Thermohydraulic tests of 3x3-rod bundle maquette
International Nuclear Information System (INIS)
The results of a 3x3-rod bundle thermohydraulic research program, performed in the Thermohydraulics Laboratory of NUCLEBRAS' Nuclear Technology Development Center, are briefly described. This program included measurements of pressure drops in one and two-phase flows, heat transfer coefficients, mixing between interconnected subchannels in one-phase flow conditions and critical heat fluxes. The measurements covered the following parameter ranges: heat fluxes from zero to the critical values, pressure ranging from 1 to 15 ata, inlet temperature from 25 to 150 sup(0)C and flow rate from 20 to 300l/min. (author)
The 3D core thermohydraulics and neutronics solution in the TRAB-SMABRE accident and transient code
International Nuclear Information System (INIS)
rod withdrawal results are compared for three different thermohydraulic solution of the PWR core, parallel coupling without cross-flow, internal coupling without cross flow and internal coupling with cross flow. The comparison proves that the internal coupling is the best solution even for the PWR transient and accident analyses. If the power differences between neighbouring fuel bundles are large, the core thermohydraulics need to be calculated by considering the 3D cross-flow as well. (author)
Transient thermohydraulic modeling of two-phase fluid systems
International Nuclear Information System (INIS)
This paper presents a transient thermohydraulic modeling, initially developed for a capillary pumped loop in gravitational applications, but also possibly suitable for all kinds of two-phase fluid systems. Using finite volumes method, it is based on Navier-Stokes equations for transcribing fluid mechanical aspects. The main feature of this 1D-model is based on a network representation by analogy with electrical. This paper also proposes a parametric study of a counterflow condenser following the sensitivity to inlet mass flow rate and cold source temperature. The comparison between modeling results and experimental data highlights a good numerical evaluation of temperatures. Furthermore, the model is able to represent a pretty good dynamic evolution of hydraulic variables.
International Nuclear Information System (INIS)
Transient thermohydraulic events often control the design of piping systems in nuclear power plants. Water hammers due to valve closure, pressure transients caused by steam collapse and pipe break all result in structural loads that are characterised by a high frequency content. What also characterises these pressures/forces is the specific spatial and time dependence that is acting on the piping system and found in the wave propagation in the contained fluid. The aim with this project has been to develop recommendations for analysis of the stress response in piping systems subjected to thermohydraulic transients. Basis for this work is that the so called two-step-method is applied and that the structural response is calculated with modal superposition. Derived analysis criteria are based on the assumption that the associated volume strain energy in the wave propagation for the contained fluid may be well defined by a parameter, here called εPN. The stress response in the piping system is assumed to be completely determined with certain accuracy for that part of the volume strain energy in the wave propagation associated with this parameter. A comprehensive work has been done to determine the accuracy in loadings calculated with RELAP5. Properties such as period elongation and associated spurious oscillations in the pressure wave transient have been investigated. Furthermore, has the characteristics of the artificial numerical damping in RELAP5 been identified. Based on desired accuracy of the thermohydraulic analysis together with knowledge about the duration of the thermohydraulic perturbation, the lowest upper frequency limit fPipe, in the modal base that is required for the structure model is calculated. With perturbation is meant such as a valve closure. According to suggested criteria and with the upper frequency limit set, the essential parameters i) largest size of the elements in the structure model and ii) the largest applicable time step in the
Transient Thermohydraulic Heat Pipe Modeling: Incorporating THROHPUT into the CAESAR Environment
Hall, Michael L.
2003-01-01
The THROHPUT code, which models transient thermohydraulic heat pipe behavior, is being incorporated into the CAESAR computational physics development environment. The CAESAR environment provides many beneficial features for enhanced model development, including levelized design, unit testing, Design by Contract™ (Meyer, 1997), and literate programming (Knuth, 1992), in a parallel, object-based manner. The original THROHPUT code was developed as a doctoral thesis research code; the current emphasis is on making a robust, verifiable, documented, component-based production package. Results from the original code are included.
Transient thermohydraulic heat pipe modeling : incorporating THROHPUT into the Caesar environment /
Energy Technology Data Exchange (ETDEWEB)
Hall, Michael L.
2002-01-01
The THROHPUT code, which models transient thermohydraulic heat pipe behavior, is being incorporated into the CAESAR computational physics development environment. The CAESAR environment provides many beneficial features for enhanced model development, including levelized design, unit testing, Design by ContractTM (Meyer, 1997), and literate programming (Knuth, 1992), in a parallel, object-based manner. The original THROHPUT code was developed as a doctoral thesis research code; the current emphasis is on making a robust, verifiable, documented, component-based production package. Results from the original code are included.
HUARPE: A thermohydraulic code for transient simulations in integrated reactors
International Nuclear Information System (INIS)
Full text: With the requirement of having a versatile calculus tool, low CPU cost, capable for parametric studies for support on conceptual reactor design step, a code is developed for simulation of integrated reactor transients. This code (HUARPE) includes coolant, steam dome, RPV structures and core modeling. The code uses a one-dimensional model for natural circulation through the circuit, where mass, momentum and energy equations are solved. A homogeneous model is used for two phase flow through the riser, drift-flux in the dome. The momentum equation is solved in an integral way, neglecting pressure variations caused by perturbations in the circuit. On the other hand, pressure variations due to hydraulic height are modeled. The dome is divided in two variable volumes to represent the steam and mixture zones. It interacts with the rest of the cooling circuit through mass interchanges with the riser and the steam generator inlet. Enthalpies distribution is computed from the energy equation. Mass flow rates are obtained from the momentum and mass equations. An integral form of the momentum equation is considered. The dome dynamic governs system pressure and mixture level evolutions. These are solved by means of the steam and mixture energy and mass equations, with a non-equilibrium model. The mixture density is obtained through the state equations, as a function of pressure and enthalpy. The core power is solved with point kinetic neutronic equations. The RPV structure is modeled in a thermal point of view, and it is divided into slabs, without interaction between each other. As the reactor primary system response is studied the steam generator is considered as a boundary heat exchange condition. Equations are solved by the finite-difference method, with an explicit algorithm for time integration and up-winding for the spatial terms. The HUARPE code has been checked against TRAC and RETRAN codes, and with experimental data
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For the analysis of transient and emergency processes during reactor operation it is necessary to have a set of codes, which calculate physical processes with a various degree of accuracy. Codes CORT and BUMT for three-dimensional thermohydraulic calculation of fast reactor core in steady state, transient and accident conditions are described in this paper. The code CORT calculates thermohydraulics of the whole fast reactor core or group of subassemblies in simplified approximation. The core is described as a set of coupled one-dimensional channels or is divided into a set of ring zones, each of those is also represented by one subassembly (S/A). The detailed three-dimensional calculation of particular S/A is carried out by code BUMT. For description of S/A thermohydraulics the authors have chosen so called 'subchannel model. In this model the S/A is split into number of channels exchanging one by one with mass, momentum and energy. The coefficients of inter channel exchange are calculated on the basis of empirical correlations. The subchannel model is supplemented by detailed (two-dimensional in each axial cross-section) calculation of fuel pin and S/A wrapper temperatures. For solution of hydrodynamic equations the full-implicit scheme is used. Code BUMT was verified using experimental data for S/A-simulators and results of calculations obtained by other codes. These codes when used in complex with neutronic code and first circuit thermohydraulic code could describe in detail the thermal state of coolant and performance of fuel pins and construction elements of reactor during steady and transient states of its operation. (author)
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TRAB-3D models the PWR and BWR reactor core using the two-group diffusion equations in homogenized fuel assembly geometry with a sophisticated nodal method. Thermohydraulics is described using four equation formulation. The stand-alone version of the code also describes thermohydraulics of the rest of the BWR circuit with one dimensional components. The SMABRE code models the thermohydraulics of light water reactors. The five equation formulation with the drift flux phase separation is modelling the two-phase behaviour. Conservation equations are solved for the phase mass, mixture momentum and phase energy. Additional equations are for the noncondensable part in gas and boron in liquid. The TRAB-3D and SMABRE codes have been coupled earlier by using the parallel coupling principle, where in the core section the 3-dimensional TRAB core, and the parallel channel coarse SMABRE core are solved in parallel, but the rest of the circulation system is solved with SMABRE. As a new development the internal coupling to meet new requirements for the PWR and BWR transient analyses is being realised. Both the circuit and core thermohydraulics are solved in SMABRE. The core thermohydraulics solution inside the core wide iterations is repeated to allow rapid power changes. These are the fast pressure changes, control rod ejection and ATWS. The numerical solution in SMABRE has been improved to allow full core simulation with separate flow channel for each fuel element of a BWR core. For the PWR plants the method is used as well by simulating the core by one-dimensional parallel channels. New development is needed for the open core calculation. (authors)
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Full text of publication follows: TRAB-3D models the PWR and BWR reactor core using the two-group diffusion equations in homogenized fuel assembly geometry with a sophisticated nodal method. Thermohydraulics is described using four-equation formulation. The stand-alone version of the code also describes thermohydraulics of the rest of the BWR circuit with one-dimensional components. The SMABRE code models the thermohydraulics of light water reactors. The five-equation formulation with the drift flux phase separation is modelling the two-phase behaviour. Conservation equations are solved for the phase mass, mixture momentum and phase energy. Additional equations are for the noncondensable in gas and boron in liquid. The TRAB-3D and SMABRE codes have been coupled earlier by using the parallel coupling principle, where in the core section the 3-dimensional TRAB core, and the parallel channel coarse SMABRE core are solved in parallel, but rest of the circulation system is solved with SMABRE. As a new development the internal coupling to meet new requirements for the PWR and BWR transient analyses is being realised. Both the circuit and core thermohydraulics are solved in SMABRE. The core thermohydraulics solution inside the core wide iterations is repeated to allow rapid power changes. These are the fast pressure changes, control rod ejection and ATWS. The numerical solution in SMABRE has been improved to allow full core simulation with separate flow channel for each fuel element of a BWR core. For the PWR plants the method is used as well by simulating the core by one-dimensional parallel channels. New development is needed for the open core calculation. In general questions could be raised, what advantages are seen with the new internal coupling in comparison with the earlier realised parallel coupling, and which advantages may be seen in building the realtor physical model on the basis of the old code, developed since 1970's. The internal coupling allows modelling
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Large system simulation codes are needed for design and safety analysis. A thermal-hydraulic simulation code for systems transient in ''Monju'' (COPD code) was developed and verified with experimental data from an experimental LMFBR ''Joyo'', 50 MWt steam generator test facility and scaled test sections of reactor vessel plenum. This paper summarizes numerical models of this code and their verifications with experimental data. Especially, a simplified analytical model to predict the transient behavior in a reactor vessel plenum is presented in detail, since this behavior has an important effect that must be taken into account in a plant thermal transient, while the reactor is tripped. The COPD is applied to design and safety analysis in ''Monju'' as follows ; (1) Safety analysis with regard to core cooling in anticipated incidents. (2) Plant thermo-hydraulic analysis for setting the design condition in thermal stress analysis and evaluation of components and pipings. (3) Control performance analysis on plant operation for design and evaluation of plant control system. Each of the above analyses requires different predictions of plant response to be analyzed. Therefore, appropriate models and input data are used in the design and evaluation according to the purpose of the analysis. This code was developed and verified under a contract with PNC. (author)
THEBES: a thermal hydraulic code for the calculation of transient two phase flow in bundle geometry
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The three dimensional thermal hydraulic code THEBES, capable to calculate transient boiling of sodium in rod bundles is described here. THEBES, derived from the transient single phase code SABRE-2A, was developed in CADARACHE by the SIES to analyse the SCARABEE N loss of flow experiments. This paper also presents the results of tests which were performed against various types of experiments: (1) transient boiling in a 7 pin bundle simulating a partial blockage at the bottom of a subassembly (rapid transient SCARABEE 7.2 experiment), (2) transient boiling in a 7 pin bundle simulating a coolant coast down (slow transient SCARABEE 7.3 experiment), (3) steady local and generalised boiling in a 19 pin bundle (GR 19 I experiment), (4) transient boiling in a 19 pin bundle simulating a coolant coast down (GR 19 I experiment), (5) steady local boiling in a 37 pin bundle with internal blockage (MOL 7C experiment). Excellent agreement was found between calculated and experimental results for these different situations. Our conclusion is that THEBES is able to calculate transient boiling of sodium in rod bundles in a quite satisfying way
International Nuclear Information System (INIS)
The safety of nuclear power plants has always been a concern when this technology is considered as an option for power generation. As a contribution to the improvement of its safety performance, a System for Identification and Classification of Transients (SICT) is being developed. This system is based in neural networks particularly Self-Organizing Maps and has as goal to assist the operation of nuclear plants. The development of this system has several phases and one of them is the demonstration of the capability of SICT to respond on time for transients being able to warn the operator. This demonstration will be achieved using experiments in a thermo-hydraulic facility - CT1 - in CDTN, having the SICT coupled to it. Before coupling the SICT with CT1 instrumentation it has to be trained to recognize different operational states possible in the installation. This training is performed using results of simulation of experiments with the RELAP5 code, in the same way as the SICT for the Nuclear Power Plant shall be preliminarily trained using results of simulations. This paper presents the description of such facility, with the coupled SICT, the carried out experiments, as well as, their simulations with RELAP5 and the overall performance of SICT. (author)
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This paper briefly describes the subchannel thermo-hydraulics code SABRE, with particular emphasis on the choice of two-phase modelling. The code has been extensively applied to the analysis of fault and severe accident situations in fast reactor cores, and in order to validate the code a range of accident simulation experiments has been analysed. The relative merits of slip models of boiling and full two-fluid representation are discussed, and results are presented comparing the two approaches. It is shown that in many situations the simpler ('3-equation') slip representation is adequate, but that there exist situations where the two-fluid ('6-equation') model is essential to even represent the physical phenomena. An example of such a situation is in wake regions, where vapour and liquid flows may be very different; in the paper this is discussed with particular reference to the flow past a blockage. In this example vapour accumulation may occur behind the blockage, and SABRE calculations of this situation are presented. (orig.)
Validity of the quasi-static assumption in transient thermo-hydraulic analysis of GCFR
International Nuclear Information System (INIS)
In reactor transient analysis, the friction factor and the heat transfer coefficient are assumed to be equal to the steady-state ones. Validity of this 'quasi-static assumption' is examined. The transient turbulent heat transfer in a circular tube is examined numerically and experimentally in step change of the pressure gradient or the heat input. Transient variations of the friction factor and the heat transfer coefficient are obtained. The times required for the flow velocity and the heat transfer coefficient to attain the steady-state values are studied. The steady state friction factor and the heat transfer coefficient are found to be applicable in transient analyses of GCFR. (auth.)
Trost, Nico; Jiménez, Javier; Imke, Uwe; Sanchez, Victor
2014-06-01
TWOPORFLOW is a thermo-hydraulic code based on a porous media approach to simulate single- and two-phase flow including boiling. It is under development at the Institute for Neutron Physics and Reactor Technology (INR) at KIT. The code features a 3D transient solution of the mass, momentum and energy conservation equations for two inter-penetrating fluids with a semi-implicit continuous Eulerian type solver. The application domain of TWOPORFLOW includes the flow in standard porous media and in structured porous media such as micro-channels and cores of nuclear power plants. In the latter case, the fluid domain is coupled to a fuel rod model, describing the heat flow inside the solid structure. In this work, detailed profiling tools have been utilized to determine the optimization potential of TWOPORFLOW. As a result, bottle-necks were identified and reduced in the most feasible way, leading for instance to an optimization of the water-steam property computation. Furthermore, an OpenMP implementation addressing the routines in charge of inter-phase momentum-, energy- and mass-coupling delivered good performance together with a high scalability on shared memory architectures. In contrast to that, the approach for distributed memory systems was to solve sub-problems resulting by the decomposition of the initial Cartesian geometry. Thread communication for the sub-problem boundary updates was accomplished by the Message Passing Interface (MPI) standard.
Uncovery boiloff transients in a 3- x 3-rod bundle. Final report
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A study was carried out of transiently boiling dry a 3- x 3-rod bundle. The location of the two-phase mixture level, bundle heat transfer, and liquid inventory were measured. The effects of injecting cold water at the top of the bundle were also studied. The report documents the test data and includes a diversion of upper tie plate and counter-current flooding phenomena on core uncovery
International Nuclear Information System (INIS)
The model developed using point kinetics and simulate various transients interactively, such as the firing of feed water turbo-pumps or the closing of the valves of main steam (MSIVs). Developed models allow to visualize, through different screens, the behavior of the whole plant as well as its control system.
RELAP-4, Transient 2 Phase Flow Thermohydraulics, LWR LOCA and Reflood
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1 - Description of problem or function: RELAP4 has been developed to describe the behavior of water-cooled nuclear reactors subjected to postulated transients, such as those resulting from loss-of-coolant, pump failure, or power excursions. The program calculates fluid conditions such as flow, pressure, mass inventory, and quality; thermal conditions such as surface temperatures, temperature profiles, and energy distributions; and heat fluxes in power generating and dissipating elements. The program also calculates reactor power, decay heat, and reactivity. In addition to describing transients in boiling-water and pressurized-water reactors, the program is sufficiently versatile to describe transients in experimental thermal-hydraulic systems. RELAP4/MOD6 was developed specifically to add a capability to the earlier RELAP codes for calculating PWR reflood phenomena. REALP-4/6-KFK is a modification of RELAP4/MOD6 update 4 which generates a file for transferring boundary conditions to the fuel rod code SSYST-2 (NEA 0684). RELAP4/MOD7/101: Performs best estimate analyses of nuclear reactors or related systems undergoing a transient. Transient thermal- hydraulic, two-phase phenomena are calculated from formulations of one-dimensional, homogeneous, equilibrium conservation equations for water mass, momentum, and energy. Heat structures are modeled using a transient one-dimensional heat conduction solution that is coupled to the fluid through heat transfer relations. Various explicit models are used to calculate nonhomogeneous, nonequilibrium behavior including a phase separation model, a vertical slip model, and a nonequilibrium model. Other models are used to represent critical flow, reactor kinetics, pressurized water reactor reflood behavior, nuclear fuel rod swelling and blockage, and components such as pumps, valves, and accumulators. 2 - Method of solution: The RELAP4 user must define the geometric features of the system to be analyzed as well as an appropriate
Solving linear systems in FLICA-4, thermohydraulic code for 3-D transient computations
International Nuclear Information System (INIS)
FLICA-4 is a computer code, developed at the CEA (France), devoted to steady state and transient thermal-hydraulic analysis of nuclear reactor cores, for small size problems (around 100 mesh cells) as well as for large ones (more than 100000), on, either standard workstations or vector super-computers. As for time implicit codes, the largest time and memory consuming part of FLICA-4 is the routine dedicated to solve the linear system (the size of which is of the order of the number of cells). Therefore, the efficiency of the code is crucially influenced by the optimization of the algorithms used in assembling and solving linear systems: direct methods as the Gauss (or LU) decomposition for moderate size problems, iterative methods as the preconditioned conjugate gradient for large problems. 6 figs., 13 refs
Modelling of transient dynamic bundle deformation using time integration scheme
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The BOW code has been examined whether its modeling capability can be extended to the simulation of interactions (i.e., fretting) between neighbouring fuel elements in a fuel bundle and between the fuel bundle and the pressure tube in a fuel channel. The current BOW code is specialized in simulating the static problems, such as the deflection of each element and interactions between neighbouring elements in a fuel bundle, and interactions between neighbouring bundles and between a bundle and the pressure tube in a fuel channel. The Wilson θ time integration scheme has been implemented in the BOW code, for the extension of its capability to modelling dynamic contact problems. As part of verification to ensure that the modification in the code functions exactly as designed, the dynamic-modelling capability of the BOW code has been applied to simple support beam cases subjected to a uniform step load at the middle of the beam. The calculation results confirmed that the modified BOW code, where the contact algorithm is implemented in the step-by-step integration manner using the Wilson θ time integration scheme, can solve the dynamic problem with unconditional convergence. This paper describes the theory and models for the new capabilities of the BOW code. (author)
A thermal-hydraulic code for transient analysis in a channel with a rod bundle
Energy Technology Data Exchange (ETDEWEB)
Khodjaev, I.D. [Research & Engineering Centre of Nuclear Plants Safety, Electrogorsk (Russian Federation)
1995-09-01
The paper contains the model of transient vapor-liquid flow in a channel with a rod bundle of core of a nuclear power plant. The computer code has been developed to predict dryout and post-dryout heat transfer in rod bundles of nuclear reactor core under loss-of-coolant accidents. Economizer, bubble, dispersed-annular and dispersed regimes are taken into account. The computer code provides a three-field representation of two-phase flow in the dispersed-annular regime. Continuous vapor, continuous liquid film and entrained liquid drops are three fields. For the description of dispersed flow regime two-temperatures and single-velocity model is used. Relative droplet motion is taken into account for the droplet-to-vapor heat transfer. The conservation equations for each of regimes are solved using an effective numerical technique. This technique makes it possible to determine distribution of the parameters of flows along the perimeter of fuel elements. Comparison of the calculated results with the experimental data shows that the computer code adequately describes complex processes in a channel with a rod bundle during accident.
Transient heat transfer behavior during reflood phase in a 2x2 ballooned rod bundle
International Nuclear Information System (INIS)
The coolability of the ballooned region is entirely different with that of the normal ones. Therefore, in this study, the transient heat transfer behavior during the reflood phase of ballooned fuel rods was experimentally investigated in a 2x2 rod bundle test facility. The coolability depends greatly on the blockage characteristics (blockage ratio, blockage length, blockage shape, and blockage configuration) and the system conditions of the test facility (flow, system pressure, and inlet temperature). Among them, the blockage ratio effect on the coolabiltiy is carefully examined varying the reflood rate in the present study, since the blockage ratio plays a significant role on the coolability under the low reflood rate condition (2.5 cm/s). The test results were analyzed with the transient temperature profiles of the fuel rods and the local heat transfer coefficient calculated using a 1-D cylindrical coordinates FVM (Finite-Volume-Method) code. Forced reflood tests with various reflood rates were performed to understand the transient heat transfer behavior and to investigate the influence of the blockage ratio on the coolability in the 2x2 rod bundle test facility. The transient temperature profiles and the local heat transfer coefficients at the upstream and downstream region of the blockage simulator were examined for non-blockage, 90% blockage, and 62% blockage conditions. In the downstream region, the coolability was greatly enhanced except for a low reflood rate (1.0 cm/s). In the upstream region, the cooling performance decreased smoothly with decreasing the reflood rate. When the reflood rate is 1.0 cm/s, the coolabilities at the both upstream and downstream region were significantly reduced regardless of the blockage ratio. As a conclusion, the coolability at the low reflood rate (1.0 cm/s) should be carefully examined with the droplet behavior as a future work
Thermohydraulic analysis of pressurized water reactors
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The computer program PANTERA is applied in the thermo-hydraulic analysis of Pressurized Water Reactor Cores (PWR). It is a version of COBRA-IIIC in which a new thermal conduction model for fuel rods was introduced. The results calculated by this program are compared with experimental data obtained from bundles of fuel rods, simulating reactor conditions. The validity of the new thermal model is checked too. The PANTERA code, through a simplified procedure of calculation, is used in the thermo-hydraulic analysis of Indian Point, Unit 2, reactor core, in stationary conditions. The results are discussed and compared with design data. (Autor)
International Nuclear Information System (INIS)
The HEXART 3D kinetics code, coupled to a Power Controller model and NPP simulator has been used to calculate a start-up transient of Unit 3 of the South Ukrainian NPP. The objectives are to test the 3D neutron kinetics and power controller ex-vessel detection models in a complex ATWS. The initiating event is simultaneous trip of 2 adjacent MCP out of 4 at 97 % of the nominal rated power. This transient is a serious validation test of core dynamics and plant component interactions with the reactor core. Modeling of the Ex-vessel Detection System and non-uniform coolant mixing is of particular importance. The computed are in satisfactory agreement with the plant measurements. (Authors)
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The main objective of the PHARE project SRR1/95 is the validation of coupled thermal-hydraulics - neutron kinetics codes, that are currently used for modelling the behaviour of the Russian pressurized water reactors VVER. Short descriptions of two transients, measured in the Loviisa-1 VVER-440 and the Balakovo-4 VVER-1000, respectively, have been presented in. DYN3D burnup and steady-state calculations for the states before the transients and their comparison with measurements have also been. The present report contains the main results of the simulation of both measured transients by the coupled code DYN3D-ATHLET. Two different versions of coupling the three-dimensional core model DYN3D to the thermal-hydraulic system code ATHLET are available and have been used for the calculations. In the external coupling, the whole core model DYN3D (neutron kinetics, thermal-hydraulics, and fuel rod model) is coupled to ATHLET by interfaces at the core bottom and top. In the internal coupling, only the neutron kinetics of DYN3D is implemented into ATHLET. (orig.)
Proceedings of the 6. National Meeting of Reactor Physics and Thermohydraulic
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The proceedings of the 6. National Meeting of Reactor Physics and Thermohydraulic - 6. ENFIR - allow to evaluate the present status of development in reactor physics and thermohydraulic fields. The mathematical models and methods for calculating neutronic of nuclear reactors, safety reactor analysis, measuring methods of neutronic parameters, computerized simulation of accidents, transients and thermohydraulic analysis are presented. (M.C.K.)
International Nuclear Information System (INIS)
The computer programme COMMIX-2 describes steady state and transient multidimensional single- and two-phase fluid flows with heat transfer in nuclear reactor components and multicomponent systems. Originally from the Argonne National Laboratory, the code has been further developed at the Kernforschungszentrum Karlsruhe. The original Point-SOR iterative method for the solution of a Poisson-like equation describing the pressure distribution in the fluid as well as the transport of enthalpy and turbulent quantities has been complemented with iterative and direct line- and block-methods. None of the newly implemented methods is original in itself but their implementation into the computer code, which can describe the most general shapes of definition domains, gave a code speed-up by a factor of 2-5, depending on the problem treated. The code capabilities are assessd by the calculation of a benchmark problem involving the numerical simulation of thermal buoyancy phenomena at a pipe/plenum interface. (orig.)
Mechanisms of heat transfer in the uncovered region of a bundle during the boil-off transient
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The small break accident which occurred at the TMI-2 plant resulted in partial uncovery of the core. To study the thermal-hydraulic phenomena in the uncovered portion of the core, tests were conducted from the EPRI/SUNY Buffalo 3 x 3 rod bundle. Observations from motion pictures and test data show that liquid entrainment and liquid fallback occur in the upper rod bundle region during the early stage of the boiling dry transient. The liquid entrainment and liquid fallback are the results of flow restrictions in the upper bundle tie-plate and spacer grids. The presence of liquid droplets during the entrainment and the fallback greatly influenced the heat transfer in the uncovered portion of the bundle
LMFR core thermohydraulics: Status and prospects
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One of the fundamental steps for a successful reactor core thermohydraulic design is the capability to predict, reliably and accurately, the temperature distribution in the core assemblies. A detailed knowledge of the assembly and fuel pin thermohydraulic behaviour in the steady state and transient conditions is an indispensable prerequisite to safe and stable operation of the reactor. Considerable experimental and theoretical studies on various aspects of LMFR core thermohydraulics are necessary to acquire such knowledge. During the last decade, there have been substantial advances in fast reactor core thermohydraulic design and operation in several countries with fast reactor programmes (notably in France, the Russian Federation, Japan, the United Kingdom, Germany and the United States of America). Chief among these has been the demonstration of reliable operation of reactor cores at a high burnup. During the last years, some additional countries such as China, India and the Republic of Korea have launched new fast reactor programmes. International exchange of information and experience on LMFR development including core thermohydraulic design is becoming of increasing importance to these countries. It is with this focus that the IAEA convened the Technical Committee on 'Methods and Codes for Calculations of Thermohydraulic Parameters for Fuel, Absorber Pins and Assemblies of LMFR's with Traditional and Burner Cores'. This meeting, attended by participants from seven countries, brought together a group of international experts to review and discuss the thermohydraulic advances and design approaches providing a reliable, safe and robust reactor core, as well as to exchange the experience accumulated in different countries of using the codes for thermohydraulic calculations and to discuss the issues requiring further research and development. A total of thirty technical papers presented covered theoretical and computational issues as well as experiments under
Physical modeling of thermohydraulic phenomena in LMFBR
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A simulation method of thermohydraulic problems in LMFBR is illustrated by a dimensional analysis of the different equations. For steady state and transient regimes on the reactor, it is shown how some experiments on small scale models with usual fluid permit a tentative solution to these problems
Energy Technology Data Exchange (ETDEWEB)
Bates, J.M.; Khan, E.U.
1980-10-01
An experimental study was performed to obtain local fluid velocity and temperature measurements in the mixed (combined free and forced) convection regime for specific flow coastdown transients. A brief investigation of steady-state flows for the purely free-convection regime was also completed. The study was performed using an electrically heated 2 x 6 rod bundle contained in a flow housing. In addition a transient data base was obtained for evaluating the COBRA-WC thermal-hydraulic computer program (a modified version of the COBRA-IV code).
International Nuclear Information System (INIS)
The lecture includes typical transients to be analyzed, the requirements put on computer codes and a description of the computer codes as well as results obtained with these codes. Transients analysis is necessary within the licensing of reactors, in risk evaluation and in other basic studies (e.g. on ATWS). The development of transient codes has been influenced by new requirements due to the extension in applications mentioned above. As examples the BWR model ALMOS and the PWR model ALMOD are described. These codes include a one-dimensional simulation of the neutron kinetics and the thermohydraulics in the coolant system. Also included in the simulation are all components of the control and safety systems, which are influencing the dynamic behaviour of the plant. Special emphasis is put on the problems of model verification (comparison with measurements). The transients behaviour of plants under extreme conditions, such as transients with a failure of the scram system, is described in detail. Examples are the loss of heat sink and the station black out for both a BWR and a PWR. (orig.)
Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Final report
International Nuclear Information System (INIS)
This project principally undertook the investigation of the thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions were emphasized. Continuing efforts are underway at MIT to complete the investigation of the mixed convection regime initiated here. A number of investigations on outlet plenum behavior were also made. The reports of these investigations are identified
Transient non-boiling heat transfer in a fuel rod bundle during accidental power excursions
International Nuclear Information System (INIS)
The physical problem studied is the transient non-boiling heat transfer of a cylindrical fuel rod consisting of fuel, gap, and cladding to a steady, fully developed turbulent flow. The fuel pin is assumed to be located in the interior region of a subassembly with regular triangular or square arrangements. The turbulent velocity field as well as turbulent transport properties are specified as functions of the coordinates normal to the axial flow direction. The heat generation within the fuel may be specified as an arbitrary function of the three spatial coordinates and time. A digital computer program has been developed. On the basis of finite-difference techniques, to solve the governing partial differential equations with their associated subsidiary conditions. Results have been obtained for a series of exponential power transients of interest to safety of liquid-metal and water cooled nuclear reactors. The general physical features of transient convective heat transfer as explored by previous investigators have qualitatively been substantiated by the present analysis. Emphasis has been devoted to investigate the differences of heat-transfer (coefficient) results from multi-region analysis including a realistic fuel rod model and single-region analysis for the coolant region only. A comparison with the engineering relationships for turbulent liquid-metal cooling by Stein, which are an extension of the heat transfer coefficient concept to account for transient heat fluxes, clearly demonstrates that, at the parameters studied, Stein's approach tends to largely overestimate the convective heat transfer at early times
International Nuclear Information System (INIS)
The results of the calculational study using the RELAP5/MOD3.2 thermalhydraulic code performed on the influence of the heat losses to the ambient and the heat accumulated in the pipelines walls upon the evolution of the thermalhydraulic processes in the primary circuit of the integral test facility ISB-WWER when simulating the transients caused by the loss of the coolant are presented in the paper. (authors)
Thermohydraulic tests in the area of reactor safety done in CDTN
International Nuclear Information System (INIS)
The main experimental works performed in the last five years at the Thermohydraulics Laboratory of the Nuclear Technology Development Center, in the field of reactor safety are briefly described. This paper cover the performing and analysis of pressure drop, heat transfer and mixing tests in 3X3 rod bundle and rewetting tests in single tube section. (autor)
Seminar on experimental thermohydraulics
International Nuclear Information System (INIS)
Considerations on reactor safety are made. Problems related to the project and assembling of facilities for experiments in steady state and transient conditions are discussed. The advantages of using model fluids are finally, analysed. (Author)
International Nuclear Information System (INIS)
Comparative analysis of thermohydraulic margins were studied of the CANDU 37 and CARA fuel bundles (FB) in Embalse power station with COBRA IV-HW code ., the geometry of the bundle laying on the channel was particularly modeled and discussing the results in comparison with former calculations with 1/6 simetry .The CARA design with enriched uranium (0.9 %) and extended burn up lets maintain the current thermohydraulic nominal margins , while compared with CANDU 37 rods FB enriched , the CARA design permits widely improve the current margins
International Nuclear Information System (INIS)
The assessment of the uncertainties of COBRA-IIIC thermal-hydraulic analyses of rod bundles is performed for a 5-by-5 bundle representing a PWR fuel assembly. In the first part of the paper the modeling uncertainties are evaluated in the term of the uncertainty of the turbulent mixing factor using the OECD NEA/NRC PSBT benchmark data. After that the uncertainties of the COBRA calculations are discussed performing Monte-Carlo type statistical analyses taking into account the modeling uncertainties and other uncertainties prescribed in the OECD NEA UAM benchmark specification. Both steady-state and transient cases are investigated. The target quantities are the uncertainties of the void distribution, the moderator density, the moderator temperature and the DNBR. We will see that - beyond the uncertainties of the geometry and the boundary conditions - it is very important to take into account the modeling uncertainties in case of bundle or sub-channel thermo-hydraulic calculations.
International Nuclear Information System (INIS)
Highlights: • A 3-D CFD is adopted to simulate transient behaviors in an SFP under the accident. • This model realistically simulates a 17 × 17 bundle, rid of porous media approach. • The loss of external cooling system accident for an SFP is assumed in this paper. • Thermal–hydraulic characteristics in a bundle are strongly influenced by grids. • The results confirm temperature rising rate used in Maanshan NPP is conservative. - Abstract: This paper develops a three-dimensional (3-D) transient computational fluid dynamics (CFD) model to simulate the thermal–hydraulic characteristics in a fuel bundle located in a spent fuel pool (SFP) under the loss of external cooling system accident. The SFP located in the Maanshan nuclear power plant (NPP) is selected herein. Without adopting the porous media approach usually used in the previous CFD works, this model uses a real-geometry simulation of a 17 × 17 fuel bundle, which can obtain the localized distributions of the flow and heat transfer during the accident. These distribution characteristics include several peaks in the axial distributions of flow, pressure, temperature, and Nusselt number (Nu) near the support grids, the non-uniform distribution of secondary flow, and the non-uniform temperature distribution due to flow mixing between rods, etc. According to the conditions adopted in the Procedure 597.1 (MNPP Plant Procedure 597.1, 2010) for the management of the loss-of-cooling event of the spent fuel pool in the Maanshan NPP, the temperature rising rate predicted by the present model can be equivalent to 1.26 K/h, which is the same order as that of 3.5 K/h in the this procedure. This result also confirms that the temperature rising rate used in the Procedure 597.1 for the Maanshan NPP is conservative. In addition, after the loss of external cooling system, there are about 44 h for the operator to repair the malfunctioning system or provide the alternative water source for the pool inventory to
Chen, Yongmin; Li, Jianpei
2015-01-01
When procuring multiple products from competing firms, a buyer may choose separate purchase, pure bundling, or mixed bundling. We show that pure bundling will generate higher buyer surplus than both separate purchase and mixed bundling, provided that trade for each good is likely to be efficient. Pure bundling is superior because it intensifies the competition between firms by reducing their cost asymmetry. Mixed bundling is inferior because it allows firms to coordinate to ...
