg of concrete waste was carried out for the evaluation of melting characteristics in the non-transferred operation mode for 20 minutes with the melter. Feeded concrete was completely melted. However, the molten bath was not easily discharged because of its high viscosity. It was found that some molten slag spat from the molten bath was coated on the surface of torch which was mounted vertically inside furnace

Application of the core-concrete interaction code Wechsl to reactor case

International Nuclear Information System (INIS)

Cenerino, G.

1986-09-01

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

Intermediate-scale tests of sodium interactions with calcite and dolomite aggregate concretes

International Nuclear Information System (INIS)

Randich, E.; Acton, R.U.

1983-09-01

Two intermediate-scale tests were performed to compare the behavior of calcite and dolomite aggregate concretes when attacked by molten sodium. The tests were performed as part of an interlaboratory comparison between Sandia National Laboratories and Hanford Engineering Development Laboratories. Results of the tests at Sandia National Laboratories are reported here. The results show that both concretes exhibit similar exothermic reactions with molten sodium. The large difference in reaction vigor suggested by thermodynamic considerations of CO 2 release from calcite and dolomite was not realized. Penetration rates of 1.4 to 1.7 mm/min were observed for short periods of time with reaction zone temperatures in excess of 800 0 C during the energetic attack. The penetration was not uniform over the entire sodium-concrete contact area. Rapid attack may be localized due to inhomogeneities in the concrete. The chemical reaction zone is less then one cm thick for the calcite concrete but is about seven cm thick for the dolomite concrete

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

Relating the structural strength of concrete sewer pipes and material properties retrieved from core samples

NARCIS (Netherlands)

Stanic, N.; Langeveld, J.G.; Salet, Theo; Clemens, F.H.L.R.

2016-01-01

Drill core samples are taken in practice for an analysis of the material characteristics of concrete pipes in order to improve the quality of the decision-making on rehabilitation actions. Earlier research has demonstrated that core sampling is associated with a significant uncertainty. In this

Heat Storage Performance of the Prefabricated Hollow Core Concrete Deck Element with Integrated Microencapsulated Phase Change Material

DEFF Research Database (Denmark)

Pomianowski, Michal Zbigniew; Heiselberg, Per; Jensen, Rasmus Lund

2012-01-01

The paper presents the numerically calculated dynamic heat storage capacity of the prefabricated hollow core concrete deck element with and without microencapsulated phase change material (PCM). The reference deck is the ordinary deck made of standard concrete material and that is broadly used...

Ex-Vessel Core Melt Modeling Comparison between MELTSPREAD-CORQUENCH and MELCOR 2.1

Energy Technology Data Exchange (ETDEWEB)

Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Farmer, Mitchell [Argonne National Lab. (ANL), Argonne, IL (United States); Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2014-03-01

System-level code analyses by both United States and international researchers predict major core melting, bottom head failure, and corium-concrete interaction for Fukushima Daiichi Unit 1 (1F1). Although system codes such as MELCOR and MAAP are capable of capturing a wide range of accident phenomena, they currently do not contain detailed models for evaluating some ex-vessel core melt behavior. However, specialized codes containing more detailed modeling are available for melt spreading such as MELTSPREAD as well as long-term molten corium-concrete interaction (MCCI) and debris coolability such as CORQUENCH. In a preceding study, Enhanced Ex-Vessel Analysis for Fukushima Daiichi Unit 1: Melt Spreading and Core-Concrete Interaction Analyses with MELTSPREAD and CORQUENCH, the MELTSPREAD-CORQUENCH codes predicted the 1F1 core melt readily cooled in contrast to predictions by MELCOR. The user community has taken notice and is in the process of updating their systems codes; specifically MAAP and MELCOR, to improve and reduce conservatism in their ex-vessel core melt models. This report investigates why the MELCOR v2.1 code, compared to the MELTSPREAD and CORQUENCH 3.03 codes, yield differing predictions of ex-vessel melt progression. To accomplish this, the differences in the treatment of the ex-vessel melt with respect to melt spreading and long-term coolability are examined. The differences in modeling approaches are summarized, and a comparison of example code predictions is provided.

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

Interaction of concretes with oxide + metal corium. The VULCANO VBS series

International Nuclear Information System (INIS)

Journeau, Christophe; Bonnet, Jean-Michel; Ferry, Lionel; Haquet, Jean-Francois; Piluso, Pascal

2009-01-01

In the hypothetical case of a severe accident, the reactor core could melt and the formed mixture, called corium, could melt through the vessel and interact with the reactor pit concrete. Corium is made from a UO 2 -rich oxidic part, in which most of the decay heat is dissipated, and a metallic part, mainly molten steel. Up to now, due to experimental constraints, most of the experiments have been performed with solely oxidic prototypic corium, or where designed so that most of the simulated decay heat was dissipated in the metallic layer. An experimental program has been set up in the VULCANO facility in which oxidic and metallic mixtures are melted in separate furnaces and poured in a concrete cavity. Induction heating is provided to the pool upper part thanks to shielding coils, so that, in case of stratification, the lighter oxidic corium-concrete mixture receives most of the power. Three experiments have been conducted: one with a limestone-rich concrete and two with a silica-rich concrete. Metal stratification has been determined from modifications of the corium electrical properties in front of the inductor and is in good accordance with calculations. Concrete ablation has been monitored. A significant vertical ablation has been observed, even in case of silica-rich concretes, for which largely radial ablation has been observed in the case of pure oxidic corium melts. Post Test Examinations have shown unexpected repartitions of metal in the pool. (author)

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

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.

Simulation experiments on the radial pool growth in gas-releasing melting system

International Nuclear Information System (INIS)

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

1983-01-01

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

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)

Concrete - Opalinus clay interaction: in-situ experiment and technique for coring undisturbed interfaces

International Nuclear Information System (INIS)

Maeder, U.; Dolder, F.; Jenni, A.; Schwyn, B.; Frieg, B.; Eul, A.

2012-01-01

undisturbed samples of the different interfaces with concrete. A first sampling campaign in 2009 used a simple stabilisation technique with a central anchor rod that was glued in before coring. It was impossible to retrieve completely undisturbed samples. A new technique was developed for the drilling campaign during February 2012, and it was successfully applied in all four sampling boreholes carried out, retrieving a total of 150 kg of core and 10 physically and chemically undisturbed interface samples. The technique comprised intersection drilling at 45 deg. inclination and 220 mm OD to within 50 cm of the vertical concrete pile. The base was reamed planar, and templates were installed to drill a circular arrangement of 6 boreholes with 46 mm OD, three at a time. These small boreholes extended across the entire pile (1.4-1.6 m), and anchor rods made of fibre glass and filled with cement were embedded with epoxy resin. A different template was subsequently used to over-core (131 mm OD / 101 mm core DM, double-barrel, acrylic liner) cutting through the reinforcements. Stabilized composite cores of 1.4-1.6 m length could be retrieved in this manner. (authors)

COMSORS: A light water reactor chemical core catcher

International Nuclear Information System (INIS)

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

1997-01-01

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

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

International Nuclear Information System (INIS)

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

1995-02-01

The WECHSL-Mod3 code is a mechanistic computer code developed for the analysis of the thermal and chemical interaction of initially molten reactor materials with concrete in a two-dimensional as well as in a one-dimensional, axisymmetrical concrete cavity. The code performs calculations from the time of initial contact of a hot molten pool over start of solidification processes until long term basemat erosion over several days with the possibility of basemat penetration. It is assumed that an underlying metallic layer exists covered by an oxidic layer or that only one oxidic layer is present which can contain a homogeneously dispersed metallic phase. Heat generation in the melt is by decay heat and chemical reactions from metal oxidation. Energy is lost to the melting concrete and to the upper containment by radiation or evaporation of sumpwater possibly flooding the surface of the melt. Thermodynamic and transport properties as well as criteria for heat transfer and solidification processes are internally calculated for each time step. Heat transfer is modelled taking into account the high gas flux from the decomposing concrete and the heat conduction in the crusts possibly forming in the long term at the melt/concrete interface. The CALTHER code (developed at CEA, France) which models the radiative heat transfer from the upper surface of the corium melt to the surrounding cavity is implemented in the present WECHSL version. The WECHSL code in its present version was validated by the BETA, ACE and SURC experiments. The test samples include a BETA and the SURC2 post test calculations and a WECHSL application to a reactor accident. (orig.) [de

Indian programme on molten salt cooled nuclear reactors

International Nuclear Information System (INIS)

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

2013-01-01

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

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

Science.gov (United States)

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

2002-01-29

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

Simulation experiment on the flooding behaviour of core melts: KATS-9

International Nuclear Information System (INIS)

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

2000-11-01

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

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

International Nuclear Information System (INIS)

Hery, M.; Lecocq, A.

1983-03-01

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

Chemistry and technology of Molten Salt Reactors - history and perspectives

International Nuclear Information System (INIS)

Uhlir, Jan

2007-01-01

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

Approximate model for calculating overall heat transfer between overlying immiscible liquid layers with bubble-induced liquid entrainment

International Nuclear Information System (INIS)

Greene, G.A.; Schwarz, C.E.

1982-01-01

In the event a commercial power reactor is subjected to a Class 9 accident resulting in gross core melting and reactor pressure vessel penetration, it has been shown that the containment integrity may subsequently be threatened by steam overpressurization, combustible gas reactions, and basemat penetration. A major contributor to these events would be the interaction of molten core debris with the structural concrete. Modeling of core-concrete interactions involves many poorly understood and complicated heat transfer phenomena for which there exists a sparse data base. One of these phenomena, which has been shown to have significant impact upon code calculations of core-concrete interactions, is the rate of heat transfer between overlying immiscible layers of core oxides and molten metals whose interface is agitated by transverse gas flow. A mathematical model is developed to analyze this heat transfer

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...

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)

Structural Behavior of Fibrous Reinforced Concrete Hollow Core One-Way Slabs Strengthening by C.F.R.P

Directory of Open Access Journals (Sweden)

وصيف مجيد

2016-02-01

Full Text Available A reinforced concrete hollow core one-way slab is one of the types of slabs used widely around the world in residential and industrial buildings to take advantage of them Economic and thermal insulation as well as to reduce the self-weight of the construction. The aim of the present study is to examine the structural behavior of the reinforced concrete hollow core one-way slabs reduce failure using the normal concrete and fibrous concrete and then strengthened using carbon fiber(CFRPThis study include molding of ( 6 specimens differ in terms of the voids volume (Vv , volumetric percentage of steel fibers (ا, and then strengthened by using fibers of carbon , with the aim of rehabilitation by fibers, carbon polymer (CFRP is to find out how efficient element structural when treated after the occurrence of the failure and the validity of its use in the event of a failure has occurred entirely or partly in the roof, and re- examined using the same method and conditions that were examined ceilings is affected through it, knowing that these ceilings have been addressed and strengthened in the same way , the results of the tests of the models that have been rehabilitated using carbon fiber (CFRP, compared with the same models before strengthening and examined reduce failure, increased very high susceptibility endurance extreme , with the increase ranging from (51.6% to (96.2%, as has been observed decrease in deflection value of models after strengthening by (CFRP.It is concluded through this study the possibility of using its concrete hollow core one-way slab as a roofing system for buildings also proved the highly efficient for this slab after rehabilitation using carbon fiber (CFRP.

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Nuclear core catchers

International Nuclear Information System (INIS)

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

1976-01-01

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

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

International Nuclear Information System (INIS)

Boussier, H.; Heuer, D.

2010-01-01

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

Fast Thorium Molten Salt Reactors Started with Plutonium

International Nuclear Information System (INIS)

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

2006-01-01

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

Effects of the presence of core debris on the behavior of sodium-concrete reactions

International Nuclear Information System (INIS)

Nguyen, D.H.; Muhlestein, L.D.

