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1.
  • Ottosen, Niels Saabye (författare)
  • Relaxation of a Rectangular Beam and Circular Shaft
  • 1982
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 1872-759X .- 0029-5493. ; 75:1, s. 67-72
  • Tidskriftsartikel (refereegranskat)abstract
    • The paper provides exact and approximate solutions to the title problems. Time-hardening creep is adopted and the stress dependence is assumed to follow the exponential expression proposed by Soderberg. Even though the exact solutions are easy to work with, some approximate solutions are discussed. These approximate solutions are obtained from the exact ones by simply ignoring certain terms; errors bounds are then directly available. The exact and approximate solutions are applied to specific problems and compared with the predictions following the exact, numerical or approximate solution of Norton's power law.
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2.
  • Bechta, Sevostian, et al. (författare)
  • Experimental studies of oxidic molten corium-vessel steel interaction
  • 2001
  • Ingår i: Nuclear Engineering and Design. - 0029-5493 .- 1872-759X. ; 210:1-3, s. 193-224
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
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3.
  • Bechta, Sevostian, et al. (författare)
  • Water boiling on the corium melt surface under VVER severe accident conditions
  • 2000
  • Ingår i: Nuclear Engineering and Design. - 0029-5493 .- 1872-759X. ; 195:1, s. 45-56
  • Tidskriftsartikel (refereegranskat)abstract
    • Experimental results are presented on the interaction of corium melt with water supplied on its surface. The tests were conducted in the `Rasplav-2' experimental facility. Corium melt was generated by induction melting in the cold crucible. The following data were obtained: heat transfer at boiling water-melt surface interaction, gas and aerosol release, post-interaction solidified corium structure. The corium melt charge had the following composition, mass%: 60% UO2+x-16% ZRO2-15% Fe2O3-6% Cr2O3-3% Ni2O3. The melt surface temperature ranged within 1920-1970 K.
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4.
  • Eriksson, Kjell (författare)
  • The effective fracture toughness of structural components obtained with the blend rule
  • 1998
  • Ingår i: Nuclear Engineering and Design. - 0029-5493 .- 1872-759X. ; 182:2, s. 123-129
  • Tidskriftsartikel (refereegranskat)abstract
    • The blend rule for the effective fracture toughness of a layered material was originally derived from the special case of a through crack in a globally elastic material and later extended to accomodate non-linear behaviour. It is now derived from a general case by considering material elements of finite size and of different toughness along and around the tip of a crack. Experimental results obtained with an inhomogeneous ordinary structural steel which support the blend rule are presented. It is shown that the effective fracture toughness governs the load-bearing capacity of a cracked full-scale structure. Some further results found in the literature for the heat-affected zone material of a high-strength microalloyed quenched and tempered structural steel and computational results for a structural steel typical of a nuclear pressure vessel are shown to support the blend rule.
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5.
  • Lindgren, Lars-Erik, et al. (författare)
  • Thermo-mechanical FE-analysis of residual stresses and stress redistribution in butt welding of a copper canister for spent nuclear fuel
  • 2002
  • Ingår i: Nuclear Engineering and Design. - 0029-5493 .- 1872-759X. ; 212:1-3, s. 401-408
  • Tidskriftsartikel (refereegranskat)abstract
    • The transient and residual temperature, stress and strain field present during electron beam welding of a plane copper end to a canister for spent nuclear fuel is calculated by the use of FEM. The subsequent stress redistribution is calculated up to 10,000 years. The canister consists of two concentric cylinders, an inner steel cylinder containing the spent nuclear fuel and an outer copper cylinder. It was found that the maximum plastic strain (plastic+creep) accumulated in the (possibly brittle) heat affected zone is ≈7%, which seems to be well below the reported ductility for the copper used.
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6.
  • Nilsson, Fred L. (författare)
  • Risk-based approach to plant life management
  • 2003
  • Ingår i: Nuclear Engineering and Design. - 0029-5493 .- 1872-759X. ; 221:03-jan, s. 293-300
  • Tidskriftsartikel (refereegranskat)abstract
    • The growing interest in risk-based management is driven by the need of developing strategies that lead to an optimal safety versus cost balance. The general philosophy behind risk-based management is briefly described here and discussed. It is formally shown that the core damage frequency can be factored into a system (PSA) part and component failure frequency part. Some of the procedures, currently applied in risk-based inspection, are discussed. The basic elements for failure frequency calculations are presented and discussed. A quantitative risk-based inspection study performed for the Oskarshamn 1 unit is briefly presented as an example of how risk-based procedures can be applied.
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7.
  • Nilsson, P., et al. (författare)
  • On the behavior of crack surface ligaments
  • 1998
  • Ingår i: Nuclear Engineering and Design. - 0029-5493 .- 1872-759X. ; 184:1, s. 145-153
  • Tidskriftsartikel (refereegranskat)abstract
    • Small ligaments connecting the fracture surfaces just behind a moving crack front are assumed to exist under certain conditions. The ligaments are rapidly torn as the crack advances. Inelastic straining of such ligaments influences the energy balance in the fracture process. The rapid tearing of a single ligament is studied both numerically and experimentally. An elastic visco-plastic material model is adopted for finite-element calculations. The results show that relatively large amounts of energy are dissipated during the tearing process. Further, the energy needed to tear a ligament increases rapidly with increasing tearing rate. The computed behavior is partly verified in a few preliminary experiments. The implications for slow stable crack tip speeds during dynamic fracture are discussed.
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8.
  • Niwa, Y., et al. (författare)
  • Integrated computerisation of operating procedures
  • 2002
  • Ingår i: Nuclear Engineering and Design. - 0029-5493 .- 1872-759X. ; 213:2-3, s. 289-301
  • Tidskriftsartikel (refereegranskat)abstract
    • The computerisation of procedures can address not only the way procedures are presented or displayed, but also how they are generated and maintained. The computerisation of procedure generation will result in an improved basis for procedure display, and also make it possible to maintain the procedures on the level of internal representation rather than on the level of display. This paper describes a project, which focused on developing a system for computerised procedure generation (CPG) based on the principles of cognitive systems engineering. The CPG system was integrated with an already existing system for computerised procedure presentation, and enhanced with a number of other functions to produce a system for integrated computerisation of operating procedures. © 2002 Elsevier Science B.V. All rights reserved.
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9.
  • Nourgaliev, R. R., et al. (författare)
  • On lattice Boltzmann modeling of phase transition in an isothermal non-ideal fluid
  • 2002
  • Ingår i: Nuclear Engineering and Design. - 0029-5493 .- 1872-759X. ; 211:2-3, s. 153-171
  • Tidskriftsartikel (refereegranskat)abstract
    • A new lattice Bolztmann BGK model for isothermal non-ideal fluid is introduced and formulated for an arbitrary lattice, composed of several D-dimensional sublattices. The model is a generalization of the free-energy-based lattice Bolztmann BGK model developed by Swift et al. (1996). We decompose the equilibrium distribution function in the BGK collision operator into ideal and non-ideal parts and employ second-order Chapman-Enskog expansion for treatment of both parts. Expansion coefficients for the non-ideal part are, in general. functions of macroscopic variables, designed to reproduce desired pressure tensor (thermodynamic aspects) and to eliminate the aphysical artifacts in the lattice Bolztmann model. The new model is shown to significantly improve quality of lattice Boltzmann modeling of interfacial phenomena. In the present model. the interface spurious velocity is orders of magnitude lower than that for existing LBE models of non-ideal fluids. A new numerical scheme for treatment of advection and collision operators is proposed to significantly extend stability limits, in comparison to existing solution methods of the 'master' lattice Bolztmann equation. Implementation of a 'multifractional stepping' procedure for advection operator allows to eliminate severe restriction CFL = 1 in traditionally used 'stream-and-collide' scheme. An implicit trapezoidal discretization of the collision operator is shown to enable excellent performance of the present model in stiff high-surface-tension regime. The proposed numerical scheme is second order accurate. both in time and space.