International Nuclear Information System (INIS)
One of the main objectives under design and development of fuel in water cooled nuclear reactors is to ensure fuel integrity during spent fuel handling operation. The on-power refuelling facility adopted in the Indian Pressurized Heavy Water Reactors (PHWRs) causes exposure of the irradiated fuel, during its unloading, to wide variations in its surroundings including exposure to dry gaseous environment. Detailed analyses have been carried out to assess the fuel pin temperature transients during the entire course of its passage from within the reactor to the outside surroundings to ascertain fuel integrity. The cases of normal as well as envisaged off-normal transport operations have been considered in these calculations. The forced air cooling provisions have also been worked out to mitigate the consequences of off-normal transport operation. The present paper deals briefly with the system description, method of calculations and the results obtained for the case of spent fuel handling in the proposed 500 MWe PHWR. (author)
Thermohydraulic and Safety Analysis for CARR Under Station Blackout Accident
International Nuclear Information System (INIS)
A thermohydraulic and safety analysis code (TSACC) has been developed using Fortran 90 language to evaluate the transient thermohydraulic behaviors and safety characteristics of the China Advanced Research Reactor(CARR) under Station Blackout Accident(SBA). For the development of TSACC, a series of corresponding mathematical and physical models were considered. Point reactor neutron kinetics model was adopted for solving reactor power. All possible flow and heat transfer conditions under station blackout accident were considered and the optional models were supplied. The usual Finite Difference Method (FDM) was abandoned and a new model was adopted to evaluate the temperature field of core plate type fuel element. A new simple and convenient equation was proposed for the resolution of the transient behaviors of the main pump instead of the complicated four-quadrant model. Gear method and Adams method were adopted alternately for a better solution to the stiff differential equations describing the dynamic behaviors of the CARR. The computational result of TSACC showed the enough safety margin of CARR under SBA. For the purpose of Verification and Validation (V and V), the simulated results of TSACC were compared with those of Relap5/Mdo3. The V and V result indicated a good agreement between the results by the two codes. Because of the adoption of modular programming techniques, this analysis code is expected to be applied to other reactors by easily modifying the corresponding function modules. (authors)
Containment severe accident thermohydraulic phenomena
International Nuclear Information System (INIS)
This report describes and discusses the containment accident progression and the important severe accident containment thermohydraulic phenomena. The overall objective of the report is to provide a rather detailed presentation of the present status of phenomenological knowledge, including an account of relevant experimental investigations and to discuss, to some extent, the modelling approach used in the MAAP 3.0 computer code. The MAAP code has been used in Sweden as the main tool in the analysis of severe accidents. The dependence of the containment accident progression and containment phenomena on the initial conditions, which in turn are heavily dependent on the in-vessel accident progression and phenomena as well as associated uncertainties, is emphasized. The report is in three parts dealing with: * Swedish reactor containments, the severe accident mitigation programme in Sweden and containment accident progression in Swedish PWRs and BWRs as predicted by the MAAP 3.0 code. * Key non-energetic ex-vessel phenomena (melt fragmentation in water, melt quenching and coolability, core-concrete interaction and high temperature in containment). * Early containment threats due to energetic events (hydrogen combustion, high pressure melt ejection and direct containment heating, and ex-vessel steam explosions). The report concludes that our understanding of the containment severe accident progression and phenomena has improved very significantly over the parts ten years and, thereby, our ability to assess containment threats, to quantify uncertainties, and to interpret the results of experiments and computer code calculations have also increased. (au)
International Nuclear Information System (INIS)
System codes incorporating a full three-dimensional (3-D) reactor core model allow best-estimate simulations of nuclear power plants; however, two main questions arise with respect to this technology. First, are full 3-D thermohydraulics necessary? Second, what net profit is gained with the 3-D neutronics? Furthermore, currently, there is only short experience with 3-D techniques; consequently, the idea has formed of the Nuclear Energy Agency Nuclear Science Committee performing a series of plant transient benchmarks in order to verify the 3-D core models. The Main-Steam-Line-Break (MSLB) Benchmark belongs to this initiative. This benchmark consists of three exercises, namely, point kinetics plant simulation, a coupled 3-D neutronics/core thermohydraulic evaluation of core response, and finally a best-estimate coupled core-plant transient simulation. This paper is based on the experience of performing the second exercise to test the 3-D and point neutron kinetics response with imposed thermohydraulic boundary conditions using TRAC/BF1 (Ref. 2), TRAC/PF1 (Ref. 3), and RETRAN-3D. The following thermohydraulic conditions are provided in the benchmark: radial distribution of mass flow rates and liquid temperatures at the core inlet, and radial distribution of pressure versus time at both the core inlet and outlet. We have developed four different reactor core models: two for TRAC/PF1 and two more for TRAC/BF1. The first one, assumed as the reference, models the core using the VESSEL component of TRAC/PF1 (3-D thermohydraulic equations). The second one models the core using the PIPE component in place of the vessel, also with the same code. The third and fourth ones are with TRAC/BF1, and both represent the core using 18 bundles with the CHANNEL component and no vessel (so no crossflow is considered). The third one has no lower plenum mixing, but the fourth one has it; it allows all the flow thermohydraulic properties to mix before the core inlet, and it is similar to
Energy Technology Data Exchange (ETDEWEB)
Escriva, A.; Munoz-cobo, J. L.; Concejal, A.; Soler, A.; Melara, J.; Albendea, M.
2013-07-01
The model developed using point kinetics and simulate various transients interactively, such as the firing of feed water turbo-pumps or the closing of the valves of main steam (MSIVs). Developed models allow to visualize, through different screens, the behavior of the whole plant as well as its control system.
Neutron-kinetic and thermo-hydraulic uncertainties in the study of Kalinin-3 benchmark
International Nuclear Information System (INIS)
The effects of nuclear data covariance on important reactor parameters are investigated. The analyses are performed on the base of the OECD/NEA coolant transient Benchmark (K-3) on measured data at Kalinin-3 Nuclear Power Plant (NPP). For this purpose the GRS uncertainty and sensitivity software package XSUSA is applied to propagate uncertainties in nuclear data libraries to the full core coupled transient calculations. Moreover, based on the previous thermo-hydraulic studies a set of most important thermo-hydraulic parameters is chosen and added to the uncertain input vector. A statistically representative set of coupled ATHLET PARCS code steady state calculations is analyzed and both integral and local output quantities are compared with the measurements available in the benchmark. The work is a step forward in establishing a ''best-estimate calculations in combination with performing uncertainty analysis'' methodology for coupled full core calculations.
Neutron-kinetic and thermo-hydraulic uncertainties in the study of Kalinin-3 benchmark
Energy Technology Data Exchange (ETDEWEB)
Pasichnyk, Ihor; Zwermann, Winfried; Velkov, Kiril [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany); Nikonov, Sergey [All-Russain Research Institute for NPP Operation (VNIIAES), Moscow (Russian Federation)
2015-09-15
The effects of nuclear data covariance on important reactor parameters are investigated. The analyses are performed on the base of the OECD/NEA coolant transient Benchmark (K-3) on measured data at Kalinin-3 Nuclear Power Plant (NPP). For this purpose the GRS uncertainty and sensitivity software package XSUSA is applied to propagate uncertainties in nuclear data libraries to the full core coupled transient calculations. Moreover, based on the previous thermo-hydraulic studies a set of most important thermo-hydraulic parameters is chosen and added to the uncertain input vector. A statistically representative set of coupled ATHLET PARCS code steady state calculations is analyzed and both integral and local output quantities are compared with the measurements available in the benchmark. The work is a step forward in establishing a ''best-estimate calculations in combination with performing uncertainty analysis'' methodology for coupled full core calculations.
Thermohydraulic relationships for advanced water cooled reactors
International Nuclear Information System (INIS)
This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes
Advanced thermohydraulic simulation code for pool-type LMFBRs (SSC-P code)
Energy Technology Data Exchange (ETDEWEB)
Madni, I.K.; Cazzoli, E.G.
1980-09-01
Models for components and processes that are needed for simulation of thermohydraulic transient in a pool-type liquid metal fast breeder reactor (LMFBR) plant are described in this report. A computer code, SSC-P, has been developed as a part of the Super System Code (SSC) development project. A user's manual is being prepared as a separate document. 27 refs., 26 figs., 1 tab.
Thermohydraulic calculation of WWER-type NPP
International Nuclear Information System (INIS)
Technique of thermohydraulic calculation of the WWER-type NPP in unsteady processes is described. Effective algorithm for solving hydrodynamics equations without regard for acoustic effects permitting to use enough large time integration step is given. Calculation of two-dimensional temperature fields in fuel element is considered. Method for calculating a pressurizer, steam generators and pumps is described as well
International Nuclear Information System (INIS)
One-dimensional mathematical models are extensively used in thermohydraulics assessment of Nuclear Power Plant (NPP) transients and accidents, because specifically 1-D system of the conservation laws allows to reduce computing time and required memory, especially in ''best estimate'' code calculations. This work is generalization of the well-known Zuber-Findley and Hancox-Nicoll methods for two-phase flow distribution parameters Cs taking into account the non-monotonous void fraction distribution in the transverse direction in terms of two superimposed monotonous profiles. The method is very useful in evaluating the saddle-shape void fraction profile effects. In this work two-phase flow distribution parameters Cs were developed for simple circular and rectangular pipes, and subchannel geometry in a rod bundle. Basic assumptions were power-mode approximations for describing the profiles of local volume flux density, phase velocity and temperature. The general analytical (quadrature) relationships for Cs were obtained and their 3-D illustrations are proposed. Also, we propose generalized formulation and simple approach to construct friction factor, heat and mass transfer coefficients within the gradient hypothesis and boundary layer assumptions. The contribution of momentum, heat and mass transfer as well as their sources and sinks in the channel cross-section are taken into account. In the same way, the friction factor, heat and mass transfer coefficients with the transversal and azimuthal variations being taken into account are proposed for subchannel geometry as well. (author)
Library thermohydraulic components for training simulators
International Nuclear Information System (INIS)
The thermohydraulic components Library was modeled in MatLab/Simulink®. This library owns Pipe type components (pump, control valve and / or heaters), storage tanks (Open, Closed and Equilibrium Water Vapor-Air) and Heat Exchangers (Co-Current, Counter-Current and U-tubes). Each component can be attached to other components through the component library Header, in order to create a more complex thermal-hydraulic system which in turn can interact with other thermal-hydraulic systems. (author)
Thermo-hydraulic and structural analysis for finger-based concept of ITER blanket first wall
International Nuclear Information System (INIS)
The blanket first wall is one of the main plasma facing components in ITER tokamak. The finger-typed first wall was proposed through the current design progress by ITER organization. In this concept, each first wall module is composed of a beam and twenty fingers. The main function of the first wall is to remove efficiently the high heat flux loading from the fusion plasma during its operation. Therefore, the thermal and structural performance should be investigated for the proposed finger-based design concept of first wall. The various case studies were performed for a unit finger model considering different loading conditions. The finite element model was made for a half of a module using symmetric boundary conditions to reduce the computational effort. The thermo-hydraulic analysis was performed to obtain the pressure drop and temperature profiles. Then the structural analysis was carried out using the maximum temperature distribution obtained in thermo-hydraulic analysis. Finally, the transient thermo-hydraulic analysis was performed for the generic first wall module to obtain the temperature evolution history considering cyclic heat flux loading with nuclear heating. After that, the thermo-mechanical analysis was performed at the time step when the maximum temperature gradient was occurred. Also, the stress analysis was performed for the component with a finger and a beam to check the residual stress of the component after thermal shrinkage assembly.
Heal, Geoffrey
2002-01-01
Biodiversity provides essential services to human societies. Many of these services are provided as public goods, so that they will typically be underprovided both by market mechanisms (because of the impossibility of excluding non-payers from using the services) and by government-run systems (because of the free rider problem). I suggest here that in some cases the public goods provided by biodiversity conservation can be bundled with private goods and their value to consumers captured in th...
Experimental round for research thermohydraulic characteristics channels with complex geometry
International Nuclear Information System (INIS)
In the paper description of experimental loop for thermohydraulic investigations of fluid flow in channel with complex geometry is given. Loop is assignment for experimental research thermic and hydrodynamic phenomena for one and two-phase one and two-component fluid flow in channels. Loop is designed and performed for void range of thermohydraulic parameters and long evaporation channels with complex geometry. (author)
Study on Unigraphics Drawing Modeling Method for 37-Element and CANFLEX Fuel Bundle
International Nuclear Information System (INIS)
The CANFLEX bundle contains 43 elements of two different diameters. It has two rings of small diameter elements on the outside, and eight elements (with diameter slightly larger than those in the standard 37-Element bundle) in the center. This larger number of small diameter elements on the outside of the CANFLEX bundle enhances thermo-hydraulic capability, resulting in a higher power capability and an improvement in operating safety margins. As a Result of advanced fuel design for CANFLEX fuel bundles, components consisting of fuel bundles are more complicated. Hence, the detailed modeling of components is inevitable in order to analyze the fuel performance by computational fluid dynamics. In this report, the basic design of the advanced fuel for CANDU reactors was carried out and the methodology for the modeling of fuel bundle were described. Firstly, the components consisting of fuel bundles were separately modeled and saved with different file names. The final feature of fuel bundle was accomplished by an assembling process of components. Since this report developed the modeling methodology based on the Unigraphics program, the basic explanations for the software were given first, and the complete modeling of 37-elements and CANFLEX fuel bundles were provided. The components of CANFLEX fuel bundles were also compared with that of 37-elements fuel bundles. Although, in this report, the modeling methodology is applied only to 37-elements and CANFLEX fuel bundles, this methodology may be applicable to the newly designed fuel bundles which are to be developed in the future
Study on Unigraphics Drawing Modeling Method for 37-Element and CANFLEX Fuel Bundle
Energy Technology Data Exchange (ETDEWEB)
Jeon, Yu Mi; Park, Joo Hwan
2010-03-15
The CANFLEX bundle contains 43 elements of two different diameters. It has two rings of small diameter elements on the outside, and eight elements (with diameter slightly larger than those in the standard 37-Element bundle) in the center. This larger number of small diameter elements on the outside of the CANFLEX bundle enhances thermo-hydraulic capability, resulting in a higher power capability and an improvement in operating safety margins. As a Result of advanced fuel design for CANFLEX fuel bundles, components consisting of fuel bundles are more complicated. Hence, the detailed modeling of components is inevitable in order to analyze the fuel performance by computational fluid dynamics. In this report, the basic design of the advanced fuel for CANDU reactors was carried out and the methodology for the modeling of fuel bundle were described. Firstly, the components consisting of fuel bundles were separately modeled and saved with different file names. The final feature of fuel bundle was accomplished by an assembling process of components. Since this report developed the modeling methodology based on the Unigraphics program, the basic explanations for the software were given first, and the complete modeling of 37-elements and CANFLEX fuel bundles were provided. The components of CANFLEX fuel bundles were also compared with that of 37-elements fuel bundles. Although, in this report, the modeling methodology is applied only to 37-elements and CANFLEX fuel bundles, this methodology may be applicable to the newly designed fuel bundles which are to be developed in the future
Study on thermo-hydraulic behavior during reflood phase of a PWR-LOCA
International Nuclear Information System (INIS)
This paper describes thermo-hydraulic behavior during the reflood phase in a postulated large-break loss-of-coolant accident (LOCA) of a PWR. In order to better predict the reflood transient in a nuclear safety analysis specific analytical models have been developed for, saturated film boiling heat transfer in inverted slung flow, the effect of grid spacers on core thermo-hydraulics, overall system thermo-hydraulic behavior, and the thermal response similarity between nuclear fuel rods and simulated rods. A heat transfer correlation has been newly developed for saturated film boiling based on a 4 x 4-rod experiment conducted at JAERI. The correlation provides a good agreement with existing experiments except in the vicinity of grid spacer locations. An analytical model has then been developed addressing the effect of grid spacers. The thermo-hydraulic behavior near the grid spacers was found to be predicted well with this model by considering the breakup of droplets in dispersed flow and water accumulation above the grid spacers in inverted slung flow. A system analysis code has been developed which couples the one-dimensional core and multi-loop primary system component models. It provides fairly good agreement with system behavior obtained in a large-scale integral reflood experiment with active primary system components. An analytical model for the radial temperature distribution in a rod has been developed and verified with data from existing experiments. It was found that a nuclear fuel rod has a lower cladding temperature and an earlier quench time than an electrically heated rod in a typical reflood condition. (author)
International Nuclear Information System (INIS)
Reduced instrument responses are presented for Thermal-Hyraulic Test Facility (THTF) Test 3.06.6B. This test was conducted by members of the Oak Ridge National Laboratory Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on August 29, 1980. The objective of the program was to investigate heat transfer phenomena believed to occur in PWR's during accidents, including small and large break loss-of-coolant accidents. Test 3.06.6B was conducted to obtain transient film boiling data in rod bundle geometry under reactor accident-type conditions. The primary purpose of this report is to make the reduced instrument responses for THTF Test 3.06.6B available. Included in the report are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers
FLICA III M - reactors or test loops thermohydraulic computer code
International Nuclear Information System (INIS)
The FLICA III M code issued from the FLICA III code, of which it is the present stage of the development. This program calculates the flow and their transfer in steady and transient state in complex geometry described by subchannels. It is particularly used for the thermal-hydraulic analysis of reactors and experimental loops with heating rod bundles. A new solution method for the hydraulic problem is developed. It gives short computer times and allows the calculation of large subchannels sets. This makes possible the detailed calculations of hot subchannels jointly with these of rod bundles set in powered reactor cores. The equations solved take into account all the significant terms of the fundamental thermal-hydraulic equations and present models for turbulence and two-phase flows. The solution method couples together all the physical variables and makes possible the detailed description of complex flows
International Nuclear Information System (INIS)
This patent describes a method of forming a fuel bundle of a nuclear reactor. The method consists of positioning the fuel rods in the bottom plate, positioning the tie rod in the bottom plate with the key passed through the receptacle to the underside of the bottom plate and, after the tie rod is so positioned, turning the tie rod so that the key is in engagement with the underside of the bottom plate. Thereafter mounting the top plate is mounted in engagement with the fuel rods with the upper end of the tie rod extending through the opening in the top plate and extending above the top plate, and the tie rod is secured to the upper side of sid top plate thus simultaneously securing the key to the underside of the bottom plate
International Nuclear Information System (INIS)
An investigation of the effects of the radial core power profile on the thermo-hydraulic behavior during the reflood phase in a PWR-LOCA has been conducted with the Slab Core Test Facility (SCTF). Since the power in an actual PWR is lower in the peripheral bundles than in the central bundles, the so called chimney effects due to radial core power profile are expected to improve the cooling of the higher power bundles. The SCTF simulates a full radius slab section of a PWR and therefore the effects of radial core power profile can be investigated. The revealed results of four forced-feed reflood tests in the SCTF are; (1) even with different radial core power profiles, flat distribution of the collapsed water level in the core are obtained for each test; (2) in the highest power bundle under the same total core power, steeper radial power profile gives higher heat transfer coefficient; and (3) redistribution of flow or cross flow between bundles is considered to be a major reason for the results described above. (author)
Characteristics of liquid and boiling sodium flows in heating pin bundles
International Nuclear Information System (INIS)
This study is related to cooling accidents which could occur in sodium cooled fast reactors. Thermo-hydraulic aspects of boiling experiments in pin bundles with helical wire-wrap spacer systems, in the case of undamaged geometries, are analyzed. Differences and analogies in the behavior of multi-rod bundle flows and one-dimensional channel flows are studied. A boiling model is developed for bundle geometries, and predictions obtained with the FLICA code using this models are presented. These predictions are compared with experimental results obtained in a water 19-rod bundle. Then, results of sodium boiling experiments through a 19-rod bundle are interpreted. Both cases of high power and reduced power are envisaged
Institute of Scientific and Technical Information of China (English)
郭小江; 赵丽莉; 汤奕; 申洪
2013-01-01
风火打捆交直流外送是未来千万千瓦级风电能源基地电力外送的重要方式之一，亟需掌握风火打捆交直流外送系统的稳定特性及其机理。文章分析了不同风电比例和不同直流控制方式下系统的暂态功角稳定特性，探讨了风电、火电和直流系统间的交互作用，并基于扩展等面积(extended equal-area criterion，EEAC)理论分析了送受端机组惯量对系统功角暂态稳定性的影响机理。结果表明：在受端电网为无穷大系统的场景下，系统暂态功角稳定性随风电比例增加而改善；在送受端机组惯量可比场景下，存在最优风电与火电配置比例，使得系统暂态功角稳定性最好。%For the 10 GW wind power bases in China, wind-thermal-bundled power transmission by AC/DC system is regarded as one of feasible ways, but the stability characteristic of this kind of transmission systems should be investigated. In the paper, the angle transient stability characteristics were studied in the scenarios with different wind power penetration and DC control modes through simulations, and the interaction among wind power, thermal power and DC systems was discussed. Besides, the effects of the inertias of sending and receiving end systems on the angle transient stability characteristics were studied based on extended equal-area criterion (EEAC). When the receiving end power grid is equated to an infinite system, the system angle transient stability is improved with the increase of wind power penetration, while the equivalent inertia of the sending end can be comparable to that of the receiving end, there is an optimal wind power penetration with which the system has the highest angle transient stability.
Annular burnout data from rod bundle experiments
International Nuclear Information System (INIS)
Burnout data for annular flow in a rod bundle are presented for both transient and steady-state conditions. Tests were performed at the Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility (THTF), a pressurized-water loop containing an electrically heated 64-rod bundle. The bundle configuration is typical of later generation pressurized-water reactors with 17 x 17 fuel arrays. Both axial and radial power profiles are flat. All experiments were carried out in upflow with subcooled inlet conditions, insuring accurate flow measurement. Conditions within the bundle were typical of those which could be encountered during a nuclear reactor loss-of-coolant accident. Level average fluid conditions within the test section were calculated using steady-state mass and energy conservation considerations for the steady-state tests and a transient, homogeneous, equilibrium computer code for the transient tests. Unlike tube dryout, burnout within a rod bundle does not necessarily occur at one distinct axial level. The location of individual rod dryout was determined by scanning rods axially and locating the position where rod superheat increased from approx. =0 to 30 K or greater. Thermocouple instrumentation within the bundle allows the location of dryout to be determined to within approximately +.5 cm for many of the tests
International Nuclear Information System (INIS)
The integral pressurized water reactor (PWR) concept, which incorporates the nuclear steam supply systems within the reactor vessel, is one of the innovative reactor types with high potential for near term deployment. An International Collaborative Standard Problem (ICSP) on Integral PWR Design, Natural Circulation Flow Stability and Thermohydraulic Coupling of Primary System and Containment during Accidents was established in 2010. Oregon State University, which made available the use of its experimental facility built to demonstrate the feasibility of the Multi-application Small Light Water Reactor (MASLWR) design, and sixteen institutes from seven Member States participated in this ICSP. The objective of the ICSP is to assess computer codes for reactor system design and safety analysis. This objective is achieved through the production of experimental data and computer code simulation of experiments. A loss of feedwater transient with subsequent automatic depressurization system blowdown and long term cooling was selected as the reference event since many different modes of natural circulation phenomena, including the coupling of primary system, high pressure containment and cooling pool are expected to occur during this transient. The power maneuvering transient is also tested to examine the stability of natural circulation during the single and two phase conditions. The ICSP was conducted in three phases: pre-test (with designed initial and boundary conditions established before the experiment was conducted), blind (with real initial and boundary conditions after the experiment was conducted) and open simulation (after the observation of real experimental data). Most advanced thermohydraulic system analysis codes such as TRACE, RELAPS and MARS have been assessed against experiments conducted at the MASLWR test facility. The ICSP has provided all participants with the opportunity to evaluate the strengths and weaknesses of their system codes in the transient
TRAWA, LWR Dynamic by Coupled Neutron Diffusion and Thermohydraulics Calculation
International Nuclear Information System (INIS)
1 - Description of problem or function: The purpose of the program is to study reactor dynamics in thermal water-cooled reactors. It treats the core as one or a few axially one-dimensional subregions. The two group neutron diffusion equations are solved simultaneously with the heat conduction equations and the two-phase hydraulic equations for one or more channels. Neither thermal nor hydraulic mixing appear between channels. Doppler, coolant density, coolant temperature, and soluble poison density feedbacks due to the thermo- hydraulics of the channels are described by using polynomial expansions for the group constants. The hydraulic circuit outside the reactor core consists of by-pass channels and risers with two- phase flow and of pump lines with incompressible flow. Various transients can be calculated by applying external disturbances. They can affect e.g. on movements of control rods, core inlet hydraulic conditions, system pressure or coefficients of neutronic shape function expansion between subregions. 2 - Method of solution: Nontrivial implicit methods are employed in the discretization of the equations to allow for sparse spatial mesh and flexible choice of time steps. The same spatial and temporal discretization is used for neutronics and thermohydraulics. 3 - Restrictions on the complexity of the problem: The dimensions of the program variable tables can easily be extended. Now the main dimensions are: 52 axial mesh points in core; 3 subregions; 10 axial regions with different fuel compositions; 7 radial mesh points in fuel rod; 6 delayed neutron groups; 6 coupled legs in pressure balance calculation; No flow reversals are allowed
International Nuclear Information System (INIS)
Highlights: • Transient thermal–hydraulic characteristics of PRHR HX were analyzed and evaluated. • Low-Re SST model with near-wall multi-sublayer grid was recommended for simulations. • New overall scaled separate effect IRWST & PRHR HX experiments were conducted. • Experimental data provided a benchmark for the validation of numerical results. • Comparison of Nu values indicated that vertical section had better heat transfer effect. - Abstract: The heat transfer effect of Passive Residual Heat Removal Heat Exchanger (PRHR HX) and buoyancy-induced flow in the In-containment Refueling Water Storage Tank (IRWST) are of great importance for the efficient and safe removal of the residual heat in the AP1000 reactor. Although some numerical studies have been conducted, only the standard k–ε model has been applied. Experimental validation of the simulation results was also not sufficient because of the lack of appropriate experimental data. In the present work, the applicability of different Reynolds Average Navier–Stokes (RANS) turbulence models and Large Eddy Simulation (LES) were examined, utilizing the commercial CFD software CFX 14.5. Further, two types of grids were built for the high/low-Reynolds turbulence models, and the y+ values as well as grids sensitivity were carefully analyzed. Meanwhile, overall scaled IRWST and PRHR HX models were built to simulate the thermal–hydraulic process in the residual heat removal accident, which was a new overall scaled separate effect IRWST&PRHR HX experiment. More than 150 thermocouples were utilized to measure the temperature in the key regions, and Particle Image Velocimetry (PIV) was utilized for the measurement of the flow velocity. Based on the validation of turbulence models in simulating the overall variations of temperature and velocity field in the IRWST model, the transient heat transfer capacity of PRHR HX was then analyzed. The results indicated that the low-Reynolds Shear Stress Transport (SST
Energy Technology Data Exchange (ETDEWEB)
Thieme, M.; Tietsch, W. [Westinghouse Electric Germany (Germany); Sanchez, V.H. [Karlsruhe Institute of Technology, Karlsruhe (Germany). Inst. fuer Reaktorsicherheit
2008-07-01
A basic requirement for the use of thermohydraulic codes is the sufficient qualification of the code, which has to be accomplished for the respective fields of application. Depending on the problem definition, codes like RELAP5/MOD3.3 (reactor loss of coolant analysis program), ATHLET (analysis of thermal-hydraulics of leaks and transients), CATHARE (code advance de thermohydraulique pour accidents de reacteur a eau, ''advanced thermohydraulic code for the simulation of accidents in light water reactors'') or TRAC B/P (transient reactor analysis code) are currently used for safety evaluation of nuclear power plants (NPP). They are well verified on the basis of extensive international verification programs. Some years ago, the U.S.-NRC has started the development of a new program system called TRACE (TRAC/RELAP advanced computational engine). TRACE is a result of long-term efforts to combine the capabilities of the NRC's four main system codes (TRAC-P, TRAC-B, RELAP5 and RAMONA) into one modernised computational tool. However, comprehensive qualification and verification efforts are still needed to validate TRACE for industrial use as well as for safety evaluations of NPP [4], [5]. Therefore, the goal of the U.S.-NRC is to improve TRACE in order to qualify it so that it will be accepted in the industry. The aim of this work is to contribute to validate the heat transfer models of the CHAN component in TRACE. For this purpose selected experimental BFBT (BWR Full-Size Fine-Mesh Bundle Tests) [1] void fraction as well as critical power steady state and transient tests of the NUPEC (Nuclear Power Engineering Cooperation) database were simulated with TRACE Version 5 RC 2. These NUPEC experiments contain unique detailed data measured not only as bundle averaged but also in a very detailed, spatial resolution up to 0.3 mm. Hence, these tests are very appropriate for such model validation. In this compact, the results of the comparison between code
International Nuclear Information System (INIS)
The increase of bundle supply has become widespread in several sectors (for instance in telecommunications and energy fields). This paper review relates strategic aspects of bundling. The main purpose of this paper is to analyze profitability of bundling strategies according to the degree of competition and the characteristics of goods. Moreover, bundling can be used as price discrimination tool, screening device or entry barriers. In monopoly case bundling strategy is efficient to sort consumers in different categories in order to capture a maximum of surplus. However, when competition increases, the profitability on bundling strategies depends on correlation of consumers' reservations values. (author)
Murray, Michael K; Stevenson, Danny; Vozzo, Raymond F
2015-01-01
We develop the theory of simplicial extensions for bundle gerbes and their characteristic classes. This formalism is used to study descent problems and equivariance for bundle gerbes. We consider in detail two examples: the basic bundle gerbe on a unitary group and a string structure for a principal bundle. We show that the basic bundle gerbe is equivariant for the conjugation action and calculate its characteristic class and that a string structure gives rise to a bundle gerbe which is equivariant for a natural action of the String 2-group.
Thermohydraulic and nuclear modeling of natural fission reactors
Viggato, Jason Charles
Experimental verification of proposed nuclear waste storage schemes in geologic repositories is not possible, however, a natural analog exists in the form of ancient natural reactors that existed in uranium-rich ores. Two billion years ago, the enrichment of natural uranium was high enough to allow a sustained chain reaction in the presence of water as a moderator. Several natural reactors occurred in Gabon, Africa and were discovered in the early 1970's. These reactors operated at low power levels for hundreds of thousands of years. Heated water generated from the reactors also leached uranium from the surrounding rock strata and deposited it in the reactor cores. This increased the concentration of uranium in the core over time and served to "refuel" the reactor. This has strong implications in the design of modern geologic repositories for spent nuclear fuel. The possibility of accidental fission events in man-made repositories exists and the geologic evidence from Oklo suggests how those events may progress and enhance local concentrations of uranium. Based on a review of the literature, a comprehensive code was developed to model the thermohydraulic behavior and criticality conditions that may have existed in the Oklo reactor core. A two-dimensional numerical model that incorporates modeling of fluid flow, temperatures, and nuclear fission and subsequent heat generation was developed for the Oklo natural reactors. The operating temperatures ranged from about 456 K to about 721 K. Critical reactions were observed for a wide range of concentrations and porosity values (9 to 30 percent UO2 and 10 to 20 percent porosity). Periodic operation occurred in the computer model prediction with UO2 concentrations of 30 percent in the core and 5 percent in the surrounding material. For saturated conditions and 30 percent porosity, the model predicted temperature transients with a period of about 5 hours. Kuroda predicted 3 to 4 hour durations for temperature transients
CASSANDRE, 2-D Reactor Dynamic FEM Program with Thermohydraulic Feedback
International Nuclear Information System (INIS)
1 - Description of program or function: CASSANDRE is a two-dimensional (x-y or r-z) finite-elements neutronics code with thermohydraulic feedback for reactor dynamics prior to the disassembly phase. The code was conceived in order to be coupled with any thermohydraulics module, although thermohydraulics feedback is only considered in r-z geometry. In the steady state, criticality search is possible either by control-rod insertion or by homogeneous poisoning of the coolant. 2 - Method of solution: The program uses multigroup diffusion theory. Its main characteristics are the use of a generalized quasi-static model, the use of a flexible multigroup point-kinetics algorithm allowing for spectral matching, and the use of a finite elements description. 3 - Restrictions on the complexity of the problem: The user must prepare a cross section library
Application of visual modelization technology to reactor thermohydraulics analysis programmes
International Nuclear Information System (INIS)
Visual modelization and XML was used to build a thermo-hydraulics security analysis model library for RELAP5 programme. The modeling system frame was constructed by the components transferred from this model library. Interfaces were described with thermo-hydraulics parameters based on windows platform. Every component was linked with the corresponding interface and input parameters, and RELAP data deck was created after checking errors. The analysis of a simple fluid flow in pipe verified that this technology could simplify the use of RELAP and improve the efficiency of study. (authors)
Observation and control system of the thermohydraulic assays laboratory
International Nuclear Information System (INIS)
The Thermohydraulic Assays Laboratory (L.E.T.) is an installation whose purpose will be the components testing and the CAREM-25 reactor thermohydraulic processes operation dynamics. This plant is located at Pilcaniyeu, province of Rio Negro. Part of the tests which will be carried out consist in the use of different control strategies. The control of the systems by digital processors (control by software) has been decided to proceed with a maximum flexibility and capacity to make changes in the algorithms. This work describes the design and implementation of a digital control system to command the three circuits of the installation. (Author)
Bundling in Telecommunications
Begoña García-Mariñoso; Xavier Martinez-Giralt; Pau Olivella
2008-01-01
The paper offers an overview of the literature on bundling in the telecommunications sector and its application in the Spanish market. We argue that the use of bundling in the provision of services is associated to technological reasons. Therefore, there appears no need to regulate bundling activities. However, this is not to say that other related telecom markets should not be scrutinized and regulated, or that the regulator should not pay attention to other bundling-related anticompetitive ...
Atucha I nuclear power plant transients analysis
International Nuclear Information System (INIS)
A program for the transients simulation thermohydraulic calculation without loss of coolant (KWU-ENACE development) to evaluate Atucha I nuclear power plant behaviour is used. The program includes systems simulation and nuclear power plants control bonds with real parameters. The calculation results show a good agreement with the output 'protocol' of various transients of the nuclear power plant, keeping the error, in general, lesser than ± 10% from the variation of the nuclear power plant's state variables. (Author)
Thermohydraulic modeling and simulation of breeder reactors
International Nuclear Information System (INIS)
This paper deals with the modeling and simulation of system-wide transients in LMFBRs. Unprotected events (i.e., the presumption of failure of the plant protection system) leading to core-melt are not considered in this paper. The existing computational capabilities in the area of protected transients in the US are noted. Various physical and numerical approximations that are made in these codes are discussed. Finally, the future direction in the area of model verification and improvements is discussed
The SABRE code for fuel rod cluster thermohydraulics
International Nuclear Information System (INIS)
This paper describes the capabilities of the SABRE code for the calculation of single phase and two phase fluid flow and temperature in fuel pin bundles, discusses the methods used in the modelling and solution of the problem, and presents some results including comparison with experiments. The SABRE code permits calculation of steady-state or transient, single or two phase flows and the geometrical options include general representation of grids, wire wraps, multiple blockages, bowed pins, etc. The derivation and solution of the difference equations is discussed. Emphasis is given to the derivation of the spatial differences in triangular subchannel geometry, and the use of central, upward or vector upwind schemes. The method of solution of the difference equations is described for both steady state and transient problems. Together with these topics we consider the problems involved in turbulence modelling and how it is implemented in SABRE. This includes supporting work with a fine scale curvilinear coordinate programme to provide turbulence source data. The problem of modelling boiling flows is discussed, with particular reference to the numerical problems caused by the rapid density change on boiling. The final part of the paper presents applications of the code to the analysis of blockage situations, the study of flow and power transients and analysis of natural circulation within clusters to demonstrate the scope of the code and compare with available experimental results. The comparisons include the calculation of a flow pressure drop characteristic of a boiling channel showing the Ledinegg instability, examples of overpower and flow rundown transients which lead to coolant boiling, and calculation of natural circulation within a rod cluster. (orig./GL)
Systems for neutronic, thermohydraulic and shielding calculation in personal computers
International Nuclear Information System (INIS)
The MTR-PC (Materials Testing Reactors-Personal Computers) system has been developed by the Nuclear Engineering Division of INVAP S.E. with the aim of providing working conditions integrated with personal computers for design and neutronic, thermohydraulic and shielding analysis for reactors employing plate type fuel. (Author)
Annular burnout data from rod-bundle experiments
International Nuclear Information System (INIS)
Burnout data for annular flow in a rod bundle are presented for both transient and steady-state conditions. Tests were performed at the Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility (THTF), a pressurized-water loop containing an electrically heated 64-rod bundle. The bundle configuration is typical of later generation pressurized-water reactors with 17 x 17 fuel arrays. Both axial and radial power profiles are flat. All experiments were carried out in upflow with subcooled inlet conditions, insuring accurate flow measurement. Conditions within the bundle were typical of those which could be encountered during a nuclear reactor loss-of-coolant accident
Nicholas Economides
2014-01-01
We discuss strategic ways that sellers can use tying and bundling with requirement conditions to extract consumer surplus. We analyze different types of tying and bundling creating (i) intra-product price discrimination; (ii) intra-consumer price discrimination; and (iii) inter-product price discrimination, and assess the antitrust liability that these practices may entail. We also discuss the impact on consumers and competition, as well as potential antitrust liability of bundling “incontest...
Development of thermohydraulic codes for modeling liquid metal boiling in LMR fuel subassemblies
International Nuclear Information System (INIS)
An investigation into the reactor core accident cooling, which are associated with the power grow up or switch off circulation pumps in the event of the protective equipment comes into action, results in the problem of liquid metal boiling heat transfer. Considerable study has been given over the last 30 years to alkaline metal boiling including researches of heat transfer, boiling patterns, hydraulic resistance, crisis of heat transfer, initial heating up, boiling onset and instability of boiling. The results of these investigations have shown that the process of liquid metal boiling has substantial features in comparison with water boiling. Mathematical modeling of two phase flows in fast reactor fuel subassemblies have been developed intensively. Significant success has been achieved in formulation of two phase flow through the pin bundle and in their numerical realization. Currently a set of codes for thermohydraulic analysis of two phase flows in fast reactor subassembly have been developed with 3D macrotransfer governing equations. These codes are used for analysis of boiling onset and liquid metals boiling in fuel subassemblies during loss-of-coolant accidents, of warming up of reactor core, of blockage of some part of flow cross section in fuel subassembly. (author)
International Nuclear Information System (INIS)
Covering the wide range of reactor safety analysis of power reactors, consisting of leak and transients, the thermohydraulic code ATHLET is being developed by the German Society for Plant and Reactor Safety (GRS). In order to extend the application range of the code to the safety analysis of low and medium flux research reactors, a model was developed and implemented permitting a description of the steam formation in the subcooled boiling regime. Considering the specific features of high flux research reactors given by both high heat flux and high flow velocity, further extension to the model of void condensation in subcooled flow has been extended and a new correlation of critical heat flux (CHF) is implemented. To validate the extended Program, the Thermal Hydraulic Test Loop (THTL) of Oak ridge National Laboratory (ORNL) was modeled and an extensive series of experiments concerning the onset of thermohydraulic flow instability (OFI) in subcooled boiling regime were calculated. The comparison between experiments and ATHLET-postcalculation shows that the extended code can accurately simulate the thermohydraulic conditions of flow instability in a wide range of heat flux up to 15 MW/m2 and inlet flow velocity up to 20 m/s. The thermohydraulic design limit characterized by the mass flux, at which the flow just becomes unstable (OFI), has been predicted in very good agreement with the experiment. However the calculated pressure drop at OFI is overestimated by a maximum deviation of about 25%. The calculated exit bulk temperature of subcooled coolant and the maximum wall temperature at OFI show a maximum deviation from experiment of 12% and 7% respectively. The extended code has been applied successfully to simulate the flow reversal in the fuel element of German high flux research reactor FRM-II. This phenomenon is expected in case of shutdown pumps failure. The results show the code's capability to simulate the flow reversal from down ward to up ward direction
Lerman, Eugene
2003-01-01
We define contact fiber bundles and investigate conditions for the existence of contact structures on the total space of such a bundle. The results are analogous to minimal coupling in symplectic geometry. The two applications are construction of K-contact manifolds generalizing Yamazaki's fiber join construction and a cross-section theorem for contact moment maps
Principal noncommutative torus bundles
DEFF Research Database (Denmark)
Echterhoff, Siegfried; Nest, Ryszard; Oyono-Oyono, Herve
2008-01-01
of bivariant K-theory (denoted RKK-theory) due to Kasparov. Using earlier results of Echterhoff and Williams, we shall give a complete classification of principal non-commutative torus bundles up to equivariant Morita equivalence. We then study these bundles as topological fibrations (forgetting the...