1984-01-01

Calculations using the SOCON model indicated the following: the temperature was increased throughout the concrete and the reaction product layer. Temperature could be raised to above sodium bp. Rate of release and accumulation of water and CO 2 gas were increased. The sodium mass transport to the reaction surface was also increased. As a consequence, more hydrogen and chemical heat were produced. The probability of concrete mechanical failure was higher. Sodium boiling inside the reaction product layer would not significantly alter the course of the reaction, unless it could reduce the rate of sodium transport. Although the chemical heat dominated during the early period, the decay heat could become the main source later. The reactions were driven by three main heat sources: the chemical heat, core debris heat and conduction heat from the hot sodium pool. The latter could become a heat sink. Even with the presence of core debris, the chemical reaction penetration was self-limiting and eventually, the reaction penetration rate decreased to a small value

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Neutron shielding studies on an advanced molten salt fast reactor design

International Nuclear Information System (INIS)

Merk, Bruno; Konheiser, Jörg

2014-01-01

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

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)

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-01

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

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

International Nuclear Information System (INIS)

Xia Bing; Jing Xingqing; Xu Xiaolin; Lv Yingzhong

2013-01-01

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

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)

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

International Nuclear Information System (INIS)

Wang Shisheng; Rineiski, Andrei; Maschek, Werner

2006-01-01

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

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

International Nuclear Information System (INIS)

Sugiyama, Ken-ichiro; Iguchi, Kentarou

2002-03-01

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

Molten Fuel Mass Assessment for Channel Flow Blockage Event in CANDU6

International Nuclear Information System (INIS)

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

2011-01-01

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

Cold crucible technique for interaction test of molten corium with structure

International Nuclear Information System (INIS)

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

2012-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-07-01

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

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

International Nuclear Information System (INIS)

Pazsit, I.; Dykin, V.

2014-01-01

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

Performance of "Waterless Concrete"

Science.gov (United States)

Toutanji, H. A.; Grugel, R. N.

2009-01-01

Waterless concrete consists of molten elementary sulfur and aggregate. The aggregates in a lunar environment will be lunar rocks and soil. Sulfur is present on the Moon in Troilite soil (FeS) and, by oxidation of the soil, iron and sulfur can be produced. Sulfur concrete specimens were cycled between liquid nitrogen (approx.]91 C) and room temperature (^21 C) to simulate exposure to a lunar environment. Cycled and control specimens were subsequently tested in compression at room temperatures (^21 C) and ^-101 C. Test results showed that due to temperature cycling, the compressive strength of cycled specimens was 20% of those non-cycled. This reduction in strength can be attributed to the large differences in thermal coefficients of expansion of the materials constituting the concrete which promoted cracking. Similar sulfur concrete mixtures were strengthened with short and long glass fibres. The lunar regolith simulant was melted in a 25 cc Pt- Rh crucible in a Sybron Thermoline high temperature MoSi2 furnace at melting temperatures of 1450 to 1600 C for times of 30 min to i hour. Glass fibres and small rods were pulled from the melt. The glass fibres were used to reinforce sulfur concrete plated to improve the flexural strength of the sulfur concrete. Beams strengthened with glass fibres showed to exhibit an increase in the flexural strength by as much as 45%.

Thermal conditions and functional requirements for molten fuel containment

International Nuclear Information System (INIS)

Kang, C.S.; Torri, A.

1980-05-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-15

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

Oxide-metal corium-concrete interaction test in the Vulcano facility

International Nuclear Information System (INIS)

Journeau, Ch.; Piluso, P.; Haquet, J.F.; Saretta, S.; Boccaccio, E.; Bonnet, J.M.

2007-01-01

Corium is likely to melt through the vessel and interact with the reactor pit concrete. Corium is made of a UO 2 -rich oxidic part, in which most of the decay heat is dissipated, and of a metallic part, mainly molten steel. An experiment has been set up in the Vulcano facility in which oxidic and metallic mixtures are molten in separate furnaces and poured in a concrete cavity. Induction heating is provided to the pool upper part thanks to shielding coils, so that, in case of stratification, the lighter oxidic corium-concrete mixture receives most of the power. Pre-calculations with the TOLBIAC-ICB corium-concrete interaction code based on the phase segregation model have provided valuable information for the dimensioning of this test: a thick metallic layer (>10 kg or 4 cm) has been chosen in order to obtain significant cavity ablation profiles depending on the selected heat transfer and stratification models. Stratification of the two liquid phases is predicted to occur in less than 10 minutes. In September 2006, the experiment was performed in the Vulcano facility. The corium was made of about 15 kg of steel at 1700 C and 30 kg of oxides (70% UO 2 , 16 % ZrO 2 and 14% concrete load) above 2000 C. It was poured in a limestone-rich concrete. This concrete type was selected for the first test, since the ablation is isotropic except for the initial transient, during oxidic corium-concrete interaction tests. 32 kW of induction power have been provided to the pool during the 4-hour test. The destruction of in-concrete thermocouples indicates that ablation was first mainly radial then became isotropic. This is quite similar to the ablation progression observed during previous tests with oxidic corium interacting with this type of concrete. Important 'volcanic activity' has been observed at the corium pool surface, compared to the previous oxidic corium experiments at Vulcano. (authors)

Oxide-metal corium-concrete interaction test in the Vulcano facility

Energy Technology Data Exchange (ETDEWEB)

Journeau, Ch.; Piluso, P.; Haquet, J.F.; Saretta, S.; Boccaccio, E.; Bonnet, J.M. [CEA Cadarache, Severe Accident Mastery experimental Lab. (DEN/DTN/STRI/LMA), 13 - Saint Paul lez Durance (France)

2007-07-01

Corium is likely to melt through the vessel and interact with the reactor pit concrete. Corium is made of a UO{sub 2}-rich oxidic part, in which most of the decay heat is dissipated, and of a metallic part, mainly molten steel. An experiment has been set up in the Vulcano facility in which oxidic and metallic mixtures are molten in separate furnaces and poured in a concrete cavity. Induction heating is provided to the pool upper part thanks to shielding coils, so that, in case of stratification, the lighter oxidic corium-concrete mixture receives most of the power. Pre-calculations with the TOLBIAC-ICB corium-concrete interaction code based on the phase segregation model have provided valuable information for the dimensioning of this test: a thick metallic layer (>10 kg or 4 cm) has been chosen in order to obtain significant cavity ablation profiles depending on the selected heat transfer and stratification models. Stratification of the two liquid phases is predicted to occur in less than 10 minutes. In September 2006, the experiment was performed in the Vulcano facility. The corium was made of about 15 kg of steel at 1700 C and 30 kg of oxides (70% UO{sub 2}, 16 % ZrO{sub 2} and 14% concrete load) above 2000 C. It was poured in a limestone-rich concrete. This concrete type was selected for the first test, since the ablation is isotropic except for the initial transient, during oxidic corium-concrete interaction tests. 32 kW of induction power have been provided to the pool during the 4-hour test. The destruction of in-concrete thermocouples indicates that ablation was first mainly radial then became isotropic. This is quite similar to the ablation progression observed during previous tests with oxidic corium interacting with this type of concrete. Important 'volcanic activity' has been observed at the corium pool surface, compared to the previous oxidic corium experiments at Vulcano. (authors)

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.

The loadings and strength of nuclear power plant structures in core damage accidents

International Nuclear Information System (INIS)

Varpasuo, P.

1994-01-01

The reactor cavity of VVER-91 NPP is a thick-walled, cylindrical reinforced concrete structure. In case of molten core-water reaction during the severe reactor accident the load carrying capacity of the cavity structure is of interest against the short impulse type loading caused by the steam explosion phenomenon. The assumed size of the impulse was 20 kPa-s and the duration was 10 ms. This investigation was divided in several phases. First, the elastic response of the cavity was determined using the ABAQUS code. Next, the static response of the cavity was evaluated using elasto-plastic properties of reinforcement and concrete and also taking into account the cracking of the concrete. This analysis was done with the aid of ABAQUS/STANDARD and ANSYS codes and the obtained results agreed reasonably with each other. In order to obtain a qualitative picture of the behaviour of the structure under the impulse load a simplified single degree of freedorn model was developed. The hoop reinforcement of the cavity was taken as an elasto-plastic spring and the wall concrete acted as a mass. Using this model the suitable amount of hoop reinforcement was determined. In next phase, the dynamic analysis of the structure was attempted using elasto-plastic material properties and concrete cracking. (13 refs., 57 figs.)

Molten salt reactors and possible scenarios for future nuclear power deployment

International Nuclear Information System (INIS)

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

2004-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2002-03-01

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

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

International Nuclear Information System (INIS)

Burson, S.B.

1987-01-01

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

Analysis of hydrogen generation according to the specific concrete composition during severe accident

International Nuclear Information System (INIS)

Seo, M. R.; Kim, M. K.

2001-01-01

The chemical composition of reactor cavity floor concrete affects the kind and amout of gases generated by MCCI and ablation of concrete. And if affects the physical and chemical characteristics of molten pool formed in the cavity. So, the specific concrete compostion is inputted in the MAAP Code used in the Level 2 PSA. and since Ulchin Unit 3 and 4 PSA, the analysis of concrete composition has been performed by the concrete mold prepared for this usage at the installation of cavity floor concrete. But, the composition of domestic concrete for construction of NPP is nearly the same as that of the standard basaltic concrete, and the effect of minor variation in composition is expected to be negligible. This report analyze the effect of the concrete composition to the generation of hydrogen due to MCCI, and discuss the necessity of analysis about the specific concrete composition for Level 2 PSA

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

International Nuclear Information System (INIS)

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

2004-11-01

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Large scale sodium interactions. Part 3. Chemical phenomena with limestone concrete

International Nuclear Information System (INIS)

Sallach, R.A.

1977-01-01

The description of the chemical processes and reaction products resulting from the exposure of concrete to molten sodium metal is important for a thorough, realistic assessment of the safety of CRBR-type reactors. Concretes are in general complex heterogenous substances whose ingredients can be derived from many sources. Consequently a wide variety of reaction processes and products might be anticipated. Initial attention has focused on a concrete in which both the aggregate and sandy components are derived from limestone. Presented are the chemical observations and experimental data from tests in which molten sodium metal at approximately 500 0 C is dropped into cold limestone concrete crucibles. Thermocouples immersed in the sodium pool indicate that the reaction proceeds in two stages. In the first stage which lasts 5 to 8 minutes, the temperature of the reacting mass hovers around 500 0 C. This stage is followed by a second stage of longer duration--greater than 100 minutes--where the temperature is 700 to 800 0 C. The main reaction product is a hard, fused, black slag which contains about 3/4 of the sodium in the initial charge. A secondary product is sodium oxide aerosol which accounts for the remaining 1/4 of the charge. It is significant that no free sodium metal is found in the slag; all sodium has completely reacted

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

PCM Concrete. [Phase Change Materials

Energy Technology Data Exchange (ETDEWEB)

Juul Andersen, T. [Danish Technological Institute, Taastrup (Denmark); Poulsen, H.-H. [BASF A/S, Roedekro (Denmark); Passov, F. [Spaencom A/S, Hedehusene (Denmark); Heiselberg, P. [Aalborg Univ..Aalborg (Denmark)

2013-04-01

PCM-Concrete was a research and development project launched in 2009 and finished in 2012. The project, which was funded by The Danish National Advanced Technology Foundation, had a total budget of 1.7 million Euros and included 4 partners: Danish Technological Institute (project manager), Aalborg University, BASF A/S and Spaencom A/S. The overall vision of the project was to reduce energy consumption for heating and cooling in buildings by developing high-performance concrete structures microencapsulated Phase Change Materials (PCM). The PCM used in the project was Micronal produced by BASF A/S. Micronal is small capsules with an acrylic shell and inside a wax with a melting point at approx. 23 deg. C equal to a comfortable indoor temperature. During the melting process thermal energy is transferred to chemical reaction (melting/solidification) depending on PCM being heated up or cooled down. Adding Micronal to concrete would theoretically increase the thermal mass of the concrete and improve the diurnal heat capacity which is the amount of energy that can be stored and released during 24 hours. Nevertheless, it is a relatively new technology that has not received much attention, yet. In the PCM-Concrete project 5 main investigations were carried out: 1) Development of concrete mix design with PCM. 2) Investigation of thermal properties of the PCM concrete: thermal conductivity, specific heat capacity, density. 3) Up-scaling the research to industrial production of PCM-concrete structures. 4) Testing energy efficiency in full scale. 5) Confronting aesthetic and acoustic barriers to full exploitation of the potential of PCM-concrete structures. The results from the test program showed: 1) That the diurnal heat storage capacity is higher for all 4 hollow core decks with tiles attached compared to the reference hollow core deck. 2) The hollow core decks with concrete tiles without PCM performs slightly better than the tiles with PCM. 3) That is was impossible to

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

Concrete portable handbook

CERN Document Server

Woodson, R Dodge

2011-01-01

Whether or not, you are on the job site or back in the office, this book will help you to avoid mistakes, code violations, and wasted time and money. The book's four part treatment begins with constituent materials followed by self contained parts on Concrete Properties, Processes, and Concrete Repair and Rehabilitation. Designed to be an ""all in one"" reference, the author includes a wealth information for the most popular types of testing. This includes: Analysis of Fresh Concrete; Testing Machines; Accelerated Testing Methods; Analysis of Hardened Concrete and Mortar; Core Sampl

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

Floors number influence on the instability parameter of reinforced concrete wall- or core-braced buildings

Directory of Open Access Journals (Sweden)

R. J. Ellwanger

Full Text Available This work aims to investigate the floors number influence on the instability parameter limit α1 of buildings braced by reinforced concrete walls and/or cores. Initially, it is showed how the Beck and König discrete and continuous models are utilized in order to define when a second order analysis is needed. The treatment given to this subject by the Brazilian code for concrete structures design (NBR 6118 is also presented. It follows a detailed analytical study that led to the derivation of equations for the limit α1 as functions of the floors number; a series of examples is presented to check their accuracy. Results are analyzed, showing the precision degree achieved and topics for continuity of research in this field are indicated.