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10.
  • Sehgal, B. R., et al. (författare)
  • Assessment of reactor vessel integrity (ARVI)
  • 2003
  • Ingår i: Nuclear Engineering and Design. - 0029-5493 .- 1872-759X. ; 221:03-jan, s. 23-53
  • Tidskriftsartikel (refereegranskat)abstract
    • The cost-shared project ARVI (assessment of reactor vessel integrity) involves a total of nine organisations from Europe and USA. The objective of the ARVI Project is to resolve the safety issues that remain unresolved for the melt vessel interaction phase of the in-vessel progression of a severe accident. The work consists of experiments and analysis development. Four tests were performed in the EC-FOREVER Programme, in which failure was achieved in-vessels employing the French pressure vessel steel. The tests were analysed with the commercial code ANSYS-Multiphysics, and the codes SYSTUS+ and PASULA, and quite good agreement was achieved for the failure location. Natural convection experiments in stratified pools have been performed in the SIMECO and the COPO facilities, which showed that much greater heat is transferred downwards for immiscible layers or before layers mix. A model for gap cooling and a set of simplified models for the system codes have been developed. MVITA code calculations have been performed for the Czech and Hungarian VVERs, towards evaluation of the in-vessel melt retention accident management scheme. Tests have been performed at the ULPU facility with organised flow for vessel external cooling. Considerable enhancement of the critical heat flux (CHF) was obtained. The ARVI Project has reached the halfway stage. This paper presents the results obtained thus far from the project.
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11.
  • Sehgal, B. R., et al. (författare)
  • Characterization of heat transfer processes in a melt pool convection and vessel-creep experiment
  • 2002
  • Ingår i: Nuclear Engineering and Design. - 0029-5493 .- 1872-759X. ; 211:2-3, s. 173-187
  • Tidskriftsartikel (refereegranskat)abstract
    • The results of an integral experiment on melt pool convection and vessel-creep deformation are presented and analyzed. The experiment is performed on a test facility, named Failure Of REactor VEssel Retention (FOREVER). The facility employs a 1/10-scaled 15Mo3-(German)-steel vessel of 400-mm diameter, 15-mm wall thickness and 750-mm height. A high-temperature ( similar or equal to1300 degreesC) oxide melt is prepared in a SiC-crucible placed in a 50 kW induction furnace and is. then, poured into the 1/10th scale vessel. A MoSi2 50 kW electric heater is employed in the melt pool to heat and maintain its temperature at 1200 degreesC, The vessel is pressurized with argon at the desired pressure. In the FOREVER/Cl experiment. the vessel wall, maintained at about 900 degreesC and pressurized to 26 bars. was subjected to creep deformation in a 24-h non-stop test. The FOREVER/Cl test is the first integral experiment, in which a decay-heated oxidic naturally-convecting melt pool was maintained in long-term contact with the hemispherical lower head of a pressurized. creeping. steel vessel. A sizeable database was obtained on melt pool temperatures. melt pool energy split, heat transfer rates. heat nux distribution on the melt (crust)-vessel contact surface, vessel temperatures and, in particular the vessel wall creep rate as a function of time. The paper provides information on the FOREVER/Cl measured thermal characteristics and analysis of the observed thermal behavior, The coupled nature of thermal and mechanical processes, as well as the effect of other system conditions (such as depressurization) on the melt pool and vessel temperature responses are analyzed.
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12.
  • Yang, Z. L., et al. (författare)
  • Numerical investigation of boiling regime transition mechanism by a Lattice-Boltzmann model
  • 2001
  • Ingår i: Nuclear Engineering and Design. - 0029-5493 .- 1872-759X. ; 204:1-3, s. 143-153
  • Tidskriftsartikel (refereegranskat)abstract
    • A numerical study has been performed to investigate the hydrodynamic aspects of the pool boilingon horizontal-, vertical- and downward-facing surfaces. The FlowLab code, which is based on a Lattice-Boltzmann (LB) model of two-phase flows, is employed. Macroscopic properties, such as surface tension (a) and contact angle (beta), are implemented through the fluid-fluid (G(sigma)) and fluid-solid (G(t)) interaction potentials. The model is found to express a linear relation between the macroscopic properties (sigma, beta) and microscopic parameters (G(sigma), G(t)). The simulation results on bubble departure diameter appear to have the same parametric dependence as the empirical correlation. Hydrodynamic aspects of two-phase flow regime transition mechanism are investigated for different surface-coolant configurations. Results of the LB simulation clearly demonstrate that not only the bubble nucleation site density (related, e.g. to the heater surface condition and heat fluxes), but also the surface position have a profound effect on the flow regime (pool boiling) characteristics. The results of the LB simulation of hydrodynamics of two-phase flow on the horizontal surface provide the pictures quite similar to the experimental observation for saturated pool boiling. Two mechanisms of flow (boiling) regime transition on the vertical surface are predicted for the local bubble coalescence at bubble generation site and the downstream bubble coalescence. On the downward-facing surfaces, friction between bubbles and the surface wall is found to significantly enlarge the bubble size prior the bubble slip upwards. This behavior is responsible for the earlier bubble coalescence, and therefore, lowers the maximum heat removal rate, in a similar regime of nucleate boiling on a downward-facing surface.
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13.
  • Adamsson, Carl, et al. (författare)
  • A reinterpretation of measurements in developing annular two-phase flow
  • 2011
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 241:11, s. 4562-4567
  • Tidskriftsartikel (refereegranskat)abstract
    • Measurements of developing films in adiabatic high pressure steam-water flow in annular geometry have been reanalyzed and compared to a linearized film-flow model. The development rate of the outer film could be determined with good accuracy in four cases. In one case it was also possible to conclude that the inner film develops faster than the outer one. When compared to the linearized model, these observations show that the deposition rate has to be almost independent of the drop concentration at the investigated conditions. Furthermore, any significant deposition by direct impaction of drops can be excluded as it would couple the development of the two films. These conclusions are quite general and do not depend on the use of any particular correlation for the deposition or entrainment rates. Finally, a rough estimate of the deposition rate was possible, confirming that deposition rates are considerably lower at high pressure steam-water flows than in air-water flows.
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14.
  • Adamsson, Carl, et al. (författare)
  • Film flow measurements for high-pressure diabatic annular flow in tubes with various axial power distributions
  • 2006
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 236:23, s. 2485-2493
  • Tidskriftsartikel (refereegranskat)abstract
    • Measurements of film flow rates in diabatic annular flow in tubes with various axial power distributions were carried out in the high-pressure two-phase flow loop at the Royal Institute of Technology (KTH), Sweden. The measurements were performed at conditions typical for boiling water reactors, i.e. 7 MPa pressure and total mass flux in a range from 750 to 1750 kg/m(2)s. Four different axial power distributions were used and the film mass flow was measured at 7 axial locations for each set of boundary conditions. The results show that the outlet peaked distribution gives less film than the inlet peaked one. This result is consistent with well known trends from measurements of dryout power. The measurements also show that the film flow at the onset of dryout is very small at investigated conditions in agreement with earlier studies. Finally it is shown that the present data is well predicted by two selected phenomenological models of annular flow.
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15.
  • Adamsson, Carl, et al. (författare)
  • Influence of Axial Power Distribution on Dryout : Film-Flow Models and Experiments
  • 2010
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 240:6, s. 1495-1505
  • Tidskriftsartikel (refereegranskat)abstract
    • The influence of axial power distributions on dryout occurrence in nuclear fuel assemblies has been studied extensively for several decades. Even though it is well known that axial power shapes which may significantly vary in nuclear reactors during their operation significantly change the dryout power level, this particular influence is rather difficult to capture with current correlations. In this paper it is shown that this influence can be captured using a phenomenological liquid film model coupled to a standard sub-channel code. The model has been applied to various geometries, including a round pipe, as well as 5 x 5 and 8 x 8 fuel rod assemblies, and highly accurate predictions have been obtained.