International Nuclear Information System (INIS)
In order to study the power transients effects on PWR fuel rod clad, ramp tests in a pressurised water loop, are carried out at OSIRIS reactor. The present thesis deals with the on-line control of the device, during power ramp and conditioning irradiation. Based on a convolution-type resolution of the kinetics equations, a dynamic compensation of the Silver self-powered neutron detector was developed. With this method, the uncertainty of the ramp end-point is lower than 1%, thus it is very suited for monitoring both transient, as well as steady state conditions. Furthermore, a thermohydraulic model of the irradiation device is described: heat transfer equations, including gamma heating in materials, are solved to obtain temperatures and thermal fluxes of steady states. Results from the model and temperature measurements of the coolant are used together for fuel power determination, in real time. The clad external temperature profile also calculated and displayed, to improve the irradiation monitoring
Thermo-Hydraulic Modelling of Buffer and Backfill
International Nuclear Information System (INIS)
The temporal evolution of saturation, liquid pressure and temperature in the components of the engineered barrier system was studied using numerical methods. A set of laboratory tests was conducted to calibrate the parameters employed in the models. The modelling consisted of thermal, hydraulic and thermo-hydraulic analysis in which the significant thermo-hydraulic processes, parameters and features were identified. CODEBRIGHT was used for the finite element modelling and supplementary calculations were conducted with analytical methods. The main objective in this report is to improve understanding of the thermo-hydraulic processes and material properties that affect buffer behaviour in the Olkiluoto repository and to determine the parametric requirements of models for the accurate prediction of this behaviour. The analyses consisted of evaluating the influence of initial canister temperature and gaps in the buffer, and the role played by fractures and the rock mass located between fractures in supplying water for buffer and backfill saturation. In the thermo-hydraulic analysis, the primary processes examined were the effects of buffer drying near the canister on temperature evolution and the manner in which heat flow affects the buffer saturation process. Uncertainties in parameters and variations in the boundary conditions, modelling geometry and thermo-hydraulic phenomena were assessed with a sensitivity analysis. The material parameters, constitutive models, and assumptions made were carefully selected for all the modelling cases. The reference parameters selected for the simulations were compared and evaluated against laboratory measurements. The modelling results highlight the importance of understanding groundwater flow through the rock mass and from fractures in the rock in order to achieve reliable predictions regarding buffer saturation, since saturation times could range from a few years to tens of thousands of years depending on the hydrogeological
Modelling of thermohydraulic emergency core cooling phenomena
International Nuclear Information System (INIS)
The codes used in the early seventies for safety analysis and licensing were based either on the homogeneous model of two-phase flow or on the so-called separate-flow models, which are mixture models accounting, however, for the difference in average velocity between the two phases. In both cases the behavior of the mixture is prescribed a priori as a function of local parameters such as the mass flux and the quality. The modern best-estimate codes used for analyzing LWR LOCA's and transients are often based on a two-fluid or 6-equation formulation of the conservation equations. In this case the conservation equations are written separately for each phase; the mixture is allowed to evolve on its own, governed by the interfacial exchanges of mass, momentum and energy between the phases. It is generally agreed that such relatively sophisticated 6-equation formulations of two-phase flow are necessary for the correct modelling of a number of phenomena and situations arising in LWR accidental situations. They are in particular indispensible for the analysis of stratified or countercurrent flows and of situations in which large departures from thermal and velocity equilibrium exist. This report will be devoted to a discussion of the need for, the capacity and the limitations of the two-phase flow models (with emphasis on the 6-equation formulations) in modelling these two-phase flow and heat transfer phenomena and/or different core cooling situations. 18 figs., 1 tab., 72 refs
International Nuclear Information System (INIS)
An investigation of the effects of the radial core power profile on the thermo-hydraulic behavior during the reflood phase of the large break LOCA of a PWR has been conducted with the Slab Core Test Facility (SCTF). Since the power in an actual PWR is lower in the peripheral bundles than in the central bundles, the so called chimney effect due to radial core power profile is expected to improve the cooling of the higher power bundles. The SCTF simulates a full radius slab section of a PWR and therefore the effects of radial core power profile can be investigated. The revealed results obtained from four forced-feed reflood tests (S1-01, S1-06, S1-08 and S1-11) in the SCTF Core-I are; (1) Two-dimensional flow in the core was induced by the radial power distribution. The direction of cross flow was from the central high power region to the peripheral low power region above the quench front and the direction was reversed below the quench front. (2) The heat transfer coefficient at the highest power bundle of the steep power profile test was higher than that of the flat power profile test under the same total core power condition. (author)
Restrictions of stable bundles
Balaji, V
2011-01-01
The Mehta-Ramanathan theorem ensures that the restriction of a stable vector bundle to a sufficiently high degree complete intersection curve is again stable. We improve the bounds for the "sufficiently high degree" and propose a possibly optimal conjecture.
International Nuclear Information System (INIS)
This paper describes works on CANDU fuel bundle fabrication in the Fuel Fabrication Development and Testing Section (FFDT) of AECL's Chalk River Laboratories. This work does not cover fuel design, pellet manufacturing, Zircaloy material manufacturing, but cover the joining of appendages to sheath tube, endcap preparation and welding, UO2 loading, end plate preparation and welding, and all inspections required in these steps. Materials used in the fabrication of CANDU fuel bundle are: 1)Ceramic UO2 Pellet 2)Zircaloy -4. Fuel Bundle Structural Material 3) Others (Zinc stearate, Colloidal graphite, Beryllium and Heium). Th fabrication of fuel element consist of three process: 1)pellet loading into the sheats, 2) endcap welding, and 3) the element profiling. Endcap welds is tested by metallography and He leak test. The endcaps of the elements are welded to the end plates to form the 37- element bundle assembly
International Nuclear Information System (INIS)
This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation at higher power densities with increased void reactivity feedback and decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of 'unexpected' instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, BWR instabilities has been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a 'new and improved' state of the art has emerged recently. (authors). 6 figs., 57 refs., 1 appendix
Subtleties Concerning Conformal Tractor Bundles
Graham, C Robin
2012-01-01
The realization of tractor bundles as associated bundles in conformal geometry is studied. It is shown that different natural choices of principal bundle with normal Cartan connection corresponding to a given conformal manifold can give rise to topologically distinct associated tractor bundles for the same inducing representation. Consequences for homogeneous models and conformal holonomy are described. A careful presentation is made of background material concerning standard tractor bundles and equivalence between parabolic geometries and underlying structures.
Comparison between new thermohydraulic one-channel models and experiments
Blender, H.; Elzmann, J.
1981-11-01
Five different thermohydraulic one-channel models, COCHA, FRANCESCA, MARMITA, STASWR and THS, were tested bu experimentally checking two-phase flows along a boiling water reactor fuel element. As regards the evolution of the vapor content along the cooling channel, the agreement between all the programs and the measurements is satisfactory for small to middle entrance undercooling in the domain of undercooled boiling. For high undercooling, only the COCHA program gives satisfactory results. For the middle part of the cooling channel, all programs are satisfactory, while in the upper part, especially for increasing outlet vapor contents, the calculated values are generally too low for all programs, and especially for FRANCESCA.
Numerical simulation of flow-induced vibrations in tube bundles
International Nuclear Information System (INIS)
Full text of publication follows: In many industrial components mechanical structures like rod cluster control assembly, fuel assembly and heat exchanger tube bundles are submitted to complex flows causing possible vibrations and damage. Fluid forces are usually split into two parts: structure motion independent forces and fluid-elastic forces coupled with tube motion and responsible for possible dynamic instability development leading to possible short term failures through high amplitude vibrations. Most classical fluid force identification methods rely on structure response experimental measurements associated with convenient data processes. Owing to recent improvements in Computational Fluid Dynamics (C.F.D.), numerical fluid force identification is now practicable in the presence of industrial configurations. The present paper is devoted to numerical simulation of flow-induced vibrations of tube bundles submitted to single-phase cross flows by using C.F.D. codes. Direct Numerical Simulation (D.N.S.), Arbitrary Lagrange Euler formulation (A.L.E.) and code coupling process are involved to predict fluid forces responsible for tube bundle vibrations in the presence of fluid structure and fluid-elastic coupling effects. In the presence of strong multi-physics coupling, simulation of flow-induced vibrations requires a fluid structure code coupling process. The methodology consists in solving in the same time thermohydraulics and mechanics problems by using an A.L.E. formulation for the fluid computation. The purpose is to take into account coupling between flow and structure motions in order to be able to capture coupling effects. From a numerical point of view, there are three steps in the computation: the fluid problem is solved on the computational domain; fluid forces acting on the moving tube are estimated; finally they are introduced in the structure solver providing the tube displacement that is used to actualize the fluid computational domain. Specific
Energy Technology Data Exchange (ETDEWEB)
Gomez Garcia-Torano, I.; Jimenez, G.
2013-07-01
The Thermo-hydraulic code GOTHIC is often used in the nuclear industry for licensing transient analysis inside containment of generation II (PWR, BWR) plants as Gen III and III + (AP1000, ESBWR, APWR). After entering the mass and energy released to the containment, previously calculated by other codes (basis, TRACE), GOTHIC allows to calculate in detail the evolution of basic parameters in the containment.
DEFF Research Database (Denmark)
Bussink, Barbara E; Holst, Anders Gaarsdal; Jespersen, Lasse;
2013-01-01
AimsTo determine the prevalence, predictors of newly acquired, and the prognostic value of right bundle branch block (RBBB) and incomplete RBBB (IRBBB) on a resting 12-lead electrocardiogram in men and women from the general population.Methods and resultsWe followed 18 441 participants included in...... men vs. 0.5%/2.3% in women, P <0.001). Significant predictors of newly acquired RBBB were male gender, increasing age, high systolic blood pressure, and presence of IRBBB, whereas predictors of newly acquired IRBBB were male gender, increasing age, and low BMI. Right bundle branch block was associated...... with significantly increased all-cause and cardiovascular mortality in both genders with age-adjusted hazard ratios (HR) of 1.31 [95% confidence interval (CI), 1.11-1.54] and 1.87 (95% CI, 1.48-2.36) in the gender pooled analysis with little attenuation after multiple adjustment. Right bundle branch...
Principal -bundles on Nodal Curves
Indian Academy of Sciences (India)
Usha N Bhosle
2001-08-01
Let be a connected semisimple affine algebraic group defined over . We study the relation between stable, semistable -bundles on a nodal curve and representations of the fundamental group of . This study is done by extending the notion of (generalized) parabolic vector bundles to principal -bundles on the desingularization of and using the correspondence between them and principal -bundles on . We give an isomorphism of the stack of generalized parabolic bundles on with a quotient stack associated to loop groups. We show that if is simple and simply connected then the Picard group of the stack of principal -bundles on is isomorphic to ⊕ , being the number of components of .
Study of Transients in an Enrichment Closed Loop
Fernandino, M
2002-01-01
In the present thesis a mathematic model is presented in order to describe the dynamic behavior inside a closed enrichment loop, the latter representing a single stage of an uranium gaseous diffusion enrichment cascade.The analytical model is turned into a numerical model, and implemented through a computational code.For the verification of the model, measurements were taken in an experimental circuit using air as the process fluid.This circuit was instrumented so as to register its characteristic thermohydraulic variables.The measured transients were simulated, comparing the numerical results with the experimental measurements.A good agreement between the characteristic setting times and the thermohydraulic parameters evolution was observed.Besides, other transients of two species separation were numerically analyzed, including setting times of each magnitude, behavior of each one of them during different transients, and redistribution of concentrations.
Study of Transients in an Enrichment Closed Loop
International Nuclear Information System (INIS)
In the present thesis a mathematic model is presented in order to describe the dynamic behavior inside a closed enrichment loop, the latter representing a single stage of an uranium gaseous diffusion enrichment cascade.The analytical model is turned into a numerical model, and implemented through a computational code.For the verification of the model, measurements were taken in an experimental circuit using air as the process fluid.This circuit was instrumented so as to register its characteristic thermohydraulic variables.The measured transients were simulated, comparing the numerical results with the experimental measurements.A good agreement between the characteristic setting times and the thermohydraulic parameters evolution was observed.Besides, other transients of two species separation were numerically analyzed, including setting times of each magnitude, behavior of each one of them during different transients, and redistribution of concentrations
Thermo-Hydraulic behaviour of dual-channel superconducting Cable-In-Conduit Conductors for ITER
International Nuclear Information System (INIS)
In an effort to optimise the cryogenics of large superconducting coils for fusion applications (ITER), dual channel Cable-In-Conduit Conductors (CICC) are designed with a central channel spiral to provide low hydraulic resistance and faster helium circulation. The qualitative and economic rationale of the conductor central channel is here justified to limit the superconductor temperature increase, but brings more complexity to the conductor cooling characteristics. The pressure drop of spirals is experimentally evaluated in nitrogen and water and an explicit hydraulic friction model is proposed. Temperatures in the cable must be quantified to guarantee superconductor margin during coil operation under heat disturbance and set adequate inlet temperature. Analytical one-dimensional thermal models, in steady state and in transient, allow to better understand the thermal coupling of CICC central and annular channels. The measurement of a heat transfer characteristic space and time constants provides cross-checking experimental estimations of the internal thermal homogenization. A simple explicit model of global inter-channel heat exchange coefficient is proposed. The risk of thermosyphon between the two channels is considered since vertical portions of fusion coils are subject to gravity. The new hydraulic model, heat exchange model and gravitational risk ratio allow the thermohydraulic improvement of CICC central spirals. (author)
International Nuclear Information System (INIS)
The operating reliability of nuclear reactors calls for a reliable strength analysis of the highly loaded core elements, one of its prerequisites being the reliable determination of the three-dimensional velocity and temperature fields. To verify thermohydraulics computer programs, extensive local temperature measurements in the rod claddings of the critical bundle zone were performed on a heated 19-rod bundle model with sodium flow and provided with spacer grids (P/D = 1.30; W/D = 1.19). The essential results are: - Outside the spacer grids, the azimuthal temperature variations of the side and corner rods are approximately 10-fold those of rods in the central bundle zone. - The spacer grids investigated give rise to great local temperature peaks and correspondingly great temperature gradients in the axial and azimuthal directions immediately around the support points. - Continuous reduction of a subchannel by rod bowing results in substantial rises of temperature which, however, are limited to adjacent cladding tubes. (orig.)
Nuclear reactors transients identification and classification system
International Nuclear Information System (INIS)
This work describes the study and test of a system capable to identify and classify transients in thermo-hydraulic systems, using a neural network technique of the self-organizing maps (SOM) type, with the objective of implanting it on the new generations of nuclear reactors. The technique developed in this work consists on the use of multiple networks to do the classification and identification of the transient states, being each network a specialist at one respective transient of the system, that compete with each other using the quantization error, that is a measure given by this type of neural network. This technique showed very promising characteristics that allow the development of new functionalities in future projects. One of these characteristics consists on the potential of each network, besides responding what transient is in course, could give additional information about that transient. (author)
International Nuclear Information System (INIS)
A hybrid bundle divertor design is presented that produces <0.3% magnetic ripple at the center of the plasma while providing adequate space for the coil shielding and structure for a tokamak fusion test reactor similar to the International Tokamak Reactor and the Engineering Test Facility (with R = 5 m, B = 5 T, and a /SUB wall/ = 1.5 m, in particular). This hybrid divertor consists of a set of quadrupole ''wing'' coils running tangent to the tokamak plasma on either side of a bundle divertor. The wing coils by themselves pull the edge of the plasma out 1.5 m and spread the thickness of the scrape-off layer from 0.1 to 0.7 m at the midplane. The clear aperture of the bundle divertor throat is 1.0 m high and 1.8 m wide. For maintenance or replacement, the hybrid divertor can be disassembled into three parts, with the bundle divertor part pulling straight out between toroidal field coils and the wing coils then sliding out through the same opening
Dynamics of flagellar bundling
Janssen, Pieter; Graham, Michael
2010-11-01
Flagella are long thin appendages of microscopic organisms used for propulsion in low-Reynolds environments. For E. coli the flagella are driven by a molecular motor, which rotates the flagella in a counter-clockwise motion (CCM). When in a forward swimming motion, all flagella bundle up. If a motor reverses rotation direction, the flagella unbundle and the cell makes a tumbling motion. When all motors turn in the same CC direction again, the flagella bundle up, and forward swimming continues. To investigate the bundling, we consider two flexible helices next to each other, as well as several flagella attached to a spherical body. Each helix is modeled as several prolate spheroids connected at the tips by springs. For hydrodynamic interactions, we consider the flagella to made up of point forces, while the finite size of the body is incorporated via Fax'en's laws. We show that synchronization occurs quickly relative to the bundling process. For flagella next to each other, the initial deflection is generated by rotlet interactions generated by the rotating helices. At longer times, simulations show the flagella only wrap once around each other, but only for flagella that are closer than about 4 helix radii. Finally, we show a run-and-tumble motion of the body with attached flagella.
Rod bundle burnout data and correlation comparisons
International Nuclear Information System (INIS)
Rod bundle burnout data from 30 steady-state and 3 transient tests were obtained from experiments performed in the Thermal Hydraulic Test Facility at the Oak Ridge National Laboratory. The tests covered a parameter range relevant to intact core reactor accidents ranging from large break to small break loss-ofcoolant conditions. Instrumentation within the 64-rod test section indicated that burnout occurred over an axial range within the bundle. The distance from the point where the first dry rod was detected to the point where all rods were dry was up to 60 cm in some of the tests. The burnout data should prove useful in developing new correlations for use in reactor thermalhydraulic codes. Evaluation of several existing critical heat flux correlations using the data show that three correlations, the Barnett, Bowring, and Katto correlations, perform similarly and correlate the data better than the Biasi correlation
Numerical model for thermal and mechanical behaviour of a CANDU 37-element bundle
International Nuclear Information System (INIS)
Prediction of transient fuel bundle deformations is important for assessing the integrity of fuel and the surrounding structural components under different operating conditions including accidents. For numerical simulation of the interactions between fuel bundle and pressure tube, a reliable numerical bundle model is required to predict thermal and mechanical behaviour of the fuel bundle assembly under different thermal loading conditions. To ensure realistic representations of the bundle behaviour, this model must include all of the important thermal and mechanical features of the fuel bundle, such as temperature-dependent material properties, thermal viscoplastic deformation in sheath, fuel-to-sheath interactions, endplate constraints and contacts between fuel elements. In this paper, we present a finite element based numerical model for predicting macroscopic transient thermal-mechanical behaviour of a complete 37-element CANDU nuclear fuel bundle under accident conditions and demonstrate its potential for being used to investigate fuel bundle to pressure tube interaction in future nuclear safety analyses. This bundle model has been validated against available experimental and numerical solutions and applied to various simulations involving steady-state and transient loading conditions. (author)
On framed quantum principal bundles
Durdevic, M
1995-01-01
A noncommutative-geometric formalism of framed principal bundles is sketched, in a special case of quantum bundles (over quantum spaces) possessing classical structure groups. Quantum counterparts of torsion operators and Levi-Civita type connections are analyzed. A construction of a natural differential calculus on framed bundles is described. Illustrative examples are presented.
Thermohydraulics of a horizontal diphasic flow of superfluid helium
International Nuclear Information System (INIS)
This study aims at characterizing helium two phase flows, and to identify the dependence of their characteristics on various thermo-hydraulic parameters: vapour velocity, liquid height, vapour density, specificities of superfluidity. Both the engineer and the physicist's points of view are taken into consideration: the first one in terms of optimization of a particular cooling scheme based on a two-phase flow, and these second one in terms of more fundamental atomization-related questions. It has been shown that for velocities around 3 to 4 m/s, the liquid phase that was initially stratified undergoes an atomization through the presence of a drop haze carried by the vapor phase.This happens for superfluid helium as well as for normal helium without main differences on atomization
Development of thermohydraulics computer programs for thermal striping phenomena
International Nuclear Information System (INIS)
Two thermohydraulics computer programs AQUA and DINUS-3, which are represented by both time- and volume-averaged transport analysis and direct numerical simulation of turbulence, respectively. were developed and validated for the evaluation of thermal striping phenomena. These codes were incorporated with higher-order difference schemes to approximate the convection terms in conservation equations and adaptive time step size control systems based on the Fuzzy theory to eliminate numerical instabilities. From validation analyses with fundamental experiments in water and sodium, it was concluded that (1) thermal striping conditions such as spatial distributions of the intensity and the frequency of the fluid temperature fluctuations can be estimated efficiently by a combined approach incorporating the AQUA code and the DINUS-3 code, and (2) the thermal striping phenomena for the in-vessel components of actual liquid metal-cooled fast reactors can be evaluated by the numerical method without conventional approaches such as large scale model experiments using sodium. (author)
Thermohydraulic stability coupled to the neutronic in a BWR
International Nuclear Information System (INIS)
In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the Laguna Verde
Thermohydraulic assessment of the RP-10 reactor core to determine the maximum power
International Nuclear Information System (INIS)
Thermohydraulic parameters assessment of the RP-10 reactor core from the most thermally demanded (hot channel). Determination of the operation thermal maximum power considering security margins and statistical treatment of uncertainty factors
Thermohydraulic investigation in justification of fast reactor core with liquid metal coolants
International Nuclear Information System (INIS)
Thermohydraulic analysis of fast reactor (FR) cores is a component of the complex of interrelated problems on FR parameters justification. These problems concern reactor physics, thermal mechanics, failure theory and other components. Thermohydraulic analysis includes determination of temperature mode of core elements, hydraulic characteristics of fuel assemblies, their nominal and maximal capacitance. The present state of the art of problem-oriented, pilot and applied thermohydraulic studies of FR cores with liquid metal coolants (LMC) is under consideration. The experimental and calculating data on thermal hydraulics of LMC reactors gathered in the Institute for Physics and Power Engineering is analyzed and generalized when comparing different calculating procedures, problem solution techniques. The methods and codes of numerical simulation of thermohydraulic processes in LMC FR core are considered. The problems and aims of further investigations are formulated
Directory of Open Access Journals (Sweden)
D. A. Robbins
1994-12-01
Full Text Available We study bundles of Banach algebras ÃÂ€:AÃ¢Â†Â’X, where each fiber Ax=ÃÂ€Ã¢ÂˆÂ’1({x} is a Banach algebra and X is a compact Hausdorff space. In the case where all fibers are commutative, we investigate how the Gelfand representation of the section space algebra ÃŽÂ“(ÃÂ€ relates to the Gelfand representation of the fibers. In the general case, we investigate how adjoining an identity to the bundle ÃÂ€:AÃ¢Â†Â’X relates to the standard adjunction of identities to the fibers.
Rudakov, A N
1990-01-01
This volume is devoted to the use of helices as a method for studying exceptional vector bundles, an important and natural concept in algebraic geometry. The work arises out of a series of seminars organised in Moscow by A. N. Rudakov. The first article sets up the general machinery, and later ones explore its use in various contexts. As to be expected, the approach is concrete; the theory is considered for quadrics, ruled surfaces, K3 surfaces and P3(C).
Hirsch, Gregory
2002-01-01
A plurality of glass or metal wires are precisely etched to form the desired shape of the individual channels of the final polycapillary optic. This shape is created by carefully controlling the withdrawal speed of a group of wires from an etchant bath. The etched wires undergo a subsequent operation to create an extremely smooth surface. This surface is coated with a layer of material which is selected to maximize the reflectivity of the radiation being used. This reflective surface may be a single layer of material, or a multilayer coating for optimizing the reflectivity in a narrower wavelength interval. The collection of individual wires is assembled into a close-packed multi-wire bundle, and the wires are bonded together in a manner which preserves the close-pack configuration, irrespective of the local wire diameter. The initial wires are then removed by either a chemical etching procedure or mechanical force. In the case of chemical etching, the bundle is generally segmented by cutting a series of etching slots. Prior to removing the wire, the capillary array is typically bonded to a support substrate. The result of the process is a bundle of precisely oriented radiation-reflecting hollow channels. The capillary optic is used for efficiently collecting and redirecting the radiation from a source of radiation which could be the anode of an x-ray tube, a plasma source, the fluorescent radiation from an electron microprobe, a synchrotron radiation source, a reactor or spallation source of neutrons, or some other source.
Bundling harvester; Nippukorjausharvesteri
Energy Technology Data Exchange (ETDEWEB)
Koponen, K. [Eko-Log Oy, Kuopio (Finland)
1996-12-31
The staring point of the project was to design and construct, by taking the silvicultural point of view into account, a harvesting and processing system especially for energy-wood, containing manually driven bundling harvester, automatizing of the harvester, and automatized loading. The equipment forms an ideal method for entrepreneur`s-line harvesting. The target is to apply the system also for owner`s-line harvesting. The profitability of the system promotes the utilization of the system in both cases. The objectives of the project were: to construct a test equipment and prototypes for all the project stages, to carry out terrain and strain tests in order to examine the usability and durability, as well as the capacity of the machine, to test the applicability of the Eko-Log system in simultaneous harvesting of energy and pulp woods, and to start the marketing and manufacturing of the products. The basic problems of the construction of the bundling harvester have been solved using terrain-tests. The prototype machine has been shown to be operable. Loading of the bundles to form sufficiently economically transportable loads has been studied, and simultaneously, the branch-biomass has been tried to be utilized without loosing the profitability of transportation. The results have been promising, and will promote the profitable utilization of wood-energy
DEFF Research Database (Denmark)
Sommer, Stefan Horst; Lauze, Francois Bernard; Nielsen, Mads; Pennec, Xavier
information to be automatically incorporated in registrations and promises to improve the standard framework in several aspects. We present the mathematical foundations of LDDKBM and derive the KB-EPDiff evolution equations, which provide optimal warps in this new framework. To illustrate the resulting......In the LDDMM framework, optimal warps for image registration are found as end-points of critical paths for an energy functional, and the EPDiff equations describe the evolution along such paths. The Large Deformation Diffeomorphic Kernel Bundle Mapping (LDDKBM) extension of LDDMM allows scale space...... diffeomorphism paths, we give examples showing the decoupled evolution across scales and how the method automatically incorporates deforma- tion at appropriate scales....
Cassou-Nogues, Ph.; Erez, B.; Taylor, M. J.
2004-01-01
We establish comparison results between the Hasse-Witt invariants w_t(E) of a symmetric bundle E over a scheme and the invariants of one of its twists E_{\\alpha}. For general twists we describe the difference between w_t(E) and w_t(E_{\\alpha}) up to terms of degree 3. Next we consider a special kind of twist, which has been studied by A. Fr\\"ohlich. This arises from twisting by a cocycle obtained from an orthogonal representation. We show how to explicitly describe the twist for representatio...
Latest developments for a computer aided thermohydraulic network
International Nuclear Information System (INIS)
Thermohydraulic networks are I-D systems characterized by a small number of basic components (pumps, valves, heat exchangers, etc) connected by pipes and limited spatially by a defined number of boundary conditions (tanks, atmosphere, etc). The network system is simulated by the well known computer program RELAPS/mod3. Information concerning the network geometry component behaviour, initial and boundary conditions are usually supplied to the RELAPS code using an ASCII input file by means of 'input cards'. CATNET (Computer Aided Thermalhydraulic NETwork) is a graphically user interface that, under specific user guidelines which completely define its range of applicability, permits a very high level of standardization and simplification of the RELAPS/mod3 input deck development process as well as of the output processing. The characteristics of the components (pipes, valves, pumps etc), defining the network system can be entered through CATNET. The CATNET interface is provided by special functions to compute form losses in the most typical bending and branching configurations. When the input of all system components is ready, CATNET is able to generate the RELAPS/mod3 input file. Finally, by means of CATNET, the RELAPS/mod3 code can be run and its output results can be transformed to an intuitive display form. The paper presents an example of application of the CATNET interface as well as the latest developments which greatly simplified the work of the users and allowed to reduce the possibility of input errors. (authors)
Thermohydraulic design of saturated temperature capsule for IASCC irradiation test
Energy Technology Data Exchange (ETDEWEB)
Ide, Hiroshi; Matsui, Yoshinori; Itabashi, Yukio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others
2002-10-01
An advanced water chemistry controlled irradiation research device is being developed in JAERI, to perform irradiation tests for irradiation assisted stress corrosion cracking (IASCC) research concerned with aging of LWR. This device enables the irradiation tests under the water chemistry condition and the temperature, which simulate the conditions for BWR core internals. The advanced water chemistry controlled irradiation research device is composed of saturated temperature capsule inserted into the JMTR core and the water chemistry control unit installed in the reactor building. Regarding the saturated temperature capsule, the Thermohydraulic design of capsule structure was done, aimed at controlling the specimen's temperature, feeding water velocity on specimen's surface to the environment of BWR nearer. As the result of adopting the new capsule structure based on the design study, it was found out that feeding water velocity at the surface of specimen's is increased to about 10 times as much as before, and nuclear heat generated in the capsule components can be removed safely even in the abnormal event such as the case of loss of feeding water. (author)
Large break LOCA experiment at reactor thermohydraulic test loop
International Nuclear Information System (INIS)
The experiments of large break LOCA in the reactor Thermohydraulic Test Loop (UUTR) has been done. The experiments were held at hot leg side by use of accumulator safety injection system and without accumulator. Two experiments were done without activating the high and low pressure safety injection system, while make up system was activated until the experiments were stopped. The test were done at 1 MWt power with about 9,4 kg/sec primary coolant flow rate, and pressure 154 bar. The phenomena were only be limited on the effect of accumulator to the system during LOCA. The results of the experiments indicated a similar system depressurization phenomena for both with and without accumulator was activated. The system trip were happened at very closely different time for both at around 10 t h cycle. The significant difference were that in the experiments were the accumulator was activated, the system depressurization was going more slowly than without the accumulator, and the rods and the fluid temperature were more lower. In the case, the water injection system from the accumulator was able to reduced the rod and the cooling temperature as long as the water inventory is available
Convergence analysis of neutronic/thermohydraulic coupling behavior of SCWR
International Nuclear Information System (INIS)
The neutronic/thermohydraulic coupling (N–T coupling) calculations play an important role in core design and stability analysis. The traditional iterative method is not applicable for some new reactors (such as supercritical water-cooled reactor) which have intense N–T coupling behavior. In this paper, the mathematical model of N–T coupling based on fixed point theory is established firstly, with the convergent criterion, which can show the real-time convergence situation of iteration. Secondly, the self-adaptive relaxation factor and corresponding algorithm are proposed. Thirdly, the convergence analysis of the method of self-adaptive relaxation factor and common relaxation iteration has been performed, based on three calculation examples of SCWR fuel assembly. The results show that the proposed algorithm can efficiently reduce the calculation time and be adapted to different coupling cases and different initial distribution. It is easy to program, providing convenience for reactor design and analysis. This research also provides the theoretical basis for further study of N–T coupling behavior of new reactors such as SCWR
International Nuclear Information System (INIS)
This report is a summary of experimental investigations describing the fuel rod behavior in the refilling and reflooding phase of a loss-of-coolant accident of a PWR. The experiments were performed with 5x5 and 7x7 rod bundles, using indirectly electrically heated fuel rod simulators of full length with original PWR-KWU-geometry, original grid spacers and Zircaloy-4-claddings (Type Biblis B). The fuel rod simulators showed a cosine shaped axial power profile in 7 steps and continuous, respectively. The results describe the influence of the different parameters such as bundle size on the maximum coolant channel blockage, that of the cooling on the size of the circumferential strain of the cladding (azimuthal temperature distribution) a cold control rod guide thimble and the flow direction (axial temperature distribution) on the resulting coolant channel blockage. The rewetting behavior of different fuel rod simulators including ballooned and burst Zircaloy claddings is discussed as well as the influence of thermocouples on the cladding temperature history and the rewetting behavior. All results prove the coolability of a PWR in the case of a LOCA. Therefore, it can be concluded that the ECC-criteria established by licensing authorities can be fulfilled. (orig./HP)
International Nuclear Information System (INIS)
To model thermal mechanical bundle deformation behaviour under high temperature conditions, several factors need to be considered. These are the sources of loads, deformation mechanisms, interactions within bundle components, bundle and pressure tube (PT) interaction, and boundary constraints on the fuel bundles under in-reactor conditions. This paper describes the modelling of the following three processes: Bundle slumping due to high temperature creep-sag of individual elements and endplates; Differential element expansion and fuel element bowing; and, Bundle distortion under axial loads. To model these processes, a number of key mechanisms for bundle deformation must be considered, which include: 1) Interaction of fuel elements in a bundle with their neighbours, 2) Endplate deformation, 3) Fuel elements lateral deformation under various loads and mechanisms, 4) Interaction within a fuel element, 5) Material property change at high temperatures, 6) Transient response of a bundle, and 7) Bundle configuration change. This paper summarises the new models needed for the mechanistic modelling of the key mechanisms mentioned above and provides an example to show how an endplate plasticity model is developed with results. (author)
Bundle Security Protocol for ION
Burleigh, Scott C.; Birrane, Edward J.; Krupiarz, Christopher
2011-01-01
This software implements bundle authentication, conforming to the Delay-Tolerant Networking (DTN) Internet Draft on Bundle Security Protocol (BSP), for the Interplanetary Overlay Network (ION) implementation of DTN. This is the only implementation of BSP that is integrated with ION.
CANFLEX fuel bundle impact test
International Nuclear Information System (INIS)
This document outlines the test results for the impact test of the CANFLEX fuel bundle. Impact test is performed to determine and verify the amount of general bundle shape distortion and defect of the pressure tube that may occur during refuelling. The test specification requires that the fuel bundles and the pressure tube retain their integrities after the impact test under the conservative conditions (10 stationary bundles with 31kg/s flow rate) considering the pressure tube creep. The refuelling simulator operating with pneumatic force and simulated shield plug were fabricated and the velocity/displacement transducer and the high speed camera were also used in this test. The characteristics of the moving bundle (velocity, displacement, impacting force) were measured and analyzed with the impact sensor and the high speed camera system. The important test procedures and measurement results were discussed as follows. 1) Test bundle measurements and the pressure tube inspections 2) Simulated shield plug, outlet flange installation and bundle loading 3) refuelling simulator, inlet flange installation and sensors, high speed camera installation 4) Perform the impact test with operating the refuelling simulator and measure the dynamic characteristics 5) Inspections of the fuel bundles and the pressure tube. (author). 8 refs., 23 tabs., 13 figs
CANFLEX fuel bundle impact test
Energy Technology Data Exchange (ETDEWEB)
Chang, Seok Kyu; Chung, C. H.; Park, J. S.; Hong, S. D.; Kim, B. D.
1997-08-01
This document outlines the test results for the impact test of the CANFLEX fuel bundle. Impact test is performed to determine and verify the amount of general bundle shape distortion and defect of the pressure tube that may occur during refuelling. The test specification requires that the fuel bundles and the pressure tube retain their integrities after the impact test under the conservative conditions (10 stationary bundles with 31kg/s flow rate) considering the pressure tube creep. The refuelling simulator operating with pneumatic force and simulated shield plug were fabricated and the velocity/displacement transducer and the high speed camera were also used in this test. The characteristics of the moving bundle (velocity, displacement, impacting force) were measured and analyzed with the impact sensor and the high speed camera system. The important test procedures and measurement results were discussed as follows. 1) Test bundle measurements and the pressure tube inspections 2) Simulated shield plug, outlet flange installation and bundle loading 3) refuelling simulator, inlet flange installation and sensors, high speed camera installation 4) Perform the impact test with operating the refuelling simulator and measure the dynamic characteristics 5) Inspections of the fuel bundles and the pressure tube. (author). 8 refs., 23 tabs., 13 figs.
Fiber Bundles and Parseval Frames
Agrawal, Devanshu; Knisley, Jeff
2015-01-01
Continuous frames over a Hilbert space have a rich and sophisticated structure that can be represented in the form of a fiber bundle. The fiber bundle structure reveals the central importance of Parseval frames and the extent to which Parseval frames generalize the notion of an orthonormal basis.