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)

Crust behavior and erosion rate prediction of EPR sacrificial material impinged by core melt jet

Energy Technology Data Exchange (ETDEWEB)

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

2017-04-01

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

Heat removal capability of core-catcher with inclined cooling channels

International Nuclear Information System (INIS)

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

2009-01-01

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

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

International Nuclear Information System (INIS)

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

1993-11-01

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

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

International Nuclear Information System (INIS)

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

1982-01-01

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

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

International Nuclear Information System (INIS)

Dalle Donne, M.; Werle, H.

1980-08-01

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

Molten salt electrorefining method

International Nuclear Information System (INIS)

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

1994-01-01

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

Anisotropic Concrete Compressive Strength

DEFF Research Database (Denmark)

Gustenhoff Hansen, Søren; Jørgensen, Henrik Brøner; Hoang, Linh Cao

2017-01-01

When the load carrying capacity of existing concrete structures is (re-)assessed it is often based on compressive strength of cores drilled out from the structure. Existing studies show that the core compressive strength is anisotropic; i.e. it depends on whether the cores are drilled parallel...

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)

Molten salt reactor concept

International Nuclear Information System (INIS)

Sood, D.D.

1980-01-01

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

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

International Nuclear Information System (INIS)

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

1994-11-01

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

The behaviour of concrete under attack of liquid steel

International Nuclear Information System (INIS)

Schneider, U.; Ehm, C.; Diederichs, U.

1983-01-01

Investigations were carried out to study the interaction between concrete and liquid steel. Different types and different forms of concrete were investigated at temperatures of liquid steel between 1.600 and 2.600 0 C. The liquid steel of 1.600 0 C was produced in an induction furnace, the liquid steel of 2.600 0 C was produced in concrete crucibles by metallothermic reactions. The reactions occuring during the interaction of concrete and liquid steel may be summarized as follows: - Concrete reacts violently upon sudden loading with high temperatures and high heat fluxes. Great quantities of steam and gases are generated. The mechanical strength decreases rapidly with increasing temperature. -At about 1.200 0 C concrete begins to melt. First the cement matrix melts, than the aggregates melt. The melts of different concretes consist of different constituents and their reactions with liquid steel vary. The temperature of the liquid steel significantly influences the intensity of the reactions and the erosion rates. - The erosion rates amounted to 30 mm/min, when liquid steel was produced in concrete crucibles. When cylindrical concrete specimens were immersed in molten steel the rate of melting off amounted up to 66 mm/min. - The dissipation of heat during the interaction brings about that the reactions between concrete and liquid steel vanish gradually, if no additional energy is fed into the system. (orig.)

Effect of Aggregate Mineralogy and Concrete Microstructure on Thermal Expansion and Strength Properties of Concrete

Directory of Open Access Journals (Sweden)

Jinwoo An

2017-12-01

Full Text Available Aggregate type and mineralogy are critical factors that influence the engineering properties of concrete. Temperature variations result in internal volume changes could potentially cause a network of micro-cracks leading to a reduction in the concrete’s compressive strength. The study specifically studied the effect of the type and mineralogy of fine and coarse aggregates in the normal strength concrete properties. As performance measures, the coefficient of thermal expansion (CTE and compressive strength were tested with concrete specimens containing different types of fine aggregates (manufactured and natural sands and coarse aggregates (dolomite and granite. Petrographic examinations were then performed to determine the mineralogical characteristics of the aggregate and to examine the aggregate and concrete microstructure. The test results indicate the concrete CTE increases with the silicon (Si volume content in the aggregate. For the concrete specimens with higher CTE, the micro-crack density in the interfacial transition zone (ITZ tended to be higher. The width of ITZ in one of the concrete specimens with a high CTE displayed the widest core ITZ (approx. 11 µm while the concrete specimens with a low CTE showed the narrowest core ITZ (approx. 3.5 µm. This was attributed to early-age thermal cracking. Specimens with higher CTE are more susceptible to thermal stress.

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.)

TECHNOLOGY FOR INSTALLATION OF REINFORCED CONCRETE FLOOR SLABS LIGHTENED BY CORE DRIVERS WITH PRELIMINARY REINFORCEMENT STRESS

OpenAIRE

S. N. Leonovich; I. I. Peredkov

2015-01-01

The paper presents technology for installation of floor slabs lightened by plastic core drivers which are preliminary stressed under construction conditions.  Efficiency of such constructive solution is justified by the action of preliminary concrete compression in the tensile zone while reducing structure dead weight due to void arrangement.  The paper provides classification of systems for preliminary stress and contains recommendations on selection of the system depending on peculiariar fe...

Study of water permeability in concrete by neutron and gamma-ray techniques

International Nuclear Information System (INIS)

Abd El-Monem, A.M.M.

2010-01-01

water infiltration in various building materials , namely concrete used for buildings basement and underwater construction is the main concern of the studies performed in this thesis. The studies aim to develop a nuclear techniques for investigation a concrete mixes with different additives capable to decrease concrete porosity and intern resist water propagation inside concrete materials without any deterioration of concrete physical and mechanical properties . These issues were achieved through the preparation of ordinary concrete mixes with different percentages of silica fume. Concrete samples of different shape and geometries were made to study water diffusion when the concrete samples are submerged in water for different periods of time. The concrete samples were first sealed by molten asphalt from all sides expect two opposite faces to ensure water migration only along one direction. Water infiltration in concrete samples with different percentages of silica fume and submerged in tap and seawater for different periods of time was studied by neutrons and gamma techniques. Also, water propagation in mortar samples with different percentages of silica fume was studied by electrical methods based on measuring the variation in electrical conductivity of these samples.

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)

Report of concrete pavement evaluation : project 105 C-4181-01 Donahoo Road, Wyandotte County.

Science.gov (United States)

2013-12-01

The physical properties of hardened concrete cores and fresh concrete test results were compared with aggregate : gradation workability differences. The concrete cores were taken from a rural two-lane concrete road in northeastern Kansas : constructe...

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

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''

Development and Performance Evaluation of High Temperature Concrete for Thermal Energy Storage for Solar Power Generation

Energy Technology Data Exchange (ETDEWEB)

Selvam, R. Panneer; Hale, Micah; Strasser, Matt

2013-03-31

Thermal energy can be stored by the mechanism of sensible or latent heat or heat from chemical reactions. Sensible heat is the means of storing energy by increasing the temperature of the solid or liquid. Since the concrete as media cost per kWh_thermal is $1, this seems to be a very economical material to be used as a TES. This research is focused on extending the concrete TES system for higher temperatures (500 °C to 600 °C) and increasing the heat transfer performance using novel construction techniques. To store heat at high temperature special concretes are developed and tested for its performance. The storage capacity costs of the developed concrete is in the range of $0.91-$3.02/kWh_thermal. Two different storage methods are investigated. In the first one heat is transported using molten slat through a stainless steel tube and heat is transported into concrete block through diffusion. The cost of the system is higher than the targeted DOE goal of $15/kWht_hermal. The increase in cost of the system is due to stainless steel tube to transfer the heat from molten salt to the concrete blocks.The other method is a one-tank thermocline system in which both the hot and cold fluid occupy the same tank resulting in reduced storage tank volume. In this model, heated molten salt enters the top of the tank which contains a packed bed of quartzite rock and silica sand as the thermal energy storage (TES) medium. The single-tank storage system uses about half the salt that is required by the two-tank system for a required storage capacity. This amounts to a significant reduction in the cost of the storage system. The single tank alternative has also been proven to be cheaper than the option which uses large concrete modules with embedded heat exchangers. Using computer models optimum dimensions are determined to have an round trip efficiency of 84%. Additionally, the cost of the structured concrete thermocline configuration provides the TES

Influence of Steel Reinforcement on In-Situ Stress Evaluation in Concrete Structures by the Core-Drilling Method

International Nuclear Information System (INIS)

McGinnis, M. J.; Pessiki, S.

2006-01-01

The core-drilling method is an emerging technique for evaluating in-situ stress in a concrete structure. A small hole is drilled into the structure, and the deformations in the vicinity of the hole are measured and related via elasticity theory to the stress. The method is similar to the ASTM hole-drilling strain-gauge method excepting that displacements rather than strains are the measured quantities. The technique may be considered nondestructive since the ability of the structure to perform its function is unaffected, and the hole is easily repaired. Displacement measurements in the current work are performed using 3D digital image correlation and industrial photogrammetry. The current paper addresses perturbations in the method caused by steel reinforcement within the concrete. The reinforcement is significantly stiffer than the surrounding concrete, altering the expected displacement field. A numerical investigation performed indicates an under-prediction of stress by as much as 18 percent in a heavily reinforced structure, although the effect is significantly smaller for more common amounts of reinforcement

Influence of Steel Reinforcement on In-Situ Stress Evaluation in Concrete Structures by the Core-Drilling Method

Science.gov (United States)

McGinnis, M. J.; Pessiki, S.

2006-03-01

The core-drilling method is an emerging technique for evaluating in-situ stress in a concrete structure. A small hole is drilled into the structure, and the deformations in the vicinity of the hole are measured and related via elasticity theory to the stress. The method is similar to the ASTM hole-drilling strain-gauge method excepting that displacements rather than strains are the measured quantities. The technique may be considered nondestructive since the ability of the structure to perform its function is unaffected, and the hole is easily repaired. Displacement measurements in the current work are performed using 3D digital image correlation and industrial photogrammetry. The current paper addresses perturbations in the method caused by steel reinforcement within the concrete. The reinforcement is significantly stiffer than the surrounding concrete, altering the expected displacement field. A numerical investigation performed indicates an under-prediction of stress by as much as 18 percent in a heavily reinforced structure, although the effect is significantly smaller for more common amounts of reinforcement.

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.)

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Experimental investigation of interface conditions between oxidic melt and ablating concrete during MCCI by means of simulating material experiments: the Artemis program

International Nuclear Information System (INIS)

Veteau, J.M.

2005-01-01

Full text of publication follows: In the frame work of R and D on Severe Accidents in PWR plants, an estimation by codes of time of basemat melt-through by Corium is required. For this, the heat flux distribution along the cavity wall must be properly modelled. Hence the knowledge of the heat transfer coefficient as well as the temperature at the interface between the melt and the solid become key issues. Phase diagram of the melt and composition governs the interface temperature which controls, at least partly, the thickness of the Corium crust formed on the molten concrete. Crust behaviour (time evolution of thickness, mechanical interaction with gas) implies a release mode of molten concrete in Corium which in turn alters the melt composition. Clearly, the molten corium-concrete interaction (MCCI) phenomenon is the result of a strong coupling between physico-chemistry and thermohydraulics. The main goal of the first test series of the Artemis program is to make a link between the interface temperature and the physico-chemistry of the melt (phase diagram) through tests conducted with simulating materials and to provide an insight on the existence, the behaviour and the composition of the crust. This test series considers 1D MCCI using a non eutectic LiCl-BaCl 2 mixture poured at 1000 deg. C in a cylindrical test section (internal diameter 0.3 m) to interact with the 0.35 m deep basemat made of the same salt mixture at the eutectic composition. This 'concrete' was especially manufactured with sintered granulates to allow gas flow from the bottom (argon), then simulating gas released by concrete in the reactor case. Constant power is applied in the pool with an helical coil and 1D MCCI is ensured by counterbalancing heat losses by controlled heating at the lateral walls and at the top of the test section. Concrete ablation is followed from the output of 45 0.5 mm diameter thermocouples. An instrumented rod periodically investigates the temperature and the position

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

Release of gases and their influence on containment integrity during a hypothetical meltdown accident

International Nuclear Information System (INIS)

Hassmann, K.; Reimann, M.