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16.
  • Adamsson, Carl, et al. (författare)
  • Modeling and Validation of a Mechanistic Tool (MEFISTO) for the Prediction of Critical Power in BWR Fuel Assemblies
  • 2011
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 241:8, s. 2843-2858
  • Tidskriftsartikel (refereegranskat)abstract
    • Westinghouse is currently developing the MEFISTO code with the main goal to achieve fast, robust, practical and reliable prediction of steady-state dryout Critical Power in Boiling Water Reactor (BWR) fuel bundle based on a mechanistic approach. A computationally efficient simulation scheme was used to achieve this goal, where the code resolves all relevant field (drop, steam and multi-film) mass balance equations, within the annular flow region, at the sub-channel level while relying on a fast and robust two-phase (liquid/steam) sub-channel solution to provide the cross-flow information. The MEFISTO code can hence provide highly detailed solution of the multi-film flow in BWR fuel bundle while enhancing flexibility and reducing the computer time by an order of magnitude as compared to a standard three-field sub-channel analysis approach. Models for the numerical computation of the one-dimensional field flowrate distributions in an open channel (e.g. a sub-channel), including the numerical treatment of field cross-flows, part-length rods, spacers grids and post-dryout conditions are presented in this paper. The MEFISTO code is then applied to dryout prediction in BWR fuel bundle using VIPRE-W as a fast and robust two-phase sub-channel driver code. The dryout power is numerically predicted by iterating on the bundle power so that the minimum film flowrate in the bundle reaches the dryout criteria. Predicted dryout powers (including trends with flow, pressure, inlet subcooling and power distribution) and predicted dryout locations (both axial and radial) are compared to experimental results, using the entire Westinghouse SVEA-96 Optima3 dryout database, and are shown to yield excellent results.
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17.
  • Adoo, N.A., et al. (författare)
  • Determination of thermal hydraulic data of GHARR-1 under reactivity insertion transients using the PARET/ANL code
  • 2011
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 241, s. 5303-5210
  • Tidskriftsartikel (refereegranskat)abstract
    • The PARET/ANL code has been adapted by the IAEA for testing transient behaviour in research reactors since it provides a coupled thermal hydrodynamic and point kinetics capability for estimating thermalhydraulic margins. A two-channel power peaking profile of the Ghana Research Reactor-1 (GHARR-1) has been developed for the PARET/ANL (Version 7.3; 2007) using the Monte Carlo N-Particle code (MCNP) to determine the thermal hydraulic data for reactivity insertion transients in the range of 2.0×10^−3k/k to 5.5×10^−3k/k. Peak clad and coolant temperatures ranged from 59.18 ◦C to 112.36 ◦C and 42.95 ◦C to 79.42 ◦C respectively. Calculated safety margins (DNBR) satisfied the MNSR thermal hydraulic design criteria for which no boiling occurs in the reactor core. The generated thermal hydraulic data demonstrated a high inherent safety feature of GHARR-1 for which the high negative reactivity feedback of the moderator limits power excursion and consequently the escalation of the clad temperature.
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18.
  • Alam, Syed Bahauddin, et al. (författare)
  • Lattice benchmarking of deterministic, Monte Carlo and hybrid Monte Carlo reactor physics codes for the soluble-boron-free SMR cores
  • 2020
  • Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 356
  • Tidskriftsartikel (refereegranskat)abstract
    • Since the use of deterministic transport code WIMS can significantly reduce the computational time compared to the Monte Carlo (MC) code Serpent and hybrid MC code MONK, one of the major objectives of this study is to observe whether deterministic code WIMS can provide accuracy in reactor physics calculations while comparing Serpent and MONK. Therefore, numerical benchmark calculations for a soluble-boron-free (SBF) small modular reactor (SMR) assembly have been performed using the WIMS, Serpent and MONK. Although computationally different in nature, these codes can solve the neutronic transport equations and calculate the required neutronic parameters. A comparison in neutronic parameters between the three codes has been carried out using two types of candidate fuels: 15% U-235 enriched homogeneously mixed all-UO2 fuel and 18% U-235 enriched micro-heterogeneous ThO2-UO2 duplex fuel in a 2D fuel assembly model using a 13x13 arrangement. The eigenvalue/ reactivity (k(infinity)) and 2D assembly pin power distribution at different burnup states in the assembly depletion are compared using three candidate nuclear data files: ENDF/B-VII, JEF2.2 and JEF3.1. A good agreement in k(infinity) values was observed among the codes for both the candidate fuels. The differences in k(infinity) between the codes are similar to 200 pcm when cross-sections based on the same nuclear data file are used. A higher difference (up to similar to 450 pcm) in the k(infinity) values is observed among the codes using cross-sections based on different data files. Finally, it can be concluded from this study that the good agreement in the results between the codes found provides enhanced confidence that modeling of SBF, SMR propulsion core systems with micro-heterogeneous duplex fuel can be performed reliably using deterministic neutronics code WIMS, offering the advantage of less expensive computation than that of the MC Serpent and hybrid MC MONK codes.
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19.
  • Alam, Syed Bahauddin, et al. (författare)
  • Neutronic feasibility of civil marine small modular reactor core using mixed D-2 O+ H2O coolant
  • 2020
  • Ingår i: Nuclear Engineering and Design. - : ELSEVIER SCIENCE SA. - 0029-5493 .- 1872-759X. ; 359
  • Tidskriftsartikel (refereegranskat)abstract
    • In an effort to decarbonize the marine sector, there are growing interests in replacing the contemporary, traditional propulsion systems with nuclear propulsion systems. The latter system allows freight ships to have longer intervals before refueling; subsequently, lower fuel costs, and minimal carbon emissions. Nonetheless, nuclear propulsion systems have remained largely confined to military vessels. It is highly desirable that a civil marine core not to use highly enriched uranium, but it is then a challenge to achieve long core lifetime while maintaining reactivity control and acceptable power distributions in the core. The objective of this study is to design a civil marine core type of single batch small modular reactor (SMR) with low enriched uranium (LEU) (20% U-235 enrichment), a soluble-boron-free (SBF) and using mixed D-2 O+ H2O coolant for operation period over a 20 years life at 333 MWth. Changing the coolant properties is the way to alter the neutron energy spectrum in order to achieve a self-sustaining core design of higher burnup. Two types of LEU fuels were used in this study: micro-heterogeneous ThO2-UO2 duplex fuel (18% U-235 enriched) and all-UO2 fuel (15% U-235 enriched). 2D Assembly designs are developed using WIMS and 3D whole-core model is developed using PANTHER code. The duplex option shows greater promise in the final burnable poison design with high thickness ZrB2 integral fuel burnable absorber (IFBA) while maintaining low, stable reactivity with minimal burnup penalty. For the final poison design with ZrB2, the duplex contributes (similar to)2.5% more initial reactivity suppression, although the all-UO2 design exhibits lower reactivity swing. Three types of candidate control rod materials: hafnium, boron carbide (B4C) and 80% silver + 15% indium + 5% cadmium (Ag-In-Cd) are examined and duplex fuel exhibits higher control rod worth with the candidate materials. B4C shows the greatest control reactivity worth for both the candidate fuels, providing (similar to)3% higher control rod worth for duplex fuel than all-UO2. Finally, 3D whole-core results from PANTHER show that the use of the mixed coolant contributes to (similar to)21.5 years core life, which is a (similar to)40% increase in core life compared to H2O coolant ((similar to)15.5 years) while using the same fuel candidates and fissile enrichment. The mixed coolant provides excellent core lifetimes comparable to those of HEU military naval vessels ((similar to)25 years vs. (similar to)21.5 years) while utilizing LEU candidate fuels.