International Nuclear Information System (INIS)
The fifth dynamic benchmark was defined at seventh AER-Symposium, held in Hoernitz, Germany in 1997. It is the first benchmark for coupled thermohydraulic system/three-dimensional hexagonal neutron kinetic core models. In this benchmark the interaction between the components of a WWER-440 NPP with the reactor core has been investigated. The initiating event is a symmetrical break of the main steam header at the end of the first fuel cycle and hot shutdown conditions with one control rod group stucking. This break causes an overcooling of the primary circuit. During this overcooling the scram reactivity is compensated and the scrammed reactor becomes re critical. The calculation was continued until the highly-borated water from the high pressure injection system terminated the power excursion. Each participant used own best-estimate nuclear cross section data. Only the initial subcriticality at the beginning of the transient was given. Solutions were received from Kurchatov Institute Russia with the code BIPR8/ATHLET, VTT Energy Finland with HEXTRAN/SMABRE, NRI Rez Czech Republic with DYN3/ATHLET, KFKI Budapest Hungary with KIKO3D/ATHLET and from FZR Germany with the code DYN3D/ATHLET.In this paper the results are compared. Beside the comparison of global results, the behaviour of several thermohydraulic and neutron kinetic parameters is presented to discuss the revealed differences between the solutions.(Authors)
Fiber bundle phase conjugate mirror
Ward, Benjamin G.
2012-05-01
An improved method and apparatus for passively conjugating the phases of a distorted wavefronts resulting from optical phase mismatch between elements of a fiber laser array are disclosed. A method for passively conjugating a distorted wavefront comprises the steps of: multiplexing a plurality of probe fibers and a bundle pump fiber in a fiber bundle array; passing the multiplexed output from the fiber bundle array through a collimating lens and into one portion of a non-linear medium; passing the output from a pump collection fiber through a focusing lens and into another portion of the non-linear medium so that the output from the pump collection fiber mixes with the multiplexed output from the fiber bundle; adjusting one or more degrees of freedom of one or more of the fiber bundle array, the collimating lens, the focusing lens, the non-linear medium, or the pump collection fiber to produce a standing wave in the non-linear medium.
Twisted Vector Bundles on Pointed Nodal Curves
Indian Academy of Sciences (India)
Ivan Kausz
2005-05-01
Motivated by the quest for a good compactification of the moduli space of -bundles on a nodal curve we establish a striking relationship between Abramovich’s and Vistoli’s twisted bundles and Gieseker vector bundles.
Energy Technology Data Exchange (ETDEWEB)
Oh, Dirk Joo; Jeong, Chang Joon; Lee, Kang Moon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1997-12-31
Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle. 4 refs., 1 fig., 4 tabs. (Author)
Thermohydraulic relationships for advanced water cooled reactors and the role of the IAEA
International Nuclear Information System (INIS)
Under the auspices of the International Atomic Energy Agency (IAEA) a Coordinated Research Program (CRP) on Thermohydraulic Relationships for Advanced Water-Cooled Reactors was carried out from 1995-1998. It was included into the IAEA's Programme following endorsement in 1995 by the International Working Group on Advanced Technologies for Water Cooled Reactors. The overall goal was to promote International Information exchange and cooperation in establishing a consistent set of thermohydraulic relationships that are appropriate for use in analyzing the performance and safety of advanced water-cooled reactors. (authors)
DRESDYN - A new platform for sodium related thermohydraulic studies and measurement developments
International Nuclear Information System (INIS)
The safe and reliable operation of liquid metal systems in innovative reactor concepts like sodium cooled fast breeder reactors or lead-bismuth targets in transmutation systems requires appropriate measuring systems and control units, both for the liquid metal single-phase flow as well as for gas bubble liquid metal two-phase flows. We report on the liquid sodium facility DRESDYN (DREsden Sodium facility for DYNamo and thermohydraulic studies), presently under construction, that will comprise experiments with geo- and astrophysical background as well as experiments for thermohydraulic studies and for the development and the test of measurement techniques for sodium flows. (author)
International Nuclear Information System (INIS)
The thermo-hydraulic characteristics of ship nuclear reactors are very important to the safety and reliability of ship voyage under the ocean conditions. Therefore, many countries have carried out plentiful investigations. This paper is based on some Asia open literature of investigations on thermo-hydraulic characteristics of ship nuclear reactors under the ocean conditions, reviews and sums up those main progresses such as the method, contents and typical results in this field, analyzes their insufficiency, and puts forward advices on the future investigation based on the known research findings. (authors)
Thermohydraulic study of a MTR fuel element aimed at the construction of an irradiation facility
International Nuclear Information System (INIS)
A thermohydraulic study of MTR fuel element is presented as a basic requirement for the development of an irradiation facility for testing fuel elements. A computer code named 'Thermo' has been developed for this purpose, which can stimulate different working conditions, such as, cooling, power elements and neutron flux, performing all pertinent thermohydraulic calculations. Thermocouples were used to measure the temperature gradients of the cooling fluid throughout the IEAR-1 reactor core. All experimental data are in good agreement with the theoretical model applied in this work. Finally, a draft of the proposed facility and its safety system is presented. (author)
Thermohydraulic relationships for advanced water cooled reactors, and the role of IAEA
International Nuclear Information System (INIS)
Under the auspices of the International Atomic Energy Agency (IAEA) a Coordinated Research Program (CRP) on Thermohydraulic Relationships for Advanced Water-Cooled Reactors was carried out from 1995-1999. It was included into the IAEA's Programme following endorsement in 1995 by the IAEA's International Working Group on Advanced Technologies for Water Cooled Reactors. The overall goal was to promote international information exchange and cooperation in establishing a consistent set of thermohydraulic relationships that are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. (authors)
The Atiyah Bundle and Connections on a Principal Bundle
Indian Academy of Sciences (India)
Indranil Biswas
2010-06-01
Let be a ∞ manifold and a Lie a group. Let $E_G$ be a ∞ principal -bundle over . There is a fiber bundle $\\mathcal{C}(E_G)$ over whose smooth sections correspond to the connections on $E_G$. The pull back of $E_G$ to $\\mathcal{C}(E_G)$ has a tautological connection. We investigate the curvature of this tautological connection.
International Nuclear Information System (INIS)
The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.)
Numerical Analysis for IFM Grid Effect on 5x5 Rods Bundle
Energy Technology Data Exchange (ETDEWEB)
Kim, Seong Jin; Cha, Jeong Hun; Seo, Kyong Won; Kim, Tae Woo; Kwon, Hyuk; Hwang, Dae Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2011-10-15
Generally, the fuel assembly consists of fuel rods, bottom and top grids, spacer grids, mixing vane, etc. The mixing vane with spacer grid is used to increase the thermal mixing between subchannels and to increase CHF(Critical Heat Flux). IFM(Intermediate Flow Mixer) grids are used to induce lateral flow between adjacent channels and are well-known as improving CHF, also. A numerical analysis using CFD code(ANSYS CFX, version 12.1) and subchannel code(MATRA-S) was conducted to investigate the influence of IFM grid on the subchannel temperature in 5x5 rods bundle with and without the IFM grid, thermohydraulically. In this study, the quantitative improvement of the mixing effect of the IFM grid is presented from the results of CFX and MATRA-S code. Moreover, capacity of predicting subchannel temperature of MATRA-S code is compared with CFX result
Numerical Analysis for IFM Grid Effect on 5x5 Rods Bundle
International Nuclear Information System (INIS)
Generally, the fuel assembly consists of fuel rods, bottom and top grids, spacer grids, mixing vane, etc. The mixing vane with spacer grid is used to increase the thermal mixing between subchannels and to increase CHF(Critical Heat Flux). IFM(Intermediate Flow Mixer) grids are used to induce lateral flow between adjacent channels and are well-known as improving CHF, also. A numerical analysis using CFD code(ANSYS CFX, version 12.1) and subchannel code(MATRA-S) was conducted to investigate the influence of IFM grid on the subchannel temperature in 5x5 rods bundle with and without the IFM grid, thermohydraulically. In this study, the quantitative improvement of the mixing effect of the IFM grid is presented from the results of CFX and MATRA-S code. Moreover, capacity of predicting subchannel temperature of MATRA-S code is compared with CFX result
DEFF Research Database (Denmark)
Risum, Niels; Strauss, David; Sogaard, Peter;
2013-01-01
The relationship between myocardial electrical activation by electrocardiogram (ECG) and mechanical contraction by echocardiography in left bundle-branch block (LBBB) has never been clearly demonstrated. New strict criteria for LBBB based on a fundamental understanding of physiology have recently...
Bundling ecosystem services in Denmark
DEFF Research Database (Denmark)
Turner, Katrine Grace; Odgaard, Mette Vestergaard; Bøcher, Peder Klith; Dalgaard, Tommy; Svenning, J.-C.
2014-01-01
We made a spatial analysis of 11 ecosystem services at a 10 km × 10 km grid scale covering most of Denmark. Our objective was to describe their spatial distribution and interactions and also to analyze whether they formed specific bundle types on a regional scale in the Danish cultural landscape....... We found clustered distribution patterns of ecosystem services across the country. There was a significant tendency for trade-offs between on the one hand cultural and regulating services and on the other provisioning services, and we also found the potential of regulating and cultural services to...... form synergies. We identified six distinct ecosystem service bundle types, indicating multiple interactions at a landscape level. The bundle types showed specialized areas of agricultural production, high provision of cultural services at the coasts, multifunctional mixed-use bundle types around urban...
Structure of the acrosomal bundle.
Schmid, Michael F; Sherman, Michael B; Matsudaira, Paul; Chiu, Wah
2004-09-01
In the unactivated Limulus sperm, a 60- micro m-long bundle of actin filaments crosslinked by the protein scruin is bent and twisted into a coil around the base of the nucleus. At fertilization, the bundle uncoils and fully extends in five seconds to support a finger of membrane known as the acrosomal process. This biological spring is powered by stored elastic energy and does not require the action of motor proteins or actin polymerization. In a 9.5-A electron cryomicroscopic structure of the extended bundle, we show that twist, tilt and rotation of actin-scruin subunits deviate widely from a 'standard' F-actin filament. This variability in structural organization allows filaments to pack into a highly ordered and rigid bundle in the extended state and suggests a mechanism for storing and releasing energy between coiled and extended states without disassembly. PMID:15343340
Locking means for fuels bundles
International Nuclear Information System (INIS)
A nuclear power reactor fuel bundle is described which has a plurality of fuel rods disposed between two end plates positioned by tie rods extending therebetween. The assembled bundle is secured by one or more locking forks which pass through slots in the tie rod ends. Springs mounted on the fuel rods and tie rods are compressed by assembling the bundle and forcing one end plate against the locking fork to maintain the fuel rods and tie rods in position between the end plates. Downward pressure on the end plate permits removal of the locking fork so that the end plates may be removed, thus giving access to the fuel rods. This construction facilitates disassembly of an irradiated fuel bundle under water
Kun, Ferenc; Zapperi, Stefano; Herrmann, Hans J.
1999-01-01
We introduce a continuous damage fiber bundle model that gives rise to macroscopic plasticity and compare its behavior with that of dry fiber bundles. Several interesting constitutive behaviors are found in this model depending on the value of the damage parameter and on the form of the disorder distribution. In addition, we compare the behavior of global load transfer models with local load transfer models and study in detail the damage evolution before failure. We emphasize the analogies be...
Energy Technology Data Exchange (ETDEWEB)
Bianchi, Paulo Henrique
2008-07-01
This work describes the study and test of a system capable to identify and classify transients in thermo-hydraulic systems, using a neural network technique of the self-organizing maps (SOM) type, with the objective of implanting it on the new generations of nuclear reactors. The technique developed in this work consists on the use of multiple networks to do the classification and identification of the transient states, being each network a specialist at one respective transient of the system, that compete with each other using the quantization error, that is a measure given by this type of neural network. This technique showed very promising characteristics that allow the development of new functionalities in future projects. One of these characteristics consists on the potential of each network, besides responding what transient is in course, could give additional information about that transient. (author)
Holomorphic bundles over elliptic manifolds
International Nuclear Information System (INIS)
In this lecture we shall examine holomorphic bundles over compact elliptically fibered manifolds. We shall examine constructions of such bundles as well as (duality) relations between such bundles and other geometric objects, namely K3-surfaces and del Pezzo surfaces. We shall be dealing throughout with holomorphic principal bundles with structure group GC where G is a compact, simple (usually simply connected) Lie group and GC is the associated complex simple algebraic group. Of course, in the special case G = SU(n) and hence GC = SLn(C), we are considering holomorphic vector bundles with trivial determinant. In the other cases of classical groups, G SO(n) or G = Sympl(2n) we are considering holomorphic vector bundles with trivial determinant equipped with a non-degenerate symmetric, or skew symmetric pairing. In addition to these classical cases there are the finite number of exceptional groups. Amazingly enough, motivated by questions in physics, much interest centres around the group E8 and its subgroups. For these applications it does not suffice to consider only the classical groups. Thus, while often first doing the case of SU(n) or more generally of the classical groups, we shall extend our discussions to the general semi-simple group. Also, we shall spend a good deal of time considering elliptically fibered manifolds of the simplest type, namely, elliptic curves
An overview of IPPE research on liquid metal fast reactor thermohydraulics
International Nuclear Information System (INIS)
The paper presents brief information on the most significant researches in the fields of liquid metal hydrodynamics and heat transfer performed in the State Scientific Center of Russian Federation 'Institute for Physics and Power Engineering' named after A.I.Leypunski applied to sodium-cooled fast reactors. Experimental methods for studying liquid metal thermohydraulics and applied measurement techniques are overviewed briefly in the paper. Some results of fundamental thermohydraulic investigations, such as quasi-universal character of velocity and temperature profile in liquid metals, if considered normally to the channel wall etc. are presented. Specific features of heat transfer in liquid metal cooled fuel subassembly are mentioned, among them there are: high level of coolant temperature; significant influence of an interchannel exchange on velocity and temperature distribution; an availability of contact thermal resistance; large azimuthal non-uniformity of velocity and temperature; 'conjugate' problem of heat transfer in combined geometry of fuel pin; an absence of stabilization of heat transfer in non-standard channels; an influence of non-uniform heat generation. Special attention is given to the temperature fields in fuel subassembly subjected to deformation because of radioactive swelling and creeping, as well as in case of blockage of a part of subassembly cross section. Some results of thermohydraulic investigation are demonstrated for intermediate heat exchangers, pressurized head collectors. Also the developed methods and codes of thermohydraulic calculations applied to fast reactor core are considered: subchannel approach, porous body model
The state of art of the methods for thermohydraulics design of LMFBR fuel elements
International Nuclear Information System (INIS)
The present (experimental and analytical) state of art of the methods for thermohydraulics design of LMFBR fuel elements is analyzed. A development program is suggested, in order to obtain a computer code for modelling the distribution of coolant enthalpy in reactor core. This computer code is in development. (Author)
Sarukhanyan, Edita; De Nicola, Antonio; Roccatano, Danilo; Kawakatsu, Toshihiro; Milano, Giuseppe
2014-03-01
The processes of CNTs bundle formation and insertion/rearrangement inside lipid bilayers, as models of cellular membranes, is described and analyzed in details using simulations on the microsecond scale. Molecular Dynamics simulations employing hybrid particle-field models (MD-SCF) show that during the insertion process lipid molecules coat bundles surfaces. The distortions of bilayers are more pronounced for systems undergoing to insertion of bundles made of longer CNTs. In particular, when the insertion occurs in perpendicular orientation, adsorption of lipids on CNTs surfaces promotes a transient poration. This result suggests mechanism of membrane disruption operated by bundles causing the formation of solvent-rich pockets.
Does size matter? : disentangling consumers' bundling preferences
Manoj K. Agarwal; Frambach, Ruud T.; Stremersch, Stefan
2000-01-01
Previous marketing literature has focused to a large extent on the effect of bundle characteristics on a consumer’s decision to buy a (fixed) bundle in a non-competitive setting. This study extends this narrow focus in four major ways. First, the authors address bundles that are customizable. Second, they distinguish between a consumer’s decision of whether to bundle (bundle choice) and the decision of how many goods or services to include in a bundle (bundle size). Third, they extend the foc...
Studies of a larger fuel bundle for the ABWR improved evolutionary reactor
International Nuclear Information System (INIS)
Studies for an Improved Evolutionary Reactor (IER) based on the Advanced Boiling Water Reactor (ABWR) were initiated in 1990. The author summarizes the current status of the core and fuel design. A core and fuel design based on a BWR K-lattice fuel bundle with a pitch larger than the conventional BWR fuel bundle pitch is under investigation. The core and fuel design has potential for improved core design flexibility and improved reactor transient response. Furthermore, the large fuel bundle, coupled with a functional control rod layout, can achieve improvement of operation and maintenance, as well as improvement of overall plant economy
State space model extraction of thermohydraulic systems – Part I: A linear graph approach
International Nuclear Information System (INIS)
Thermohydraulic simulation codes are increasingly making use of graphical design interfaces. The user can quickly and easily design a thermohydraulic system by placing symbols on the screen resembling system components. These components can then be connected to form a system representation. Such system models may then be used to obtain detailed simulations of the physical system. Usually this kind of simulation models are too complex and not ideal for control system design. Therefore, a need exists for automated techniques to extract lumped parameter models useful for control system design. The goal of this first paper, in a two part series, is to propose a method that utilises a graphical representation of a thermohydraulic system, and a lumped parameter modelling approach, to extract state space models. In this methodology each physical domain of the thermohydraulic system is represented by a linear graph. These linear graphs capture the interaction between all components within and across energy domains – hydraulic, thermal and mechanical. These linear graphs are analysed using a graph-theoretic approach to derive reduced order state space models. These models capture the dominant dynamics of the thermohydraulic system and are ideal for control system design purposes. The proposed state space model extraction method is demonstrated by considering a U-tube system. A non-linear state space model is extracted representing both the hydraulic and thermal domain dynamics of the system. The simulated state space model is compared with a Flownex® model of the U-tube. Flownex® is a validated systems thermal-fluid simulation software package. - Highlights: • A state space model extraction methodology based on graph-theoretic concepts. • An energy-based approach to consider multi-domain systems in a common framework. • Allow extraction of transparent (white-box) state space models automatically. • Reduced order models containing only independent state variables
International Nuclear Information System (INIS)
The operating reliability of nuclear reactors calls for a reliable strength analysis of the highly loaded core elements, one of its prerequisites being the reliable determination of the three-dimensional velocity and temperature fields. To verify thermohydraulic computer programs, extensive local temperature measurements in the rod claddings of the critical bundle zone were performed on a heated 19-rod bundle model with sodium flow and provided with spacer grids (P/D = 1.30; W/D = 1.19). These are the essential results obtained: outside the spacer grids the azimuthal temperature variations of the side and corner rods are greater by approximately the factor 10 in the bundle geometry under consideration as compared to rods in the central bundle zone. The spacer grids investigated give rise to great local temperature peaks and correspondingly great temperature gradients in the axial and azimuthal directions immediately around the support points. Continuous reduction of a subchannel by rod bowing results in subtantial rises of temperature which, however, are limited to the adjacent cladding tube zones
Numerical modeling of secondary side thermohydraulics of horizontal steam generator
Energy Technology Data Exchange (ETDEWEB)
Melikhov, V.I.; Melikhov, O.I.; Nigmatulin, B.I. [Research and Engineering Centre of LWR Nuclear Plants Safety, Moscow (Russian Federation)
1995-12-31
A mathematical model for the transient three-dimensional secondary side thermal hydraulics of the horizontal steam generator has been developed. The calculations of the steam generator PGV-1000 and PGV-4 nominal regimes and comparison of numerical and experimental results have been carried out. 7 refs.
Advanced Fuel Bundles for PHWRS
International Nuclear Information System (INIS)
The fuel used by NPCIL presently is natural uranium dioxide in the form of 19- element fuel bundles for 220 MWe PHWRs and 37-element fuel bundles for the TAPP-3&4 540 MWe units. The new 700 MWe PHWRs also use 37-element fuel bundles. These bundles are of short 0.5 m length of circular geometry. The cladding is of collapsible type made of Zircaloy-4 material. PHWRs containing a string of short length fuel bundles and the on-power refueling permit flexibility in using different advanced fuel designs and in core fuel management schemes. Using this flexibility, alternative fuel concepts are tried in Indian PHWRs. The advances in PHWR fuel designs are governed by the desire to use resources other than uranium, improve fuel economics by increasing fuel burnup and reduce overall spent nuclear fuel waste and improve reactor safety. The rising uranium prices are leading to a relook into the Thorium based fuel designs and reprocessed Uranium based and Plutonium based MOX designs and are expected to play a major role in future. The requirement of synergism between different type of reactors also plays a role. Increase in fuel burnup beyond 15 000 MW∙d/TeU in PHWRs, using higher fissile content materials like slightly enriched uranium, Mixed Oxide and Thorium Oxide in place of natural uranium in fuel elements, was studied many PHWR operating countries. The work includes reactor physics studies and test irradiation in research reactors and power reactors. Due to higher fissile content these bundles will be capable of delivering higher burnup than the natural uranium bundles. In India the fuel cycle flexibility of PHWRs is demonstrated by converting this type of technical flexibility to the real economy by irradiating these different types of advanced fuel materials namely Thorium, MOX, SEU, etc. The paper gives a review of the different advanced fuel design concepts studied for Indian PHWRs. (author)
Cohomology of line bundles: Applications
Blumenhagen, Ralph; Jurke, Benjamin; Rahn, Thorsten; Roschy, Helmut
2012-01-01
Massless modes of both heterotic and Type II string compactifications on compact manifolds are determined by vector bundle valued cohomology classes. Various applications of our recent algorithm for the computation of line bundle valued cohomology classes over toric varieties are presented. For the heterotic string, the prime examples are so-called monad constructions on Calabi-Yau manifolds. In the context of Type II orientifolds, one often needs to compute cohomology for line bundles on finite group action coset spaces, necessitating us to generalize our algorithm to this case. Moreover, we exemplify that the different terms in Batyrev's formula and its generalizations can be given a one-to-one cohomological interpretation. Furthermore, we derive a combinatorial closed form expression for two Hodge numbers of a codimension two Calabi-Yau fourfold.
Principal bundles the classical case
Sontz, Stephen Bruce
2015-01-01
This introductory graduate level text provides a relatively quick path to a special topic in classical differential geometry: principal bundles. While the topic of principal bundles in differential geometry has become classic, even standard, material in the modern graduate mathematics curriculum, the unique approach taken in this text presents the material in a way that is intuitive for both students of mathematics and of physics. The goal of this book is to present important, modern geometric ideas in a form readily accessible to students and researchers in both the physics and mathematics communities, providing each with an understanding and appreciation of the language and ideas of the other.
Methodology to assess plant thermal-hydraulic behavior during transients and accidents
International Nuclear Information System (INIS)
The safety evaluation approach proposed is based on strict criteria regarding fuel and thermohydraulics. Models used in the assessment are based on separate effect tests using large rigs. Computer codes have been developed incorporating the models - for transients, LOCA and containment. System tests have been conducted to validate the code system. The codes are used to investigate accident management of LWRs. The report is in a poster form. 20 figs
Development of a model for the primary system CAREM reactor's stationary thermohydraulic calculation
International Nuclear Information System (INIS)
The ESCAREM program oriented to CAREM reactors' stationary thermohydraulic calculation is presented. As CAREM gives variations in relation to models for BWR (Boiling Water Reactors)/PWR (Pressurized Water Reactors) reactors, it was decided to develop a suitable model which allows to calculate: a) if the Steam Generator design is adequate to transfer the power required; b) the circulation flow that occurs in the Primary System; c) the temperature at the entrance (cool branch) and d) the contribution of each component to the pressure drop in the circulation connection. Results were verified against manual calculations and alternative numerical models. An experimental validation at the Thermohydraulic Essays Laboratory is suggested. A parametric analysis series is presented on CAREM 25 reactor, demonstrating operating conditions, at different power levels, as well as the influence of different design aspects. (Author)
International Nuclear Information System (INIS)
This report explains the numerical methods and the set-up method of input data for a single-phase multi-dimensional thermohydraulics direct numerical simulation code DINUS-3 (Direct Numerical Simulation using a 3rd-order upwind scheme). The code was developed to simulate non-stationary temperature fluctuation phenomena related to thermal striping phenomena, developed at Power Reactor and Nuclear Fuel Development Corporation (PNC). The DINUS-3 code was characterized by the use of a third-order upwind scheme for convection terms in instantaneous Navier-Stokes and energy equations, and an adaptive control system based on the Fuzzy theory to control time step sizes. Author expect this report is very useful to utilize the DINUS-3 code for the evaluation of various non-stationary thermohydraulic phenomena in reactor applications. (author)
International Nuclear Information System (INIS)
The save operation of liquid metal systems in innovative reactor concepts requires appropriate measuring systems and instrumentation, both for the liquid metal single-phase flow as well as for gas bubble liquid metal two-phase flows. At HZDR the large-scale liquid sodium facility DRESDYN (DREsden Sodium facility for DYNamo and thermohydraulic studies) is under construction that will comprise experiments with geo- and astrophysical background as well as experiments for thermohydraulic studies of sodium flows. The development of flow measurement techniques has a long tradition at HZDR. It covers contactless flow-rate sensors, local velocity measurements such as the Ultrasound Doppler Velocimetry (UDV), the Contactless Inductive Flow Tomography (CIFT), as well as X-ray visualizations of liquid metal two-phase flows, which all will be exploited and further developed at an In-Service-Inspection experiment in the framework of DRESDYN. (author)
Influence of the outlet air temperature on the thermohydraulic behaviour of air coolers
Directory of Open Access Journals (Sweden)
Đorđević Emila M.
2003-01-01
Full Text Available The determination of the optimal process conditions for the operation of air coolers demands a detailed analysis of their thermohydraulic behaviour on the one hand, and the estimation of the operating costs, on the other. One of the main parameters of the thermohydraulic behaviour of this type of equipment, is the outlet air temperature. The influence of the outlet air temperature on the performance of air coolers (heat transfer coefficient overall heat transfer coefficient, required surface area for heat transfer air-side pressure drop, fan power consumption and sound pressure level was investigated in this study. All the computations, using AirCooler software [1], were applied to cooling of the process fluid and the condensation of a multicomponent vapour mixture on two industrial devices of known geometries.
Exploring Bundling Theory with Geometry
Eckalbar, John C.
2006-01-01
The author shows how instructors might successfully introduce students in principles and intermediate microeconomic theory classes to the topic of bundling (i.e., the selling of two or more goods as a package, rather than separately). It is surprising how much students can learn using only the tools of high school geometry. To be specific, one can…
Bundled Discounts and EC Judicial Review
Christian Roques
2008-01-01
The Community Courts' case law is rich with cases relating to tying or bundling practices in their classical economic form. However, the same cannot be said for the second acceptance of bundled discounts.
Neutronics and thermohydraulics of the reactor C.E.N.E. Part II
International Nuclear Information System (INIS)
In this report the analysis of neutronics thermohydraulics and shielding of the 10 HWt swimming pool reactor C.E.N.E is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs
State space model extraction of thermohydraulic systems Part I: a linear graph approach
Uren, Kenneth Richard; Schoor, George van
2013-01-01
This second paper in a two part series presents the application of a developed state space model extraction methodology applied to a Brayton cycle-based PCU (power conversion unit) of a PBMR (pebble bed modular reactor). The goal is to investigate if the state space extraction methodology can cope with larger and more complex thermohydraulic systems. In Part I the state space model extraction methodology for the purpose of control was described in detail and a state space represen...
Coupling of the thermohydraulic code ATHLET with the neutron kinetic core model DYN3D
International Nuclear Information System (INIS)
The coupling of advanced thermohydraulic codes with 3-dimensional neutron kinetic codes corresponds to the effort to replace conservative estimations by best estimate calculations. ATHLET is an advanced thermohydraulic code, developed by the German Gesellschaft fur Anlagen- und Reaktorsicherheit (GRS). Up to now only point kinetics and 1-dimensional neutron kinetics have been included. The DYN3D code, developed in the Research Centre Rossendorf (RCR) for the improvement of the simulation of reactivity initiated accidents in nuclear reactors with hexagonal fuel elements comprises 3-dimensional neutron kinetics, models for the thermohydraulics of the core including heat transfer from the fuel to the coolant and a fuel rod behavior model. The reactor core model DYN3D was coupled with the ATHLET code on two basically different ways. The first way of coupling uses only the neutron kinetics part of the DYN3D code (internal coupling). This coupling along the core is very close and demands an high effort of programming due to the high number of coupling parameters. In the second way the whole core is cut out from the ATHLET plant model. The core is completely modeled by the DYN3D code (external coupling). In this case the interfaces are located at the bottom and at the top of the core. At this interfaces the pressures, mass flow rates, enthalpies and concentrations of boron acid have to be transferred. This way of coupling is easy to realize by interconnection of an interface routine. It is effectively supported by the General Control and Simulation Modul (GCSM) of the ATHLET code. Almost no changes of the single programs are necessary. Another advantage of this coupling is that the complete DYN3D model can be used. The disadvantage of this method is the splitting of the thermohydraulics
Failure properties of fiber bundle models
Pradhan, Srutarshi; Chakrabarti, Bikas K.
2003-01-01
We study the failure properties of fiber bundles when continuous rupture goes on due to the application of external load on the bundles. We take the two extreme models: equal load sharing model (democratic fiber bundles) and local load sharing model. The strength of the fibers are assumed to be distributed randomly within a finite interval. The democratic fiber bundles show a solvable phase transition at a critical stress (load per fiber). The dynamic critical behavior is obtained analyticall...
Bundling Information Goods: Pricing, Profits, and Efficiency
Yannis Bakos; Erik Brynjolfsson
1999-01-01
We study the strategy of bundling a large number of information goods, such as those increasingly available on the Internet, and selling them for a fixed price. We analyze the optimal bundling strategies for a multiproduct monopolist, and we find that bundling very large numbers of unrelated information goods can be surprisingly profitable. The reason is that the law of large numbers makes it much easier to predict consumers' valuations for a bundle of goods than their valuations for the indi...
Quantum principal bundles and corresponding gauge theories
Durdevic, M
1995-01-01
A generalization of classical gauge theory is presented, in the framework of a noncommutative-geometric formalism of quantum principal bundles over smooth manifolds. Quantum counterparts of classical gauge bundles, and classical gauge transformations, are introduced and investigated. A natural differential calculus on quantum gauge bundles is constructed and analyzed. Kinematical and dynamical properties of corresponding gauge theories are discussed.
Strategic and welfare implications of bundling
DEFF Research Database (Denmark)
Martin, Stephen
1999-01-01
A standard oligopoly model of bundling shows that bundling by a firm with a monopoly over one product has a strategic effect because it changes the substitution relationships between the goods among which consumers choose. Bundling in appropriate proportions is privately profitable, reduces rivals...
On Volumes of Arithmetic Line Bundles
Yuan, Xinyi
2008-01-01
We show an arithmetic generalization of the recent work of Lazarsfeld-Mustata which uses Okounkov bodies to study linear series of line bundles. As applications, we derive a log-concavity inequality on volumes of arithmetic line bundles and an arithmetic Fujita approximation theorem for big line bundles.
Simulation of flow across complicated domain between tube bundles by the discrete vortex method
Institute of Scientific and Technical Information of China (English)
无
2003-01-01
On the basis of the analysis of numerical simulation methods for the complicated domain between tube bundles, an improved Lagragian discrete vortex method (DVM) and corresponding algorithm are put forward to solve the practical difficulties of flow across tube bundles. With this method the amount of vortices can be reduced considerably, which makes quick calculation possible. Applied to the practical configuration of horizontal tube bundles, the DVM simulation is carried out and compared with the experimental results. Both the transient flow field and the profile of mean velocity and fluctuations are in good agreement with experimental results, which indicate that the DVM is suitable for the simulation of single-phase flow across tube bundles.
Steady state test on PWR steam generator thermohydraulics
International Nuclear Information System (INIS)
Experimental activity on U-tube steam generator thermal hydraulics is under way at CISE and SIET in the framework of ENEA's LWR safety research programme. The test section includes 9 tubes. Hot side and cold side can be separated simulated, with primary and secondary fluid in full thermalhydraulic conditions. The experimental matrix includes: steady state tests (in both adiabatic and diabatic conditions); transients tests that simulate various accidents. Some steady state tests are reported. The secondary side average density, measured by the quick closing valve technique can be accurately calculated by the Zuber-Dix and Zuber-Rohuani correlations. Continuous pressure drops can be very well predicted by an adapted version of Thom correlation and CISE DIF-3 correlation: the development of an empirical correlation was, instead, necessary for assessment of the local pressure drops across spacer grids
International Nuclear Information System (INIS)
The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in
Simplicial principal bundles in parametrized spaces
Roberts, David M
2012-01-01
In this paper, motivated by recent interest in higher gauge theory, we prove that the fiberwise geometric realization functor takes a certain class of simplicial principal bundles in a suitable category of spaces over a fixed space $B$ to fiberwise principal bundles. As an application we show that the fiberwise geometric realization of the universal simplicial principal bundle for a simplicial group $G$ in the category of spaces over $B$ gives rise to a fiberwise principal bundle with structure group $|G|$. An application to classifying theory for fiberwise principal bundles is described.
Multipath packet switch using packet bundling
DEFF Research Database (Denmark)
Berger, Michael Stubert
The basic concept of packet bundling is to group smaller packets into larger packets based on, e.g., quality of service or destination within the packet switch. This paper presents novel applications of bundling in packet switching. The larger packets created by bundling are utilized to extend...... switching capacity by use of parallel switch planes. During the bundling operation, packets will experience a delay that depends on the actual implementation of the bundling and scheduling scheme. Analytical results for delay bounds and buffer size requirements are presented for a specific scheduling...
Model of turbine blades bundles
Czech Academy of Sciences Publication Activity Database
Půst, Ladislav; Pešek, Luděk
Prague : Institute of Thermomechanics, Academy of Sciences of the Czech Republic, v. v. i., 2013 - (Zolotarev, I.), s. 467-477 ISBN 978-80-87012-47-5. ISSN 1805-8256. [Engineering Mechanics 2013 /19./. Svratka (CZ), 13.05.2013-16.05.2013] R&D Projects: GA ČR GA101/09/1166 Institutional support: RVO:61388998 Keywords : free and forced vibrations * eigenmodes * mathematical model * bundle of blades Subject RIV: BI - Acoustics
Model of turbine blades bundles
Czech Academy of Sciences Publication Activity Database
Půst, Ladislav; Pešek, Luděk
Praha : Insitute of Thermomechanics ASCR, v. v. i., 2013 - (Zolotarev, I.). s. 125-126 ISBN 978-80-87012-46-8. ISSN 1805-8248. [Engineering Mechanics 2013 /19./. 13.05.2013-16.05.2013, Svratka] R&D Projects: GA ČR GA101/09/1166 Institutional support: RVO:61388998 Keywords : free and forced vibrations * eigenmodes * bundle of blades Subject RIV: BI - Acoustics
Competitive nonlinear pricing and bundling
Armstrong, Mark; Vickers, John
2006-01-01
We examine the impact of multiproduct nonlinear pricing on profit, consumer surplus and welfare in a duopoly. When consumers buy all their products from one firm (the one-stop shopping model), nonlinear pricing leads to higher profit and welfare, but often lower consumer surplus, than linear pricing. By contrast, in a unit-demand model where consumers may buy one product from one firm and another product from another firm, bundling generally acts to reduce profit and welfare and to boost cons...
Maximum allowable heat flux for a submerged horizontal tube bundle
International Nuclear Information System (INIS)
For application to industrial heating of large pools by immersed heat exchangers, the socalled maximum allowable (or open-quotes criticalclose quotes) heat flux is studied for unconfined tube bundles aligned horizontally in a pool without forced flow. In general, we are considering boiling after the pool reaches its saturation temperature rather than sub-cooled pool boiling which should occur during early stages of transient operation. A combination of literature review and simple approximate analysis has been used. To date our main conclusion is that estimates of q inch chf are highly uncertain for this configuration
Quantum bundles and their symmetries
International Nuclear Information System (INIS)
Wave functions in the domain of observables such as the Hamiltonian are not always smooth functions on the classical configuration space Q. Rather, they are often best regarded as functions on a G bundle EG over Q or as sections of an associated bundle. If H is a classical group which acts on Q, its quantum version HG, which acts on EG, is not always H, but an extension of H by G. A powerful and physically transparent construction of EG and HG, where G = U(1) and H1(Q,Z) = 0, has been developed using the path space P. (P consists of paths on Q from a fixed point). In this paper the authors show how to construct EG and HG when G is U(1) or U(1) x π1(Q) and there is no restriction on H1(Q,Z). The method is illustrated with concrete examples, such as a system of charges and monopoles. The method is illustrated with concrete examples, such as a system of charges and monopoles. The authors argue also that P is a sort of superbundle from which a large variety of bundles can be obtained by imposing suitable equivalence relations
Photonic bandgap fiber bundle spectrometer
Hang, Qu; Syed, Imran; Guo, Ning; Skorobogatiy, Maksim
2010-01-01
We experimentally demonstrate an all-fiber spectrometer consisting of a photonic bandgap fiber bundle and a black and white CCD camera. Photonic crystal fibers used in this work are the large solid core all-plastic Bragg fibers designed for operation in the visible spectral range and featuring bandgaps of 60nm - 180nm-wide. 100 Bragg fibers were chosen to have complimentary and partially overlapping bandgaps covering a 400nm-840nm spectral range. The fiber bundle used in our work is equivalent in its function to a set of 100 optical filters densely packed in the area of ~1cm2. Black and white CCD camera is then used to capture spectrally "binned" image of the incoming light at the output facet of a fiber bundle. To reconstruct the test spectrum from a single CCD image we developed an algorithm based on pseudo-inversion of the spectrometer transmission matrix. We then study resolution limit of this spectroscopic system by testing its performance using spectrally narrow test peaks (FWHM 5nm-25nm) centered at va...