1981-01-01

The sequence of a hypothetical core melt down accident has been subdivided into four phases. Heating up of the core until failure of the core support structure is the first phase. It starts at a certain water level in the reactor pressure vessel (RPV) and ends with the failure of the grid plate. The second phase is characterized by the evaporation of the water in the lower plenum of the RPV. The second phase lasts until a molten core debris is formed. The third phase is concerned with the heating up of the pressure vessel after formation of a molten pool in the lower plenum of the RPV. After pressure vessel failure, the molten corium will interact in the fourth phase with the concrete structure beneath the pressure vessel. In this paper the gas release during all four accident phases and the resulting pressure-time history within the containment of a German standard PWR is given, taking into account violent combustion of hydrogen. In particular, the differences caused by dsestruction of concrete with silicious and with calcareous aggregates has been analyzed. The basis for the results in the 4th phase is the WECHSL code. Long term containment calculations have been performed with the COCMEL-code

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.

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.

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Preliminary safety analysis of molten salt breeder reactor

International Nuclear Information System (INIS)

Cheng Maosong; Dai Zhimin

2013-01-01

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

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

International Nuclear Information System (INIS)

Yamamoto, Takahisa; Mitachi, Koshi

2004-01-01

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

Report of concrete pavement evaluation : project 105 C-4181-01 Donahoo Road, Wyandotte County : [technical summary].

Science.gov (United States)

2013-12-01

The physical properties of hardened concrete cores and fresh concrete test results were compared with aggregate gradation workability differences. The concrete cores were taken from a rural two-lane concrete road in northeastern Kansas constructed in...

Numerical analysis on seismic behavior of reinforced concrete beam to concrete filled steel tubular column connections with ring-beam

Energy Technology Data Exchange (ETDEWEB)

Zhao, Yi., E-mail: zhaoyi091218@163.com [School of Civil and Architectural Engineering, Zhongyuan University of Technology,Zhengzhou 450000 (China); Xu, Li. Hua. [School of Civil Engineering, Wuhan University, No.8, Donghu Road, WuHan 430072 (China)

2016-06-08

This paper presents numerical study of the seismic behavior of reinforced concrete beam to concrete filled steel tube column connections with ring-beam. The material stress-strain relations, element type and boundary condition are selected, which are consistent with actual situation. Then the seismic behavior of this type of joint are researched by ABAQUS, and finite element analyses are conducted under cyclic loading. Its parameters are discussed including thickness of steel tubular column wall, sectional dimension of the ring-beam and strength of the core concrete. The results show that the ultimate capacity of the connections is improved with sectional dimension of the ring-beam increased. In the meanwhile, the influence on skeleton curve of the joints is slight of which included thickness of steel tubular column wall and strength of the core concrete.

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)

Detection and removal of molten salts from molten aluminum alloys

Energy Technology Data Exchange (ETDEWEB)

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

1999-08-02

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Core-concrete interactions with overlying water pools

International Nuclear Information System (INIS)

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

1993-11-01

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

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.)

Reliability of core test – Critical assessment and proposed new approach

OpenAIRE

Shafik Khoury; Ali Abdel-Hakam Aliabdo; Ahmed Ghazy

2014-01-01

Core test is commonly required in the area of concrete industry to evaluate the concrete strength and sometimes it becomes the unique tool for safety assessment of existing concrete structures. Core test is therefore introduced in most codes. An extensive literature survey on different international codes’ provisions; including the Egyptian, British, European and ACI Codes, for core analysis is presented. All studied codes’ provisions seem to be unreliable for predicting the in-situ concrete ...

Results of and prospects for studies on molten salt nuclear reactors

International Nuclear Information System (INIS)

Hery, M.; Lecocq, A.

1983-04-01

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

Molten fluoride fuel salt chemistry

International Nuclear Information System (INIS)

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

1995-01-01

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

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

Effectiveness of Horizontal Rebar on Concrete Block Retaining Wall Strength

OpenAIRE

Krishpersad Manohar; Rikhi Ramkissoon

2016-01-01

The effectiveness of including a horizontal rebar compared to only a vertical rebar in concrete filled core interlocking concrete block retaining wall sections was investigated with respect to the horizontal retaining force. Experimental results for three specimens of interlocking blocks with vertical rebar and concrete filled cores showed an average horizontal retaining force of 24546 N ± 5.7% at an average wall deflection of 13.3 mm. Experimental results for three wall specimens of interloc...

Evaluation of Sulfur 'Concrete' for Use as a Construction Material on the Lunar Surface

Science.gov (United States)

Grugel, R. N.

2008-01-01

Combining molten sulfur with any number of aggregate materials forms, when solid, a mixture having attributes similar, if not better, to conventional water-based concrete. As a result the use of sulfur "concrete" on Earth is well established, particularly in corrosive environments. Consequently, discovery of troilite (FeS) on the lunar surface prompted numerous scenarios about its reduction to elemental sulfur for use, in combination with lunar regolith, as a potential construction material; not requiring water, a precious resource, for its manufacture is an obvious advantage. However, little is known about the viability of sulfur concrete in an environment typified by extreme temperatures and essentially no atmosphere. The experimental work presented here evaluates the response of pure sulfur and sulfur concrete subjected to laboratory conditions that approach those expected on the lunar surface, the results suggesting a narrow window of application.

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Improvement to molten salt reactors

International Nuclear Information System (INIS)

Bienvenu, Claude.

1975-01-01

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

The advanced containment experiments (ACE) Project

International Nuclear Information System (INIS)

Sehgal, B.R.; Ritzman, R.; Merilo, M.; Rahn, F.; Machiels, A.

1992-01-01

The overall structure and content of the ACE Project, which has been obtaining experimental data in four key areas of LWR severe accident technology are described. The key areas consist of filtration systems for vented containment concepts, radioiodine behavior in containment, the interaction of molten core material with structural concrete, and the use of water to terminate the core-concrete interaction process. Experiment procedures used in each phase of the work are summarized and the principal results and conclusions developed to date are discussed

Simulation of VVER MCCI reactor test case with ASTEC V2/MEDICIS computer code

International Nuclear Information System (INIS)

Stefanova, A.; Grudev, P.; Gencheva, R.

2011-01-01

This paper presents an application of the ASTEC v2, module MEDICIS for simulation of VVER Molten core concrete interaction test (MCCI) case without water injection. The main purpose of performed calculation is verification and improvement of module MEDICIS/ASTECv2 for better simulation of core concrete interaction processes. The VVER-1000 reference nuclear power plant was chosen as SARNET2 benchmark MCCI test-case. The initial conditions for MCCI test are taken after SBO scenario calculated with ASTEC version 1.3R2 by INRNE. (authors)

Experimental investigation of interface conditions between oxidic melt and ablating concrete during MCCI by means of simulating material experiments: the Artemis program

Energy Technology Data Exchange (ETDEWEB)

Veteau, J.M. [Commissariat a l' Energie Atomique, DEN/DTN/SE2T/LPTM, 17 rue des Martyrs 38 - Grenoble cedex 9 (France)

2005-07-01

Full text of publication follows: In the frame work of R and D on Severe Accidents in PWR plants, an estimation by codes of time of basemat melt-through by Corium is required. For this, the heat flux distribution along the cavity wall must be properly modelled. Hence the knowledge of the heat transfer coefficient as well as the temperature at the interface between the melt and the solid become key issues. Phase diagram of the melt and composition governs the interface temperature which controls, at least partly, the thickness of the Corium crust formed on the molten concrete. Crust behaviour (time evolution of thickness, mechanical interaction with gas) implies a release mode of molten concrete in Corium which in turn alters the melt composition. Clearly, the molten corium-concrete interaction (MCCI) phenomenon is the result of a strong coupling between physico-chemistry and thermohydraulics. The main goal of the first test series of the Artemis program is to make a link between the interface temperature and the physico-chemistry of the melt (phase diagram) through tests conducted with simulating materials and to provide an insight on the existence, the behaviour and the composition of the crust. This test series considers 1D MCCI using a non eutectic LiCl-BaCl{sub 2} mixture poured at 1000 deg. C in a cylindrical test section (internal diameter 0.3 m) to interact with the 0.35 m deep basemat made of the same salt mixture at the eutectic composition. This 'concrete' was especially manufactured with sintered granulates to allow gas flow from the bottom (argon), then simulating gas released by concrete in the reactor case. Constant power is applied in the pool with an helical coil and 1D MCCI is ensured by counterbalancing heat losses by controlled heating at the lateral walls and at the top of the test section. Concrete ablation is followed from the output of 45 0.5 mm diameter thermocouples. An instrumented rod periodically investigates the temperature

An investigation of tendon sheathing filler migration into concrete

International Nuclear Information System (INIS)

Naus, D.J.; Oland, C.B.

1998-03-01

During some of the inspections at nuclear power plants with prestressed concrete containments, it was observed that the containments has experienced leakage of the tendon sheathing filler (i.e., streaks). The objective of this activity was to provide an indication of the extent of tendon sheathing filler leakage into the concrete and its affects on concrete properties. Literature was reviewed and concrete core samples were obtained from the Trojan Nuclear Plant and tested. The literature primarily addressed effects of crude or lubricating oils that are known to cause concrete damage. However, these materials have significantly different characteristics relative to the materials used as tendon sheathing fillers. Examination and testing of the concrete cores indicated that the appearance of tendon sheathing filler on the concrete surface was due to leakage from the conduits and its subsequent migration through cracks that were present. Migration of the tendon sheathing filler was confined to the cracks and there was no perceptible movement into the concrete. Results of compressive strength testing indicated that the concrete quality was consistent in the containment and that the strength had increased over 40% in 25.4 years relative to the average compressive strength at 28-days age

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)

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

International Nuclear Information System (INIS)

Ogasawara, H.

1998-01-01

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

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)

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.

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.

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

Phase 1 sampling and analysis plan for the 304 Concretion Facility closure activities

International Nuclear Information System (INIS)

Adler, J.G.

1994-01-01

This document provides guidance for the initial (Phase 1) sampling and analysis activities associated with the proposed Resource Conservation and Recovery Act of 1976 (RCRA) clean closure of the 304 Concretion Facility. Over its service life, the 304 Concretion Facility housed the pilot plants associated with cladding uranium cores, was used to store engineering equipment and product chemicals, was used to treat low-level radioactive mixed waste, recyclable scrap uranium generated during nuclear fuel fabrication, and uranium-titanium alloy chips, and was used for the repackaging of spent halogenated solvents from the nuclear fuels manufacturing process. The strategy for clean closure of the 304 Concretion Facility is to decontaminate, sample (Phase 1 sampling), and evaluate results. If the evaluation indicates that a limited area requires additional decontamination for clean closure, the limited area will be decontaminated, resampled (Phase 2 sampling), and the result evaluated. If the evaluation indicates that the constituents of concern are below action levels, the facility will be clean closed. Or, if the evaluation indicates that the constituents of concern are present above action levels, the condition of the facility will be evaluated and appropriate action taken. There are a total of 37 sampling locations comprising 12 concrete core, 1 concrete chip, 9 soil, 11 wipe, and 4 asphalt core sampling locations. Analysis for inorganics and volatile organics will be performed on the concrete core and soil samples. Separate concrete core samples will be required for the inorganic and volatile organic analysis (VOA). Analysis for inorganics only will be performed on the concrete chip, wipe, and asphalt samples

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-09-01

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

Fundamentals of molten-salt thermal technology

International Nuclear Information System (INIS)

1980-08-01

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

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

International Nuclear Information System (INIS)

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

1976-01-01

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

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)

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)

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

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

International Nuclear Information System (INIS)

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

1983-01-01

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

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

International Nuclear Information System (INIS)

Storr, G.J.

1989-08-01

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

New ADAC centre cools and heats environmental-friendly. Concrete core tempering; Neue ADAC-Zentrale kuehlt und heizt umweltvertraeglich. Betonkerntemperierung

Energy Technology Data Exchange (ETDEWEB)

Anon,

2010-12-15

At present, the new centre of the General German Automobile Association (ADAC) arises in Munich's suburb Sendling/Westpark. In summer 2011, nearly 2,400 coworkers of ADAC so far distributed over six locations in Munich shall work together in the new office building. So that they feel well in this new building, Rehau AG (Rehau, Federal Republic of Germany) supplies an environmental-friendly heating and cooling of the building by means of concrete core tempering.