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20.
  • Alam, Syed Bahauddin, et al. (författare)
  • Small modular reactor core design for civil marine propulsion using micro-heterogeneous duplex fuel. Part I: Assembly-level analysis
  • 2019
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 346, s. 157-175
  • Tidskriftsartikel (refereegranskat)abstract
    • In an effort to de-carbonise commercial freight shipping, there is growing interest in the possibility of using nuclear propulsion systems. In this reactor physics study, we seek to design a soluble-boron-free (SBF) and low-enriched uranium (LEU) (<20% U-235 enrichment) civil nuclear marine propulsion small modular reactor (SMR) core that provides at least 15 effective full-power-years (EFPY) life at 333 MWth using 18% U-235 enriched micro-heterogeneous ThO2-UO2 duplex fuel and 15% U-235 enriched homogeneously mixed all-UO2 fuel. We use WIMS to develop subassembly designs and PANTHER to examine whole-core arrangements.The assembly-level behaviours of candidate burnable poison (BP) materials and control rods are investigated. We examine gadolinia (Gd2O3), erbia (Er2O3) and ZrB2 integral fuel burnable absorber (IFBA) as BPs. We arrive at a design with the candidate fuels loaded into 13 x 13 assemblies using IFBA pins for reactivity control. Taking advantage of self-shielding effects, this design maintains low and stable assembly reactivity with relatively little burnup penalty. Thorium-based duplex fuel offers better performance than all-UO2 fuel with all BP options considered. Duplex fuel has similar to 20% lower reactivity swing and, in consequence, lower initial reactivity than all-UO2 fuel. The lower initial reactivity and smaller reactivity swing make the task of reactivity control through BP design easier in the thorium-rich duplex core. For control rod design, we examine boron carbide (B4C), hafnium, and Ag-In-Cd alloy. All the candidate materials exhibit greater rod worth for the duplex design. For both fuels, B4C has the highest rod worth. In particular, one of the major objectives of this study is to offer/explore a thorium-based candidate alternative fuel platform for the proposed marine core. It is proven by literature reviews that the ability of the duplex fuel was never explored in the context of a single-batch, LEU, SBF, long-life SMR core. In this regard, the motivation of this paper is to observe the neutronic performance of the proposed duplex fuel with respect to the UO2 fuel and 'open the option' of designing the functional cores with both the duplex and UO2 fuel cores.
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21.
  • Alam, Syed Bahauddin, et al. (författare)
  • Small modular reactor core design for civil marine propulsion using micro-heterogeneous duplex fuel. Part II: whole-core analysis
  • 2019
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 346, s. 176-191
  • Tidskriftsartikel (refereegranskat)abstract
    • Civil marine reactors face a unique set of design challenges. These include requirements for a small core size and long core lifetime, a 20% cap on fissile loading, and limitations on using soluble neutron absorbers. In this reactor physics study, we seek to design a core that meets these requirements over a 15 effective full-power-years (EFPY) life at 333 MWth using homogeneously mixed all-UO2 and micro-heterogeneous ThO2-UO2 duplex fuels. In a companion (Part I) paper, we found assembly designs using 15% and 18% U-235 for UO2 and duplex fuels, respectively, loaded into 13 x 13 pin arrays. High thickness (150 mu m) ZrB2 integral fuel burnable absorber (IFBA) pins and boron carbide (B4C) control rods are used for reactivity control. Taking advantage of self-shielding effects, these designs maintain low and stable assembly reactivity with little burnup penalty.In this paper (Part II), whole-core design analyses are performed for small modular reactor (SMR) to determine whether the core remains critical for at least 15 EFPY with a reactivity swing of less than 4000 pcm, subject to appropriate constraints. The main challenge is to keep the radial form factor below its limit (1.50). Burnable poison radial-zoning is examined in the quest for a suitable arrangement to control power peaking. Optimized assemblies are loaded into a 3D reactor model in PANTHER. The PANTHER results confirm that the fissile loadings of both fuels are well-designed for the target lifetime: at the end of the (similar to)15-year cycle, the cores are on the border of criticality. The duplex fuel core can achieve (similar to)4% longer core life, has a (similar to)3% lower initial reactivity and (similar to)30% lower reactivity swing over life than the final UO2 core design. The duplex core is therefore the more successful design, giving a core life of (similar to)16 years and a reactivity swing of less than 2500 pcm, while satisfying all the neutronic safety parameters. In particular, one of the major objectives of this study is to offer/explore a thorium-based candidate alternative fuel platform for the proposed marine core. It is proven by literature reviews that the ability of the duplex fuel was never explored in the context of a single-batch, LEU, SBF, long-life SMR core. In this regard, the motivation of this paper is to understand the underlying physics of the duplex fuel and 'open the option' of designing the functional cores with both the duplex and UO2 fuel cores.
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22.
  • Almyashev, V. I., et al. (författare)
  • Oxidation effects during corium melt in-vessel retention
  • 2016
  • Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 305, s. 389-399
  • Tidskriftsartikel (refereegranskat)abstract
    • In the in-vessel corium retention studies conducted on the Rasplav-3 test facility within the ISTC METCOR-P project and OECD MASCA program, experiments were made to investigate transient processes taking place during the oxidation of prototypic molten corium. Qualitative and quantitative data have been produced on the sensitivity of melt oxidation rate to the type of oxidant, melt composition, molten pool surface characteristics. The oxidation rate is a governing factor for additional heat generation and hydrogen release; also for the time of secondary inversion of oxidic and metallic layers of corium molten pool.
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23.
  • Alsmeyer, H, et al. (författare)
  • Ex-vessel core melt stabilization research (ECOSTAR)
  • 2005
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 235:2-4, s. 271-284
  • Tidskriftsartikel (refereegranskat)abstract
    • The project on ex-vessel core melt stabilization research (ECOSTAR) started in January 2000 to be concluded by end of 2003. The project is performed by 14 partner institutions from five European countries and involves a large number of experiments with low- and high-temperature simulant melts and real corium at different scales. Model development and scaling analysis allows application of the research results to existing and to future LWRs in the area of reactor design and accident mitigation. The project is oriented toward the analysis and mitigation of severe accident sequences that could occur in the ex-vessel phase of a postulated core melt accident. The issues are: (1) the release of melt form the pressure vessel, (2) the transfer and spreading of the melt on the basement, (3) the analysis of the physical-chemical processes that are important for corium behavior especially during concrete erosion with onset of solidification, and (4) stabilization of the melt by cooling through direct water contact. The results achieved so far resolve a number of important issues: the amount of melt that could be transferred at RPV failure from the RPV into the containment can be substantially reduced by lowering the residual pressure in the primary circuit. It is found that melt dispersion also strongly depends on the location of the RPV failure, and that lateral failure results in substantially less melt dispersion. During melt release, the impinging melt jet could erode parts of the upper basement surface. Jet experiments and a derived heat transfer relation allow estimation of its contribution to concrete erosion. Spreading of the corium melt on the available basement surface is an important process, which defines the initial conditions for concrete attack or for the efficiency of cooling in case of water contact, respectively. Validation of the spreading codes based on a large-scale benchmark experiment is underway and will allow determination of the initial conditions, for which a corium melt can be assumed to spread homogeneously over the available surface. Experiments with UO(2)-based corium melts highlight the role of phase segregation during onset of melt solidification and during concrete erosion. To cool the spread corium melt, the efficacy of top flooding and bottom flooding is investigated in small-scale and in large-scale experiments, supported by model developments. Project assessment is continuing to apply the results to present and future reactors.
  •  
24.