Thermohydraulics and fuel rod behavior during reflood/quench situations
International Nuclear Information System (INIS)
At the Forschungszentrum Karlsruhe (FZKA) the Institute for Reactor Safety (IRS) supports the experimental activities focused on the safety features of existing and advanced light water reactors by plant calculations using SCDAP/RELAP5 (S/R5) mod3.2 and RELAP5 (R5) mod 3.3. These codes are continuously validated by comparison with various experiments. In the area of design basis safety R5 mod3.3.b code validation and code application for the high performance LWR (HPLWR, 5th European Framework Programme is ongoing), extending the capability of RELAP5 to simulate conditions of supercritical water. Best estimate safety analyses for severe accidents with the in-house version S/R5 mod 3.2.irs were performed successfully and documented for the European Pressurized Water Reactor. (FZK-6299, -6315, -6567, http://bibliothek.fzk.de/zb/berichte). Analytical support for bundle experiments in the QUENCH facility at FZK is ongoing with S/R5 mod3.2.irs to define test parameters and to support experimental analyses. A fine discretization with 5 cm meshes has been developed to better simulate reflood conditions. The improved shattering model works fairly well for an intact rod geometry, but is not adequate for damaged fuel rods. Furthermore, S/R5 is used for test analyses especially for QUENCH-06, basis for the OECD/NEA/CSNI International Standard Problem No. 45 ISP-45, and for QUENCH-07 as contribution to EU-program COLOSS. A code extension to simulate B4C oxidation is planned, but postponed until 2003 due to limited man power caused by the OECD/NEA ISP-45 benchmark. Besides, ICARE2V3 is used to support S/R5 analyses with respect to QUENCH and Phebus analyses. The blind phase of the OECD/NEA/CSNI International Standard Problem No. 45 ISP-45 (QUENCH-06) is finished, the preparation of the OECD Interpretation and Comparison report is under way. The current state of code to code comparison reveals strong variation within the same code family probably attributed to user effects. The
Thermohydraulics and fuel rod behavior during reflood/quench situations
Energy Technology Data Exchange (ETDEWEB)
Hering, W.; Homann, C.; Sanchez, V.; Sengpiel, W.; Struwe, D. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Reaktorsicherheit; Lamy, J.S. [EDF, Paris (France)
2001-07-01
At the Forschungszentrum Karlsruhe (FZKA) the Institute for Reactor Safety (IRS) supports the experimental activities focused on the safety features of existing and advanced light water reactors by plant calculations using SCDAP/RELAP5 (S/R5) mod3.2 and RELAP5 (R5) mod 3.3. These codes are continuously validated by comparison with various experiments. In the area of design basis safety R5 mod3.3.b code validation and code application for the high performance LWR (HPLWR, 5{sup th} European Framework Programme is ongoing), extending the capability of RELAP5 to simulate conditions of supercritical water. Best estimate safety analyses for severe accidents with the in-house version S/R5 mod 3.2.irs were performed successfully and documented for the European Pressurized Water Reactor. (FZK-6299, -6315, -6567, http://bibliothek.fzk.de/zb/berichte). Analytical support for bundle experiments in the QUENCH facility at FZK is ongoing with S/R5 mod3.2.irs to define test parameters and to support experimental analyses. A fine discretization with 5 cm meshes has been developed to better simulate reflood conditions. The improved shattering model works fairly well for an intact rod geometry, but is not adequate for damaged fuel rods. Furthermore, S/R5 is used for test analyses especially for QUENCH-06, basis for the OECD/NEA/CSNI International Standard Problem No. 45 ISP-45, and for QUENCH-07 as contribution to EU-program COLOSS. A code extension to simulate B4C oxidation is planned, but postponed until 2003 due to limited man power caused by the OECD/NEA ISP-45 benchmark. Besides, ICARE2V3 is used to support S/R5 analyses with respect to QUENCH and Phebus analyses. The blind phase of the OECD/NEA/CSNI International Standard Problem No. 45 ISP-45 (QUENCH-06) is finished, the preparation of the OECD Interpretation and Comparison report is under way. The current state of code to code comparison reveals strong variation within the same code family probably attributed to user effects
Thermohydraulics of emergency core cooling in light water reactors
International Nuclear Information System (INIS)
This report, by a group of experts of the OECD-NEA Committee on the Safety of Nuclear Installations, reviews the current state-of-knowledge in the field of emergency core cooling (ECC) for design-basis, loss-of-coolant accidents (LOCA) and core uncover transients in pressurized- and boiling-water reactors. An overview of the LOCA scenarios and ECC phenomenology is provided for each type of reactor, together with a brief description of their ECC systems. Separate-effects and integral-test facilities, which contribute to understanding and assessing the phenomenology, are reviewed together with similarity and scaling compromises. All relevant LOCA phenomena are then brought together in the form of tables. Each phenomenon is weighted in terms of its importance to the course of a LOCA, and appraised for the adequacy of its data base and analytical modelling. This qualitative procedure focusses attention on the modelling requirements of dominant LOCA phenomena and the current capabilities of the two-fluid models in two-phase flows. This leads into the key issue with ECC: quantitative code assessment and the application of system codes to predict with a well defined uncertainty the behaviour of a nuclear power plant. This issue, the methodologies being developed for code assessment and the question of how good is good enough are discussed in detail. Some general conclusions and recommendations for future research activities are provided
A Thermo-Hydraulic Tool for Automatic Virtual Hazop Evaluation
Directory of Open Access Journals (Sweden)
Pugi L.
2014-12-01
Full Text Available Development of complex lubrication systems in the Oil&Gas industry has reached high levels of competitiveness in terms of requested performances and reliability. In particular, the use of HazOp (acronym of Hazard and Operability analysis represents a decisive factor to evaluate safety and reliability of plants. The HazOp analysis is a structured and systematic examination of a planned or existing operation in order to identify and evaluate problems that may represent risks to personnel or equipment. In particular, P&ID schemes (acronym of Piping and Instrument Diagram according to regulation in force ISO 14617 are used to evaluate the design of the plant in order to increase its safety and reliability in different operating conditions. The use of a simulation tool can drastically increase speed, efficiency and reliability of the design process. In this work, a tool, called TTH lib (acronym of Transient Thermal Hydraulic Library for the 1-D simulation of thermal hydraulic plants is presented. The proposed tool is applied to the analysis of safety relevant components of compressor and pumping units, such as lubrication circuits. Opposed to the known commercial products, TTH lib has been customized in order to ease simulation of complex interactions with digital logic components and plant controllers including their sensors and measurement systems. In particular, the proposed tool is optimized for fixed step execution and fast prototyping of Real Time code both for testing and production purposes. TTH lib can be used as a standard SimScape-Simulink library of components optimized and specifically designed in accordance with the P&ID definitions. Finally, an automatic code generation procedure has been developed, so TTH simulation models can be directly assembled from the P&ID schemes and technical documentation including detailed informations of sensor and measurement system.
International Nuclear Information System (INIS)
This work explores different possible sequences at the loss of a train of the DTH when the plant is lowering power. The study of the different possible trajectories has been done through the collapse tool and study thermo-hydraulic each of these paths is done by the code TRACE Thermo-hydraulic.
Study on the transient behaviours of MNSR reactor for control rod withdrawal
International Nuclear Information System (INIS)
The transient behaviours of Miniature Neutron Source Reactor MNSR are analyzed and calculated with the reactor thermohydraulics RETRAN-02 program and the reactor physics MARIA program. The obtained event sequence and consequence from the calculation are compared with the experiments. The effective resonance integral for study on Doppler effect is taken into account. The reactivity temperature coefficient weighting factors are computed. The transient parameters related to reactor power peaking, coolant inlet temperatures, outlet temperatures and coolant mass flow, etc. are computed and compared with the experimental results. (6 refs., 2 figs., 5 tabs.)
Thermohydraulic calculations in rectangular channels for RA-6 type reactors with transition regime
International Nuclear Information System (INIS)
In August 2000 and within the framework of the RA-6 core conversion from high to low enrichment (20%), a preliminary analysis was performed to evaluate the maximum power that the reactor could operate with the new kernel without makeing substantial changes. This meant keeping intact, for example, the concrete shield of the pool and the nucleus inlet and outlet pipes embedded in the walls. Preliminary results indicated that for these boundary conditions a maximum power of about 3 MWt could be achieved. In August 2005 the project was resumed and new calculations performed taking as a starting point the ECBE plate fuel element(U3O8-Al). A core was developed with cooling channle widths of 2.6 mm for the control fuel elements and 2.7 mm for standard fuel elements. The thermo-hydraulic calculation puts in evidence that coolant flow into the core was in the transitional regime for the vast majority of configurations. While TERMIC code, used for thermo-hydraulic design, has been extensively tested and validated for use in research reactors under turbulent and laminar flows, this is not so for transition conditions. The transition regime is strongly dependent on conditions such as flow inlet characteristics, channel geometry, etc.. and therefore there are no reliable correlations for general use. For this reason we found it convenient to carry out experiments simulating the working conditions in order to adjust the code results with experimental data. In the present work we show the experimental results, the simulation of the experiences using the TERMIC code, and the adjustments made to the correlations used by the code so that it can be applied to the thermo-hydraulic design of the new core.
A two-dimensional thermohydraulic calculation on the core of pressurized water reactor
International Nuclear Information System (INIS)
The design of pressurized water reactors with natural coolant circulation and passive safety features, enables one to meet increased requirements for nuclear power station safety. However, in the design there are problems associated in part with the thermohydraulic calculations on the circulation loop as a whole and on the individual parts: core, power section, overflow zone, and descending part. There are numerous programs for calculating these cores, which enable one to analyze the fuel-pin assemblies and core as a whole. Those programs have been developed mainly for the VVER reactors, and their application to other pressurized water reactors requires testing. COBSIM is a modification of the common COBRA-3C program, which was set up for calculating the cores of foreign reactors with square cross section fuel-pin assemblies. Nevertheless, it has been used for the thermohydraulic calculation on a VVER with hexagonal fuel-pin assemblies. COBSIM is intended for cellwise analysis of VVER fuel-pin assemblies. The physical models are based on the assumption that the conservation equation for the two-phase flow can be written in the same way as for a single-phase flow with coolant properties averaged over the control volume. The true bulk steam content in bulk boiling can be calculated in various ways: from the homogeneous model, with a given phase slip coefficient, by means of a modified Armand relation, or with a mass flow steam content specified by the user. The boiler during underheating can be calculated from the Levy model. In COBSIM, one can specify the coolant flow at the core inlet over the channels or the mean flow over all the channels, which can be distributed in such a way as to satisfy the condition for equal pressure differences in all channels. COBSIM can be adapted to boiling-water reactors, but one then has to consider the applicability of cellwise analysis in the thermohydraulic calculation on the entire core
Some asymmetric thermohydraulic behaviors of liquid metal-gas two-phase MHD flows
International Nuclear Information System (INIS)
In this paper magnetohydrodynamic effect on liquid-metal two-phase flow and heat transfer are summarized based on the measurements made by the present author in NaK-nitrogen flow in a vertical round tube in the presence of a transverse magnetic field. This study covered a wide range of two-phase flow patterns from bubbly flow to annular-dispersed flow, including flow pattern observation, measurements of phase distributions, liquid film behavior, and heat transfer coefficient. Particular emphases are directed towards describing asymmetric thermohydraulic structures induced by the applied magnetic field
Reduction of fluid property errors of various thermohydraulic codes for supercritical water systems
International Nuclear Information System (INIS)
Various thermohydraulic codes (like WAHA, ATHLET, RELAP) work perfectly for pressurized water and steam, but fail to give reliable results for supercritical water. This might be a surprise, because theoretically Supercritical Water (SCW) should be a simpler system than normal water, due to the lack of phase transition and two-phase flow in the supercritical region. Some of the problem is caused by low accuracy of the fluid properties due to the presence of the pseudo-critical line. In this presentation we would like to address this pseudocriticality-related problem and to establish a method for the error-reduction. (orig.)
Technique of analysis and error detection for thermo-hydraulic system data
International Nuclear Information System (INIS)
Statistical techniques based on estimation theory were developed for the analysis of steady-state data from thermo-hydraulic systems, which could be either experimental loops or operating power plants. The method seeks to resolve errors in the component heat balances which describe the system, to obtain system parameter estimates which are more accurate than the raw data, and to flag possible faulty sensors. Sample results are given for the analysis of test data from the Sodium Loop Safety Faciltiy (SLSF) P3 experiment
International Nuclear Information System (INIS)
The operation of passive safety systems is based on the use of gravitation, natural circulation processes, compressed gases energy. The passive systems ensure shutdown, reactor shutdown cooling and continuous after-heat removal. The results of investigations of thermohydraulic processes, during which the operability of WWER NPP passive safety systems have been justified, are considered. The processes are blowdown of subcooled liquid into opposing steam flow at GE-2 system starting-up, undeveloped boiling of subcooled liquid on horizontal tubes under condensation operation of WWER steam generator, heat transfer in air-air heat exchanger of WWER passive filtration system
International Nuclear Information System (INIS)
The influence of different types of Ignalina NPP structure modeling on the results of thermo-hydraulic evaluation of ALS (accident localization system) compartments is analyzed. There are five types of structure models applied for calculations. The results showed that in the case of maximum design-based accident the difference between maximum overpressure in an accident compartment did not exceed 2.7%. In the future it is intended to apply the model with equivalent reinforced (concrete) material considering the paint layer covering the metal lining of ALS structures, because this model provides the most conservative results, and consumes less computer time. (author)
Validation of thermohydraulic codes by comparison of experimental results with computer simulations
International Nuclear Information System (INIS)
The results obtained by simulation of three cases from CANON depressurization experience, using the TRAC-PF1 computer code, version 7.6, implanted in the VAX-11/750 computer of Brazilian CNEN, are presented. The CANON experience was chosen as first standard problem in thermo-hydraulic to be discussed at ENFIR for comparing results from different computer codes with results obtained experimentally. The ability of TRAC-PF1 code to prevent the depressurization phase of a loss of primary collant accident in pressurized water reactors is evaluated. (M.C.K.)
Energy Technology Data Exchange (ETDEWEB)
Bjoerndahl, Olof; Letzter, Adam; Marcinkiewicz, Jerzy; Segle, Peter (Inspecta Nuclear AB, 104 25 Stockholm (Sweden))
2007-03-15
Transient thermohydraulic events often control the design of piping systems in nuclear power plants. Water hammers due to valve closure, pressure transients caused by steam collapse and pipe break all result in structural loads that are characterised by a high frequency content. What also characterises these pressures/forces is the specific spatial and time dependence that is acting on the piping system and found in the wave propagation in the contained fluid. The aim with this project has been to develop recommendations for analysis of the stress response in piping systems subjected to thermohydraulic transients. Basis for this work is that the so called two-step-method is applied and that the structural response is calculated with modal superposition. Derived analysis criteria are based on the assumption that the associated volume strain energy in the wave propagation for the contained fluid may be well defined by a parameter, here called epsilon{sub PN}. The stress response in the piping system is assumed to be completely determined with certain accuracy for that part of the volume strain energy in the wave propagation associated with this parameter. A comprehensive work has been done to determine the accuracy in loadings calculated with RELAP5. Properties such as period elongation and associated spurious oscillations in the pressure wave transient have been investigated. Furthermore, has the characteristics of the artificial numerical damping in RELAP5 been identified. Based on desired accuracy of the thermohydraulic analysis together with knowledge about the duration of the thermohydraulic perturbation, the lowest upper frequency limit f{sub Pipe}, in the modal base that is required for the structure model is calculated. With perturbation is meant such as a valve closure. According to suggested criteria and with the upper frequency limit set, the essential parameters i) largest size of the elements in the structure model and ii) the largest applicable time
Quadratic bundle and nonlinear equations
International Nuclear Information System (INIS)
The paper is aimed at giving an exhaustive description of the nonlinear evolution equations (NLEE), connected with the quadratic bundle (the spectral parameter lambda, which enters quadratically into the equations) and at describing Hamiltonian structure of these equations. The equations are solved through the inverse scattering method (ISM). The basic formulae for the scattering problem are given. The spectral expansion of the integrodifferential operator is used so that its eigenfunctions are the squared solutions of the equation. By using the notions of Hamiltonian structure hierarchy and gauge transformations it is shown how to single out physically interesting NLEE
Static stress analysis of CANFLEX fuel bundles
International Nuclear Information System (INIS)
The static stress analysis of CANFLEX bundles is performed to evaluate the fuel structural integrity during the refuelling service. The structure analysis is carried out by predicting the drag force, stress and displacements of the fuel bundle. By the comparison of strength tests and analysis results, the displacement values are well agreed within 15%. The analysis shows that the CANFLEX fuel bundle keep its structural integrity. 24 figs., 6 tabs., 12 refs. (Author) .new
Damping Properties of the Hair Bundle
Baumgart, Johannes; Kozlov, Andrei S.; Risler, Thomas; Hudspeth, A. James
2015-01-01
The viscous liquid surrounding a hair bundle dissipates energy and dampens oscillations, which poses a fundamental physical challenge to the high sensitivity and sharp frequency selectivity of hearing. To identify the mechanical forces at play, we constructed a detailed finite-element model of the hair bundle. Based on data from the hair bundle of the bullfrog's sacculus, this model treats the interaction of stereocilia both with the surrounding liquid and with the liquid in the narrow gaps b...
Tying, Bundling, and Loyalty/Requirement Rebates
Nicholas Economides
2011-01-01
I discuss the impact of tying, bundling, and loyalty/requirement rebates on consumer surplus in the affected markets. I show that the Chicago School Theory of a single monopoly surplus that justifies tying, bundling, and loyalty/requirement rebates on the basis of efficiency typically fails. Thus, tying, bundling, and loyalty/requirement rebates can be used to extract consumer surplus and enhance profit of firms with market power. I discuss the various setups when this occurs.
Bundling and Competition on the Internet
Yannis Bakos; Erik Brynjolfsson
2000-01-01
The Internet has signi.cantly reduced the marginal cost of producing and distributing digital information goods. It also coincides with the emergence of new competitive strategies such as large-scale bundling. In this paper, we show that bundling can create “economies of aggregation” for information goods if their marginal costs are very low, even in the absence of network externalities or economies of scale or scope. We extend the Bakos-Brynjolfsson bundling model (1999) to settings with sev...
Bundling and joint marketing by rival firms
Jeitschko, Thomas D.; Jung, Yeonjei; Kim, Jaesoo
2014-01-01
We study joint marketing arrangements by competing firms who engage in price discrimination between consumers who patronize only one firm (single purchasing) and those who purchase from both competitors (bundle purchasers). Two types of joint marketing are considered. Firms either commit to a component-price that applies to bundle-purchasers and then firms set stand-alone prices for single purchasers; or firms commit to a rebate off their stand alone price that will be applied to bundle-purch...
Statistical Constitutive Equation of Aramid Fiber Bundles
Institute of Scientific and Technical Information of China (English)
熊杰; 顾伯洪; 王善元
2003-01-01
Tensile impact tests of aramid (Twaron) fiber bundles were carried om under high strain rates with a wide range of 0. 01/s～1000/s by using MTS and bar-bar tensile impact apparatus. Based on the statistical constitutive model of fiber bundles, statistical constitutive equations of aramid fiber bundles are derived from statistical analysis of test data at different strain rates. Comparison between the theoretical predictions and experimental data indicates statistical constitutive equations fit well with the experimental data, and statistical constitutive equations of fiber bundles at different strain rates are valid.
Hydraulic characteristics of HANARO fuel bundles
Energy Technology Data Exchange (ETDEWEB)
Cho, S.; Chung, H. J.; Chun, S. Y.; Yang, S. K.; Chung, M. K. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1997-12-31
This paper presents the hydraulic characteristics measured by using LDV (Laser Doppler Velocimetry) in subchannels of HANARO, KAERI research reactor, fuel bundle. The fuel bundle consists of 18 axially finned rods with 3 spacer grids, which are arranged in cylindrical configuration. The effects of the spacer grids on the turbulent flow were investigated by the experimental results. Pressure drops for each component of the fuel bundle were measured, and the friction factors of fuel bundle and loss coefficients for the spacer grids were estimated from the measured pressure drops. Implications regarding the turbulent thermal mixing were discussed. Vibration test results measured by using laser vibrometer were presented. 9 refs., 12 figs. (Author)
Evaluation of nonequilibrium effects in bundle dispersed-flow film boiling
International Nuclear Information System (INIS)
The effects of thermodynamic nonequilibrium in dispersed flow film boiling heat transfer are examined. Steady-state and transient rod-bundle data are used to evaluate several empirical heat-transfer models commonly employed to predict post-CHF behavior. The models that account for thermodynamic nonequilibrium perform adequately, while those that ignore nonequilibrium effects incur errors in wall superheat as high as 1900K. Nonequilibrium effects can also be treated by explicitly modeling the phenomena. The thermal-hydraulic code COBRA-TF employs this approach. Using bundle data, the models in the code are evaluated. Analysis suggests that the interfacial heat transfer is overpredicted
Extension of holomorphic bundles to the disc (and Serre's Problem on Stein bundles)
Rosay, Jean-Pierre
2006-01-01
We show how to extend some holomorphic bundles with fifer C^2 and base an open set in C, to bundles on the Riemann Sphere, by an extremely simple technique. In particular, it applies to examples of non-Stein bundles constructed by Skoda and Demailly. It gives an example on C, with polynomial transition automorphisms.
Neutronics and thermo-hydraulic design of supercritical-water cooled solid breeder TBM
Energy Technology Data Exchange (ETDEWEB)
Cheng, Jie; Wu, Yingwei, E-mail: wyw810@mail.xjtu.edu.cn; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng
2015-03-15
Highlights: • A supercritical-water cooled solid breeder test blanket module (SWCB TBM) was designed. • The neutronics calculations show that the tritium breeding ratio (TBR) of SWCB TBM is 1.17. • The outlet temperature of SWCB TBM can reach as high as 500 °C. • Both thermal stress and deformation of the SWCB TBM design are within safety limits. - Abstract: In this paper, the supercritical-water cooled solid breeder test blanket module (SWCB TBM), using the supercritical water as the coolant, Li{sub 4}SiO{sub 4} lithium ceramic pebbles as a breeder, and beryllium pebbles as a neutron multiplier, was designed and analyzed for ITER. The results of neutronics, thermo-hydraulic and thermo-mechanical analysis are presented for the SWCB TBM. Neutronics calculations show that the proposed TBM has high tritium breeding ratio and power density. The tritium breeding ratio (TBR) of the proposed design is 1.17, which is greater than that of 1.15 required for tritium self-sufficiency. The thermo-hydraulic calculation proved that the TBM components can be effectively cooled to the allowable temperature with the temperature of outlet reaching 500 °C. According to thermo-mechanics calculation results, the first wall with the width of 17 mm is safe and the deformation of first wall is far below the limited value. All the results showed that the current TBM design was reasonable under the ITER normal condition.
Investigation of the thermohydraulic systems in MATLAB & Simulink using developed library
Directory of Open Access Journals (Sweden)
Lavicka D.
2013-04-01
Full Text Available Recent developments in the designing of the thermohydraulic systems have heightened the need for special software tools. Possibility of incorporating the measured data, easy and quick rearrangement of the system and short simulation time play a key role in successful progress in designing of the thermohydraulic systems. Commercial software sometimes does not fit to our special tasks or it lacks some utilities. This paper attempts to show usefulness of MATLAB and Simulink in the development of the “drag and drop” software. In the Simulink it is possible to develop specialized libraries containing graphical blocks representing parts of the real systems. Easy cooling circuit was designed using developed graphical blocks. The main parts of the circuit consist of the heat source and heat exchanger/cooler. Another parts such as pumps, pipes and properties of the cooling medium etc. are taken into account in the investigated system. These results obtained by the developed software are presented for similar circuits with different heat exchangers/coolers. Simulations for different heat exchangers were solved with different initial conditions. These results are consequently compared. Characteristic properties (heating/cooling power and pressure drop according to flow rate of the heat exchangers/coolers were obtained experimentally in our laboratory and by the numerical CFD simulation in ANSYS/FLUENT software.
EBaLM-THP - A neural network thermohydraulic prediction model of advanced nuclear system components
International Nuclear Information System (INIS)
In lieu of the worldwide energy demand, economics and consensus concern regarding climate change, nuclear power - specifically near-term nuclear power plant designs are receiving increased engineering attention. However, as the nuclear industry is emerging from a lull in component modeling and analyses, optimization for example using ANN has received little research attention. This paper presents a neural network approach, EBaLM, based on a specific combination of two training algorithms, error-back propagation (EBP), and Levenberg-Marquardt (LM), applied to a problem of thermohydraulics predictions (THPs) of advanced nuclear heat exchangers (HXs). The suitability of the EBaLM-THP algorithm was tested on two different reference problems in thermohydraulic design analysis; that is, convective heat transfer of supercritical CO2 through a single tube, and convective heat transfer through a printed circuit heat exchanger (PCHE) using CO2. Further, comparison of EBaLM-THP and a polynomial fitting approach was considered. Within the defined reference problems, the neural network approach generated good results in both cases, in spite of highly fluctuating trends in the dataset used. In fact, the neural network approach demonstrated cumulative measure of the error one to three orders of magnitude smaller than that produce via polynomial fitting of 10th order
International Nuclear Information System (INIS)
Analyses of the integrity of the reactor pressure vessel (RPV) of NPP Stade (KKS) under emergency core cooling conditions have been carried out using experimental and theoretical/numerical investigations to improve modelling of the complex thermohydraulic processes during loss of coolant accidents and to gain more realistic input data for an updated fracture mechanics analysis. The results presented for RPV-KKS take into account the effect of austenitic cladding and include a reevaluation and application of the warm prestress effect. The results of thermohydraulic analyses were used as input to finite-element calculations to evaluate temperature and stresses in the vessel wall. Fracture mechanics analyses have proven the exclusion of crack initiation for crack sizes below and even above the detection threshold of the continuously improved NDE method. With this result, the safety against brittle fracture according to the first KTA/ASME criterion - exclusion of crack initiation - is proven. In addition, finite-element arrest analyses were performed, showing that all postulated cracks were arrested within the allowable crack depth of 3/4 of the wall thickness. By that, the safety against brittle fracture according to the second KTA/ASME criterion - crack arrest - is confirmed, also if the NDE results were totally neglected. Therefore, no critical crack depth exists as a possible reason for a failure of the pressure vessel. (J.S.). 12 refs., 8 figs
Analysis of Westinghouse MB2 test using the Steam-generator Thermohydraulics Analysis code STAF
International Nuclear Information System (INIS)
Highlights: • Westinghouse MB2 test is analyzed by the STAF code. • Measured data from MB2 test agree well with calculated results. • Distribution of temperature, heat flux and heat transfer coefficient is analyzed. • Parameters for FIV analysis are presented. - Abstract: In the present study, we develop a Steam-generator Thermohydraulics Analysis code based on Fluent (STAF) for predicting the three-dimensional localized thermal–hydraulic characteristics in the primary and secondary sides of steam generator. STAF code is developed based on the porous media model in Fluent. The flow resistances caused by the tubes, support plates, downcomer and separators are introduced to the momentum equation as additional source terms of shell side fluid; the heat transfer from primary to secondary side fluid is considered as the source term of energy equation of secondary side fluid. The flow and heat transfer in primary side, as well as the tube-to-shell-side heat transfer are solved by the user-defined functions in Fluent. STAF is used to simulate the Westinghouse MB2 test, and localized thermohydraulics parameters are obtained. The numerical results show good agreement with experimental results, demonstrating the ability of STAF to model the three-dimensional flow and heat transfer characteristics in primary and secondary side of steam generator. Besides, parameters associated with flow-induced vibration are also analyzed
International Nuclear Information System (INIS)
Highlights: • A new Th–U mixed fuel assembly for SCWR has been introduced and investigated. • Neutronic and thermohydraulic characteristics of the new assembly have been studied. • The new fuel assembly satisfies design rules of SCWR. • The introduced fuel assembly can fulfill the sustainable breeding Th–U cycle. • The new fuel assembly also has advantages with respect to lower generation of minor actinides and reactor safety. - Abstract: The exploitation of thorium fuel is a promising way to overcome the pressing problems of nuclear fuel supply, nuclear waste and nuclear proliferation. In this paper, a novel conceptual design of a breeding thorium–uranium (Th–U) mixed fuel assembly in SCWR is proposed, which is aimed to achieve the breeding ratio bigger than 1.0, so as to fulfill the sustainable breeding thorium–uranium cycle. Through the calculations of neutronics and neutronic/thermohydraulic (N–T) coupling, the results indicate that the introduced conceptual design of a breeding Th–U mixed fuel assembly in SCWR satisfies design rules of SCWR, with considerable advantages with respect to breeding performance, lower minor actinide generation and reactor safety
Neutronics and thermo-hydraulic design of supercritical-water cooled solid breeder TBM
International Nuclear Information System (INIS)
Highlights: • A supercritical-water cooled solid breeder test blanket module (SWCB TBM) was designed. • The neutronics calculations show that the tritium breeding ratio (TBR) of SWCB TBM is 1.17. • The outlet temperature of SWCB TBM can reach as high as 500 °C. • Both thermal stress and deformation of the SWCB TBM design are within safety limits. - Abstract: In this paper, the supercritical-water cooled solid breeder test blanket module (SWCB TBM), using the supercritical water as the coolant, Li4SiO4 lithium ceramic pebbles as a breeder, and beryllium pebbles as a neutron multiplier, was designed and analyzed for ITER. The results of neutronics, thermo-hydraulic and thermo-mechanical analysis are presented for the SWCB TBM. Neutronics calculations show that the proposed TBM has high tritium breeding ratio and power density. The tritium breeding ratio (TBR) of the proposed design is 1.17, which is greater than that of 1.15 required for tritium self-sufficiency. The thermo-hydraulic calculation proved that the TBM components can be effectively cooled to the allowable temperature with the temperature of outlet reaching 500 °C. According to thermo-mechanics calculation results, the first wall with the width of 17 mm is safe and the deformation of first wall is far below the limited value. All the results showed that the current TBM design was reasonable under the ITER normal condition
Principal Bundles on the Projective Line
Indian Academy of Sciences (India)
V B Mehta; S Subramanian
2002-08-01
We classify principal -bundles on the projective line over an arbitrary field of characteristic ≠ 2 or 3, where is a reductive group. If such a bundle is trivial at a -rational point, then the structure group can be reduced to a maximal torus.
The Verlinde formula for Higgs bundles
Andersen, Jørgen Ellegaard; Pei, Du
2016-01-01
We propose and prove the Verlinde formula for the quantization of the Higgs bundle moduli spaces and stacks for any simple and simply-connected group. This generalizes the equivariant Verlinde formula for the case of $SU(n)$ proposed previously by the second and third author. We further establish a Verlinde formula for the quantization of parabolic Higgs bundle moduli spaces and stacks.
CANFLEX fuel bundle strength tests (test report)
Energy Technology Data Exchange (ETDEWEB)
Chang, Seok Kyu; Chung, C. H.; Kim, B. D.
1997-08-01
This document outlines the test results for the strength tests of the CANFLEX fuel bundle. Strength tests are performed to determine and verify the amount of the bundle shape distortion which is against the side-stops when the bundles are refuelling. There are two cases of strength test; one is the double side-stop test which simulates the normal bundle refuelling and the other is the single side-stop test which simulates the abnormal refuelling. the strength test specification requires that the fuel bundle against the side-stop(s) simulators for this test were fabricated and the flow rates were controlled to provide the required conservative hydraulic forces. The test rig conditions of 120 deg C, 11.2 MPa were retained for 15 minutes after the flow rate was controlled during the test in two cases, respectively. The bundle loading angles of number 13- number 15 among the 15 bundles were 67.5 deg CCW and others were loaded randomly. After the tests, the bundle shapes against the side-stops were measured and inspected carefully. The important test procedures and measurements were discussed as follows. (author). 5 refs., 22 tabs., 5 figs.
CANFLEX fuel bundle strength tests (test report)
International Nuclear Information System (INIS)
This document outlines the test results for the strength tests of the CANFLEX fuel bundle. Strength tests are performed to determine and verify the amount of the bundle shape distortion which is against the side-stops when the bundles are refuelling. There are two cases of strength test; one is the double side-stop test which simulates the normal bundle refuelling and the other is the single side-stop test which simulates the abnormal refuelling. the strength test specification requires that the fuel bundle against the side-stop(s) simulators for this test were fabricated and the flow rates were controlled to provide the required conservative hydraulic forces. The test rig conditions of 120 deg C, 11.2 MPa were retained for 15 minutes after the flow rate was controlled during the test in two cases, respectively. The bundle loading angles of number 13- number 15 among the 15 bundles were 67.5 deg CCW and others were loaded randomly. After the tests, the bundle shapes against the side-stops were measured and inspected carefully. The important test procedures and measurements were discussed as follows. (author). 5 refs., 22 tabs., 5 figs
k-Gerbes, Line Bundles and Anomalies
Ekstrand, C
2000-01-01
We use sets of trivial line bundles for the realization of gerbes. For1-gerbes the structure arises naturally for the Weyl fermion vacuum bundle at afixed time. The Schwinger term is an obstruction in the triviality of a1-gerbe.
k-Gerbes, Line Bundles and Anomalies
International Nuclear Information System (INIS)
We use sets of trivial line bundles for the realization of gerbes. For 1-gerbes the structure arises naturally for the Weyl fermion vacuum bundle at a fixed time. The Schwinger term is an obstruction in the triviality of a 1-gerbe. (author)
k-Gerbes, Line Bundles and Anomalies
Ekstrand, C.
2000-01-01
We use sets of trivial line bundles for the realization of gerbes. For 1-gerbes the structure arises naturally for the Weyl fermion vacuum bundle at a fixed time. The Schwinger term is an obstruction in the triviality of a 1-gerbe.
Heights for line bundles on arithmetic surfaces
Jahnel, Joerg
1995-01-01
For line bundles on arithmetic varieties we construct height functions using arithmetic intersection theory. In the case of an arithmetic surface, generically of genus g, for line bundles of degree g equivalence is shown to the height on the Jacobian defined by the Theta divisor.
Damping Properties of the Hair Bundle
Baumgart, Johannes; Kozlov, Andrei S.; Risler, Thomas; Hudspeth, A. J.
2011-11-01
The viscous liquid surrounding a hair bundle dissipates energy and dampens oscillations, which poses a fundamental physical challenge to the high sensitivity and sharp frequency selectivity of hearing. To identify the mechanical forces at play, we constructed a detailed finite-element model of the hair bundle. Based on data from the hair bundle of the bullfrog's sacculus, this model treats the interaction of stereocilia both with the surrounding liquid and with the liquid in the narrow gaps between the individual stereocilia. The investigation revealed that grouping stereocilia in a bundle dramatically reduces the total drag. During hair-bundle deflections, the tip links potentially induce drag by causing small but very dissipative relative motions between stereocilia; this effect is offset by the horizontal top connectors that restrain such relative movements at low frequencies. For higher frequencies the coupling liquid is sufficient to assure that the hair bundle moves as a unit with a low total drag. This work reveals the mechanical characteristics originating from hair-bundle morphology and shows quantitatively how a hair bundle is adapted for sensitive mechanotransduction.
Fock modules and noncommutative line bundles
Landi, Giovanni
2016-09-01
To a line bundle over a noncommutative space there is naturally associated a Fock module. The algebra of corresponding creation and annihilation operators is the total space algebra of a principal U(1) -bundle over the noncommutative space. We describe the general construction and illustrate it with examples.
TRAWA, a transient analysis code for water reactions
International Nuclear Information System (INIS)
TRAWA is a transient analysis code for water reactors. It solves the two-group neutron diffusion equations simultaneously with the heat conduction equations and the two-phase hydraulic equations for one or more channels. At most one-dimensional submodels are used. Neither thermal nor hydraulic mixing appear between channels. Doppler, coolant density, coolant temperature, and soluble poison density feedbacks due to the thermohydraulics of the channels are described by using polynomial expansions for the group constants. The hydraulic circuit outside the reactor core consists of by-pass channel and risers with two-phase flow and of pump lines with incompressible flow. Nontrivial implicit methods are employed in the discretization of the equations to allow for sparse spatial mesh and flexible choice of time steps. Various transients can be calculated by applying external disturbances. The code is extensively supplied by input and output capabilities. TRAWA is written in FORTRAN V for UNIVAC 1108 computer. (author)
International Nuclear Information System (INIS)
The problems of numerical modeling of thermohydraulics in assembly of fuel elements of fast reactors with the partial blockage of cross-section under the coolant are considered. The information about existing codes constructed on use of subchannel technique and model of porous body are presented. The results of calculation obtained by these codes are presented. (author)
Tone, Florentina
2011-01-01
Pursuing our work in [18], [17], [20], [5], we consider in this article the two-dimensional thermohydraulics equations. We discretize these equations in time using the implicit Euler scheme and we prove that the global attractors generated by the numerical scheme converge to the global attractor of the continuous system as the time-step approaches zero.
Dirac structures and Dixmier-Douady bundles
Alekseev, A
2009-01-01
A Dirac structure on a vector bundle V is a maximal isotropic subbundle E of the direct sum of V with its dual. We show how to associate to any Dirac structure a Dixmier-Douady bundle A, that is, a Z/2Z-graded bundle of C*-algebras with typical fiber the compact operators on a Hilbert space. The construction has good functorial properties, relative to Morita morphisms of Dixmier-Douady bundles. As applications, we show that the `spin' Dixmier-Douady bundle over a compact, connected Lie group (as constructed by Atiyah-Segal) is multiplicative, and we obtain a canonical `twisted Spin-c-structure' on spaces with group valued moment maps.