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.

UO2/magnetite concrete interaction and penetration study

International Nuclear Information System (INIS)

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

1983-01-01

The concrete structure represents a line of defense in safety assessment of containment integrity and possible minimization of radiological releases following a reactor accident. The penetration study of hot UO 2 particles into limestone concrete and basalt concrete highlighted some major differences between the two concretes. These included penetration rate, melting and dissolution phenomena, released gases, pressurization of the UO 2 chamber, and characteristics of post-test concrete. The present study focuses on the phenomena associated with core debris interaction with and penetration into magnetite type concrete. The real material experiment was carried out with UO 2 particles and magnetite concrete in a test apparatus similar to the one utilized in the UO 2 /limestone experiment

Temperature field in concrete when in contact with hot liquids

International Nuclear Information System (INIS)

Andrade Lima, F.R. de.

1981-09-01

In an HCDA (Hypothetical Core Disruptive Accident) it is postulated that liquid metal coolants and core materials come in contact with the retaining concrete structure. A mathematical model and an associated computer program was previously developed to describle the transient heat and mass transfer in the concrete. Implementations on the original program-USINT- are included to consider the variations of the thermal conductivity as a function of the temperature. Also a subroutine - PLOTTI - is incorporated to the program for the plotting of the results. The new program - USINTG - is used to calculate the temperature and pressure fields and the water released from concrete structures during a sodium leak simulation and with the concrete structures in contact with liquid sodium. No consideration about chemical reactions involving the sodium when in contact with concrete is considered. (Author) [pt

Characterization of basin concrete in support of structural integrity demonstration for extended storage

International Nuclear Information System (INIS)

Duncan, A.

2014-01-01

Concrete core samples from C basin were characterized through material testing and analysis to verify the design inputs for structural analysis of the L Basin and to evaluate the type and extent of changes in the material condition of the concrete under extended service for fuel storage. To avoid the impact on operations, core samples were not collected from L area, but rather, several concrete core samples were taken from the C Basin prior to its closure. C basin was selected due to its similar environmental exposure and service history compared to L Basin. The microstructure and chemical composition of the concrete exposed to the water was profiled from the water surface into the wall to evaluate the impact and extent of exposure. No significant leaching of concrete components was observed. Ingress of carbonation or deleterious species was determined to be insignificant. No evidence of alkali-silica reactions (ASR) was observed. Ettringite was observed to form throughout the structure (in air voids or pores); however, the sulfur content was measured to be consistent with the initial concrete that was used to construct the facility. Similar ettringite trends were observed in the interior segments of the core samples. The compressive strength of the concrete at the mid-wall of the basin was measured, and similar microstructural analysis was conducted on these materials post compression testing. The microstructure was determined to be similar to near-surface segments of the core samples. The average strength was 4148 psi, which is well-above the design strength of 2500 psi. The analyses showed that phase alterations and minor cracking in a microstructure did not affect the design specification for the concrete

Molten salt reactors

International Nuclear Information System (INIS)

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

2014-01-01

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

Summary of treat experiments on oxide core-disruptive accidents

International Nuclear Information System (INIS)

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

1979-02-01

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

Étude thermodynamique du corium en cuve - Application à l'interaction corium/béton

OpenAIRE

Quaini , Andrea

2015-01-01

During a severe accident in a pressurised water reactor, the nuclear fuel can interact with the Zircaloy cladding, the neutronic absorber and the surrounding metallic structure forming a partially or completely molten mixture. The molten core can then interact with the reactor steel vessel forming a mixture called in-vessel corium. In the worst case, this mixture can pierce the vessel and pour onto the concrete underneath the reactor, leading the formation of the ex-vessel corium. Furthermore...

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)

Method of detecting construction faults in concrete pressure vessels

International Nuclear Information System (INIS)

Robertson, S.A.; Duhoux, M.; Dawance, G.; Carrie, C.; Morel, D.

1976-01-01

A major problem in the design and construction of concrete pressure vessels for nuclear power stations is the risk of excessive air leaks through the concrete itself, due to faulty construction. The 'sonic coring' method of non-destructive concrete testing has been used successfully in pile and diaphragm wall construction control for several years, and the potential use of this method to control the presence of faults in concrete pressure vessels is here described. (author)

Concrete core tempering in a passive house. Simulation behaviour of a single-family house; Betonkerntemperierung im Passivhaus. Simulationsverhalten eines Einfamilienhauses

Energy Technology Data Exchange (ETDEWEB)

Mahler, B; Seiwald, H [Steinbeis-Transferzentrum Energie-, Gebaeude- und Solartechnik; Koch, K P [Rehau AG und Co., Erlangen-Eltersdorf (Germany)

2004-07-01

New office buildings often have concrete core tempering systems for combined heating and cooling. The technology may be applied to well-insulated domestic buildings as well, provided that certain boundary conditions are met. [German] Beim Neubau von Buerogebaeuden findet die Betonkerntemperierung fuer kombinierte Heiz- und Kuehlsysteme immer mehr Verbreitung. Im Zuge des stark reduzierten Heizleistungsbedarfs hochgedaemmter Wohngebaeude bietet sich die Betonkerntemperierung unter bestimmten Voraussetzungen auch hier als Heizsystem an. (orig.)

Molten fuel studies at Winfrith

International Nuclear Information System (INIS)

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

1988-01-01

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

Sampling device for radioactive molten salt

International Nuclear Information System (INIS)

Shindo, Masato

1998-01-01

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

Thermal performances of molten salt steam generator

International Nuclear Information System (INIS)

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

2016-01-01

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

Post-accident core coolability of light water reactors

International Nuclear Information System (INIS)

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

1983-01-01

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

Molten salt reactor type

International Nuclear Information System (INIS)

1977-01-01

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

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

Science.gov (United States)

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

2000-01-01

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

Study on some factors affecting the results in the use of MIP method in concrete research

International Nuclear Information System (INIS)

Kumar, Rakesh; Bhattacharjee, B.

2003-01-01

Effects of rate of pressure application and forms and type of sample on porosity and pore size distribution of concrete estimated through mercury intrusion porosimetry (MIP) are presented in this experimental work. Two different forms of concrete sample, namely, crushed chunks of concrete and small core drilled out from the concrete beam specimens, were used for this study. The results exhibit that the rate of pressure application in mercury porosimetry has little effect on porosity and pore size distribution of concrete. It is also demonstrated that small cores drilled out from large concrete specimens are preferable as samples for performing porosimetry test on concrete

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

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.

Analysis of production factors in high performance concrete

Directory of Open Access Journals (Sweden)

Gilberto Carbonari

2003-01-01

Full Text Available The incorporation of silica fume and superplasticizers in high strength and high performance concrete, along with a low water-cement ratio, leads to significant changes in the workability and the energy needed to homogenize and compact the concrete. Moreover, several aspects of concrete production that are not critical for conventional concrete are important for high strength concrete. This paper will discuss the need for controlling the humidity of the aggregates, optimizing the mixing sequence used in the fabrication, and the slump loss. The application of a silica fume concrete in typical building columns will be analyzed considering the required consolidation, the variability of the material strength within the structural element and the relation between core and molded specimen strength. Comparisons will also be made with conventional concrete.

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)

Effects of Basalt Fibres on Mechanical Properties of Concrete

Directory of Open Access Journals (Sweden)

El-Gelani A. M.

2018-01-01

Full Text Available This paper presents the results of an experimental program carried out to investigate the effects of Basalt Fibre Reinforced Polymers (BFRP on some fundamental mechanical properties of concrete. Basalt fibres are formed by heating crushed basalt rocks and funnelling the molten basalt through a spinneret to form basalt filaments. This type of fibres have not been widely used till recently. Two commercially available chopped basalt fibres products with different aspect ratios were investigated, which are dry basalt (GeoTech Fibre and basalt pre-soaked in an epoxy resin (GeoTech Matrix .The experimental work included compression tests on 96 cylinders made of multiple batches of concrete with varying amounts of basalt fibre additives of the two mentioned types, along with control batches containing no fibres. Furthermore, flexural tests on 24 prisms were carries out to measure the modulus of rupture, in addition to 30 prisms for average residual strength test. Results of the research indicated that use of basalt fibres has insignificant effects on compressive strength of plain concrete, where the increase in strength did not exceed about 5%. On the other hand, results suggest that the use of basalt fibres may increase the compressive strength of concrete containing fly as up top 40%. The rupture strength was increased also by 8% to 28% depending on mix and fibre types and contents. Finally, there was no clear correlation between the average residual strength and ratios of basalt fibres mixed with the different concrete batches.

Accelerator molten-salt breeding and thorium fuel cycle

International Nuclear Information System (INIS)

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

1990-01-01

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

Thermal hydraulic study of a corium molten pool

International Nuclear Information System (INIS)

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

1993-01-01

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

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

Science.gov (United States)

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

2017-11-01

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

Transformation and fragmentation behavior of molten metal drop in sodium pool

International Nuclear Information System (INIS)

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

2007-01-01

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

Influence of corium oxidation on fission product release from molten pool

Energy Technology Data Exchange (ETDEWEB)

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

2010-05-15

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

Core catcher cooling for a gas-cooled fast breeder

International Nuclear Information System (INIS)

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

1976-01-01

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

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.)

Corium crust strength measurements

Energy Technology Data Exchange (ETDEWEB)

Lomperski, S. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439-4840 (United States)], E-mail: lomperski@anl.gov; Farmer, M.T. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439-4840 (United States)], E-mail: farmer@anl.gov

2009-11-15

Corium strength is of interest in the context of a severe reactor accident in which molten core material melts through the reactor vessel and collects on the containment basemat. Some accident management strategies involve pouring water over the melt to solidify it and halt corium/concrete interactions. The effectiveness of this method could be influenced by the strength of the corium crust at the interface between the melt and coolant. A strong, coherent crust anchored to the containment walls could allow the yet-molten corium to fall away from the crust as it erodes the basemat, thereby thermally decoupling the melt from the coolant and sharply reducing the cooling rate. This paper presents a diverse collection of measurements of the mechanical strength of corium. The data is based on load tests of corium samples in three different contexts: (1) small blocks cut from the debris of the large-scale MACE experiments, (2) 30 cm-diameter, 75 kg ingots produced by SSWICS quench tests, and (3) high temperature crusts loaded during large-scale corium/concrete interaction (CCI) tests. In every case the corium consisted of varying proportions of UO{sub 2}, ZrO{sub 2}, and the constituents of concrete to represent a LWR melt at different stages of a molten core/concrete interaction. The collection of data was used to assess the strength and stability of an anchored, plant-scale crust. The results indicate that such a crust is likely to be too weak to support itself above the melt. It is therefore improbable that an anchored crust configuration could persist and the melt become thermally decoupled from the water layer to restrict cooling and prolong an attack of the reactor cavity concrete.

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...

Mechanical Properties and Durability of "Waterless Concrete"

Science.gov (United States)

Toutanji, Houssam; Grugel, Richard N.

2008-01-01

Waterless concrete consists of molten elementary sulfur and aggregate. The aggregates in lunar environment will be lunar rocks and soil. Sulfur is present on the Moon in Troilite soil (FeS) and by oxidation soil iron and sulfur can be produced. Iron can be used to reinforce the sulfur concrete. Sulfur concrete specimens were cycled between liquid nitrogen (approximately 191 C) and room temperature (approximately 21 C) to simulate exposure to a lunar environment. Cycled and control specimens were subsequently tested in compression at room temperatures (approximately 21 C) and approximately 101 C. Test results showed that due to temperature cycling, compressive strength of cycled specimens was 20% of those non-cycled. Microscopic examination of the fracture surfaces from the cycled samples showed clear de-bonding of the sulfur from the aggregate material whereas it was seen well bonded in those non-cycled. This reduction in strength can be attributed to the large differences in thermal coefficients of expansion of the materials constituting the concrete which promoted cracking. Similar sulfur concrete mixtures were strengthened with short and long glass fibers. The glass fibers from lunar regolith simulant was melted in a 25 cc Pt-Rh crucible in a Sybron Thermoline high temperature MoSi2 furnace at melting temperatures of 1450 to 1600 C for times of 30 min to 1 hour. Glass fibers were cast from the melt into graphite crucibles and were annealed for a couple of hours at 600 C. Glass fibers and small rods were pulled from the melt. The glass melt wets the ceramic rod and long continuous glass fibers were easily hand drawn. The glass fibers were immediately coated with a protective polymer to maintain the mechanical strength. The glass fibers were used to reinforce sulfur concrete plated to improve the flexural strength of the sulfur concrete. Prisms beams strengthened with glass fibers were tested in 4-point bending test. Beams strengthened with glass fiber showed to

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.).