  • Anderson, Patrick (författare)
  • Analytic study of the steel liner near the equipment hatch in a 1 : 4 scale containment model
  • 2008
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 1872-759X .- 0029-5493. ; 238:7, s. 1641-1650
  • Tidskriftsartikel (refereegranskat)abstract
    • A containment scale-model test, performed at Sandia National Laboratories, was loaded by overpressurization and the first leak was concluded to be caused by tears in the steel liner found near the equipment hatch. These tears were located in the vicinity of the vertical fold in between the general curved part and the embossment (vertical bend line). A 3D finite element analysis of the region near the equipment hatch, shows that high localized strains will develop in the vicinity of the bend line. It is shown that the liner separates from the concrete wall near the bend line when the containment expands. The tensioned liner will be in contact with the surface of the concrete wall in general, but near the vertical bend line the liner tends to be straightened out. This flexural behaviour cause high strains in the weld located in the bend line. The actual peak strain level is depending on the detailed geometry in the bend line and the failure strain level of a welded biaxial stressed zone is difficult to define. However, the analysis presented in this paper shows that the flexural behaviour in the bend line most likely contributed to the liner tears found in the scale-model test. A general conclusion from the study presented in this paper is that, the non-linear plastic behaviour of the liner is very sensitive to the detailed design and the interaction between the liner and the concrete.
  •  
25.
  • Anderson, Patrick, et al. (författare)
  • Average force along unbonded tendons: a field study at nuclear reactor containments in Sweden
  • 2005
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 1872-759X .- 0029-5493. ; 235:1, s. 91-100
  • Tidskriftsartikel (refereegranskat)abstract
    • The tightness and integrity of all Swedish reactor containments depend directly and indirectly on the function of the post-tensioned system. The tendon force in containments with unbonded tendons is followed up at regular in-service inspections (ISI) to ensure that the remaining force is sufficient. At the inspections, the tendon force is measured with so-called lift-off technique, where a jack is used to lift the end anchor. The interpretation of the measuring results is not obvious. One difficulty, which affects all tendons to different extents, is the influence of friction between the tendon and duct. This influence can cause a redistribution of force along the tendons after the original tensioning. The ordinary lift-off method only measures the force at the end of the tendon, which not always represents the change of force in the rest of the tendon. A method for measuring average force along a tendon is presented in this article. In this method, the elongation of the tendon is measured when the jack force is increased to a reference level with known force distribution. Measuring results from the latest ISI in Sweden show that the end force has decreased more than the average force, especially for long tendons with high influence of friction.
  •  
26.
  • Anderson, Patrick (författare)
  • Thirty years of measured prestress at Swedish nuclear reactor containments
  • 2005
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 1872-759X .- 0029-5493. ; 235:21, s. 2323-2336
  • Tidskriftsartikel (refereegranskat)abstract
    • The main function of the reactor containment, i.e. to ensure tightness at a major internal accident, depends directly on the prestressing system. To secure that the prestress level is sufficient, the tendon force has been measured during the whole time of operation. The general results from these measurements show that the loss of prestress 30 years after tensioning is between 5 and 10%. This is much lower loss than predicted initially at the design stage. More advanced and today commonly used models for predicting prestress loss show better agreement with the results. The main reasons for the relatively low loss are assumed to be: (1) the confirmed slow drying process of the concrete and (2) the high concrete age at the initial tensioning. The results also indicate that the temperature has a major influence on the loss of prestress.
  •  
27.
  •  
28.
  • Anghel, Ionut, 1971-, et al. (författare)
  • Experimental investigation of post-dryout heat transfer in annuli with flow obstacles
  • 2012
  • Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 246, s. 82-90
  • Tidskriftsartikel (refereegranskat)abstract
    • An experimental study on post-dryout heat transfer was conducted in the High-pressure WAter Test (HWAT) loop at the Royal Institute of Technology in Stockholm, Sweden. The objective of the experiments was to investigate the influence of flow obstacles on the post-dryout heat transfer. The investigated operational conditions include mass flux equal to 500 kg/m2 s, inlet sub-cooling 10 K and system pressure 5 and 7 MPa. The experiments were performed in annuli in which the central rod was supported with five pin spacers. Two additional types of flow obstacles were placed in the exit part of the test section: a cylinder supported on the central rod only and a typical BWR grid spacer cell. The measurements indicate that flow obstacles improve heat transfer in the boiling channel. It has been observed that the dryout power is higher when additional obstacles are present. In addition the wall temperature in post-dryout heat transfer regime is reduced due to increased turbulence and drop deposition. The present data can be used for validation of computational models of post-dryout heat transfer in channels with flow obstacles.
  •  
29.
  • Anghel, Ionut, et al. (författare)
  • On post-dryout heat transfer in channels with flow obstacles
  • 2014
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 270, s. 351-358
  • Tidskriftsartikel (refereegranskat)abstract
    • This paper describes a new approach to predict post-dryout heat transfer in channels with flow obstacles. Using experimental data obtained in annular test sections at prototypical BWR conditions, the existing Saha correlation for post-dryout heat transfer has been modified to account for the presence of obstacles. The obstacle effect is taken into account in the following way: (a) the critical quality downstream of an obstacle is found; (b) an exponential function of equilibrium quality is applied to describe the decrease of heat transfer coefficient in the developing post-dryout region; (c) an additional heat transfer enhancement term is applied downstream of the obstacle. The new approach is applied to annular test sections with various flow obstacles and a significant improvement of accuracy of wall temperature prediction, as compared to reference methods, is shown.
  •  
30.
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31.
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32.
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33.
  • Anglart, Henryk, 1954- (författare)
  • Progress in understanding and modelling of annular two-phase flows with heat transfer
  • 2019
  • Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 345, s. 166-182
  • Tidskriftsartikel (refereegranskat)abstract
    • Annular two-phase flows with heat transfer play important role in many industrial applications. In particular, thermal margins of Boiling Water Reactors (BWR) are entirely determined by this type of flow and heat transfer conditions. To avoid dryout, a liquid film must be present on heated rods of BWR fuel assemblies during normal operation. The present paper describes the recent progress in understanding and modelling of the governing phenomena of annular two-phase flow and heat transfer. A special attention has been devoted to experimental observations that have the most significant influence on the adopted modelling approach. The primary goal is to pave a path to mechanistic modelling of dryout and post-dryout heat transfer applicable to nuclear fuel assemblies. Current Computational Fluid Dynamics (CFD) approaches to model the governing phenomena are presented and their further improvements are suggested.
  •  
34.
  • Anglart, Henryk, et al. (författare)
  • Study of spray cooling of a pressure vessel head of a boiling water reactor
  • 2010
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 240:2, s. 252-257
  • Tidskriftsartikel (refereegranskat)abstract
    • The present paper deals with a theoretical analysis of the spray cooling of a Reactor Pressure Vessel (RPV) head in a Boiling Water Reactor (BWR). To this end a detailed computational model has been developed. The model predicts the trajectories, diameters and temperatures of subcooled droplets moving in saturated vapor. The model has been validated through comparison with experimental data, in which droplet temperatures were measured as functions of the distance that they cover in saturated vapor from the moment they leave the sprinkler outlet to the moment they impact on the RPV head inner wall. The calculations are in very good agreement with measurements, confirming the model adequacy for the present study. The model has been used for a parametric study to investigate the influence of several parameters on the cooling efficiency of the spray system. Based on the study it has been shown that one of the main parameters that govern the temperature increase in a subcooled droplet is its initial diameter. Comparisons are also made between conclusions from the theoretical model and observations made through flow and temperature measurements in the plant (Forsmark 1 and 2). One of these observations is that the rate at which the RPV head temperature decreases on the way down from hot to cold standby is constant and independent of the sprinkling flow rate as long as the flow rate is above a certain minimum value. Accordingly, the theoretical model shows that if one assumes that the cooling of the RPV head is through a water film built on the inner wall due to sprinkling, the heat removal rate is only very weakly dependent on the sprinkling flow rate.