Bringing the CANFLEX fuel bundle to market
International Nuclear Information System (INIS)
CANFLEX is a 43-element CANDU fuel bundle, under joint development by AECL and KAERI, to facilitate the use of various advanced fuel cycles in CANDU reactors through the provision of enhanced operating margins. The bundle uses two element diameters (13.5 and 11.5 mm ) to reduce element ratings by 20%, and includes the use of critical-heat-flux (CHF) enhancing appendages to increase the minimum CHF ratio or dryout margin of the bundle. Test programs are underway to demonstrate: the irradiation behaviour, hydraulic characteristics and reactor physics properties of the bundle, along with a test program to demonstrate the ability of the bundle to be handled by CANDU-6 fuelling machines. A fuel design manual and safety analysis reports have been drafted, and both analyses, plus discussions with utilities are underway for a demonstration irradiation in a CANDU-6 reactor. (author)
Line bundle embeddings for heterotic theories
Nibbelin, Stefan Groot; Ruehle, Fabian
2016-04-01
In heterotic string theories consistency requires the introduction of a non-trivial vector bundle. This bundle breaks the original ten-dimensional gauge groups E8 × E8 or SO(32) for the supersymmetric heterotic string theories and SO(16) × SO(16) for the non-supersymmetric tachyon-free theory to smaller subgroups. A vast number of MSSM-like models have been constructed up to now, most of which describe the vector bundle as a sum of line bundles. However, there are several different ways of describing these line bundles and their embedding in the ten-dimensional gauge group. We recall and extend these different descriptions and explain how they can be translated into each other.
CANDU fuel bundle skin friction factor
International Nuclear Information System (INIS)
Single-phase, incompressible fluid flow skin friction factor correlations, primarily for CANDU 37-rod fuel bundles, were reviewed. The correlations originated from curve-fits to flow test data, mostly with new fuel bundles in new pressure tubes (flow tubes), without internal heating. Skin friction in tubes containing fuel bundles (noncircular flow geometry) was compared to that in equivalent diameter smooth circular tubes. At Reynolds numbers typical of normal flows in CANDU fuel channels, the skin friction in tubes containing bundles is 8 to 15% higher than in equivalent diameter smooth circular tubes. Since the correlations are based on scattered results from measurements, the skin friction with bundles may be even higher than indicated above. The information permits over- or under-prediction of the skin friction, or choosing an intermediate value of friction, with allowance for surface roughnesses, in thermal-hydraulic analyses of CANDU heat transport systems. (author) 9 refs., 2 figs
Line bundle embeddings for heterotic theories
Nibbelink, Stefan Groot
2016-01-01
In heterotic theories consistency requires the introduction of a non-trivial vector bundle. This bundle breaks the original ten-dimensional gauge groups E_8 x E_8 or SO(32) for the supersymmetric heterotic theories and SO(16) x SO(16) for the non-supersymmetric tachyon-free theory to smaller subgroups. A vast number of MSSM-like models have been constructed up to now, most of which describe the vector bundle as a sum of line bundles. However, there are several different ways of describing these line bundles and their embedding in the ten-dimensional gauge group. We recall and extend these different descriptions and explain how they can be translated into each other.
Canonical singular hermitian metrics on relative logcanonical bundles
Tsuji, Hajime
2010-01-01
This supersedes 0704.0566. We prove the invariance of logarithmic plurigenera for a projective family of KLT pairs and the adjoint line bundle of KLT line bundles. The proof uses the canonical singular hermitian metrics on relative logcanonical bundles.
On Harder–Narasimhan Reductions for Higgs Principal Bundles
Indian Academy of Sciences (India)
Arijit Dey; R Parthasarathi
2005-05-01
The existence and uniqueness of – reduction for the Higgs principal bundles over nonsingular projective variety is shown. We also extend the notion of – reduction for (, )-bundles and ramified -bundles over a smooth curve.
Effect of bundle size on BWR fuel bundle critical power performance
International Nuclear Information System (INIS)
Effect of the bundle size on the BWR fuel bundle critical power performance was studied. For this purpose, critical power tests were conducted with both 6 x 6 (36 heater rods) and 12 x 12 (144 heater rods) size bundles in the GE ATLAS heat transfer test facility located in San Jose, California. All the bundle geometries such as rod diameter, rod pitch and rod space design are the same except size of flow channel. Two types of critical power tests were performed. One is the critical power test with uniform local peaking pattern for direct comparison of the small and large bundle critical power. Other is the critical power test for lattice positions in the bundle. In this test, power of a group of four rods (2 x 2 array) in a lattice region was peaked higher to probe the critical power of that lattice position in the bundle. In addition, the test data were compared to the COBRAG calculations. COBRAG is a detailed subchannel analysis code for BWR fuel bundle developed by GE Nuclear Energy. Based on these comparisons the subchannel model was refined to accurately predict the data obtained in this test program, thus validating the code capability of handling the effects of bundle size on bundle critical power for use in the study of the thermal hydraulic performance of the future advance BWR fuel bundle design. The author describes the experimental portion of the study program
Gauge symmetries and fibre bundles
International Nuclear Information System (INIS)
The matter is organized as follows. After a brief introduction to the concept of gauge invariance and its relationship to determinism, we introduce in chapters 3 and 4 the notion of fibre bundles in the context of a discussion on spinning point particles and Dirac monopoles. Chapter 3 deals with a non relativistic treatment of the spinning particle. The non trivial extension to relativistic spinning particles is dealt with in Chapter 5. The free particle system as well as interactions with external electromagnetic and gravitational fields are discussed in detail. In chapter 5 we also elaborate on a remarkable relationship between the charge-monopole system and the system of a massless particle with spin. The classical description of Yang-Mills particles with internal degrees of freedom, such as isospin or colour, is given in chapter 6. We apply the above in a discussion of the classical scattering of particles off a 't Hooft-Polyakov monopole. In chapter 7 we elaborate on a Kaluza-Klein description of particles with internal degrees of freedom. The canonical formalism and the quantization of most of the preceeding systems are discussed in chapter 8. The dynamical systems given in chapters 3-7 are formulated on group manifolds. The procedure for obtaining the extension to super-group manifolds is briefly discussed in chapter 9. In chapter 10, we show that if a system admits only local Lagrangians for a configuration space Q, then under certain conditions, it admits a global Lagrangian when Q is enlarged to a suitable U(1) bundle over Q. Conditions under which a symplectic form is derivable from a Lagrangian are also found. (orig./HSI)
An investigation of highly pressurized transient fluid flow in pipelines
International Nuclear Information System (INIS)
This paper discusses transient processes in natural gas pipelines. The method of characteristics (MOC) is applied for the analysis of two transient categories, where the governing one-dimensional, hyperbolic conservation equations are linearized and solved without neglecting any of their term. First, we present a parametric study of the pressurized flow encountered when pipelines are utilized for the transportation or the temporary storage of natural gas. The non-ideal compressibility of natural gas is included in the model and its impact on the thermo-hydraulic processes is elucidated. Second, we model the hydrodynamics of a pipeline whose downstream boundary is a periodic discharge rate. The results show that, in response to these boundary conditions, the pressure distribution in the pipeline also undergoes periodic variations. Furthermore, our simulation results confirm the usefulness of MOC for numerical simulation of flow phenomena in pipelines. - Highlights: ► The goal of this work was an investigation of high pressure fluid transients through long pipelines. ► The MOC method was used for the simulation of the hydrodynamic processes in response to transients. ► The first transient type represented the line packing and subsequent pressurization. ► Special attention was given to the effect of non-ideal compressibility of natural gas flow. ► The second type of transients dealt with the effect of periodic downstream boundary conditions.
Methodology of study of the boiling crisis in a nuclear fuel rod bundle
International Nuclear Information System (INIS)
The boiling crisis is one of the phenomena limiting the available power from a nuclear power plant. It has been widely studied for decades, and numerous data, models, correlations or tables are now available in the literature. If we now try to obtain a general view of previous work in this field, we may note that there are several ways of tackling the subject. The mechanistic models try to modelize the two-phase flow topology and the interaction between different sublayers, and must be validated by comparison with basic experiments, such as DEBORA, where we try to get some detailed informations on the two-phase flow pattern in a pure and simple geometry. This allows us to obtain a better knowledge of the so-called 'intrinsic effect'. Up to now, these models are not yet acceptable for a nuclear use. As the geometry of the rod bundles and grids has a tremendous importance for the actual Critical Heat Flux (CHF), it is compulsory to have more precise results for a given fuel rod bundle in a restricted range of parameter: this leads to the empirical approach, using empirical CHF predictors (tables, correlations, splines, ...). One of the key points of such a method is the obtention of the local thermohydraulic values, that is to say the evaluation of the so-called 'mixing effect'. This is done by a subchannel analysis code or equivalent, which can be qualified on two kinds of experiments: overall flow measurements in a subchannel, such as HYDROMEL in single-phase flow or GRAZIELLA in two-phase flow, or detailed measurements inside a subchannel, such as AGATE. Nevertheless, the final qualification of a specific nuclear fuel, i.e. the synthesis of these mechanistic and empirical approaches, intrinsic and mixing effects, ..., must be achieved on a global test such as OMEGA. This is the strategy used in France by CEA and his partners FRAMATOME and EdF. (author)
Preliminary report: NIF laser bundle review
International Nuclear Information System (INIS)
As requested in the guidance memo 1, this committe determined whether there are compelling reasons to recommend a change from the NIF CDR baseline laser. The baseline bundle design based on a tradeoff between cost and technical risk, which is replicated four times to create the required 192 beams. The baseline amplifier design uses bottom loading 1x4 slab and flashlamp cassettes for amplifier maintenance and large vacuum enclosures (2.5m high x 7m wide in cross-section for each of the two spatial filters in each of the four bundles. The laser beams are arranged in two laser bays configured in a u-shape around the target area. The entire bundle review effort was performed in a very short time (six weeks) and with limited resources (15 personnel part-time). This should be compared to the effort that produced the CDR design (12 months, 50 to 100 personnel). This committee considered three alternate bundle configurations (2x2, 4x2, and 4x4 bundles), and evaluated each bundle against the baseline design using the seven requested issues in the guidance memo: Cost; schedule; performance risk; maintainability/operability; hardware failure cost exposure; activation; and design flexibility. The issues were reviewed to identify differences between each alternate bundle configuration and the baseline
Prioritary omalous bundles on Hirzebruch surfaces
Aprodu, Marian; Marchitan, Marius
2016-01-01
An irreducible algebraic stack is called unirational if there exists a surjective morphism, representable by algebraic spaces, from a rational variety to an open substack. We prove unirationality of the stack of prioritary omalous bundles on Hirzebruch surfaces, which implies also the unirationality of the moduli space of omalous H-stable bundles for any ample line bundle H on a Hirzebruch surface (compare with Costa and Miro-Ŕoig, 2002). To this end, we find an explicit description of the duals of omalous rank-two bundles with a vanishing condition in terms of monads. Since these bundles are prioritary, we conclude that the stack of prioritary omalous bundles on a Hirzebruch surface different from P1 ×P1 is dominated by an irreducible section of a Segre variety, and this linear section is rational (Ionescu, 2015). In the case of the space quadric, the stack has been explicitly described by N. Buchdahl. As a main tool we use Buchdahl's Beilinson-type spectral sequence. Monad descriptions of omalous bundles on hypersurfaces in P4, Calabi-Yau complete intersection, blowups of the projective plane and Segre varieties have been recently obtained by A.A. Henni and M. Jardim (Henni and Jardim, 2013), and monads on Hirzebruch surfaces have been applied in a different context in Bartocci et al. (2015).
Singular hermitian metrics on vector bundles
De Cataldo, M A A
1997-01-01
We introduce a notion of singular hermitian metrics (s.h.m.) for holomorphic vector bundles and define positivity in view of $L^2$-estimates. Associated with a suitably positive s.h.m. there is a (coherent) sheaf 0-th kernel of a certain $d''$-complex. We prove a vanishing theorem for the cohomology of this sheaf. All this generalizes to the case of higher rank known results of Nadel for the case of line bundles. We introduce a new semi-positivity notion, $t$-nefness, for vector bundles, establish some of its basic properties and prove that on curves it coincides with ordinary nefness. We particularize the results on s.h.m. to the case of vector bundles of the form $E=F \\otimes L$, where $F$ is a $t$-nef vector bundle and $L$ is a positive (in the sense of currents) line bundle. As applications we generalize to the higher rank case 1) Kawamata-Viehweg Vanishing Theorem, 2) the effective results concerning the global generation of jets for the adjoint to powers of ample line bundles, and 3) Matsusaka Big Theor...
Deformations of the generalised Picard bundle
International Nuclear Information System (INIS)
Let X be a nonsingular algebraic curve of genus g ≥ 3, and let Mξ denote the moduli space of stable vector bundles of rank n ≥ 2 and degree d with fixed determinant ξ over X such that n and d are coprime. We assume that if g = 3 then n ≥ 4 and if g = 4 then n ≥ 3, and suppose further that n0, d0 are integers such that n0 ≥ 1 and nd0 + n0d > nn0(2g - 2). Let E be a semistable vector bundle over X of rank n0 and degree d0. The generalised Picard bundle Wξ(E) is by definition the vector bundle over Mξ defined by the direct image pMξ *(Uξ x pX*E) where Uξ is a universal vector bundle over X x Mξ. We obtain an inversion formula allowing us to recover E from Wξ(E) and show that the space of infinitesimal deformations of Wξ(E) is isomorphic to H1(X, End(E)). This construction gives a locally complete family of vector bundles over Mξ parametrised by the moduli space M(n0,d0) of stable bundles of rank n0 and degree d0 over X. If (n0,d0) = 1 and Wξ(E) is stable for all E is an element of M(n0,d0), the construction determines an isomorphism from M(n0,d0) to a connected component M0 of a moduli space of stable sheaves over Mξ. This applies in particular when n0 = 1, in which case M0 is isomorphic to the Jacobian J of X as a polarised variety. The paper as a whole is a generalisation of results of Kempf and Mukai on Picard bundles over J, and is also related to a paper of Tyurin on the geometry of moduli of vector bundles. (author)
Prediction of void fraction in a subchannel and bundle geometry with FLICA4 and TRACE
International Nuclear Information System (INIS)
In order to encourage advancement in subchannel analyses of fluid flow in rod bundles, an international benchmark program, namely the OECD/NRC PWR Subchannel and Bundle Test (PSBT) benchmark, has been organized. The PSBT benchmark aims at assessing the capabilities of subchannel analysis codes, system codes, and computational fluid dynamics (CFD) codes for the prediction of detailed void distributions in subchannels, including departure from nucleate boiling, on the basis of experimental data measured at a full scale prototypical PWR rod bundle. Within the framework of the PSBT benchmark, analyses of the void distribution in a subchannel and bundle geometry have been performed at the Paul Scherrer Institut by means of the subchannel analysis code FLICA4, and the thermal-hydraulic system code TRACE. Steady-state scenarios are analyzed by using both FLICA4 and TRACE and transient analyses are performed with FLICA4. In particular, the TRACE calculations for bundle geometry have been carried out employing a three-dimensional vessel component to model cross flows between subchannels. The analysis aims at evaluating the applicability of a three-dimensional vessel component of TRACE for subchannel analyses, as well as validating the subchannel code FLICA4. The calculated void fractions are compared to the experimental data, and the accuracies of the predictions by both codes are appraised by means of a statistical analysis. (author)
Geometry of quantum principal bundles, 1
Durdevic, M
1995-01-01
A theory of principal bundles possessing quantum structure groups and classical base manifolds is presented. Structural analysis of such quantum principal bundles is performed. A differential calculus is constructed, combining differential forms on the base manifold with an appropriate differential calculus on the structure quantum group. Relations between the calculus on the group and the calculus on the bundle are investigated. A concept of (pseudo)tensoriality is formulated. The formalism of connections is developed. In particular, operators of horizontal projection, covariant derivative and curvature are constructed and analyzed. Generalizations of the first structure equation and of the Bianchi identity are found. Illustrative examples are presented.
Weak equivalence classes of complex vector bundles
Hông-Vân Lê
2006-01-01
For any complex vector bundle Ek of rank k over a manifold Mm with Chern classes ci Î H2i(Mm, Z) and any non-negative integers l1, . . ., lk we show the existence of a positive number p(m, k) and the existence of a complex vector bundle Êk over Mm whose Chern classes are p(m, k) × li × ci Î H2i(Mm, Z). We also discuss a version of this statement for holomorphic vector bundles over projective algebraic manifolds.
Assembly mechanism for nuclear fuel bundles
International Nuclear Information System (INIS)
A description is given of a nuclear power reactor fuel bundle having tie rods fastened to a lower tie plate and passing through openings in the upper tie plate with the assembled bundle secured by rotatable locking sleeves which engage slots provided in the upper tie plate. Pressure exerted by helical springs mounted around each of the fuel rods urge the upper tie plate against the locking sleeves. The bundle may be disassembled after depressing the upper tie plate and rotating the locking sleeves to the unlocked position
Vector supersymmetry in the universal bundle
International Nuclear Information System (INIS)
We present a vector supersymmetry for Witten-type topological gauge theories, and examine its algebra in terms of a superconnection formalism. When covariant constraints on the supercurvature are chosen, a correspondence is established with the universal bundle construction of Atiyah and Singer. The vector supersymmetry represents a certain shift operator in the curvature of the universal bundle, and can be used to generate the hierarchy of observables in these theories. This formalism should lead to the construction of vector supergravity theories, and perhaps to the gravitational analogue of the universal bundle. (orig.)
Bundle duct interaction studies for fuel assemblies
International Nuclear Information System (INIS)
It is known that the wire-wrapped rods and duct in an LMFBR are undergoing a gradual structural distortion from the initially uniform geometry under the combined effects of thermal expansion and irradiation induced swelling and creep. These deformations have a significant effect on flow characteristics, thus causing changes in thermal behavior such as cladding temperature and temperature distribution within a bundle. The temperature distribution may further enhance or retard irradiation induced deformation of the bundle. This report summarizes the results of the continuing effort in investigating the bundle-duct interaction, focusing on the need for the large development plant
Overview of recent deterministic thermohydraulic analyses of operational events for Slovak NPPs
International Nuclear Information System (INIS)
Based on review of cooperation of the VUJE, Inc. with Slovak NPP operator within the last three years, the paper describes selected operational occurrences, which were required to be analyzed from thermohydraulic point of view. For each event a short problem description is given, followed by information of analytical methodology and approach, as well as overview of the most important findings. The events, described in detail include problems with mechanical fatigue of the pressurizer surge line due to the thermal stratification in the upper part of the line (CFD simulation using FLUENT code), leakage from the primary circuit due to the thermal stress and fatigue of the ECCS pipelines (RELAP5 simulation of the ECCS line section with two back valves) and low-level leakage through the reactor flange sealing (RELAP5 simulation together with activity analysis). (author)
Thermo-hydraulic Quench Propagation at the LHC Superconducting Magnet String
Rodríguez-Mateos, F; Serio, L
1998-01-01
The superconducting magnets of the LHC are protected by heaters and cold by-pass diodes. If a magnet quenches, the heaters on this magnet are fired and the magnet chain is de-excited in about two minu tes by opening dump switches in parallel to a resistor. During the time required for the discharge, adjacent magnets might quench due to thermo-hydraulic propagation in the helium bath and/or heat con duction via the bus bar. The number of quenching magnets depends on the mechanisms for the propagation. In this paper we report on quench propagation experiments from a dipole magnet to an adjacent ma gnet. The mechanism for the propagation is hot helium gas expelled from the first quenching magnet. The propagation changes with the pressure opening settings of the quench relief valves.
Analysis of natural circulation stability in a low pressure thermohydraulic test loop
International Nuclear Information System (INIS)
This paper discusses an instability study of a natural circulation (NC) loop performed with the aid of Relap5 thermal-hydraulic system code. This loop has been designed and constructed for the analysis of relevant thermohydraulic parameters of a nuclear reactor. In this study, the main parameters for the stability of NC are identified and characterized through the execution of proper code runs. The obtained stability boundary (SB) in the dimensionless Zuber- Sub-cooling plane is compared with the SB reported in referenced literature. The agreement of predicted NC stability boundaries with the results of independent studies demonstrates both the capability of the mentioned code in assessing NC loop stability and the quality of the performed calculations.(author)
Particle and thermo-hydraulic maldistribution of nanofluids in parallel microchannel systems
Maganti, Lakshmi Sirisha; sundararajan, T; Das, Sarit K
2016-01-01
Fluidic maldistribution in microscale multichannel devices requires deep understanding to achieve optimized flow and heat transfer characteristics. A thorough computational study has been performed to understand the concentration and thermohydraulic maldistribution of nanofluids in parallel microchannel systems using an Eulerian Lagrangian twin phase model. The study reveals that nanofluids cannot be treated as homogeneous single phase fluids in such complex flow domains and effective property models fail drastically to predict the performance parameters. To comprehend the distribution of the particulate phase, a novel concentration maldistribution factor has been proposed. It has been observed that distribution of particles need not essentially follow the flow pattern, leading to higher thermal performance than expected from homogeneous models. Particle maldistribution has been conclusively shown to be due to various migration and diffusive phenomena like Stokesian drag, Brownian motion, thermophoretic drift...
International Nuclear Information System (INIS)
A short review is given for models using in thermohydraulic code HYDRA-IBRAE/LM and the results of verification of calculational code on the problems of liquid metal coolant flow and heat transfer. It is shown that developed version of code HYDRA-IBRAE/LM simulates one-phase flow of lead, sodium and lead-bismuth coolants with high accuracy and the processes of sodium boiling with good one. The results of applied calculations of once-through steam generators are considered. It is pointed out that code HYDRA-IBRAE/LM represents correctly physics of the processes and phenomena taking place in the steam generator. The results of cross-verification calculations of lead steam generator by codes HYDRA-IBRAE/LM and TRIANA-4 show satisfactory agreement of results on temperatures of coolants and materials of channel walls in Field tube
Thermo-hydraulic test of the moderator cell of LH2 cold neutron source at BNC
International Nuclear Information System (INIS)
Complete text of publication follows. Thermo-hydraulic experiment was carried out in order to test the performance of the direct cooled liquid hydrogen moderator cell to be installed at the research reactor of the Budapest Neutron Center (BNC). Two electric heaters up to 300 W each imitated the nuclear heat release in the liquid hydrogen as well as in the construction material. The test moderator cell was also equipped with temperature gauges to measure the hydrogen temperature at different positions as well as the inlet and outlet temperature of cooling He gas. The hydrogen pressure in the connected buffer volume was also controlled. At 140 W expected total heat load the moderator cell was filled with liquid hydrogen within 4 hours. The heat load and hydrogen pressure characteristics of the moderator cell are also presented. (author)
Thermohydraulic responses of a water-cooled tokamak fusion DEMO to loss-of-coolant accidents
Nakamura, M.; Tobita, K.; Someya, Y.; Utoh, H.; Sakamoto, Y.; Gulden, W.
2015-11-01
Major in- and ex-vessel loss-of-coolant accidents (LOCAs) of a water-cooled tokamak fusion DEMO reactor have been analysed. Analyses have identified responses of the DEMO systems to these accidents and pressure loads to confinement barriers for radioactive materials. As for the in-VV LOCA, we analysed the multiple double-ended break of the first wall cooling pipes around the outboard toroidal circumference. As for the ex-VV LOCA, we analysed the double-ended break of the primary cooling pipe. The thermohydraulic analysis results suggest that the in- and ex-vessel LOCAs crucially threaten integrity of the primary and final confinement barriers, respectively. Mitigations of the loads to the confinement barriers are also discussed.
Fundamental study on thermo-hydraulic phenomena concerning passive safety of advanced marine reactor
International Nuclear Information System (INIS)
The objective of this study is to investigate the thermo-hydraulic behavior of a fluid region confined in a rectangular parallelepiped cavity equipped with a heater and a cooler. The motivation of this study is to clarify a thermal buffer effect for an innovative marine nuclear reactor to realize passive safety. In the present study, experiments were carried out with conditions of laminar convection. Temperature and flow behavior was visualized by the liquid-crystal suspension method, by which the temperature distribution in liquid can be observed as a colored map. Thermal plumes from the heater and the cooler, global natural circulation in the cavity and thermal stratification were observed as elements of the complicated phenomena. Using a code which solves the Navier-Stokes and energy equations, numerical simulations under steady and unsteady condition were carried out to predict the experimental results for two-dimensional, laminar situations, and a good agreement was obtained. (author)
RFSP simulations of Darlington FINCH refuelling transient
International Nuclear Information System (INIS)
Immediately after refuelling of a channel, the fresh bundles are free of fission products. Xenon-135, the most notable of the saturating fission products, builds up to an equilibrium level in about 30 h. The channel power of the refuelled channel would therefore initially peak and then drop to a steady-state level. The RFSP code can track saturating-fission-product transients and power transients. The Fully INstrumented CHannels (FINCHs) in Darlington NGS provides channel power data on the refuelling power transients. In this paper, such data has been used to identify the physical evidence of the fission-product transient effect on channel power, and to validate RFSP fission-product-driver calculation results. (author)
Full Scale Thermo-hydraulic Simulation of a Helium-Helium Printed Circuit Heat Exchanger
International Nuclear Information System (INIS)
In this paper, the thermo-hydraulic full scale simulation is performed to study the temperature distributions, thermal stress, pressure drop and outlet temperature in a Helium-Helium printed circuit heat exchanger (PCHE) in a VHTR simulate helium loop. The entire PCHE is composed of 40 stacks of rectangular shaped micro-channels for helium gas [type A] (inlet temperature, 400 .deg. C) and 40 stacks of semi-ellipse shaped micro-channels for helium [type B] (inlet temperature, 300 .deg. C). The experimental result is compared to that of computer simulation, COMSOL multi-physics software. The Helium-Helium PCHE is considered a prototype of the newly developed PCHE by Korea Atomic Energy Research Institute (KAERI). The full scale thermo-hydraulic simulation was successfully performed to obtain temperature distribution, pressure drop and thermal stress in 40 sets of flow channel stacks in a helium-helium printed circuit heat exchanger in a VHTR simulate helium loop. We obtained a quite similar temperature distribution with the 3D measured infrared temperature distribution. To our knowledge, this is the first full scale numerical study on the PCHE, which considers all microchannels, that the convection effect on the outside surfaces of the PCHE is applied. The very high-temperature reactor (VHTR) or high-temperature gas-cooled reactor(HTGR) is a fourth-generation nuclear power reactor that uses the ceramic coated fuel, TRISO, in which the fission gas does not leak even at temperatures higher than 1600 .deg. C. The VHTR necessarily requires an intermediate loop composed of a hot gas duct (HGD), an intermediate heat exchanger (IHX) and a process heat exchanger (PHE). The IHX is one of the important components of VHTR system because the IHX transfers the 950 .deg. C of high temperature massive heat to a hydrogen production plant or power conversion unit at high system pressure
A subchannel and CFD analysis of void distribution for the BWR fuel bundle test benchmark
Energy Technology Data Exchange (ETDEWEB)
In, Wang-Kee; Hwang, Dae-Hyun [Korea Atomic Energy Research Institute (KAERI), 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Jeong, Jae Jun, E-mail: jjjeong@pusan.ac.kr [School of Mechanical Engineering, Pusan National University, Jangjeon-dong, Geumjeong-gu, Busan 609-735 (Korea, Republic of)
2013-05-15
Highlights: ► We analyzed subchannel void distributions using subchannel, system and CFD codes. ► The mean error and standard deviation at steady states were compared. ► The deviation of the CFD simulation was greater than those of the others. ► The large deviation of the CFD prediction is due to interface model uncertainties. -- Abstract: The subchannel grade and microscopic void distributions in the NUPEC (Nuclear Power Engineering Corporation) BFBT (BWR Full-Size Fine-Mesh Bundle Tests) facility have been evaluated with a subchannel analysis code MATRA, a system code MARS and a CFD code CFX-10. Sixteen test series from five different test bundles were selected for the analysis of the steady-state subchannel void distributions. Four test cases for a high burn-up 8 × 8 fuel bundle with a single water rod were simulated using CFX-10 for the microscopic void distribution benchmark. Two transient cases, a turbine trip without a bypass as a typical power transient and a re-circulation pump trip as a flow transient, were also chosen for this analysis. It was found that the steady-state void distributions calculated by both the MATRA and MARS codes coincided well with the measured data in the range of thermodynamic qualities from 5 to 25%. The results of the transient calculations were also similar to each other and very reasonable. The CFD simulation reproduced the overall radial void distribution trend which produces less vapor in the central part of the bundle and more vapor in the periphery. However, the predicted variation of the void distribution inside the subchannels is small, while the measured one is large showing a very high concentration in the center of the subchannels. The variations of the void distribution between the center of the subchannels and the subchannel gap are estimated to be about 5–10% for the CFD prediction and more than 20% for the experiment.
Institute of Scientific and Technical Information of China (English)
YU Wei-dong; YAN Hao-jing; Ron Postle; Yang Shouren
2002-01-01
Due to the effects of samples and testing conditions on fibre-bundle tensile behaviour, it is necessary to investigate the relationships between experimental factors and tensile properties for the fibre-bumdle tensile tester (TENSOR). The effects of bundle sample preparation, fibre bundle mass and fibre alignment have been tested. The experimental results indicated that (1) the low damage in combing and no free-end fibres in the cut bundle are most important for the sample preparation; (2) the reasonable bundle mass is 400- 700tex, but the tensile properties measured should bemodified with the bundle mass because a small amount of bundle mass causes the scatter results, while the larger is the bundle mass, the more difficult to comb fibres parallel and to clamp fibre evenly; and (3) the fibre irregular arrangement forms a slack bundle resulting in interaction between fibres, which will affect the reproducibility and accuracy of the tensile testing.
Self-mapping degrees of torus bundles and torus semi-bundles
Sun, Hongbin; Wang, Shicheng; Wu, Jianchun
2010-01-01
Each closed oriented 3-manifold $M$ is naturally associated with a set of integers $D(M)$, the degrees of all self-maps on $M$. $D(M)$ is determined for each torus bundle and torus semi-bundle $M$. The structure of torus semi-bundle is studied in detail. The paper is a part of a project to determine $D(M)$ for all 3-manifolds in Thurston's picture.
In-pool damaged fuel bundle recovery
International Nuclear Information System (INIS)
While preparing to rerack the Oyster Creek Nuclear Generating Station, GPU Nuclear had need to move a damaged fuel bundle. This bundle had no upper tie plate and could not be moved in the normal manner. GPU Nuclear formed a small, dedicated project team to disassemble, package, and move this damaged bundle. The team was composed of key personnel from GPU Nuclear Fuels Projects, OCNGS Operations and Proto-Power/Bisco, a specialty contractor who has fuel bundle reconstitution and rod consolidation experience, remote tooling, underwater video systems and experienced technicians. Proven tooling, clear procedures and a simple approach were important, but the key element was the spirit of teamwork and leadership exhibited by the people involved. In spite of several emergent problems which a task of this nature presents, this small, close knit utility/vendor team completed the work on schedule and within the exposure and cost budgets
In-pool damaged fuel bundle recovery
International Nuclear Information System (INIS)
While preparing to rerack the Oyster Creek Nuclear Generating Station, GPU Nuclear had need to move a damaged fuel bundle. This bundle had no upper tie plate and could not be moved in the normal manner. GPU Nuclear formed a small, dedicated project team to disassemble, package and move this damaged bundle. The team was composed of key personnel from GPU Nuclear Fuels Projects, OCNGS Operations and Proto-Power / Bisco, a specialty contractor who has fuel bundle reconstitution and rod consolidation experience, remote tooling, underwater video systems and experienced technicians. Proven tooling, clear procedures and a simple approach were important, but the key element was the spirit of teamwork and leadership exhibited by the people involved
Nuclear fuel bundle disassembly and assembly tool
International Nuclear Information System (INIS)
A nuclear power reactor fuel bundle is described which has a plurality of tubular fuel rods disposed in parallel array between two transverse tie plates. It is secured against disassembly by one or more locking forks which engage slots in tie rods which position the transverse plates. Springs mounted on the fuel and tie rods are compressed when the bundle is assembled thereby maintaining a continual pressure against the locking forks. Force applied in opposition to the springs permits withdrawal of the locking forks so that one tie plate may be removed, giving access to the fuel rods. An assembly and disassembly tool facilitates removal of the locking forks when the bundle is to be disassembled and the placing of the forks during assembly of the bundle. (U.S.)
Quantum Bundle Description of Quantum Projective Spaces
Ó Buachalla, Réamonn
2012-12-01
We realise Heckenberger and Kolb's canonical calculus on quantum projective ( N - 1)-space C q [ C p N-1] as the restriction of a distinguished quotient of the standard bicovariant calculus for the quantum special unitary group C q [ SU N ]. We introduce a calculus on the quantum sphere C q [ S 2 N-1] in the same way. With respect to these choices of calculi, we present C q [ C p N-1] as the base space of two different quantum principal bundles, one with total space C q [ SU N ], and the other with total space C q [ S 2 N-1]. We go on to give C q [ C p N-1] the structure of a quantum framed manifold. More specifically, we describe the module of one-forms of Heckenberger and Kolb's calculus as an associated vector bundle to the principal bundle with total space C q [ SU N ]. Finally, we construct strong connections for both bundles.
Twin tori for a new bundle divertor
International Nuclear Information System (INIS)
A new bundle divertor system using the straight stagnation axis in toroidal field together with the uniform field along the axis is discussed in detail. We call this type of divertor as the ''muffler divertor'' because of its shape. (author)
Noncommutative principal bundles through twist deformation
Aschieri, Paolo; Pagani, Chiara; Schenkel, Alexander
2016-01-01
We construct noncommutative principal bundles deforming principal bundles with a Drinfeld twist (2-cocycle). If the twist is associated with the structure group then we have a deformation of the fibers. If the twist is associated with the automorphism group of the principal bundle, then we obtain noncommutative deformations of the base space as well. Combining the two twist deformations we obtain noncommutative principal bundles with both noncommutative fibers and base space. More in general, the natural isomorphisms proving the equivalence of a closed monoidal category of modules and its twist related one are used to obtain new Hopf-Galois extensions as twists of Hopf-Galois extensions. A sheaf approach is also considered, and examples presented.
Crosstalk analysis of carbon nanotube bundle interconnects
Zhang, Kailiang; Tian, Bo; Zhu, Xiaosong; WANG, FANG; Wei, Jun
2012-01-01
Carbon nanotube (CNT) has been considered as an ideal interconnect material for replacing copper for future nanoscale IC technology due to its outstanding current carrying capability, thermal conductivity, and mechanical robustness. In this paper, crosstalk problems for single-walled carbon nanotube (SWCNT) bundle interconnects are investigated; the interconnect parameters for SWCNT bundle are calculated first, and then the equivalent circuit has been developed to perform the crosstalk analys...
A Geometric Approach to Noncommutative Principal Bundles
Wagner, Stefan
2011-01-01
From a geometrical point of view it is, so far, not sufficiently well understood what should be a "noncommutative principal bundle". Still, there is a well-developed abstract algebraic approach using the theory of Hopf algebras. An important handicap of this approach is the ignorance of topological and geometrical aspects. The aim of this thesis is to develop a geometrically oriented approach to the noncommutative geometry of principal bundles based on dynamical systems and the representation theory of the corresponding transformation group.
Parahoric bundles on a compact Riemann surface
Balaji, V
2010-01-01
Let $X$ be a compact Riemann surface of genus $g \\geq 2$. The aim of this paper is to study homomorphisms of certain discrete subgroups of $PSL(2, {\\mathbb R})$ into maximal compact subgroups of semisimple simply connected algebraic groups and relate them to torsors under a Bruhat-Tits group scheme. We also construct the moduli spaces of semistable parahoric bundles. These results generalize the theorem of Mehta and Seshadri on parabolic vector bundles.