221-U Facility concrete and reinforcing steel evaluations specification for the canyon disposition initiative (CDI)

International Nuclear Information System (INIS)

Baxter, J.T.

1998-01-01

This describes a test program to establish the in-situ material properties of the reinforced concrete in Building 221-U for comparison to the original design specifications. Field sampling and laboratory testing of concrete and reinforcing steel structural materials in Building 221-U for design verification will be undertaken. Forty seven samples are to be taken from radiologically clean exterior walls of the canyon. Laboratory testing program includes unconfined compressive strength of concrete cores, tensile strength of reinforcing steel, and petrographic examinations of concrete cores taken from walls below existing grade

The molten salt reactor adventure

International Nuclear Information System (INIS)

MacPherson, H.G.

1985-01-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Boussier, H.; Heuer, D.

2010-07-01

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

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

International Nuclear Information System (INIS)

Farahani, A.A.

1997-01-01

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

Structure and thermodynamics of molten salts

International Nuclear Information System (INIS)

Papatheodorou, G.N.

1983-01-01

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

Applications of ASTEC integral code on a generic CANDU 6

Energy Technology Data Exchange (ETDEWEB)

Radu, Gabriela, E-mail: gabriela.radu@nuclear.ro [Institute for Nuclear Research, Campului 1, 115400 Mioveni, Arges (Romania); Prisecaru, Ilie [Power Engineering Department, University “Politehnica” of Bucharest, 313 Splaiul Independentei, Bucharest (Romania)

2015-05-15

Highlights: • Short overview of the models included in the ASTEC MCCI module. • MEDICIS/CPA coupled calculations for a generic CANDU6 reactor. • Two cases taking into account different pool/concrete interface models. - Abstract: In case of a hypothetical severe accident in a nuclear power plant, the corium consisting of the molten reactor core and internal structures may flow onto the concrete floor of containment building. This would cause an interaction between the molten corium and the concrete (MCCI), in which the heat transfer from the hot melt to the concrete would cause the decomposition and the ablation of the concrete. The potential hazard of this interaction is the loss of integrity of the containment building and the release of fission products into the environment due to the possibility of a concrete foundation melt-through or containment over-pressurization by the gases produced from the decomposition of the concrete or by the inflammation of combustible gases. In the safety assessment of nuclear power plants, it is necessary to know the consequences of such a phenomenon. The paper presents an example of application of the ASTECv2 code to a generic CANDU6 reactor. This concerns the thermal-hydraulic behaviour of the containment during molten core–concrete interaction in the reactor vault. The calculations were carried out with the help of the MEDICIS MCCI module and the CPA containment module of ASTEC code coupled through a specific prediction–correction method, which consists in describing the heat exchanges with the vault walls and partially absorbent gases. Moreover, the heat conduction inside the vault walls is described. Two cases are presented in this paper taking into account two different heat transfer models at the pool/concrete interface and siliceous concrete. The corium pool configuration corresponds to a homogeneous configuration with a detailed description of the upper crust.

Definition of breeding gain for molten salt reactors - 147

International Nuclear Information System (INIS)

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

2010-01-01

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

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

International Nuclear Information System (INIS)

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

1985-01-01

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

Simulation of the ACE L2 and ACE L5 MCCI experiment under dry surface conditions with ASTEC MEDICIS using an effective heat transfer model

Energy Technology Data Exchange (ETDEWEB)

Agethen, Kathrin; Koch, Marco K. [Bochum Univ. (Germany). Reactor Simulation and Safety Group

2013-07-01

In a postulated severe accident the loss of cooling can lead to a melting of the core and to a failure of the vessel. The molten core material discharges to the containment cavity and interacts with the concrete basemat. The heat up of the concrete leads to the release of sparing gases (H{sub 2}, CO{sub 2}, SiO), which stir the pool und causes chemical reactions. Especially the metals (Zr, Fe, Ni, Cr) in the corium are oxidized und the exothermic energy is released to the melt, which raises the melt temperature further. The release of combustible gases (H{sub 2}, CO) and fission products to the containment atmosphere occurs as a result. In the long time (>10 h) containment failure and basemat penetration may occur, which can lead to fission product release to the environment. For further development and validation, simulations of experiments in which molten core concrete interaction (MCCI) is investigated, are necessary. In this work the new available effective heat transfer model in MEDICIS is used to calculate experiments of the ACE program, in which generic corium material is heated up and interacts with the concrete basemat. Here, especially the ACE L2 experiment with siliceous concrete and the ACE L5 experiment with limestone common sand (LCS) concrete will be presented. These experiments enable to analyze the heat transfer from the interior of the melt to the upper surface under dry conditions. Secondary the modeling in ASTEC version 2.p2 with the effective heat transfer module in MEDICIS is described. Results of MEDICIS simulations will be discussed by means of phenomena like ablation behavior and erosions depth, layer temperature and surface heat loss. Finally the issue of an effective heat transfer coefficient for the surface under dry conditions without top flooding is figured out. (orig.)

Dynamic Heat Storage and Cooling Capacity of a Concrete Deck with PCM and Thermally Activated Building System

DEFF Research Database (Denmark)

Pomianowski, Michal Zbigniew; Heiselberg, Per; Jensen, Rasmus Lund

2012-01-01

This paper presents a heat storage and cooling concept that utilizes a phase change material (PCM) and a thermally activated building system (TABS) implemented in a hollow core concrete deck. Numerical calculations of the dynamic heat storage capacity of the hollow core concrete deck element...... in the article highlight the potential of using TABS and PCM in a prefabricated concrete deck element....

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Interaction between a fluid at high temperature and a concrete: contribution to the modeling of heat and mass transfers

International Nuclear Information System (INIS)

Introini, C.

2010-01-01

In the late phases of some scenario of hypothetical severe accident in Pressurized Water Reactors, a molten mixture of core and vessel structures, called corium, comes to interact with the concrete basemat. The safety numerical tools are lumped parameter codes. They are based on a large averaged description of heat and mass transfers which raises some uncertainties about the multi-scale description of the exchanges but also about the adopted boundary layer structure in the vicinity of the ablation front. In this context, the aim of this work is to tackle the problem of the boundary layer structure by means of direct numerical simulation. This work joins within the more general framework of a multi-scale description and a multi-scale modeling, namely from the local scale associated with the vicinity of the ablation front to the scale associated with the lumped parameter codes. Such a multi-scale description raises not only the problem of the local description of the multiphase multicomponent flow but also the problem of the up-scaling between the local- and the macro-scale which is associated with the convective structures within the pool of corium. Here, we are particularly interested in the building of effective boundary conditions or wall laws for macro-scale models. The difficulty of the multiphase multicomponent problem at the local scale leads us to consider a relatively simplified problem. Effective boundary conditions are built in the frame of a domain decomposition method and numerical experiments are performed for a natural convection problem in a stamp shaped cavity to assess the validity of the proposed wall laws. Even if the treated problem is still far from the target applications, this contribution can be viewed as a first step of a multi-scale modeling of the exchanges for the molten core concrete issue. In the more complicated case of multiphase multicomponent flows, it is necessary to have a direct numerical simulation tool of the flow at the local

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

NARCIS (Netherlands)

Rai, N.; van Westrenen, W.

2014-01-01

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-09-15

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

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.

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-07-01

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

Aluminum titanate crucible for molten uranium

International Nuclear Information System (INIS)

Asbury, J.J.

1975-01-01

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

Bearing capacity and rigidity of short plastic-concrete-tubal vertical columns under transverse load

Science.gov (United States)

Dolzhenko, A. V.; Naumov, A. E.; Shevchenko, A. E.

2018-03-01

The results of mathematical modeling in determining strain-stress distribution parameters of a short plastic-concrete-tubal vertical column under horizontal load as those in vertical constructions are described. Quantitative parameters of strain-stress distribution during vertical and horizontal loads and horizontal stiffness were determined by finite element modeling. The internal stress in the concrete column core was analyzed according to equivalent stress in Mohr theory of failure. It was determined that the bearing capacity of a short plastic- concrete-tubal vertical column is 25% higher in resistibility and 15% higher in rigidness than those of the caseless concrete columns equal in size. Cracks formation in the core of a short plastic-concrete-tubal vertical column happens under significantly bigger horizontal loads with less amount of concrete spent than that in caseless concrete columns. The significant increase of bearing capacity and cracking resistance of a short plastic-concrete-tubal vertical column under vertical and horizontal loads allows recommending them as highly effective and highly reliable structural wall elements in civil engineering.

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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

International Nuclear Information System (INIS)

Rogers, J.T.

1984-06-01

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

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.

Use of combined destructive and non-destructive test methods to assess the strength of concrete

Energy Technology Data Exchange (ETDEWEB)

Arioz, O. [Optimizing Consultancy, Izmir (Turkey); Kilinc, K. [Kirklareli University, Department of Civil Engineering, Kirklareli (Turkey); Ramyar, K. [Ege University, Department of Civil Engineering, Ismir (Turkey); Tuncan, M.; Tuncan, A. [Anadolu University, Department of Civil Engineering, Eskişehir (Turkey)

2013-07-01

The compressive strength test applied on standard samples is one of the most important tests indicating the quality of concrete in structures. The results of the standard tests are compared with the values used in design calculations and the quality of concrete is controlled. Although the standard tests are well accepted by the construction industry, they may not represent the in-situ strength of concrete due to the differences between the degree of compaction and curing conditions of concrete and those of standard samples. In-situ strength is also important for the efficient planning of the construction works in huge projects. In the present study, the results obtained from standard tests, core tests, ultrasonic pulse velocity tests, and rebound hammer tests were extensively analysed for the assessment of concrete strength. Key words: Concrete strength, standard tests, core test, ultrasonic pulse velocity, rebound number.

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)

LIFE Materails: Molten-Salt Fuels Volume 8

Energy Technology Data Exchange (ETDEWEB)

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

2008-12-11

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

LIFE Materails: Molten-Salt Fuels Volume 8

International Nuclear Information System (INIS)

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

2008-01-01

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

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

Science.gov (United States)

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

2017-04-01

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

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)

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

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.

Behavior of concrete cylinders confined by a ferro-geopolymer jacket in axial compression

Directory of Open Access Journals (Sweden)

Kothay Heng

2017-06-01

Full Text Available It is beneficial to utilize geopolymers for their potential properties to rehabilitate concrete structures. These properties include high adhesion to Ordinary Portland Cement (OPC concrete even at low degrees of interfacial roughness, high durability and good fire resistance. This paper introduces use of a ferro-geopolymer jacket to strengthen concrete columns. It is a kind of jacket constructed with a geopolymer mortar reinforced with a wire mesh. This study was conducted to investigate the behavior of concrete cylinders confined with a ferro-geopolymer jacket in axial compression. OPC concrete cylinders with 100 mm diameter and 200 mm height were fabricated. High calcium fly ash-based geopolymer mortar, activated with sodium hydroxide (NaOH and sodium silicate (Na2SiO3, cured at a temperature of 25 ºC was used. Ferro-geopolymer jackets with a25 mm thickness, were reinforced with 1, 2 and 3 layers of expanded metal mesh and cast around concrete cylinders. The study results revealed that the compressive load carrying capacity and axial stiffness of concrete cylinders were improved. A monolithic failure mode was obtained as a result of a strong adhesion between the geopolymer and the concrete core. Enhancement of compressive load carrying capacity of the jacketed concrete cylinders was caused by a combination of a confinement effect and the compressive load resistance of the jacket transferred from concrete core through bonding.

Molten salt processes in special materials preparation

International Nuclear Information System (INIS)

Krishnamurthy, N.; Suri, A.K.

2013-01-01

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

Studies on components for a molten salt reactor

International Nuclear Information System (INIS)

Nejedly, M.; Matal, O.