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35.
  • Audouin, L., et al. (författare)
  • Quantifying differences between computational results and measurements in the case of a large-scale well-confined fire scenario
  • 2011
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 1872-759X .- 0029-5493. ; 241:1, s. 18-31
  • Tidskriftsartikel (refereegranskat)abstract
    • The objective of this work was to quantify comparisons between several computational results and measurements performed during a pool fire scenario in a well-confined compartment. This collaborative work was initiated under the framework of the OECD fire research program and involves the most frequently used fire models in the fire community, including field and zone models. The experimental scenario was conducted at the French Institut de Radioprotection et de Surete Nucleaire (IRSN) and deals with a full-scale liquid pool fire in a confined and mechanically ventilated compartment representative for nuclear plants. The practical use of different metric operators and their ability to report the capabilities of fire models are presented. The quantitative comparisons between measurements and numerical results obtained from "open" calculations concern six important quantities from a safety viewpoint: gas temperature, oxygen concentration, wall temperature, total heat flux, compartment pressure and ventilation flow rate during the whole fire duration. The results indicate that it is important to use more than one metric for the validation process in order to get information on the uncertainties associated with different aspects of fire safety. (C) 2010 Elsevier B.V. All rights reserved.
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36.
  • Bandaru, S V Ravikumar, et al. (författare)
  • Multi-nozzle spray cooling of a reactor pressure vessel steel plate for the application of ex-vessel cooling
  • 2021
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 375
  • Tidskriftsartikel (refereegranskat)abstract
    • Spray cooling is a versatile technology for various cooling applications involving high surface heat fluxes. Experimental facility was built to study heat transfer performance of an upward multi-nozzle array of water sprays impacting a surface of heated plate made of reactor vessel grade steel. The effect of inclination angles of the steel surface on the cooling performance was investigated to assess heat transfer in complex semispherical/ semielliptical geometry of large reactor lower head and to address possible application of spray cooling in severe accident management (SAM) of light water reactors (LWRs) based on In-vessel melt retention with external reactor vessel cooling (IVR-ERVC). Joule heating of SA302B steel foil of 0.15 mm thickness and surface area of 96 cm2 enabled prototypic heat fluxes to be evacuated during reactor accident. A 2×3 array of full jet narrow-coned pressure-swirl spray nozzles was used to reproduce multi-nozzle cooling. The tests were conducted as a series of consequent steady states realized at stepwise increasing power and surface heat fluxes up to the maximum values of 29 kW and 2.97 MW/m2 limited in the specific facility design. Seven surface inclinations, between 0o and 90o were tested and no significant variations in spray cooling performance with the inclination of the heated surface was found. The results indicated a promising prospect of using a multi-nozzle array system for cooling of large surface area of reactor lower head. Much higher heat fluxes can be safely extracted by spray cooling in comparison with the critical heat fluxes that appeared at RPV water pool cooling at natural convection. The maximum value of heat flux at direct spray impact zones and its drop-off slightly from the center to the periphery of the spray cone was detected in the tests. The water flow rate and liquid subcooling significantly influenced maximum steel surface temperature but had no noticeable effects on surface temperature uniformity. The spray-to-spray interaction had no observable effects on local surface temperatures, however, the colliding zones where four spray cones have visible effects on local surface temperatures due to poor liquid momentum. The results also showed that more uniform liquid film distribution could be obtained for some inclinations because of improved liquid drainage, which in turn leads to maintaining low surface temperatures. 
  •  
37.
  • Bandini, G., et al. (författare)
  • Assessment of systems codes and their coupling with CFD codes in thermal-hydraulic applications to innovative reactors
  • 2015
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 281, s. 22-38
  • Tidskriftsartikel (refereegranskat)abstract
    • The THINS project of the 7th Framework EU Program on nuclear fission safety is devoted to the investigation of crosscutting thermal hydraulic issues for innovative nuclear systems. A significant effort in the project has been dedicated to the qualification and validation of system codes currently employed in thermal hydraulic transient analysis for nuclear reactors. This assessment is based either on already available experimental data, or on the data provided by test campaigns carried out in the frame of THINS project activities. Data provided by TALL and CIRCE facilities were used in the assessment of system codes for HLM reactors, while the PHENIX ultimate natural circulation test was used as reference for a benchmark exercise among system codes for sodium-cooled reactor applications. In addition, a promising grid-free pool model based on proper orthogonal decomposition is proposed to overcome the limits shown by the thermal hydraulic system codes in the simulation of pool-type systems. Furthermore, multi-scale system-CFD solutions have been developed and validated for innovative nuclear system applications. For this purpose, data from the PHENIX experiments have been used, and data are provided by the tests conducted with new configuration of the TALL-3D facility, which accommodates a 3D test section within the primary circuit. The TALL-3D measurements are currently used for the validation of the coupling between system and CFD codes.
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38.
  • Basso, Simone, et al. (författare)
  • Empirical closures for particulate debris bed spreading induced by gas-liquid flow
  • 2016
  • Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 297, s. 19-25
  • Tidskriftsartikel (refereegranskat)abstract
    • Efficient removal of decay heat from the nuclear reactor core debris is paramount for termination of severe accident progression. One of the strategies is based on melt fragmentation, quenching and cooling in a deep pool of water under the reactor vessel. Geometrical configuration of the debris bed is among the important factors which determine possibility of removing the decay heat from the debris bed by natural circulation of the coolant. For instance, a tall mound-shape debris bed can be non-coolable, while the same debris can be coolable if spread uniformly. Decay heat generates a significant amount of thermal energy which goes to production of steam inside the debris bed. Two-phase flow escaping through the top layer of the bed becomes a source of mechanical energy which can move the particulate debris along the slope of the bed. The motion of the debris will lead to flattening of the bed. Such process is often called "self-leveling" phenomenon. Spreading of the debris bed by the self-leveling process can take significant time, depending on the initial debris bed configuration and other parameters. There is a competition between the time scales for reaching (i) a coolable configuration of the bed, and (ii) onset of dryout and re-melting of the debris. In the previous work we have demonstrated that the rate of particulate debris spreading is determined by local gas velocity and local slope angle of the bed. In this work we develop a scaling approach and a closure for prediction of debris spreading rate based on generalization of available experimental data. We demonstrate that introduced scaling criteria are universal for particles of different shapes and size distributions.
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39.
  • Basso, Simone, et al. (författare)
  • Preliminary Risk assessment of ex-vessle debris bed coolability for a Nordic BWR
  • Ingår i: Nuclear Engineering and Design. - 0029-5493 .- 1872-759X.