Evaluation on BDI of large diameter pin bundles by out-of-pile bundle compression test
International Nuclear Information System (INIS)
Bundle-duct interaction (BDI) in core fuel subassemblies in fast reactors (FRs) is a limiting factor for fuel burnup. Since the large diameter fuel pin is generally believed to be a measure to improve FR fuel performance, the out-of-pile bundle compression test with large diameter pins (φ8.5mm and (φ 10.4mm) was performed to evaluate BDI in these bundles. In the compression test, bundle cross-sectional images (CT images) under BDI condition were obtained by using the X-ray computer tomography. In the main study, the CT images were numerically analyzed to evaluate deformation of the large diameter pin bundle due to BDI. The CT image analysis results revealed that pin-to-duct contact did not occur when the flat-to-flat bundle compression level reached one wire diameter (BDI level of 1dw), which indicates that BDI in large diameter pin bundles was mitigated similarly to the currently used small diameter pin bundles. In addition, the mitigation mechanism for BDI, which delays initiation of pin-to-duct contact, was investigated by using the computer code analysis. The code analysis results showed that cladding oval-distortion acted as a major mitigation mechanism for BDI as in the case of small pin diameter bundles. (author)
Analytical and CFD investigation of ex-core cooling of the nuclear fuel rod bundle in a water pool
International Nuclear Information System (INIS)
The efficiency of ex-core cooling of nuclear fuel assemblies under decay heat generation is influenced by many conditions, among them being coolant flow rate, position of fuel assemblies in a water pool, and position of coolant inlets and outlets. A combination of unacceptable thermal-hydraulic conditions occurred at the Nuclear Power Plant PAKS in Hungary in April 2003, during the process of nuclear fuel assembly chemical cleaning in a specially designed tank. The cooling of the nuclear fuel rod bundles in the tank was not efficient under low coolant flow rates through the cleaning tank, and after several hours the boiling of cooling water occurred with subsequent dry-out of nuclear fuel rod bundles. The thermal-hydraulic conditions in the cleaning tank that led to the unexpected event are analysed both analytically and with a CFD approach for idealized conditions of one nuclear fuel rod bundle with the bottom by-pass opening. The analytical analysis is based on a pressure balance of low Reynolds number upward water coolant flow through the bundle, downward water flow in the pool around the bundle, flow across the by-pass opening and outlet flow from the cleaning vessel. The transient CFD simulations are performed in order to demonstrate multidimensional effects of the event. The water density dependence on the temperature is taken into account in both analytical and CFD investigation, as the dominant effect that influences the buoyancy forces between the water flow streams inside and outside the vertically positioned bundle in the water pool. The influence of the bundle bottom by-pass area on the water pool thermal-hydraulic conditions and on the efficiency of the nuclear fuel rods cooling is analysed. Both analytical and CFD results show that the continuous cooling of the fuel rods can not be achieved for higher values of the bundle bottom by-pass areas. The averaged coolant temperature in the water pool outside the bundle becomes higher than the average
International Nuclear Information System (INIS)
Thermo-hydraulic instabilities of a boiling natural circulation loop with a chimney under high pressure were investigated using linear stability analysis. Drift-flux model was used for two-phase flow model. The instability regions as well as the thermo-hydraulic characteristics in the chimney such as wavy feature were examined, which were compared with the characteristics in low pressure. Instability could occur when exit quality was relatively low, which was the same manner as the characteristics in low pressure. In high-pressure, void was generated near channel exit, and void wave propagated in the chimney. In low pressure, steam was generated only near the chimney exit due to gravity induced flashing, and single-phase enthalpy wave, that is, temperature wave propagated in single-phase flow region. Though flow could be very stable in the high pressure and high power condition, the decay ratio of higher mode could be larger than that of lower mode. (author)
K-Theories for Certain Infinite Rank Bundles
Larrain-Hubach, Andres
2011-01-01
Several authors have recently constructed characteristic classes for classes of infinite rank vector bundles appearing in topology and physics. These include the tangent bundle to the space of maps between closed manifolds, the infinite rank bundles in the families index theorem, and bundles with pseudodifferential operators as structure group. In this paper, we construct the corresponding K-theories for these types of bundles. We develop the formalism of these theories and use their Chern ch...
Effect of left bundle branch block on TIMI frame count
Hatice Tolunay; Ahmet Kasapkara; İsa Öner Yüksel; Nurcan Başar; Ayşe Saatcı Yaşar; Mehmet Bilge
2010-01-01
Aim: Left bundle branch block is an independent risk factorfor cardiac mortality. In this study we aimed to evaluatecoronary blood flow with TIMI frame count in patients with left bundle branch block and angiographically proven normal coronary arteries.Materials and methods: We retrospectively studied 17 patients with left bundle branch block and as a control group 16 patients without left bundle branch block. All patientshad angiographically proven normal coronary arteries.Left bundle branch...
Product-bundling and Incentives for Merger and Strategic Alliance
Sue Mialon
2009-01-01
This paper analyzes firms' choice between a merger and a strategic alliance in bundling their product with other complementary products. We consider a framework in which firms can improve profits only from product-bundling. While mixed bundling is not profitable, pure bundling is because pure bundling reduces consumers' choices, and thus softens competition among firms. Firms benefit the most from this reduced competition if they form an alliance. Firms do not gain as much from a merger becau...
Seiler, J. M.; Rameau, B.
Bundle sodium boiling in nominal geometry for different accident conditions is reviewed. Voiding of a subassembly is controlled by not only hydrodynamic effects but mainly by thermal effects. There is a strong influence of the thermal inertia of the bundle material compared to the sodium thermal inertia. Flow instability, during a slow transient, can be analyzed with numerical tools and estimated using simplified approximations. Stable boiling operational conditions under bundle mixed convection (natural convection in the reactor) can be predicted. Voiding during a fast transient can be approximated from single channel calculations. The phenomenology of boiling behavior for a subassembly with inlet completely blocked, submitted to decay heat and lateral cooling; two-phase sodium flow pressure drop in a tube of large hydraulic diameter under adiabatic conditions; critical flow phenomena and voiding rate under high power, slow transient conditions; and onset of dry out under local boiling remains problematical.
Mechanism of Actin Filament Bundling by Fascin
Energy Technology Data Exchange (ETDEWEB)
Jansen, Silvia; Collins, Agnieszka; Yang, Changsong; Rebowski, Grzegorz; Svitkina, Tatyana; Dominguez, Roberto (UPENN); (UPENN-MED)
2013-03-07
Fascin is the main actin filament bundling protein in filopodia. Because of the important role filopodia play in cell migration, fascin is emerging as a major target for cancer drug discovery. However, an understanding of the mechanism of bundle formation by fascin is critically lacking. Fascin consists of four {beta}-trefoil domains. Here, we show that fascin contains two major actin-binding sites, coinciding with regions of high sequence conservation in {beta}-trefoil domains 1 and 3. The site in {beta}-trefoil-1 is located near the binding site of the fascin inhibitor macroketone and comprises residue Ser-39, whose phosphorylation by protein kinase C down-regulates actin bundling and formation of filopodia. The site in {beta}-trefoil-3 is related by pseudo-2-fold symmetry to that in {beta}-trefoil-1. The two sites are {approx}5 nm apart, resulting in a distance between actin filaments in the bundle of {approx}8.1 nm. Residue mutations in both sites disrupt bundle formation in vitro as assessed by co-sedimentation with actin and electron microscopy and severely impair formation of filopodia in cells as determined by rescue experiments in fascin-depleted cells. Mutations of other areas of the fascin surface also affect actin bundling and formation of filopodia albeit to a lesser extent, suggesting that, in addition to the two major actin-binding sites, fascin makes secondary contacts with other filaments in the bundle. In a high resolution crystal structure of fascin, molecules of glycerol and polyethylene glycol are bound in pockets located within the two major actin-binding sites. These molecules could guide the rational design of new anticancer fascin inhibitors.
International Nuclear Information System (INIS)
By inputting the experimental data, information and others on thermo-hydraulic characteristics of integrated ship propulsion reactor accumulated hitherto by the Ship Research Institute and some recent cooperation results into the nuclear ship engineering simulation system, it was conducted not only to contribute an improvement study on next ship reactor by executing general analysis and evaluation on motion characteristics under ship body motion conditions, safety at accidents, and others of the integrated ship reactor but also to investigate and prepare some measures to apply fundamental experiment results based on obtained here information to safety countermeasure of the nuclear ships. In 1997 fiscal year, on safety of the integrated ship propulsion reactor loading nuclear ship, by adding experimental data on unstable flow analysis and information on all around of the analysis to general data base fundamental program, development to intellectual data base program was intended; on effect of pulsation flow on thermo-hydraulic characteristics of ship propulsion reactor; after pulsation flow visualization experiment, experimental equipment was reconstructed into heat transfer type to conduct numerical analysis of pulsation flow by confirming validity of numerical analysis code under comparison with the visualization experiment results; and on thermo-hydraulic behavior in storage container at accident of active safety type ship propulsion reactor; a flashing vibration test using new apparatus finished on its higher pressurization at last fiscal year to examine effects of each parameter such as radius and length of exhausting nozzle and pool water temperature. (G.K.)
International Nuclear Information System (INIS)
The computer codes used in construction of nuclear reactors projects, specifically with regard to the thermo-hydraulics concepts part of your core, whose main goal reproduce actual operating conditions in order to predict quantitatively the limiting conditions of operation so that the safety limit is not exceeded. Computational methods for studies of fluid flow are developed around the world, including Brazil. With the evolution of computers application of numerical methods greatly reduced response time results, and tends to further decrease the extent that the computers and processors develop, making it feasible to use programming accident simulations of heat transfer reactors. The software developed in this paper presents a method for analyzing the thermo-hydraulic behavior of the Brazilian Multipurpose Reactor (RMB) after its shutdown. The software solves the conservations equations applied to the core and also the lower and upper regions of the RMB. The thermo-hydraulics characteristics studied are: the temperatures of the core, cladding and refrigerant, the mass flow and the heat transfer. The numerical resolution was performed using the Matlab language and the outputs are presented in graphs and tables forms. (author)
International Nuclear Information System (INIS)
KWU keeps a file on transients for the PWR plants. This file stores test data and significant graphs and curves of readings which are sent to the manufacturer for analysis. The data are used to produce analyses of the dynamics and incidents/deviations in operation. In addition, comprehensive simulation models are fed with readings from practical operation. The paper outlines the development aims, the historical evolution, the hardware and software concepts, the acquisition/detection method, applications and examples. (DG)
Energy Technology Data Exchange (ETDEWEB)
Nam, Seung Hyun; Choi, Jae Young; Venneria, Paolo F.; Jeong, Yong Hoon; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of)
2015-05-15
NTR engines have continued as a main stream based on the mature technology. The typical core design of the NERVA derived engines uses hexagonal shaped fuel elements with circular cooling channels and structural tie-tube elements for supporting the fuel elements, housing moderator and regeneratively cooling the moderator. The state-of-the-art NTR designs mostly use a fast or epithermal neutron spectrum core utilizing a HEU fuel to make a high power reactor with small and simple core geometry. Nuclear propulsion is the most promising and viable option to achieve challenging deep space missions. Particularly, the attractions of a NTR include excellent thrust and propellant efficiency, bimodal capability, proven technology, and safe and reliable performance. The KANUTER-HEU and -LEU are the innovative and futuristic NTR engines to reduce the reactor size and to implement a LEU fuel in the reactor by using thermal neutron spectrum. The KANUTERs have some features in the reactor design such as the integrated fuel element and the regeneratively cooling channels to increase room for moderator and heat transfer in the core, and ensuing rocket performance. To study feasible design points in terms of thermo-hydraulics and to estimate rocket performance of the KANUTERs, the NSES is under development. The model of the NSES currently focuses on thermo-hydraulic analysis of the peculiar and complex EHTGR design during the propulsion mode in steady-state. The results indicate comparable performance for future applications, even though it uses the heavier LEU fuel. In future, the NSES will be modified to obtain temperature distribution of the entire reactor components and then more extensive design analysis of neutronics, thermohydraulics and their coupling will be conducted to validate design feasibility and to optimize the reactor design enhancing the rocket performance.
International Nuclear Information System (INIS)
a main stream based on the mature technology. The typical core design of the NERVA derived engines uses hexagonal shaped fuel elements with circular cooling channels and structural tie-tube elements for supporting the fuel elements, housing moderator and regeneratively cooling the moderator. The state-of-the-art NTR designs mostly use a fast or epithermal neutron spectrum core utilizing a HEU fuel to make a high power reactor with small and simple core geometry. Nuclear propulsion is the most promising and viable option to achieve challenging deep space missions. Particularly, the attractions of a NTR include excellent thrust and propellant efficiency, bimodal capability, proven technology, and safe and reliable performance. The KANUTER-HEU and -LEU are the innovative and futuristic NTR engines to reduce the reactor size and to implement a LEU fuel in the reactor by using thermal neutron spectrum. The KANUTERs have some features in the reactor design such as the integrated fuel element and the regeneratively cooling channels to increase room for moderator and heat transfer in the core, and ensuing rocket performance. To study feasible design points in terms of thermo-hydraulics and to estimate rocket performance of the KANUTERs, the NSES is under development. The model of the NSES currently focuses on thermo-hydraulic analysis of the peculiar and complex EHTGR design during the propulsion mode in steady-state. The results indicate comparable performance for future applications, even though it uses the heavier LEU fuel. In future, the NSES will be modified to obtain temperature distribution of the entire reactor components and then more extensive design analysis of neutronics, thermohydraulics and their coupling will be conducted to validate design feasibility and to optimize the reactor design enhancing the rocket performance
Temperature Distributions in LMR Fuel Pin Bundles as Modeled by COBRA-IV-I
Wright, Steven A.; Stout, Sherry
2005-02-01
Most pin type reactor designs for space power or terrestrial applications group the fuel pins into a number of relatively large fuel pin bundles or subassemblies. Fuel bundles for terrestrial liquid metal fast breeders reactors typically use 217 - 271 pins per sub-assembly, while some SP100 designs use up to 331 pins in a central subassembly that was surrounded by partial assemblies. Because thermal creep is exponentially related to temperature, small changes in fuel pin cladding temperature can make large differences in the lifetime in a high temperature liquid metal reactor (LMR). This paper uses the COBRA-IV-I computer code to determine the temperature distribution within LMR fuel bundles. COBRA-IV-I uses the sub-channel analysis approach to determine the enthalpy (or temperature) and flow distribution in rod bundles for both steady-state and transient conditions. The COBRA code runs in only a few seconds and has been benchmarked and tested extensively over a wide range of flow conditions. In this report the flow and temperature distributions for two types of lithium cooled space reactor core designs were calculated. One design uses a very tight fuel pin packing that has a pitch to diameter ratio of 1.05 (small wire wrap with a diameter of 392 μm) as proposed in SP100. The other design uses a larger pitch to diameter ratio of 1.09 with a larger more conventional sized wire wrap diameter of 1 mm. The results of the COBRA pin bundle calculations show that the larger pitch-to-diameter fuel bundle designs are more tolerant to local flow blockages, and in addition they are less sensitive to mal-flow distributions that occur near the edges of the subassembly.
Safety analysis report of the irradiation test of Type-B bundle
Energy Technology Data Exchange (ETDEWEB)
Lee, Choong Sung; Lim, I. C.; Lee, B. C.; Ryu, J. S.; Kim, H. R
2000-06-01
The HANARO fuel, U{sub 3}Si-A1, has been developed by AECL and tested in NRU reactor. In the course of the fuel qualification tests, only one case was performed under the higher power condition than maximum linear power which was expected in the design stage. The Korea regulatory body, KINS imposed that HANARO shall be operated at the power level less than 24MW which is 80% of the design full power until HANARO shows the repetitive performance of the fuel at the power condition abov e 112.8KW/m. To resolve this imposition, KAERI designed two types of special test bundles: two non-instrumented(Type-A) and one instrumented(Type-B) test bundles. Two Type-A bundles were irradiated in HANARO: one of them has finished PIE and the other is under PIE. Type-B bundle was loaded in the core during 1.32 day at 1996, but outstanding FIV(flow induced vibration) was observed at the pool top because of long guide tube attached to the top of the bundle. The successful installation of the chimney fastener to fix the guide tube resulted in conducting the irradiation test of Type-B bundle again. The test will start at mid- July, 2000. In order to safely do the Type-B irradiation test, the safety analysis for the nuclear, mechanical and thermal-hydraulic aspects was performed. The reactivity worth and the maximum 1 near power predicted by VENTURE are 6.3mk/k and 121.6kW/m, respectively. Thermal margins for normal and transient conditions using MATRA-h, are assessed to satisfy the safety criteria.
Cooperative retraction of bundled type IV pili enables nanonewton force generation.
Directory of Open Access Journals (Sweden)
Nicolas Biais
2008-04-01
Full Text Available The causative agent of gonorrhea, Neisseria gonorrhoeae, bears retractable filamentous appendages called type IV pili (Tfp. Tfp are used by many pathogenic and nonpathogenic bacteria to carry out a number of vital functions, including DNA uptake, twitching motility (crawling over surfaces, and attachment to host cells. In N. gonorrhoeae, Tfp binding to epithelial cells and the mechanical forces associated with this binding stimulate signaling cascades and gene expression that enhance infection. Retraction of a single Tfp filament generates forces of 50-100 piconewtons, but nothing is known, thus far, on the retraction force ability of multiple Tfp filaments, even though each bacterium expresses multiple Tfp and multiple bacteria interact during infection. We designed a micropillar assay system to measure Tfp retraction forces. This system consists of an array of force sensors made of elastic pillars that allow quantification of retraction forces from adherent N. gonorrhoeae bacteria. Electron microscopy and fluorescence microscopy were used in combination with this novel assay to assess the structures of Tfp. We show that Tfp can form bundles, which contain up to 8-10 Tfp filaments, that act as coordinated retractable units with forces up to 10 times greater than single filament retraction forces. Furthermore, single filament retraction forces are transient, whereas bundled filaments produce retraction forces that can be sustained. Alterations of noncovalent protein-protein interactions between Tfp can inhibit both bundle formation and high-amplitude retraction forces. Retraction forces build over time through the recruitment and bundling of multiple Tfp that pull cooperatively to generate forces in the nanonewton range. We propose that Tfp retraction can be synchronized through bundling, that Tfp bundle retraction can generate forces in the nanonewton range in vivo, and that such high forces could affect infection.
International Nuclear Information System (INIS)
PNC has been developing a computer code 'BAMBOO' to analyze the wire spaced FBR fuel pin bundle deformation under the BDI (Bundle Duct Interaction) condition by means of the three dimensional F.E.M. This code analyzes fuel pins' bowing and oval deformations which are dominant deformation behaviors of the fuel pin bundle under the BDI condition. In this study the 'BAMBOO' code is validated on the out-of-pile compression test of the FBR bundle (compression test) by comparing the results of the code analysis with the compression test results, and the highly irradiated (≥2.1x1027 n/m2, E > 0.1 MeV) bundle deformation behaviors are investigated from the viewpoint of the similarity to those in the compression test based on the analytical results of the code. (1) The calculated pin-to-duct minimum clearances as a function of the BDI levels in the compression test analysis agree with the experimental values evaluated from the CT image analysis of the bundle cross-section in the compression test within ±0.2 mm. And the calculated values of the fuel pins' oval deformations agree with the experimental values based on the pin diameter measurements done after the compression test within ±0.05 mm. (2) By comparing the irradiation induced bundle deformation with the bundle deformation in the compression test based on the code analysis, it is confirmed that the changes of the pin-to-duct minimum clearances with the BDI levels show equivalent trends between the both bundle deformations. And in this code analysis of the irradiation induced bundle deformation, contact loads between the fuel pins and the pacer wires are extremely small (below 10 kgf) even at about 3 dw of the BDI level compared to those in the compression test analysis. (J.P.N.)
NIF laser bundle review. Final report
International Nuclear Information System (INIS)
We performed additional bundle review effort subsequent to the completion of the preliminary report and are revising our original recommendations. We now recommend that the NIF baseline laser bundle size be changed to the 4x2 bundle configuration. There are several 4x2 bundle configurations that could be constructed at a cost similar to that of the baseline 4x12 (from $11M more to about $11M less than the baseline; unescalated, no contingency) and provide significant system improvements. We recommend that the building cost estimates (particularly for the in-line building options) be verified by an architect/engineer (A/E) firm knowledgeable about building design. If our cost estimates of the in-line building are accurate and therefore result in a change from the baseline U-shaped building layout, the acceptability of the in-line configuration must be reviewed from an operations viewpoint. We recommend that installation, operation, and maintenance of all laser components be reviewed to better determine the necessity of aisles, which add to the building cost significantly. The need for beam expansion must also be determined since it affects the type of bundle packing that can be used and increases the minimum laser bay width. The U-turn laser architecture (if proven viable) offers a reduction in building costs since this laser design is shorter than the baseline switched design and requires a shorter laser bay
International Nuclear Information System (INIS)
The objectives of this paper are to make clear the thermo-hydraulic behaviors of boiling two-phase flow under transient conditions and its effects on the burnout phenomena. First, the critical heat flux in slug flow pattern, which is regarded as one kind of transient states microscopically, has been obtained. Also the critical heat flux for parallel channels has been studied considering the instability of flow. Next, the critical heat fluxes have been investigated for increasing heat input, for rapid depressurization and for loss of flow. Also the critical heat flux for increasing heat input has been made clear at elevated pressure in which the heat input has been loaded rampwise or stepwise. Through those studies the burnout behaviors and critical heat fluxes under transient conditions have been clarified fairly well. (author)
Transient response of a high-capacity heat pipe for Space Station Freedom
Ambrose, J. H.; Holmes, H. R.
1991-01-01
High-capacity heat pipe radiator panels have been proposed as the primary means of heat rejection for Space Station Freedom. In this system, the heat pipe would interface with the thermal bus condensers. Changes in system heat load can produce large temperature and heat load variations in individual heat pipes. Heat pipes could be required to start from an initially cold state, with heat loads temporarily exceeding their low-temperature transport capacity. The present research was motivated by the need for accurate prediction of such transient operating conditions. In this work, the cold startup of a 6.7-meter long high-capacity heat pipe is investigated experimentally and analytically. A transient thermohydraulic model of the heat pipe was developed which allows simulation of partially-primed operation. The results of cold startup tests using both constant temperature and constant heat flux evaporator boundary conditions are shown to be in good agreement with predicted transient response.
Transients analysis by reactivity insertion in research reactors
International Nuclear Information System (INIS)
PARET code was used to simulate accidental situations arising from positive reactivity insertions, in order to analyze the behavior of RP-10 reactor. The simulations considered three different cases: First is for the reactor operating at 10 Mw nominal power with 3 pumps in use, the second, at 6.6 Mw with only one pump working. In all cases the reactor trip was assumed when a 12 Mw power level is reached. An additional simulation for the reactor operating at 50w before the reactivity insertion, showed to be the worst accidental situation of all cases because of the higher temperature and power rise. Hot channel thermohydraulic and kinetic parameters have been evaluated at each axial mesh point and transient time step. None of the cases showed melting of fuel plates
Turbulent flow through two asymmetric rod bundles
International Nuclear Information System (INIS)
Measurements of the mean velocity, of the wall shear stresses, and of the turbulence have been performed in four wall subchannels of rod bundles of four parallel rods enclosed in a rectangular channel. The pitch-to-diameter ratio was P/D=1.148 and the wall-to-diameter ratios ranged from 1.045 to 1.252. The full Reynolds stress tensor has been determined by hot-wire technique. The results of the turbulences intensities show that the flow through rod bundles differs widely from flow through circular tubes. More sophisticated analytical tools than presently available are required to predict turbulent flow through rod bundles with sufficient accuracy
Assembly mechanism for nuclear fuel bundles
International Nuclear Information System (INIS)
In a nuclear power reactor fuel bundle having tie rods fastened to a lower tie plate and passing through openings in the upper tie plate, the assembled bundle is secured by locking lugs fixed to rotatable locking sleeves which engage the upper tie plate. Pressure exerted by helical springs mounted around each of the tie rods urge retaining lugs fixed to a retaining sleeve associated with respective tie rods into a position with respect to the locking sleeve to prevent accidental disengagement of the upper plate from the locking lugs. The bundle may be disassembled by depressing the retaining sleeves and rotating the locking lugs to the disengaged position, and then removing the upper tie plate
Porous Silicon and Denim Fiber Bundle Characterization
Deuro, Randi Ellen
My thesis research aims to characterize and exploit materials in an efficient, rapid, non-destructive manner. Part I of this document summarizes my research on porous silicon (pSi) design, fabrication, and surface modification for use as a novel chemical sensor. The optimization of fabrication process parameters (etching time, etching solution, electrode shape, and the fixing process) on pSi photoluminescence (PL) is presented. I have also investigated the effects of analyte vapors (acetonitrile, toluene, methanol, acetone) on the pSi PL and surface chemistry using luminescence and Fourier-transform infrared (FT-IR) spectroscopy and microscopy methods. The mechanism and benefits of one method of pSi surface modification and protection (ultraviolet (UV) hydrosilylation) will also be presented. Finally, high thorough-put methods of pSi sensor production are described. In Part II of this document, I introduce a novel technique for analyzing and discriminating among denim fiber bundles. An investigation into the benefits of luminescence-based multispectral imaging (LMSI) for denim fiber bundle identification has been conducted. I explore the power of nitromethane (CH 3NO2) based quenching in fiber bundle classification and identify the quenching mechanism. The luminescence spectra (450 - 850 nm) and images from the denim fiber bundles were obtained while exciting at 325 nm or 405 nm. Here, LMSI data were recorded in < 10 s and subsequently assessed by principal component analysis (PCA) and rendered red, green, blue (RGB) component histograms. The results show that LMSI data can be used to rapidly and uniquely classify all the fiber bundle types studied in this research. These non-destructive techniques eliminate extensive sample preparation and allow for rapid multispectral image collection, analysis, and assessment. The quenching data also revealed that the dye molecules within the individual fiber bundles exhibited dramatically different accessibilities to CH 3NO2.
Chung, Peter J.; Song, Chaeyeon; Deek, Joanna; Miller, Herbert P.; Li, Youli; Choi, Myung Chul; Wilson, Leslie; Feinstein, Stuart C.; Safinya, Cyrus R.
2016-01-01
Tau, an intrinsically disordered protein confined to neuronal axons, binds to and regulates microtubule dynamics. Although there have been observations of string-like microtubule fascicles in the axon initial segment (AIS) and hexagonal bundles in neurite-like processes in non-neuronal cells overexpressing Tau, cell-free reconstitutions have not replicated either geometry. Here we map out the energy landscape of Tau-mediated, GTP-dependent ‘active' microtubule bundles at 37 °C, as revealed by synchrotron SAXS and TEM. Widely spaced bundles (wall-to-wall distance Dw–w≈25–41 nm) with hexagonal and string-like symmetry are observed, the latter mimicking bundles found in the AIS. A second energy minimum (Dw–w≈16–23 nm) is revealed under osmotic pressure. The wide spacing results from a balance between repulsive forces, due to Tau's projection domain (PD), and a stabilizing sum of transient sub-kBT cationic/anionic charge–charge attractions mediated by weakly penetrating opposing PDs. This landscape would be significantly affected by charge-altering modifications of Tau associated with neurodegeneration. PMID:27452526
International Nuclear Information System (INIS)
The APACHE code (Automatic Analysis of Failures of Hydraulic and Thermohydraulic Circuits more particularly of Water) situates in an important program of computer codes development in the field of studies on reliability and safety of systems in nuclear power plants. APACHE is an automatic generation code of failure pattern and of their effects. After a presentation of the theoretical basis, the methodological principles of the theory of networks are developed. Then, the model of the code is developed: model of individual behavior of each classical model component of normal behavior and model of failure pattern with specifications. The global model of hydraulic systems and the resolution systems are then developed. More particularly, some aspects of the theory of graphs, and the algorithms developed for the automatic construction of the equation systems and especially the algorithm of the research of meshes are presented. The computer aspect of the code and the programming of the code with its limits and some specifications are described. The practical aspect of utilization is finally presented
Thermohydraulic behavior of the liquid metal target of a spallation neutron source
Energy Technology Data Exchange (ETDEWEB)
Takeda, Y.
1996-06-01
The author presents work done on three main problems. (1) Natural circulation in double coaxial cylindircal container: The thermohydraulic behaviour of the liquid metal target of the spallation neutron source at PSI has been investigated. The configuration is a natural-circulation loop in a concentric double-tube-type container. The results show that the natural-circulation loop concept is valid for the design phase of the target construction, and the current specified design criteria will be fulfilled with the proposed parameter values. (2) Flow around the window: Water experiments were performed for geometry optimisation of the window shape of the SINQ container for avoiding generating recirculation zones at peripheral area and the optimal cooling of the central part of the beam entrance window. Flow visualisation technique was mainly used for various window shapes, gap distance between the window and the guide tube edge. (3) Flow in window cooling channels: Flows in narrow gaps of cooling channels of two different types of windows were studied by flow visualisation techniques. One type is a slightly curved round cooling channel and the other is hemispherical shape, both of which have only 2 mm gap distance and the water inlet is located on one side and flows out from the opposite side. In both cases, the central part of the flow area has lower velocity than peripheral area.
Thermo-hydraulic analysis for SCWR during power-raising phase of startup
International Nuclear Information System (INIS)
The study of thermal characteristics during startup is one of the most important aspects for safety analysis of supercritical water-cooled reactor (SCWR). According to the given sliding pressure mode of SCWR, thermal analysis on temperature-raising phase and power-raising phase of startup are carried out. Considering the radial heterogeneity of power distribution,thermal characteristics for different assemblies during startup are also put forward. The results show that,during temperature-raising phase with core power increased only, the temperature of moderator, coolant and fuel cladding in inner assemblies are increased with little amplitude. During power-raising phase with core power and feed-water flow rate increased, the coolant temperature keeps unchanged, but the moderator temperature is decreased. With a greater variation of power, fuel cladding temperature shows a greater increase. Furthermore, considering the uneven distribution of radial power, thermo-hydraulic characteristics with uneven cladding temperature distribution shows a certain horizontal heterogeneity for different fuel assemblies, which becomes serious as flow rate and power increase. By adjusting flow rate distribution in different fuel assemblies or changing power setting during startup, the cladding temperature difference could be effectively reduced, which provides a certain reference for startup optimization of SCWR. (authors)
International Nuclear Information System (INIS)
The objective of the project was to draw up an instrumentation plan for the French core melting programme PHEBUS FP. This instrumentation plan essentially was to include proven and reliable instruments for recording various thermohydraulic, aerosol and hydrogen phenomena. The candidate measuring methods, which are known mainly from reactor safety programmes, have been described and examined for their usefulness in PHEBUS. Each method and instrument has been described in detail under various aspects such as measuring principle, measuring range, technical design, evaluation model, calibration procedure, accuracy, previous experience, commercial availability, etc. Special attention has been paid to the behaviour of the measuring transducers when exposed to radiation. First, the performance of the instruments was compared with the requirements of PHEBUS. The results of this comparison served as the basis for a measuring concept in tabular form, giving the locations of the measurements, the measuring tasks, and the number and kind of instruments that are recommended. Redundancy and cost-benefit aspects have been taken into account in qualitative terms
Bundling in semiflexible polymers: A theoretical overview.
Benetatos, Panayotis; Jho, YongSeok
2016-06-01
Supramolecular assemblies of polymers are key modules to sustain the structure of cells and their function. The main elements of these assemblies are charged semiflexible polymers (polyelectrolytes) generally interacting via a long(er)-range repulsion and a short(er)-range attraction. The most common supramolecular structure formed by these polymers is the bundle. In the present paper, we critically review some recent theoretical and computational advances on the problem of bundle formation, and point a few promising directions for future work. PMID:26813628
A bundle of sticks in my garden
Farran, Sue
2012-01-01
The English law of property is often described as a ‘bundle of sticks’ in which each ‘stick’ represents a particular right. Gardens challenge these rights and wreak havoc on the ‘bundle of sticks’. This paper looks at the twenty-first century manifestations of community engagement with ground and explores how ‘gardening’ is undermining concepts of ownership, possession and management of land and how the fence between what is private and what is public is being encroached and challenged by com...
Characteristic classes of quantum principal bundles
Durdevic, M
1995-01-01
A noncommutative-geometric generalization of classical Weil theory of characteristic classes is presented, in the conceptual framework of quantum principal bundles. A particular care is given to the case when the bundle does not admit regular connections. A cohomological description of the domain of the Weil homomorphism is given. Relations between universal characteristic classes for the regular and the general case are analyzed. In analogy with classical geometry, a natural spectral sequence is introduced and investigated. The appropriate counterpart of the Chern character is constructed, for structures admitting regular connections. Illustrative examples and constructions are presented.
TRIGA spent fuel bundles safe storage
Energy Technology Data Exchange (ETDEWEB)
Negut, G.; Covaci, St. [Institute for Nuclear Research, Research Reactor Dept., Pitesti (Romania); Prisecaru, I.; Dupleac, D. [Bucharest Univ. Politehnica, Power and Nuclear Engineering Dept., Bucharest (Romania)
2007-07-01
TRIGA-SSR is a steady state research and material test reactor that has been in operation since 1980. The original TRIGA fuel was HEU (highly enriched uranium) with a U{sup 235} enrichment of 93 per cent. Almost all TRIGA HEU fuel bundles are now burned-up. Part of the spent fuel was loaded and transferred to US, in a Romania - DOE arrangement. The rest of the TRIGA fuel bundles have to be temporarily stored in the TRIGA facility. As the storage conditions had to be established with caution, neutron and thermal hydraulic evaluations of the storage conditions were required. Some criticality evaluations were made based on the SAR (Safety Analysis Report) data. Fuel constant axial temperature approximation effect is usual for criticality computations. TRIGA-SSR fuel bundle geometry and materials model for SCALE5-CSAS module allows the introduction of a fuel temperature dependency for the entire fuel active height, using different materials for each fuel bundle region. Previous RELAP5 thermal hydraulic computations for an axial and radial power distribution in the TRIGA fuel pin were done. Fuel constant temperature approximation overestimates pin factors for every core operating at high temperatures. From the thermal hydraulic point of view the worst condition of the storage grid occurs when the transfer channel is accidentally emptied of water from the pool, or the bundle is handled accidentally to remain in air. All the residual heat from the bundles has to be removed without fuel overheating and clad failure. RELAP5 computer code for residual heat removal was used in the assessment of residual heat removal. We made a couple of evaluations of TRIGA bundle clad temperatures in air cooling conditions, with different residual heat levels. The criticality computations have shown that the spent TRIGA fuel bundles storage grid is strongly sub-critical with k(eff) = 0.5951. So, there is no danger for a criticality accident for this storage grid type. The assessment is done
Scaling Shift in Multicracked Fiber Bundles
Manca, Fabio; Giordano, Stefano; Palla, Pier Luca; Cleri, Fabrizio
2014-12-01
Bundles of fibers, wires, or filaments are ubiquitous structures in both natural and artificial materials. We investigate the bundle degradation induced by an external damaging action through a theoretical model describing an assembly of parallel fibers, progressively damaged by a random population of cracks. Fibers in our model interact by means of a lateral linear coupling, thus retaining structural integrity even after substantial damage. Monte Carlo simulations of the Young's modulus degradation for increasing crack density demonstrate a remarkable scaling shift between an exponential and a power-law regime. Analytical solutions of the model confirm this behavior, and provide a thorough understanding of the underlying physics.
Safe Harbors for Quantity Discounts and Bundling
Dennis W. Carlton; Michael Waldman
2008-01-01
The courts and analysts continue to struggle to articulate safe harbors for a wide variety of common business pricing practices in which either a single product is sold at a discount if purchased in bulk or in which multiple products are bundled together at prices different from the ones that would emerge if the products were purchased separately. The phenomenon of tying in which the sale of one product is conditioned on the purchase of another is closely related to bundling. Its analysis rel...
TRIGA spent fuel bundles safe storage
International Nuclear Information System (INIS)
TRIGA-SSR is a steady state research and material test reactor that has been in operation since 1980. The original TRIGA fuel was HEU (highly enriched uranium) with a U235 enrichment of 93 per cent. Almost all TRIGA HEU fuel bundles are now burned-up. Part of the spent fuel was loaded and transferred to US, in a Romania - DOE arrangement. The rest of the TRIGA fuel bundles have to be temporarily stored in the TRIGA facility. As the storage conditions had to be established with caution, neutron and thermal hydraulic evaluations of the storage conditions were required. Some criticality evaluations were made based on the SAR (Safety Analysis Report) data. Fuel constant axial temperature approximation effect is usual for criticality computations. TRIGA-SSR fuel bundle geometry and materials model for SCALE5-CSAS module allows the introduction of a fuel temperature dependency for the entire fuel active height, using different materials for each fuel bundle region. Previous RELAP5 thermal hydraulic computations for an axial and radial power distribution in the TRIGA fuel pin were done. Fuel constant temperature approximation overestimates pin factors for every core operating at high temperatures. From the thermal hydraulic point of view the worst condition of the storage grid occurs when the transfer channel is accidentally emptied of water from the pool, or the bundle is handled accidentally to remain in air. All the residual heat from the bundles has to be removed without fuel overheating and clad failure. RELAP5 computer code for residual heat removal was used in the assessment of residual heat removal. We made a couple of evaluations of TRIGA bundle clad temperatures in air cooling conditions, with different residual heat levels. The criticality computations have shown that the spent TRIGA fuel bundles storage grid is strongly sub-critical with k(eff) = 0.5951. So, there is no danger for a criticality accident for this storage grid type. The assessment is done for
Impact of bundle deformation on CHF: ASSERT-PV assessment of extended burnup Bruce B bundle G85159W
International Nuclear Information System (INIS)
This paper presents a subchannel thermalhydraulic analysis of the effect on critical heat flux (CHF) of bundle deformation such as element bow and diametral creep. The bundle geometry is based on the post-irradiation examination (PIE) data of a single bundle from the Bruce B Nuclear Generating Station, Bruce B bundle G85159W, which was irradiated for more than two years in the core during reactor commissioning. The subchannel code ASSERT-PV IST is used to assess changes in CHF and dryout power due to bundle deformation, compared to the reference, undeformed bundle. (author)
Abelian conformal field theory and determinant bundles
DEFF Research Database (Denmark)
Andersen, Jørgen Ellegaard; Ueno, K.
2007-01-01
Following [10], we study a so-called bc-ghost system of zero conformal dimension from the viewpoint of [14, 16]. We show that the ghost vacua construction results in holomorphic line bundles with connections over holomorphic families of curves. We prove that the curvature of these connections are...
Optimization of a bundle divertor for FED
International Nuclear Information System (INIS)
Optimal double-T bundle divertor configurations have been obtained for the Fusion Engineering Device (FED). On-axis ripple is minimized, while satisfying a series of engineering constraints. The ensuing non-linear optimization problem is solved via a sequence of quadratic programming subproblems, using the VMCON algorithm. The resulting divertor designs are substantially improved over previous configurations
Capacity efficiency of recovery request bundling
DEFF Research Database (Denmark)
Ruepp, Sarah Renée; Dittmann, Lars; Berger, Michael Stübert; Stidsen, Thomas Riis; Lagakos, Stephen; Perlovsky, Leonid; Jha, Manoi; Covaci, Brindusa; Zaharim, Azarni; Mastorakis, Nikos
2010-01-01
This paper presents a comparison of recovery methods in terms of capacity efficiency. In particular, a method where recovery requests are bundled towards the destination (Shortcut Span Protection) is evaluated against traditional recovery methods. Our simulation results show that Shortcut Span Pr...... Protection uses more capacity than the unbundled related methods, but this is compensated by easier control and management of the recovery actions....