2003-01-01

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

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

International Nuclear Information System (INIS)

Chi Chenwei; Ferng Yuhming; Pei Baushei; Liang Jenqhorng

2011-01-01

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Comparison of the performance of concrete-filled steel tubular and hollow steel diagrid buildings

Science.gov (United States)

Peter, Minu Ann; S, Sajith A.; Nagarajan, Praveen

2018-03-01

In the recent construction scenario, diagrid structures are becoming a popular high-rise building structural system. Diagrid structures consist of diagonals in the perimeter and an interior core. The corner and interior vertical columns are not required due to the structural efficiency of diagrid structural systems. Steel and concrete are commonly used material for diagrid. An alternate material for diagrid is concrete-filled steel tube (CFST). CFST incorporates the advantages of both steel and concrete. In CFST, the inward buckling of the steel tube is effectively prevented by the filled concrete. The compressive strength of concrete increases due to the tri-axial state of stress in concrete induced by the steel tube. The longitudinal as well as lateral reinforcement to the concrete core is also provided by the steel tube. This paper compares the performance of CFST and steel diagrid buildings using linear static analysis. For this purpose, a 12 storey and 36 storey building are analysed using finite element method and CFST diagrid building is found to perform better.

Radiant heat evaluation of concrete: a study of the erosion of concrete due to surface heating

International Nuclear Information System (INIS)

Chu, T.Y.

1978-01-01

Experiments were carried out to investigate the erosion of concrete under high surface heat flux in connection with the core-melt/concrete interaction studies. The dominate erosion mechanism was found to be melting at the surface accompanied by chemical decomposition of the concrete beneath the melt-solid interface. The erosion process reaches a steady state after an initial transient. The steady state is characterized by an essentially constant erosion rate at the surface and a nonvarying (with respect to the moving melt interface) temperature distribution within the concrete. For the range of incident heat flux 64 W/cm 2 to 118 W/cm 2 , the corresponding steady state erosion rate varies from approximately 8 cm/hr to 23 cm/hr. A simple ablation/melting model is proposed for the erosion process. The model was found to be able to correlate all temperature responses at various depths from all tests at large times and for temperatures above approximately 250 0 C

Waste treatment using molten salt oxidation

International Nuclear Information System (INIS)

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

1996-01-01

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

Molten fuel-moderator interaction

International Nuclear Information System (INIS)

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

1987-02-01

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Preliminary Design of Optimized Reactor Insulator for Severe Accident Mitigation of APR1400

International Nuclear Information System (INIS)

Heo, Sun; Lee, Jae-Gon; Kang, Yong-Chul

2007-01-01

APR1400, a Korean evolutionary advance light water reactor, has many advanced safety feature to prevent and mitigate of design basis accident (DBA) and severe accident. When reactor cooling system (RCS) fails to cooling its core, the core melted down and the molten core gathers together on bottom of reactor vessel. The molten core hurts reactor vessel and is released to containment, which raises the release of radioactive isotopes and the heating of the containment atmosphere. Finally, the corium is accumulated in the bottom of reactor cavity and it also raises the Molten Core and Concrete Interaction (MCCI) and the heating of containment atmosphere. There are two strategies to cooling molten core. Those are in-vessel retention and ex-vessel cooling. At the early stage of APR1400 design, only ex-vessel cooling which is cooling of the molten core outside the vessel after vessel failure is considered based on EPRI Utility Requirement Document (URD) for Evolutionary LWR. However, a need has been arisen to reflect current research findings on severe accident phenomena and mitigation technologies to Korean URD and IVRERVC (In-Vessel corium Retention using Ex-Reactor Vessel Cooling) was adopted APR1400. The ERVC is not considered as a licensing design basis but based on the defense-in-depth principle and safety margin basis, which is the top-tier requirement of the severe accident mitigation design as stated in the KURD. The Severe Accident Management strategy for APR1400 is intended to aid the plant operating staff to secure reactor vessel integrity in the early stage of the severe accident. As a part of a design implementation of IVR-ERVC for APR1400, we developed the preliminary design requirement, design specification and conceptual design

Metal Production by Molten Salt Electrolysis

DEFF Research Database (Denmark)

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

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

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

International Nuclear Information System (INIS)

Dai, Zhimin; Liu, Wei

2013-01-01

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

Inertia-confining thermonuclear molten salt reactors

International Nuclear Information System (INIS)

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

1984-01-01

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

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

International Nuclear Information System (INIS)

Yoshioka, Ritsuo; Mitachi, Koshi

2013-01-01

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

Molten salt: Corrosion problems and electrometallurgy in nuclear applications

International Nuclear Information System (INIS)

Santarini, G.

1981-01-01

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

The role of high temperature heterogeneous reaction kinetics in the rate of radionuclide vaporisation during core-concrete interactions

International Nuclear Information System (INIS)

Raymond, D.P.; Clough, P.N.

1989-09-01

Heterogeneous reactions may cause enhanced release of radionuclides during the core-concrete interaction (CCl) stage of a PWR severe accident. The VANESA computer code models these CCI releases using chemical equilibrium assumptions; however, the possibility that chemical kinetics could prevent equilibrium from being achieved is considered in this report. Direct experimental evidence is lacking on these reactions. Therefore, some analogues studies are reviewed, including examples of Eyring's surface reaction rate theory; sequential vaporisation-oxidation processes; iron and steelmaking chemistry; radionuclide evaporation from solid UO 2 . This circumstantial evidence appeared to agree with the current assumptions, in VANESA and some UK modelling studies, that mass transfer, rather than chemical kinetics will limit the rate at which equilibrium is attained. (author)

The foundation mass concrete construction technology of Hongyun Building B tower raft

Science.gov (United States)

Liu, Yu; Yin, Suhua; Wu, Yanli; Zhao, Ying

2017-08-01

The foundation of Hongyun building B tower is made of raft board foundation which is 3300mm in the thickness and 2800mm beside side of the core tube. It is researched that the raft foundation mass concrete construction technology is expatiated from temperature and cracks of the raft foundation and the temperature control and monitoring of the concrete base slab construction and concrete curing.

Cylindrical prestressed concrete pressure vessel for a nuclear power plant

International Nuclear Information System (INIS)

Horner, M.; Hodzic, A.; Haferkamp, D.

1976-01-01

A prestressed concrete pressure vessel for a HTGR is proposed which encloses, in addition to the reactor core, not only the heat-exchanging facilities but also the turbine unit. The reinforcement of the cylindrical concrete body is to be carried out with special care, it is provided for horizontal tendons, the prestressed concrete pressure vessel has a wire-winding device, while the longitudinal reinforcement is achieved by tendous guided in parallel to the vesses axes through the interspaces between the pods. (UWI) [de

Ceramics for Molten Materials Transfer

Science.gov (United States)

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

2009-01-01

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

Development of a safety analysis code for molten salt reactors

International Nuclear Information System (INIS)

Zhang Dalin; Qiu Suizheng; Su Guanghui

2009-01-01

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

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

Science.gov (United States)

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

2017-12-01

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

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

International Nuclear Information System (INIS)

Furukawa, K.

1985-01-01

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

EPRTM engineered features for core melt mitigation in severe accidents

International Nuclear Information System (INIS)

Fischer, Manfred; Henning, Andreas

2009-01-01

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

Transfer characteristics of a lithium chloride–potassium chloride molten salt

Directory of Open Access Journals (Sweden)

Eve Mullen

2017-12-01

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

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

International Nuclear Information System (INIS)

Keiser, J.R.

1977-05-01

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

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)

Conributions of the VULCANO experimental programme to ther understaing of MCCI phenomena

Energy Technology Data Exchange (ETDEWEB)

Journeau, Christophe; Piluso, Pascal; Correggio, Patricia; Ferry, Lionel; Fritz, Gerald; Haquet, Jean Francois; Monerris, Jose; Ruggieri, Jean Michel; Sanchez- Brusset, Mathieu; Parga, Clemente [CEA, DEN, Cadarache, STRI/LMA, lez Durance (France)

2012-04-15

Molten Core Concrete Interaction (MCCI) is a complex process characterized by concrete ablation and volatile generation; Thermal and solutal convection in a bubble-agitated melt; Physico-chemical evolution of the corium pool with a wide solidification range (of the order of 1000 K). Twelve experiments have been carried out in the VULCANO facility with prototypic corium and sustained heating. The dry oxidic corium tests have contributed to show that silica-rich concrete experience an anisotropic ablation. This unexpected ablation pattern is quite reproducible and can be recalculated, provided an empirical anisotropy factor is assumed. Dry tests with oxide and metal liquid phases have also yielded unexpected results: a larger than expected steel oxidation and unexpected topology of the metallic phase (at the bottom of the cavity and also on the vertical concrete walls). Finally, VULCANO has proved its interest for the study of mitigation solutions such as the COMET bottom flooding core catcher.

Conributions of the VULCANO experimental programme to ther understaing of MCCI phenomena

International Nuclear Information System (INIS)

Journeau, Christophe; Piluso, Pascal; Correggio, Patricia; Ferry, Lionel; Fritz, Gerald; Haquet, Jean Francois; Monerris, Jose; Ruggieri, Jean Michel; Sanchez- Brusset, Mathieu; Parga, Clemente

2012-01-01

Molten Core Concrete Interaction (MCCI) is a complex process characterized by concrete ablation and volatile generation; Thermal and solutal convection in a bubble-agitated melt; Physico-chemical evolution of the corium pool with a wide solidification range (of the order of 1000 K). Twelve experiments have been carried out in the VULCANO facility with prototypic corium and sustained heating. The dry oxidic corium tests have contributed to show that silica-rich concrete experience an anisotropic ablation. This unexpected ablation pattern is quite reproducible and can be recalculated, provided an empirical anisotropy factor is assumed. Dry tests with oxide and metal liquid phases have also yielded unexpected results: a larger than expected steel oxidation and unexpected topology of the metallic phase (at the bottom of the cavity and also on the vertical concrete walls). Finally, VULCANO has proved its interest for the study of mitigation solutions such as the COMET bottom flooding core catcher.

Electrochemistry of plutonium in molten halides

International Nuclear Information System (INIS)

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

1987-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-09-15

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

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.

Development of Severe Accident Containment Analysis Package

Energy Technology Data Exchange (ETDEWEB)

Park, Chang-Hwan; Kim, Dong-Min; Seo, Jea-Uk; Lee, Dea-Young; Park, Soon-Ho; Lee, Jae-Gwon; Lee, Jin-Yong; Lee, Byung-Chul [FNC Technology Co., Yongin (Korea, Republic of)

2016-10-15

In safety viewpoint, the pressure and temperature of the containment is the important parameters, of course, the local hydrogen concentration is also the parameter of the major concern because of its flammability and the risk of the detonation. In addition, there are possibilities of occurrence of other relevant phenomena following the reactor core melting such as DCH(direct containment heating) due to HPME(high pressure melt ejection), steam explosion due to fuel-coolant interaction in the reactor cavity and molten core concrete interaction at the late stage. It is important to predict the containment responses during a severe accident by a reasonable accuracy for establishing of effective mitigation strategies and preparation of the safety features required. In this paper, the overview of the SACAP development status is presented. SACAP is developed so as to be able to analyze, so called, Ex-Vessel severe accident phenomena including thermal-hydraulics, combustible gas burn, direct containment heating, steam explosion and molten core-concrete interaction. At the parallel time, SACAP and In-Vessel analysis module named CSPACE are processed for integration through MPI communication coupling. Development of the integrated severe accident analysis code system will be completed in following one year to make the code revision zero to be released.

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.

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

The use of Impact-Echo and Spectral-Analysis-of-Surface-Waves methods for the concrete investigation of Rogers Dam spillway structure

International Nuclear Information System (INIS)

Olson, L.D.; Sack, D.A.; Chan, Y.F.; Gilmore, R.T.; Christy, J.T.; Dumont, M.F.

1994-01-01

IE and SASW NDT methods were employed to investigate the concrete conditions of the Rogers Hydro Station's concrete spillway. The results showed that the surface gunite/shotcrete on the majority of the piers and walls was delaminated from the interior sound concrete and that there was a significant amount of sound concrete in the interior cores of the piers and walls. The NDT results correspond well in general trend with those of the concrete coring performed in 1988 and 1991; however, because of the greatly varying concrete conditions and additional concrete deterioration since 1988 when the majority of the horizontal cores were taken, a direct point-per-point comparison cannot be made. The NDT results correspond well with the as-found conditions on Pier 1 and on the east abutment wall except that the deep degradation (SASW results) conditions on the west face of Pier 1 did not exist. The highly deteriorated nature of the concrete could have contributed to the lower wave velocities. The cracking conditions below the demolition line were not verified. Some of the IE echoes could have been caused by the boundary conditions, in view of the highly fractured nature of the surface concrete/shotcrete and the very low strength and deteriorated concrete at the construction joints

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)

Wireless Concrete Strength Monitoring of Wind Turbine Foundations.