  • Tidskriftsartikel (refereegranskat)abstract
    • In Nordic design of boiling water reactors (BWRs) a deep water pool under the reactor vessel is employed as a severe accident management strategy for the core melt fragmentation and the long term cooling of corium debris. The height and shape of the debris bed are among the most important factors that determine if decay heat can be removed from the porous debris bed by natural circulation of water. The debris bed geometry is formed as a result of melt release, fragmentation, sedimentation and settlement on the containment basemat. After settlement, the shape can change with time due to movement of particles promoted by the coolant flow (debris bed self-leveling process). Both aleatory (accident scenario, stochastic) and epistemic (modeling, lack of knowledge) uncertainties are important for assessing the risks. The present work describes a preliminary risk analysis of debris bed coolability for Nordic BWRs under severe accident conditions. It was assumed that once debris remelting starts containment failure becomes imminent. Such assumption allows to estimate the containment failure probability by calculating the probability that the time necessary for the spreading debris bed to achieve a coolable configuration will be shorter than the onset time of debris bed re-melting. An artificial neural network was employed as a surrogate model (SM) for the mechanistic full model (FM) of the debris spreading in order to achieve computationally efficient propagation of uncertainties. The effect of uncertainty in the ranges and probability density functions (PDFs) of the input parameters was addressed. Parameters defining shapes of the PDFs were varied for three different distribution families (beta, truncated normal and triangular). The results of the risk analysis were reported as complementary cumulative distribution functions (CCDFs) of the conditional containment failure probability (CCFP). It is demonstrated that CCFP can vary in wide ranges depending on the randomly selected combinations of the PDFs of the input parameters. Given the selected ranges of the input parameters, sensitivity analyses identified: the effective particle diameter and the debris bed porosity as the largest contributors to the CCFP uncertainty. It was shown that the self-leveling phenomenon reduces sensitivity of debris coolability to the initial shape of the bed. However, the initial shape remains an important uncertainty factor for the most likely values of the particle size and porosity. Importance of the initial shape increases when the effectiveness of the self-leveling is small (e.g. in case of high initial temperature or heat up rate of the debris). Findings of this work in combination with consideration of the necessary efforts can be used for prioritization of the future research on obtaining new information on the uncertain parameters.
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40.
  • Basso, Simone, et al. (författare)
  • The effect of self-leveling on debris bed coolability under severe accident conditions
  • 2016
  • Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 305, s. 246-259
  • Tidskriftsartikel (refereegranskat)abstract
    • Nordic-type boiling water reactors employ melt fragmentation, quenching, and long term cooling of the debris bed in a deep pool of water under the reactor vessel as a severe accident (SA) mitigation strategy. The height and shape of the bed are among the most important factors that determine if decay heat can be removed from the porous debris bed by natural circulation of water. The debris bed geometry depends on its formation process (melt release, fragmentation, sedimentation and settlement on the containment basemat), but it also changes with time afterwards, due to particle redistribution promoted by coolant flow (self-leveling). The ultimate goal of this work is to develop an approach to the assessment of the probability that debris in such a variable-shape bed can reach re-melting (which means failure of SA mitigation strategy), i.e. the time necessary for the slumping debris bed to reach a coolable configuration is larger than the time necessary for the debris to reach the re-melting temperature. For this purpose, previously developed models for particulate debris spreading by self-leveling and debris bed dryout are combined to assess the time necessary to reach a coolable state and evaluate its uncertainty. Sensitivity analysis was performed to screen out less important input parameters, after which Monte Carlo simulation was carried out in order to collect statistical characteristics of the coolability time. The obtained results suggest that, given the parameters ranges typical of Nordic BWR5, only a small fraction of debris beds configurations exhibits the occurrence of dryout. Of the initially non-coolable configurations, a significant portion becomes coolable due to debris bed self-leveling.
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41.
  • Bechta, Sevostian, et al. (författare)
  • Corium phase equilibria based on MASCA, METCOR and CORPHAD results
  • 2008
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 238:10, s. 2761-2771
  • Tidskriftsartikel (refereegranskat)abstract
    • Experimental data on component partitioning between suboxidized corium melt and steel in the invessel melt retention (IVR) conditions are compared. The data are produced within the OECD MASCAprogram and the ISTC CORPHAD project under close-to-isothermal conditions and in the ISTC METCORproject under thermal gradient conditions. Chemical equilibrium in the U–Zr–Fe(Cr,Ni,. . .)–O system isreached in all experiments. In MASCA tests the molten pool formed under inert atmosphere has twoimmiscible liquids, oxygen-enriched (oxidic) and oxygen-depleted (metallic), resulting of the miscibilitygap of the mentioned system. Sub-system data of the U–Zr–Fe(Cr,Ni,. . .)–O phase diagram investigatedwithin the ISTC CORPHAD project are interpreted in relation with the MASCA results. In METCOR teststhe equilibrium is established between oxidic liquid and mushy metallic part of the system. Results ofcomparison are discussed and the implications for IVR noted.
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42.
  • Bechta, Sevostian, et al. (författare)
  • Corrosion of vessel steel during its interaction with molten corium : Part 2. Model development
  • 2006
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 236:13, s. 1362-1370
  • Tidskriftsartikel (refereegranskat)abstract
    • An experimental examination of the cooled vessel steel corrosion during the interaction with molten corium is presented. The experiments havebeen conducted on “Rasplav-2” test facility and followed up with physico-chemical and metallographic analyses of melt samples and coriumspecimeningots. The results discussed in the first part of the paper have revealed specific corrosion mechanisms for air and inert atmosphere abovethe melt. Models have been proposed based on this information and approximate curves constructed for the estimation of the corrosion rate orcorrosion depth of vessel steel in conditions simulated by the experiments.
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43.
  • Bechta, Sevostian, et al. (författare)
  • Corrosion of vessel steel during its interaction with molten corium : Part 1. Experimental
  • 2006
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 236:17, s. 1810-1829
  • Tidskriftsartikel (refereegranskat)abstract
    • This paper is concerned with corrosion of a cooled vessel steel structure interacting with molten corium in air and neutral (nitrogen) atmospheresduring an in-vessel retention scenario. The data on corrosion kinetics at different temperatures on the heated steel surface, heat flux densities andoxygen potential in the system are presented. The post-test physico-chemical and metallographic analyses of melt samples and the corium–specimeningot have clarified certain mechanisms of steel corrosion taking place during the in-vessel melt interaction.
  •  
44.
  • Bechta, Sevostian, et al. (författare)
  • Influence of corium oxidation on fission product release from molten pool
  • 2010
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 240:5, s. 1229-1241
  • Tidskriftsartikel (refereegranskat)abstract
    • Qualitative and quantitative determination of the release of low-volatile fission products and core materialsfrom molten oxidic corium was investigated in the EVAN project under the auspices of ISTC. Theexperiments 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 oxidationkinetics, critical influence of melt surface temperature and oxidation index on the fission productrelease rate, aerosol particle composition and size distribution. The relevance of measured high releaseof Sr from the molten pool for the reactor application is highlighted. Comparisons of the experimentaldata with those from the COLIMA CA-U3 test and the VERCORS tests, as well as with predictions fromIVTANTHERMO and GEMINI/NUCLEA codes are made. Recommendations for further investigations areproposed following the major observations and discussions.
  •  
45.
  • Bechta, Sevostian, et al. (författare)
  • VVER vessel steel corrosion at interaction with molten corium in oxidizing atmosphere
  • 2009
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 239:6, s. 1103-1112
  • Tidskriftsartikel (refereegranskat)abstract
    • The long-term in-vessel corium retention (IVR) in the lower head bears a risk of the vessel wall deterioration caused by steel corrosion. The ISTC METCOR Project has studied physicochemical impact of prototypic coria having different compositions in air and steam and has generated valuable experimental data on vessel steel corrosion. It is found that the corrosion rate is sensitive to corium composition, but the composition of oxidizing above-melt atmosphere (air, steam) has practically no influence on it. A model of the corrosion process that integrates the experimental data, is proposed and used for development of correlations.
  •  
46.
  • Bergagio, Mattia, et al. (författare)
  • Experimental investigation of mixing of non-isothermal water streams at BWR operating conditions
  • 2017
  • Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 317, s. 158-176
  • Tidskriftsartikel (refereegranskat)abstract
    • In this experimental investigation, wall surface temperatures have been measured during mixing of three water streams in the annular gap between two coaxial stainless-steel tubes. The inner tube, with an outer diameter of 35 mm and a thickness of 5 mm, holds six K-type, ungrounded thermocouples with a diameter of 0.5 mm, which measured surface temperatures with a sampling rate of either 100 Hz or 1000 Hz. The tube was rotated from 0 to 360° and moved in a range of 387 mm in the axial direction to allow measurements of surface temperatures in the whole mixing region. The outer tube has an inner diameter of 80 mm and a thickness of 10 mm to withstand a water pressure of 9 MPa. A water stream at a temperature of either 333 K or 423 K and a Reynolds number between 1657 and 8410 rose vertically in the annular gap and mixed with two water streams at a temperature of 549 K and a Reynolds number between 3.56E5 and 7.11E5. These two water streams entered the annulus radially on the same axial level, 180° apart. Water pressure was kept at 7.2 MPa. Temperature recordings were performed at five axial and eight azimuthal locations, for each set of boundary conditions. Each recording lasted 120 s to provide reliable data on the variance, intermittency and frequency of the surface temperature time series at hand. Thorough calculations indicate that the uncertainty in the measured temperature is of 1.58 K. The mixing region extends up to 0.2 m downward of the hot inlets. In most cases, measurements indicate non-uniform mixing in the azimuthal direction, because of asymmetries in either geometry or mass flow rates at the hot inlets. Due to the measurement accuracy and a relatively simple geometry, an experimental database has been obtained for validation of computational methods to predict thermal mixing and fatigue. Furthermore, these data can provide new insight into turbulent mixing at BWR operating conditions and, more generally, into mixing coupled to the dynamics, also termed level-2 mixing.
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47.
  • Bergagio, Mattia, et al. (författare)
  • Large eddy simulation of thermal mixing with conjugate heat transfer at BWR operating conditions
  • 2020
  • Ingår i: Nuclear Engineering and Design. - : ELSEVIER SCIENCE SA. - 0029-5493 .- 1872-759X. ; 356
  • Tidskriftsartikel (refereegranskat)abstract
    • Thermal fatigue occurs in most metals under cyclic heat loads and can threaten the structural integrity of metal parts. Detailed knowledge of these loads is of utter importance to prevent such issues. In this study, a large eddy simulation (LES) with wall-adapting local eddy viscosity (WALE) subgrid model is performed to better understand turbulent thermal mixing in an annulus with a pair of opposing cold inlets at a low axial level (z = 0.15 m) and with a pair of opposing hot inlets at a higher axial level (z = 0.80 m). Each inlet pair is 90 degrees from each other in the azimuthal direction. Conjugate heat transfer between fluid and structure is accounted for. The geometry simplifies a control-rod guide tube (CRGT) in a boiling water reactor (BWR). LES results are compared with measurement data. This is one of the first times BWR conditions are met in both experiments and LES: pressure equals 7.2 MPa, while the temperature difference between hot and cold inlets reaches 216 K. LES temperatures at the fluid-structure interface are fairly correlated with their experimental equivalents, with regard to mean values, local variances, and dangerous oscillation modes in fatigue-prone areas (z = 0.65 - 0.67 m). An elastic analysis of the structure is performed to evaluate stress intensities there. From them, cumulative fatigue usage factors (CUFs) are estimated and used as screening criteria in the subsequent frequency analysis of temperature time series at the fluid-structure interface. The likelihood of initiating a fatigue crack is linked to the maximum CUF, which is 3.2 x 10(-5) for a simulation time of similar to 10 s.
  •  
48.
  • Bian, Boshen, et al. (författare)
  • Direct numerical simulation of molten pool convection in a 3D semicircular slice at different Prandtl numbers
  • 2022
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 393, s. 111772-
  • Tidskriftsartikel (refereegranskat)abstract
    • In this paper, a Direct Numerical Simulation (DNS) of an internally heated (IH) natural convection in a 3D semicircular slice molten pool is conducted using Nek5000, a CFD solver with spatial discretization based on the spectral element method. The mesh requirements in the bulk and boundary layers are both fulfilled using known correlations. A calculation of a simplified internally heated box is first established with an excellent agreement to existing data. Next, simulation of the 3D semi-circular is performed showing qualitative agreement with the general flow observations from the BALI experiments. The velocity field shows that the flow domain is divided into three regions, i.e., intensive turbulent eddies in the upper domain, weak flow motion in the lower domain, and the descending flow along the curved boundary. Correspondingly, the temperature field in the upper domain is relatively homogenous, while that in the lower domain is characterized by stratified layers. Further, the heat flux distribution along the boundaries shows that the heat fluxes fluctuate along the top wall due to turbulent eddies, and the heat fluxes at the curved wall increase nonlinearly from the bottom to the top. Finally, the influence of Prandtl number indicates that smaller Prandtl number will lead to more turbulence eddies, deeper descending flow, and more even redistribution of heat thereby lowering the maximum heat flux to the curved walls.
  •  
49.
  • Bottomley, D., et al. (författare)
  • Severe accident research in the core degradation area : An example of effective international cooperation between the European Union (EU) and the Commonwealth of Independent States (CIS) by the International Science and Technology Center
  • 2012
  • Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 252, s. 226-241
  • Tidskriftsartikel (refereegranskat)abstract
    • The International Science and Technology Center (ISTC) was set up in Moscow to support non-proliferation of sensitive knowledge and technologies in biological, chemical and nuclear domains by engaging scientists in peaceful research programmes with a broad international cooperation. The paper has two following objectives: to describe the organization of complex, international, experimental and analytical research of material processes under extreme conditions similar to those of severe accidents in nuclear reactors and, to inform briefly about some results of these studies. The main forms of ISTC activity are Research Projects and Supporting Programs. In the Research Projects informal contact expert groups (CEGs) were set up by ISTC to improve coordination between adjacent projects and to encourage international collaboration. The European Commission was the first to use this. The CEG members - experts from the national institutes and industry - evaluated and managed the projects' scientific results from initial stage of proposal formulation until the final reporting. They were often involved directly in the project's details by joining the Steering Committees of the project. The Contact Expert Group for Severe Accidents and Management (CEG-SAM) is one of these groups, five project groups from this area from the total of 30 funded projects during 10 years of activity are detailed to demonstrate this: (1) QUENCH-VVER from RIAR, Dimitrovgrad and IBRAE, Moscow, and PARAMETER projects (SF1-SF4) from LUCH, Podolsk and IBRAE, Moscow; these concerned a detailed study of bundle quenching from high temperature; (2) Reactor Core Degradation; a modelling project simulating the fuel rod degradation and loss of geometry from IBRAE, Moscow; (3) METCOR projects from NITI, St. Petersburg on the interaction of core melt with reactor vessel steel; (4) INVECOR project, NNE Kurchatov City, Kazakhstan; this is a large-scale facility to examine the vessel steel retention of 60 kg corium during the decay heat; and finally, (5) CORPHAD and PRECOS projects, NITI, St. Petersburg undertook a systematic examination of refractory ceramics relevant to in-vessel and ex-vessel coria, particularly examining poorly characterised, limited data or experimentally difficult systems.
  •  
50.
  • Bubelis, E., et al. (författare)
  • System codes benchmarking on a low sodium void effect SFR heterogeneous core under ULOF conditions
  • 2017
  • Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 320, s. 325-345
  • Tidskriftsartikel (refereegranskat)abstract
    • This paper discusses system codes benchmarking activities on an ASTRID-like heterogeneous fast core under a representative design basis accident condition: the unprotected loss of flow accident (ULOF). The paper provides evidence that all the system codes used in this exercise are capable to simulate the transient behavior of heterogeneous SFR cores up to the initiation of sodium boiling. As a proof of this, a comparison of steady-state results and dynamic simulation results for a ULOF transient (simulated using system codes in combination with neutron point kinetics) are provided and discussed in this paper. The paper contains a brief description of the system codes (TRACE, CATHARE, SIM-SFR, SAS-SFR, ATHLET, SPECTRA, SAS4A) used by the participants (PSI, CEA, EDF, KIT, GRS, UPVLC, NRG, KTH), assumptions made during the simulations, as well as results obtained.
  •  
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