Science.gov (United States)

Perry, Marcus; Fusiek, Grzegorz; Niewczas, Pawel; Rubert, Tim; McAlorum, Jack

2017-12-16

Wind turbine foundations are typically cast in place, leaving the concrete to mature under environmental conditions that vary in time and space. As a result, there is uncertainty around the concrete's initial performance, and this can encourage both costly over-design and inaccurate prognoses of structural health. Here, we demonstrate the field application of a dense, wireless thermocouple network to monitor the strength development of an onshore, reinforced-concrete wind turbine foundation. Up-to-date methods in fly ash concrete strength and maturity modelling are used to estimate the distribution and evolution of foundation strength over 29 days of curing. Strength estimates are verified by core samples, extracted from the foundation base. In addition, an artificial neural network, trained using temperature data, is exploited to demonstrate that distributed concrete strengths can be estimated for foundations using only sparse thermocouple data. Our techniques provide a practical alternative to computational models, and could assist site operators in making more informed decisions about foundation design, construction, operation and maintenance.

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-05-15

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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.

Effect of water on the triaxial response under monotonic loading of asphalt concrete used in dams

Science.gov (United States)

Gaxiola Hernández, Alberto; Ossa López, Alexandra

2018-01-01

Embankment dams with asphalt concrete cores have been constructed on practically all continents with satisfactory results. Nowadays many advantages, such as the mechanical strength, are known that makes asphalt concrete a competitive alternative for the construction of the impervious elements of dams. However, the current available information does not describe the effect of prolonged contact between asphalt concrete and water on the structure of an embankment dam. In this research cylindrical asphalt concrete specimens with a void content similar to that used in impervious barriers of dams were fabricated and submerged in water for a prolonged period to simulate the conditions experienced by asphalt concrete placed inside an embankment dam as its core material. Subsequently, triaxial compression tests were conducted on the specimens. The results indicated that the asphalt concrete exhibited a reduction in strength because of the saturation process to which the material was subjected. However, no changes were observed in the mechanical response to prolonged contact with water for periods of up to 12 months.

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)

Core melt retention and cooling concept of the ERP

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-01

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Selective Adsorption of Sodium Aluminum Fluoride Salts from Molten Aluminum

Energy Technology Data Exchange (ETDEWEB)

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

2007-08-16

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

Historic Concrete : From Concrete Repair to Concrete Conservation

NARCIS (Netherlands)

Heinemann, H.A.

2013-01-01

Concrete like materials were already applied during the Roman Empire. After the decline of the Roman Empire, a wide scale application of concrete only reappeared in the 19th century. Here lies also the origin of modern (reinforced) concrete. Since then, both concrete application and composition have

A Simplified Method for Stationary Heat Transfer of a Hollow Core Concrete Slab Used for TABS

DEFF Research Database (Denmark)

Yu, Tao; Heiselberg, Per Kvols; Lei, Bo

2014-01-01

Thermally activated building systems (TABS) have been an energy efficient way to improve the indoor thermal comfort. Due to the complicated structure, heat transfer prediction for a hollow core concrete used for TABS is difficult. This paper proposes a simplified method using equivalent thermal...... resistance for the stationary heat transfer of this kind of system. Numerical simulations are carried out to validate this method, and this method shows very small deviations from the numerical simulations. Meanwhile, this method is used to investigate the influence of the thickness of insulation on the heat...... transfer. The insulation with a thickness of more than 0.06 m can keep over 95 % of the heat transferred from the lower surface, which is beneficial to the radiant ceiling cooling. Finally, this method is extended to involve the effect of the pipe, and the numerical comparison results show that this method...

Adhesive liquid core optical fibers for crack detection and repairs in polymer and concrete matrices

Science.gov (United States)

Dry, Carolyn M.

1995-04-01

This work is an investigation into the feasibility of using liquid core optical fibers for the detection and self repair of cracking in cement or polymer materials generated by dynamic or static loading. These experiments rely on our current research sponsored by the National Science Foundation. It combines that work on the concept of internal adhesive delivery from hollow fibers for repair with nondestructive fiber optic analysis of the crack localization and volume within the same system. The need to monitor the internal state of civil structures and materials is great. Existing instrumentation techniques that mainly rely on magnetism, electricity, or stress gauges are limited if used for remote measurements in concrete or composites. They are sensitive to electrical magnetic noises and they degrade in the environment over time. Optical fibers are attractive because they are immune to electromagnetic interference and are sensitive over long distances. The combination of the ability to remotely measure crack occurrence in real time and determine the location and volume of crack damage in the matrix is unique in the field of optic sensors (or any sensors in general). The combination of this with crack repair, rebonding of any detached or broken fibers, and replenishment of liquid core chemicals, when necessary, make this a potentially powerful sensing and repair tool. Work on this research topic of the combination sponsored by the University of Illinois, looks very promising as a rapid innovative advance.

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

Nonlinear core deflection in injection molding

Science.gov (United States)

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

2018-05-01

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

Experimental studies of oxidic molten corium-vessel steel interaction

International Nuclear Information System (INIS)

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

2001-01-01

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

Experimental studies of oxidic molten corium-vessel steel interaction

Energy Technology Data Exchange (ETDEWEB)

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

2001-12-01

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

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

Science.gov (United States)

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

2018-02-01

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

Reactors with pressure vessel in pre-stressed concrete

International Nuclear Information System (INIS)

Devillers, Christian; Lafore, Pierre

1964-12-01

After having proposed a general description of the evolution of the general design of reactors with a vessel in pre-stressed concrete, this report outlines the interest of this technical solution of a vessel in pre-stressed concrete with integrated exchangers, which is to replace steel vessel. This solution is presented as much safer. The authors discuss the various issues related to protection: inner and outer biological protection of the vessel, material protection (against heating, steel irradiation, Wigner effect, and moderator radiolytic corrosion). They report the application of calculation methods: calculation of vessel concrete heating, study of the intermediate zone in integrated reactors, neutron spectrum and flows in the core of a graphite pile

Dynamics of the Molten Contact Line

Science.gov (United States)

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

1999-01-01

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

Fission product removal from molten salt using zeolite

International Nuclear Information System (INIS)

Pereira, C.; Babcock, B.D.

1996-01-01

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

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)

Electrochemical ion separation in molten salts

Science.gov (United States)

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

2017-12-19

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

Molten salt reactor as asymptotic safety nuclear system

International Nuclear Information System (INIS)

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

1989-01-01

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

Method for the construction of a nuclear reactor with a prestressed concrete pressure vessel

International Nuclear Information System (INIS)

Schoening, J.; Schwiers, H.G.

1981-01-01

Method for the construction of nuclear reactors with prestressed concrete pressure vessel, providing during the initial stage of construction of the prestressed concrete pressure vessel a support structure around the liner. This enables an early mounting of core components in clean conditions as well as load reductions for final concreting in layers of the prestressed concrete pressure vessel. By applying the support structure, the overall assembly time of these nuclear power plant is considerably reduced without extra cost. (orig.) [de

46 CFR 151.50-55 - Sulfur (molten).

Science.gov (United States)

2010-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-04-01

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

Chloride Ingress in Concrete with Different Age at Time of First Chloride Exposure

DEFF Research Database (Denmark)

Hansen, Esben Østergaard; Iskau, Martin Riis; Hasholt, Marianne Tange

2016-01-01

Concrete structures cast in spring have longer time to hydrate and are therefore denser and more resistant to chloride ingress when first subjected to deicing salts in winter than structures cast in autumn. Consequently, it is expected that a spring casting will have a longer service life....... This hypothesis is investigated in the present study by testing drilled cores from concrete cast in 2012 and 2013 on the Svendborgsund Bridge. The cores are subject to petrographic examination and mapping of chloride profiles. Moreover, chloride migration coefficients have been measured. The study shows...

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)

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

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.

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

1993-11-01

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

Numerical analysis of concrete-filled tubes with stiffening plates under large deformation axial loading

OpenAIRE

Albareda Valls, Albert

2013-01-01

Concrete-filled tubes have been increasingly used these recent decades thanks to their improved structural behavior, especially under compression.Concrete filling in these sections improves ¡ts compressive strength thanks to lateral pressure coming from confinement effect provided by the steel tube. At elevated percentages of loading,concrete suffers an important volumetric expansion, which is clearly restricted by the tube. Therefore, the core is subjected to a severe lateral pressure tha...

Application of lithium in molten-salt reduction processes

International Nuclear Information System (INIS)

Gourishankar, K. V.

1998-01-01

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

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

International Nuclear Information System (INIS)

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

1976-01-01

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

Simulationsexperimente zum Ausbreitungsverhalten von Kernschmelzen: KATS-8 bis KATS-17

International Nuclear Information System (INIS)

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

2001-03-01

In future Light Water Reactors special devices (core catchers) might be required to prevent containment failure by basement erosion after reactor pressure vessel meltthrough during a core meltdown accident. Quick freezing of the molten core masses is desirable to reduce release of radioactivity. Several concepts of core catcher devices have been proposed based on the spreading of corium melt onto flat surfaces with subsequent water cooling. Therefore, a series of experiments to investigate high temperature melt spreading on flat surfaces has been carried out using alumina-iron thermite melts as a simulant. The oxidic thermite melt is conditioned by adding other oxides to simulate a realistic corium melt as close as possible. Spreading of oxidic and metallic melts have been performed in one- and two-dimensional geometry. Substrates were inert ceramical layer, dry concrete and concrete with a water layer of several millimeters. The influence of a shallow water layer on the surface onto the spreading behaviour has also been studied. (orig.) [de

Molten salt extractive distillation process for zirconium-hafnium separation

International Nuclear Information System (INIS)

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

1989-01-01

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

Advances in molten salt electrochemistry towards future energy systems

International Nuclear Information System (INIS)

Ito, Yasuhiko

2005-01-01

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

Partially molten magma ocean model

International Nuclear Information System (INIS)

Shirley, D.N.

1983-01-01

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

Molten salts processes and generic simulation

International Nuclear Information System (INIS)

Ogawa, Toru; Minato, Kazuo

2001-01-01

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

Development of MPS Method for Analyzing Melt Spreading Behavior and MCCI in Severe Accidents

Science.gov (United States)

Yamaji, Akifumi; Li, Xin

2016-08-01

Spreading of molten core (corium) on reactor containment vessel floor and molten corium-concrete interaction (MCCI) are important phenomena in the late phase of a severe accident for assessment of the containment integrity and managing the severe accident. The severe accident research at Waseda University has been advancing to show that simulations with moving particle semi-implicit (MPS) method (one of the particle methods) can greatly improve the analytical capability and mechanical understanding of the melt behavior in severe accidents. MPS models have been developed and verified regarding calculations of radiation and thermal field, solid-liquid phase transition, buoyancy, and temperature dependency of viscosity to simulate phenomena, such as spreading of corium, ablation of concrete by the corium, crust formation and cooling of the corium by top flooding. Validations have been conducted against experiments such as FARO L26S, ECOKATS-V1, Theofanous, and SPREAD for spreading, SURC-2, SURC-4, SWISS-1, and SWISS-2 for MCCI. These validations cover melt spreading behaviors and MCCI by mixture of molten oxides (including prototypic UO2-ZrO2), metals, and water. Generally, the analytical results show good agreement with the experiment with respect to the leading edge of spreading melt and ablation front history of concrete. The MPS results indicate that crust formation may play important roles in melt spreading and MCCI. There is a need to develop a code for two dimensional MCCI experiment simulation with MPS method as future study, which will be able to simulate anisotropic ablation of concrete.

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

International Nuclear Information System (INIS)

Csom, Gyula; Feher, Sandor; Szieberthj, Mate

2002-01-01

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

Determination of thermal characteristics of standard and improved hollow concrete blocks using different measurement techniques

DEFF Research Database (Denmark)

Caruana, C.; Yousif, C.; Bacher, Peder

2017-01-01

The lighter weight, improved thermal properties and better acoustic insulation of hollow-core concrete blocks are few of the characteristics that one encounters when comparing them to traditional Maltese globigerina limestone solid blocks. As a result, hollow concrete blocks have recently been...

Refractory thermowell for continuous high temperature measurement of molten metal

International Nuclear Information System (INIS)

Thiesen, T.J.

1992-01-01

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

A basic study on fluoride-based molten salt electrolysis technology

Energy Technology Data Exchange (ETDEWEB)

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

2001-04-01

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

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

International Nuclear Information System (INIS)

Calderoni, Pattrick

2010-01-01

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