Line bundles on moduli and related spaces
Huebschmann, Johannes
2009-01-01
Let G be a Lie goup, let M and N be smooth connected G-manifolds, let f be a smooth G-map from M to N, and let P denote the fiber of f. Given a closed and equivariantly closed relative 2-form for f with integral periods, we construct the principal G-circle bundles with connection on P having the given relative 2-form as curvature. Given a compact Lie group K, a biinvariant Riemannian metric on K, and a closed Riemann surface S of genus s, when we apply the construction to the particular case where f is the familiar relator map from a product of 2s copies of K to K we obtain the principal K-circle bundles on the associated extended moduli spaces which, via reduction, then yield the corresponding line bundles on possibly twisted moduli spaces of representations of the fundamental group of S in K, in particular, on moduli spaces of semistable holomorphic vector bundles or, more precisely, on a smooth open stratum when the moduli space is not smooth. The construction also yields an alternative geometric object, d...
Bundle Gerbes Applied to Quantum Field Theory
Carey, A L; Murray, M; Carey, Alan; Mickelsson, Jouko; Murray, Michael
2000-01-01
This paper reviews recent work on a new geometric object called a bundle gerbe and discusses some new examples arising in quantum field theory. One application is to an Atiyah-Patodi-Singer index theory construction of the bundle of fermionic Fock spaces parametrized by vector potentials in odd space dimensions and a proof that this leads in a simple manner to the known Schwinger terms (Mickelsson-Faddeev cocycle) for the gauge group action. This gives an explicit computation of the Dixmier-Douady class of the associated bundle gerbe. The method works also in other cases of fermions in external fields (external gravitational field, for example) provided that the APS theorem can be applied; however, we have worked out the details only in the case of vector potentials. Another example, in which the bundle gerbe curvature plays a role, arises from the WZW model on Riemann surfaces. A further example is the `existence of string structures' question. We conclude by showing how global Hamiltonian anomalies fit with...
Quantum field theories on Hilbert bundles
International Nuclear Information System (INIS)
We investigate whether it is possible to maintain the computational features of QED while avoiding some of its mathematical difficulties by formulating QFTs on Hilber bundles. This encounters two problems: 1) Haag's theorem persists, and 2) admissible fields do not generate motions on the base space. To do the latter, the coupling constant has to be a vector field upon the base space. (orig.)
Assembly mechanism for nuclear fuel bundles
International Nuclear Information System (INIS)
This invention relates to an assembly mechanism for nuclear power reactor fuel bundles using a novel, simple and inexpensive means. The mechanism is readily operable remotely, avoids separable parts and is applicable to fuel assemblies in which the upper tie plate is rigidly mounted on the tie rods which hold it in place. (UK)
Capacity efficiency of recovery request bundling
DEFF Research Database (Denmark)
Ruepp, Sarah Renée; Dittmann, Lars; Berger, Michael Stübert; Stidsen, Thomas Riis; Lagakos, Stephen; Perlovsky, Leonid; Jha, Manoi; Covaci, Brindusa; Zaharim, Azarni; Mastorakis, Nikos
2010-01-01
This paper presents a comparison of recovery methods in terms of capacity efficiency. In particular, a method where recovery requests are bundled towards the destination (Shortcut Span Protection) is evaluated against traditional recovery methods. Our simulation results show that Shortcut Span...
Riemann Surfaces: Vector Bundles, Physics, and Dynamics
DEFF Research Database (Denmark)
Sikander, Shehryar
the monodromy with respect to the pulled back connection. The formula for the representation includes a series with coefficients as iterated integrals. This series is closely related to the cyclotomic version of the Drinfel'd associator. The geodesic flow in the unit the tangent bundle of this Teichmueller...
In-pile test of Qinshan PWR fuel bundle
International Nuclear Information System (INIS)
In-pile test of Qinshan Nuclear Power Plant PWR fuel bundle has been conducted in HWRR HTHP Test loop at CIAE. The test fuel bundle was irradiated to an average burnup of 25000 Mwd/tU. The authors describe the structure of (3 x 3-2) test fuel bundle, structure of irradiation rig, fuel fabrication, irradiation conditions, power and fuel burnup. Some comments on the in-pile performance for fuel bundle, fuel rod and irradiation rig were made
Holomorphic Vector Bundle on Hopf Manifolds with Abelian Fundamental Groups
Institute of Scientific and Technical Information of China (English)
Xiang Yu ZHOU; Wei Ming LIU
2004-01-01
Let X be a Hopf manifolds with an Abelian fundamental group. E is a holomorphic vector bundle of rank r with trivial pull-back to W = Cn - {0}. We prove the existence of a non-vanishing section of L(×) E for some line bundle on X and study the vector bundles filtration structure of E. These generalize the results of D. Mall about structure theorem of such a vector bundle E.
Anatomic Double-Bundle Posterior Cruciate Ligament Reconstruction
Chahla, Jorge; Nitri, Marco; Civitarese, David; Dean, Chase S.; Moulton, Samuel G.; LaPrade, Robert F.
2016-01-01
The posterior cruciate ligament (PCL) is known to be the main posterior stabilizer of the knee. Anatomic single-bundle PCL reconstruction, focusing on reconstruction of the larger anterolateral bundle, is the most commonly performed procedure. Because of the residual posterior and rotational tibial instability after the single-bundle procedure and the inability to restore the normal knee kinematics, an anatomic double-bundle PCL reconstruction has been proposed in an effort to re-create the n...
Existence of vector bundles and global resolutions for singular surfaces
Vezzosi, G; S. SCHROER
2002-01-01
Abstract- We prove two results about vector bundles on singular algebraic surfaces. First, on proper surfaces there are vector bundles of rank two with arbitrarily large second Chern number and fixed determinant. Second, on separated normal surfaces any coherent sheaf is the quotient of a vector bundle. As a consequence, for such surfaces the Quillen K-theory of vector bundles coincides with the Waldhausen K-theory of perfect complexes. Examples show that, on non-separated schemes, usually...
CANFLEX - an advanced fuel bundle for CANDU
International Nuclear Information System (INIS)
The performance of CANDU pressurized heavy-water reactors, in terms of lifetime load factors, is excellent. More than 600 000 bundles containing natural-uranium fuel have been irradiated, with a low defect rate; reactor unavailability due to fuel incidents is typically zero. To maintain and improve CANDU's competitive position, Atomic Energy of Canada Limited (AECL) has an ongoing program comprising design, safety and availability improvements, advanced fuel concepts and schemes to reduce construction time. One key finding is that the introduction of slightly-enriched uranium (SEU, less than 1.5 wt% U-235 in U) offers immediate benefits for CANDU, in terms of fuelling and back-end disposal costs. The use of SEU places more demands on the fuel because of extended burnup, and an anticipated capability to load-follow also adds to the performance requirements. To ensure that the duty-cycle targets for SEU and load-following are achieved, AECL is developing a new fuel bundle, termed CANFLEX (CANdu FLEXible), where flexible refers to the versatility of the bundle with respect to operational and fuel-cycle options. Though the initial purpose of the new 43-element bundle is to introduce SEU into CANDU, CANFLEX is extremely versatile in its application, and is compatible with other fuel cycles of interest: natural uranium in existing CANDU reactors, recycled uranium and mixed-oxides from light-water reactors, and thoria-based fuels. Capability with a variety of fuel cycles is the key to future CANDU success in the international market. The improved performance of CANFLEX, particularly at high burnups, will ensure that the full economic benefits of advanced fuels cycles are achieved. A proof-tested CANFLEX bundle design will be available in 1993 for large-scale commercial-reactor demonstration
Interplanetary Overlay Network Bundle Protocol Implementation
Burleigh, Scott C.
2011-01-01
The Interplanetary Overlay Network (ION) system's BP package, an implementation of the Delay-Tolerant Networking (DTN) Bundle Protocol (BP) and supporting services, has been specifically designed to be suitable for use on deep-space robotic vehicles. Although the ION BP implementation is unique in its use of zero-copy objects for high performance, and in its use of resource-sensitive rate control, it is fully interoperable with other implementations of the BP specification (Internet RFC 5050). The ION BP implementation is built using the same software infrastructure that underlies the implementation of the CCSDS (Consultative Committee for Space Data Systems) File Delivery Protocol (CFDP) built into the flight software of Deep Impact. It is designed to minimize resource consumption, while maximizing operational robustness. For example, no dynamic allocation of system memory is required. Like all the other ION packages, ION's BP implementation is designed to port readily between Linux and Solaris (for easy development and for ground system operations) and VxWorks (for flight systems operations). The exact same source code is exercised in both environments. Initially included in the ION BP implementations are the following: libraries of functions used in constructing bundle forwarders and convergence-layer (CL) input and output adapters; a simple prototype bundle forwarder and associated CL adapters designed to run over an IPbased local area network; administrative tools for managing a simple DTN infrastructure built from these components; a background daemon process that silently destroys bundles whose time-to-live intervals have expired; a library of functions exposed to applications, enabling them to issue and receive data encapsulated in DTN bundles; and some simple applications that can be used for system checkout and benchmarking.
Experimental studies on heat transfer to supercritical water in 2 × 2 rod bundle with two channels
Energy Technology Data Exchange (ETDEWEB)
Gu, H.Y., E-mail: guhanyang@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Dongchuan Road 800, 200240, Shanghai (China); Hu, Z.X.; Liu, D.; Xiao, Y. [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Dongchuan Road 800, 200240, Shanghai (China); Cheng, X. [Institute of Fusion and Reactor Technologies (IFRT), Karlsruhe Institute of Technologies (KIT), Karlsruhe, 76131 (Germany)
2015-09-15
Highlights: • Heat transfer to supercritical water in a 2 × 2 rod bundle is investigated. • Effects of system parameters on heat transfer in bundle are analyzed. • The test data were compared with twenty heat transfer correlations. - Abstract: The experiment of heat transfer to supercritical water in 2 × 2 rod bundle is performed at Shanghai Jiao Tong University. The test section consists of two channels separated by a square steel assembly box with rounded corners. Water flows downward in the first channel and then turns upward in the second channel to cool the 2 × 2 rod bundle installed inside the assembly box. The bundle consists of four heated rods of 10 mm in O.D. and 1.18 in pitch-to-diameter ratio. The fluid enthalpy in the first channel increases due to the heat transfer through the assembly box when flowing downward. The minimum fluid enthalpy increase in the first channel appears at the pseudo-critical region due to the small temperature difference between the two channels. Effects of various parameters on heat transfer behavior inside the 2 × 2 rod bundle are similar to those observed in tube or annuli. No special phenomenon in heat transfer is observed during the mass flux and power transient. The steady-state heat transfer correlation is applicable to predict the heat transfer in the mass or power transient sequence. In addition, the importance of several dimensionless numbers and the accuracy of 20 heat transfer correlations are assessed. It is concluded that the buoyancy parameter proposed by Cheng et al. (2009) shows unique effect on heat transfer coefficient. Among the 20 selected heat transfer correlations, the correlations of Jackson and Fewster (1975) and Bishop et al. (1964) give the best predictions when compared with the experimental data.
Compactifications of reductive groups as moduli stacks of bundles
DEFF Research Database (Denmark)
Martens, Johan; Thaddeus, Michael
Let G be a reductive group. We introduce the moduli problem of "bundle chains" parametrizing framed principal G-bundles on chains of lines. Any fan supported in a Weyl chamber determines a stability condition on bundle chains. Its moduli stack provides an equivariant toroidal compactification of ...
VECTOR BUNDLE, KILLING VECTOR FIELD AND PONTRYAGIN NUMBERS
Institute of Scientific and Technical Information of China (English)
周建伟
1991-01-01
Let E be a vector bundle over a compact Riemannian manifold M. We construct a natural metric on the bundle space E and discuss the relationship between the killing vector fields of E and M. Then we give a proof of the Bott-Baum-Cheeger Theorem for vector bundle E.
Noncommutative principal torus bundles via parametrised strict deformation quantization
Hannabuss, Keith; Mathai, Varghese
2009-01-01
In this paper, we initiate the study of a parametrised version of Rieffel's strict deformation quantization. We apply it to give a classification of noncommutative principal torus bundles, in terms of parametrised strict deformation quantization of ordinary principal torus bundles. The paper also contains a putative definition of noncommutative non-principal torus bundles.
Laser cutting for dismantling of PHWR fuel bundles
International Nuclear Information System (INIS)
Detailed investigation was carried out on laser cutting of zircaloy-2 PHWR fuel pin bundles. Initially, trials were done to standardize ten parameters for cutting of tie plates to which individual fuel pins are welded in a bundle. Using these parameters, the tie plates were cut into several pieces so that each fuel pin is individually separated out from the bundle. (author)
Geometry of torus bundles in integrable Hamiltonian systems
Lukina, Olga
2008-01-01
Thesis is concerned with global properties of Lagrangian bundles, i.e. symplectic n-torus bundles, as these occur in integrable Hamiltonian systems. It treats obstructions to triviality and concerns with classification of such bundles, as well as with manifestations of global invariants in real-worl
Stability of Picard Bundle Over Moduli Space of Stable Vector Bundles of Rank Two Over a Curve
Indian Academy of Sciences (India)
Indranil Biswas; Tomás L Gómez
2001-08-01
Answering a question of [BV] it is proved that the Picard bundle on the moduli space of stable vector bundles of rank two, on a Riemann surface of genus at least three, with fixed determinant of odd degree is stable.
Numerical simulation of the flow in a tight lattice SFR rod-bundle with grid spacers
International Nuclear Information System (INIS)
The accurate prediction of the flow in rod bundles is crucial for both the design and the safe operation of nuclear reactor systems. However the geometry complexity and the non-uniformity of the flow introduce peculiar features that can be reproduced only with three-dimensional CFD. In fact, even in bare rod bundles, i.e., bundles that do not contain any spacing device, the structure of the flow is inherently complex, due to the presence of a large-scale instability. Moreover, experimental analysis has clearly shown that when reducing the pitch-to-diameter ratio (P/D) the turbulence field in rod bundles deviates significantly from that in a circular tube. For extremely tight configurations the existence of large-scale coherent structures has been shown, which is responsible for the high inter-sub-channel heat and momentum exchange. While there is a fairly extensive literature on the presence of these structures in bare rod-bundles and simplified geometries, there are no available studies of their presence in geometry containing grid spacers or wires. A series of fully transient simulations of turbulence have been performed for an infinite tight triangular lattice (typical of current SFR designs) with and without a grid spacer. The simulations have been performed using Large Eddy Simulation (LES) with the spectral element code Nek5000 and unsteady Reynolds Averaged Navier-Stokes (URANS) for P/D=1.08. Several structure recognition techniques and statistical methods have been applied in order to investigate the flow field and the three-dimensional pattern of the coherent structures. The results prove that, for the configuration studied, coherent structures are indeed present and contribute significantly to the flow dynamics. (author)
International Nuclear Information System (INIS)
In a liquid metal fast reactor (LMFR), temperature fluctuations in the fluid close to a structure occur in many areas: core outlet zone, lower part of hot pool, free surface of pool, IHX outlet, secondary circuit, water steam interface in steam generators. In some conditions, these temperature fluctuations can lead to mechanical damage to structures. Consequently, knowledge of temperature fluctuations and induced thermomechanical damage to structures is essential to support design and maintenance during the plant life-time. In response to a recommendation from the IWGFR, the IAEA convened a Specialist Meeting on 'Correlation between material properties and thermohydraulics conditions in LMFRs' in November 1994. The purpose of the meeting was to exchange information on the state of the art on thermalhydraulic aspects of temperature fluctuations (mixing jet phenomena, temperature gradient fluctuations, transfer of fluctuations from the fluid to the wall), and associated thermomechanical studies (thermal striping, thermal ratchetting, high strain fatigue) as well as design criteria to avoid damage. The main areas discussed by the delegates were: thermalhydraulics and thermomechanics. The objective of thermalhydraulic activities is the characterization of the temperature fluctuations on the wall. Three main items can be identified, for which both the experimental and calculational approaches were considered: identification of the areas where the fluctuations may occur; characterization of the fluctuations in the fluid; and transfer of the fluid fluctuations to the walls. For thermomechanical studies, which cover the effect of the fluctuations in the structures, the following subjects are of great importance: determination of the damage modes induced by thermal loadings in structures (thermal striping, ratchetting, high strain fatigue), and study of all damage modes so as to take them into account in the design criteria and to provide rules for avoiding failure of the
International Nuclear Information System (INIS)
A large-scale sodium-cooled fast breeder reactor in the feasibility studies on commercialized fast reactors has a feature of consideration of thorough simplified and compacted systems and components design to realize drastic economical improvements. Therefore, special attentions should be paid to thermohydraulic designs for gas entrainment behavior from free surface, flow-induced vibration of in-vessel components, thermal stratification in the plenum, thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. A numerical analysis was carried out with a multi-dimensional code AQUA to confirm an applicability to the evaluations for the in-vessel thermohydraulic phenomena using a 1/10 scaled water experiment simulating the large-scale fast breeder reactor in the feasibility studies. From the analysis, the following results were obtained. (1) In-vessel thermohydraulics characterized by a radiated flow pattern to the reactor vessel wall and a strong upward flow through a slit of the upper core structures were evaluated. These characteristics agreed approximately with the water experiment. (2) The upward velocity values at the slit agreed well with the experimental data under a condition of γz = 0.3 and ξz = 0.5, though overall evaluations of the in-vessel thermohydraulics were failed to predict quantitatively. (3) The AQUA code is applicable to the in-vessel thermohydraulics evaluations in the feasibility studies, though it is necessary to make further modifications of the calculational models for accurate evaluations. On the one hand, it was confirmed that calculated results for the 1/10 water experimental model and the 1/1 actual-scaled model agreed quantitatively for the in-vessel thermohydraulics characteristics indicated above. (author)
Flow and heat transfer thermohydraulic modelisation during the reflooding phase of a P.W.R.'s core
International Nuclear Information System (INIS)
Some generalities about L.O.C.A. are first recalled. The French experimental studies about Emergency Core Cooling System are briefly described. The different heat transfer mechanisms to take into account, according to the flow pattern in the dry zone, and the correlations or methods to calculate them, are defined. Then the Thermohydraulic code computer: FLIRA, which describe the reflooding phase, and a modelisation taking into account the different flow patterns are setted. A first interpretation of ERSEC experiments with a tubular test section shows that it is possible, with this modelisation and some classical heat transfer correlations, to describe the reflooding phase.
CFD study on coolant mixing in VVER-440 fuel rod bundles and fuel assembly heads
International Nuclear Information System (INIS)
A CFD model of VVER-440 fuel assembly heads was developed based on the technical documentation of a full-scale test facility built in the Kurchatov Institute, Russia. Steady-state and transient calculations were performed to validate the model with a measurement set. Effects of the spatial resolution, turbulence models, difference schemes and different inlet boundary conditions were investigated. Inlet boundary conditions were determined with both the COBRA subchannel code and a fuel rod bundle CFD model that was built for this special purpose. The results were compared against experimental data. The sensitivity studies showed that a grid of about 8 million cells, high resolution scheme and BSL Reynolds stress model are suitable sets to provide accurate prediction for the signal of the in-core thermocouple. The best prediction was achieved with transient calculation using inlet boundary conditions generated with the CFD fuel rod bundle model. The results indicated that the coolant mixing is intensive but not perfect in the assembly head. Besides, the significant role of the outflow from the central tube was also proven. The transient runs revealed relatively large temperature fluctuations near the in-core thermocouple housing.
Productivity and costs of slash bundling in Nordic conditions
Energy Technology Data Exchange (ETDEWEB)
Kaerhae, K.; Vartiamaeki, T. [Metsaeteho Oy, P.O. Box 101, FI-00171 Helsinki (Finland)
2006-12-15
The number of slash bundlers and the volume of slash bundling have been rapidly increasing during the last few years in Finland. However, no comprehensive time or follow-up studies have been carried out on slash bundling technology in Finland or in any other country. Metsateho Oy carried out studies on the productivity and costs of slash bundling in different Nordic recovering conditions. The study methods included both time and follow-up studies. Data were collected during the summer and winter period primarily in Norway spruce (Picea abies L. Karst.) dominated clear cutting sites. The bundling techniques performed by different types of bundler (Fiberpac 370, Timberjack 1490D, Pika RS 2000, Valmet WoodPac) were studied. The average productivity of slash bundling was 18.1 bundles per operating (E{sub 15}, including delays shorter than 15min) hour with the Timberjack 1490D and Fiberpac 370 bundlers in the follow-up study. The operator of the slash bundler had the greatest effect on the productivity of bundling. The prerequisite for increased bundling volumes is a reduction in the costs of the most expensive sub-stage of the bundling supply chain, i.e. bundling itself. This requires improved recovery conditions at bundling sites, increased bundling productivity, larger sized bundles, and the execution of bundling operations in two work shifts using an efficient bundler and effective operator working methods. Implementation of these development measures will bring the bundling supply chain up to a speed that makes it the most competitive supply chain for forest chips in terms of total supply costs for long-distance transportation distances of more than 60km. (author)
Energy Technology Data Exchange (ETDEWEB)
Kaipainen, H.; Seppaenen, V.; Rinne, S.
1996-12-31
The conditions on which the bundling of the harvesting residues from spruce regeneration fellings would become profitable were studied. The calculations showed that one of the most important features was sufficient compaction of the bundle, so that the portion of the wood in the unit volume of the bundle has to be more than 40 %. The tests showed that the timber grab loader of farm tractor was insufficient for production of dense bundles. The feeding and compression device of the prototype bundler was constructed in the research and with this device the required density was obtained.The rate of compaction of the dry spruce felling residues was about 40 % and that of the fresh residues was more than 50 %. The comparison between the bundles showed that the calorific value of the fresh bundle per unit volume was nearly 30 % higher than that of the dry bundle. This means that the treatment of the bundles should be done of fresh felling residues. Drying of the bundles succeeded well, and the crushing and chipping tests showed that the processing of the bundles at the plant is possible. The treatability of the bundles was also excellent. By using the prototype, developed in the research, it was possible to produce a bundle of the fresh spruce harvesting residues, the diameter of which was about 50 cm and the length about 3 m, and the rate of compaction over 50 %. By these values the reduction target of the costs is obtainable
International Nuclear Information System (INIS)
At the Novo-Voronezh Nuclear Power Plant, the fifth VVER-1000 unit, which was operated at nominal power from February 1980, completed nine fuel cycles in July 1990. The first unit of the Kalinin Nuclear Power Plant has operated from April 1984; in October 1990 the sixth fuel loading was completed. To data these power units are operating in steady-state in three-year fuel cycles (from June 1986 and from September 1989, respectively). By the end of 1988, operational experience had been accumulated on 1407 fuel element bundles on the third to the sixth fuel loading at Kalinin and the fifth to the ninth at Novo-Voronezh, which are in the transient and steady-state regimes of a three-year cycle. Of the 561 fuel element bundles monitored for gamma radiation, 14 were designated as leaking, which was 2.5% of the total bundles or 0.008% of the total number of fuel elements. Thus, a high degree of reliability was attained with enriched fuel elements. Here the authors analyze the reliability of fuel element bundles in taking the VVER-1000s to a three-year fuel cycle, and also generalize and systematize information on the fundamental characteristics of a group of fuel element bundles in going to to steady-state conditions of the three-year fuel cycle
Cobra-IE Evaluation by Simulation of the NUPEC BWR Full-Size Fine-Mesh Bundle Test (BFBT)
Energy Technology Data Exchange (ETDEWEB)
Burns, C. J. and Aumiler, D. L.
2006-04-26
The COBRA-IE computer code is a thermal-hydraulic subchannel analysis program capable of simulating phenomena present in both PWRs and BWRs. As part of ongoing COBRA-IE assessment efforts, the code has been evaluated against experimental data from the NUPEC BWR Full-Size Fine-Mesh Bundle Tests (BFBT). The BFBT experiments utilized an 8 x 8 rod bundle to simulate BWR operating conditions and power profiles, providing an excellent database for investigation of the capabilities of the code. Benchmarks performed included steady-state and transient void distribution, single-phase and two-phase pressure drop, and steady-state and transient critical power measurements. COBRA-IE effectively captured the trends seen in the experimental data with acceptable prediction error. Future sensitivity studies are planned to investigate the effects of enabling and/or modifying optional code models dealing with void drift, turbulent mixing, rewetting, and CHF.
Cobra-IE Evaluation by Simulation of the NUPEC BWR Full-Size Fine-Mesh Bundle Test (BFBT)
International Nuclear Information System (INIS)
The COBRA-IE computer code is a thermal-hydraulic subchannel analysis program capable of simulating phenomena present in both PWRs and BWRs. As part of ongoing COBRA-IE assessment efforts, the code has been evaluated against experimental data from the NUPEC BWR Full-Size Fine-Mesh Bundle Tests (BFBT). The BFBT experiments utilized an 8 x 8 rod bundle to simulate BWR operating conditions and power profiles, providing an excellent database for investigation of the capabilities of the code. Benchmarks performed included steady-state and transient void distribution, single-phase and two-phase pressure drop, and steady-state and transient critical power measurements. COBRA-IE effectively captured the trends seen in the experimental data with acceptable prediction error. Future sensitivity studies are planned to investigate the effects of enabling and/or modifying optional code models dealing with void drift, turbulent mixing, rewetting, and CHF
International Nuclear Information System (INIS)
This report describes the results of the steady state thermohydraulic analysis of upgraded JRR-3 core under natural convective cooling mode, using COOLOD-N code. In the code, function to calculate flow-rate under natural convective cooling mode, and a heat transfer package have been newly added to the COOLOD code which has been developed in JAERI. And this report describes outline of the COOLOD-N code. The results of analysis show that the thermohydraulics of upgraded JRR-3 core, under natural convective cooling mode have enough margine to ONB temperature, DNB heat flux and occurance of blisters in fuel meats, which are design criterion of upgraded JRR-3. (author)
International Nuclear Information System (INIS)
Requirements of neutron, thermohydraulic and safety analysis calculation are very important because of issuing new version of SAR for DNRR, research on construction of new research reactor and nuclear power plant. Research on application of system of neutron, thermohydraulic and safety analysis codes in order to simulation of the Dalat Nuclear Research Reactor has been done in the frame work of research theme in the year 2002-2003. The purposes of the research are maintaining safety operation of the DNRR and enhancement of man power and calculation and safety analysis tool potential. (author)
DYN3D/M2 - a Code for Calculation of Reactivity Transients in Cores with Hexagonal Geometry
Rohde, Ulrich; Grundmann, Ulrich
2010-01-01
The code DYN3D/M2 consists of a the 3-dimensional neutron kinetic model of the code HEXDYN3D and the thermohydraulic model of the code FLOCAL. The neutron kinetics of DYN3D/M2 is calculated by using a nodal expansion method (NEM) for hexagonal geometry. The developed method solves the neutron diffusion equation for two energy groups. Stationary state and transient behaviour can be calculated. By help of the code PREPAR-EC parameterizid neutron physical constants of given burnup distribution c...
Comparison of ASSERT subchannel code with Marviken bundle data
International Nuclear Information System (INIS)
In this paper ASSERT predictions are compared with the Marviken 6-rod bundle and 36+1 rod bundle. The predictions are presented for two experiments in the 6-rod bundle and four experiments in the 36+1 rod bundle. For low inlet subcooling, the void predictions are in good agreement with the experimental data. For high inlet subcooling, however, the agreement is not as good. This is attributed to the fact that in the high inlet subcooling experiments, single phase turbulent mixing plays a more important role in determining flow conditions in the bundle
Multiwalled carbon nanotube reinforced biomimetic bundled gel fibres.
Kim, Young-Jin; Yamamoto, Seiichiro; Takahashi, Haruko; Sasaki, Naruo; Matsunaga, Yukiko T
2016-08-19
This work describes the fabrication and characterization of hydroxypropyl cellulose (HPC)-based biomimetic bundled gel fibres. The bundled gel fibres were reinforced with multiwalled carbon nanotubes (MWCNTs). A phase-separated aqueous solution with MWCNT and HPC was transformed into a bundled fibrous structure after being injected into a co-flow microfluidic device and applying the sheath flow. The resulting MWCNT-bundled gel fibres consist of multiple parallel microfibres. The mechanical and electrical properties of MWCNT-bundled gel fibres were improved and their potential for tissue engineering applications as a cell scaffold was demonstrated. PMID:27200527
Effectiveness of Hair Bundle Motility as the Cochlear Amplifier
Sul, Bora; Iwasa, Kuni H.
2009-01-01
The effectiveness of hair bundle motility in mammalian and avian ears is studied by examining energy balance for a small sinusoidal displacement of the hair bundle. The condition that the energy generated by a hair bundle must be greater than energy loss due to the shear in the subtectorial gap per hair bundle leads to a limiting frequency that can be supported by hair-bundle motility. Limiting frequencies are obtained for two motile mechanisms for fast adaptation, the channel re-closure mode...
Anatomic Double-Bundle Posterior Cruciate Ligament Reconstruction.
Chahla, Jorge; Nitri, Marco; Civitarese, David; Dean, Chase S; Moulton, Samuel G; LaPrade, Robert F
2016-02-01
The posterior cruciate ligament (PCL) is known to be the main posterior stabilizer of the knee. Anatomic single-bundle PCL reconstruction, focusing on reconstruction of the larger anterolateral bundle, is the most commonly performed procedure. Because of the residual posterior and rotational tibial instability after the single-bundle procedure and the inability to restore the normal knee kinematics, an anatomic double-bundle PCL reconstruction has been proposed in an effort to re-create the native PCL footprint more closely and to restore normal knee kinematics. We detail our technique for an anatomic double-bundle PCL reconstruction using Achilles and anterior tibialis tendon allografts. PMID:27284530
Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments
International Nuclear Information System (INIS)
The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results
Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments
Energy Technology Data Exchange (ETDEWEB)
Mendler, O J; Takeuchi, K; Young, M Y
1986-10-01
The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results.
The turbulent flow in rod bundles
International Nuclear Information System (INIS)
Experimental studies have shown that the axial and azimuthal turbulence intensities in the gap regions of rod bundles increase strongly with decreasing rod spacing; the fluctuating velocities in the axial and azimuthal directions have a quasi-periodic behaviour. To determine the origin of this phenomenon, an its characteristics as a function of the geometry and the Reynolds number, an experimental investigation was performed on the turbulent in several rod bundles with different aspect ratios (P/D, W/D). Hot-wires and microsphones were used for the measurements of velocity and wall pressure fluctuations. The data were evaluated to obtain spectra as well as auto and cross correlations. Based on the results, a phenomenological model is presented to explain this phenomenon. By means of the model, the mass exchange between neighbouring subchannels is explained
Reactor application of an improved bundle divertor
International Nuclear Information System (INIS)
A Bundle Divertor was chosen as the impurity control and plasma exhaust system for the beam driven Demonstration Tokamak Hybrid Reactor - DTHR. In the context of a preconceptual design study of the reactor and associated facility a bundle divertor concept was developed and integrated into the reactor system. The overall system was found feasible and scalable for reactors with intermediate torodial field strengths on axis. The important design characteristics are: the overall average current density of the divertor coils is 0.73 kA for each tesla of toroidal field on axis; the divertor windings are made from super-conducting cables supported by steel structures and are designed to be maintainable; the particle collection assembly and auxiliary cryosorption vacuum pump are dual systems designed such that they can be reactivated alterntively to allow for continuous reactor operation; and the power requirement for energizing and operating the divertor is about 5 MW
Venereau polynomials and related fiber bundles
Kaliman, Shulim; ZAIDENBERG, MIKHAIL
2003-01-01
The Venereau polynomials v-n:=y+x^n(xz+y(yu+z^2)), n>= 1, on A4 have all fibers isomorphic to the affine space A3. Moreover, for all n>= 1 the map (v-n, x) : A4 -> A2 yields a flat family of affine planes over A2. In the present note we show that over the punctured plane A2\\0, this family is a fiber bundle. This bundle is trivial if and only if v-n is a variable of the ring C[x][y,z,u] over C[x]. It is an open question whether v1 and v2 are variables of the polynomial ring C[x,y,z,u]. S. Vene...
A fibre bundle formulation of quantum geometry
International Nuclear Information System (INIS)
Quantum geometries whose points are stochastic and serve as seats for quantum space-time excitons are formulated as fibre bundles over base spaces of mean values with a Minkowski or general relativistic structure. The fibres contain the proper wave functions of all exciton states in a given model. The notion of covariance and propagation in quantum space-times constituting such fibre bundles is investigated. Maxwell and Yang-Mills gauge degrees of freedom are introduced by appropriately enlarging the structure group, which in all cases contains phase-space representations of the Poincare group corresponding to the exciton wave function sample space specific to a given model. It is shown that these formulations give rise in a natural manner to certain realizations of the relativistic canonical commutation relations in terms of covariant derivatives involving internal as well as external degrees of freedom of space-time excitons
Heterotic String Compactification and New Vector Bundles
Lin, Hai; Wu, Baosen; Yau, Shing-Tung
2016-07-01
We propose a construction of Kähler and non-Kähler Calabi-Yau manifolds by branched double covers of twistor spaces. In this construction we use the twistor spaces of four-manifolds with self-dual conformal structures, with the examples of connected sum of n {mathbb{P}2}s. We also construct K3-fibered Calabi-Yau manifolds from the branched double covers of the blow-ups of the twistor spaces. These manifolds can be used in heterotic string compactifications to four dimensions. We also construct stable and polystable vector bundles. Some classes of these vector bundles can give rise to supersymmetric grand unified models with three generations of quarks and leptons in four dimensions.
Client Provider Collaboration for Service Bundling
Directory of Open Access Journals (Sweden)
LETIA, I. A.
2008-04-01
Full Text Available The key requirement for a service industry organization to reach competitive advantages through product diversification is the existence of a well defined method for building service bundles. Based on the idea that the quality of a service or its value is given by the difference between expectations and perceptions, we draw the main components of a frame that aims to support the client and the provider agent in an active collaboration meant to co-create service bundles. Following e3-value model, we structure the supporting knowledge around the relation between needs and satisfying services. We deal with different perspectives about quality through an ontological extension of Value Based Argumentation. The dialog between the client and the provider takes the form of a persuasion whose dynamic object is the current best configuration. Our approach for building service packages is a demand driven approach, allowing progressive disclosure of private knowledge.
Radiological evidence for the triple bundle anterior cruciate ligament.
MacKay, James W; Whitehead, Harry; Toms, Andoni P
2014-10-01
The anterior cruciate ligament (ACL) has traditionally been described as having two bundles--one anteromedial and one posterolateral. This has been challenged by studies proposing the existence of a third, intermediate, bundle with distinct functional significance, an arrangement that has been described in a number of domesticated animal species. No radiological evidence for the triple bundle ACL has previously been described. A prevalence study was carried out on 73 consecutive human knee magnetic resonance (MR) studies to determine the number of visible bundles, excluding individuals with a history of ACL injury or mucoid degeneration. A triple bundle ACL was demonstrated in 15 out of 73 human knees (20.5%, 95% confidence interval 12.9-31.2%). This is the first radiological description of the human triple bundle ACL. There was MR imaging evidence of a triple bundle ACL in approximately one fifth of human knees in this study. PMID:24890455
Transient pseudohypoaldosteronism
Directory of Open Access Journals (Sweden)
Stajić Nataša
2011-01-01
Full Text Available Introduction. Infants with urinary tract malformations (UTM presenting with urinary tract infection (UTI are prone to develop transient type 1 pseudohypoaldosteronism (THPA1. Objective. Report on patient series with characteristics of THPA1, UTM and/or UTI and suggestions for the diagnosis and therapy. Methods. Patients underwent blood and urine electrolyte and acid-base analysis, serum aldosterosterone levels and plasma rennin activity measuring; urinalysis, urinoculture and renal ultrasound were done and medical and/or surgical therapy was instituted. Results. Hyponatraemia (120.9±5.8 mmol/L, hyperkalaemia (6.9±0.9 mmol/L, metabolic acidosis (plasma bicarbonate, 11±1.4 mmol/L, and a rise in serum creatinine levels (145±101 μmol/L were associated with inappropriately high urinary sodium (51.3±17.5 mmol/L and low potassium (14.1±5.9 mmol/L excretion. Elevated plasma aldosterone concentrations (170.4±100.5 ng/dL and the very high levels of the plasma aldosterone to potassium ratio (25.2±15.6 together with diminished urinary K/Na values (0.31±0.19 indicated tubular resistance to aldosterone. After institution of appropriate medical and/or surgical therapy, serum electrolytes, creatinine, and acid-base balance were normalized. Imaging studies showed ureteropyelic or ureterovesical junction obstruction in 3 and 2 patients, respectively, posterior urethral valves in 3, and normal UT in 1 patient. According to our knowledge, this is the first report on THPA1 in the Serbian literature. Conclusion. Male infants with hyponatraemia, hyperkalaemia and metabolic acidosis have to have their urine examined and the renal ultrasound has to be done in order to avoid both, the underdiagnosis of THPA1 and the inappropriate medication.
RAP-2A Computer code for transients analysis in fast reactors
International Nuclear Information System (INIS)
The RAP-2A computer code is designed for analyzing thermohydraulic transients and/or steady state problems for large LMFBR cores. Physical and mathematical models, main input-output data, the flow chart of the code and a sample problem are given. RAP-2A calculates the power and the thermoydraulic transients initiated by a flow or reactivity changes, from a normal operating state of the reactor up to core disassembly. In this analysis a representative fuel pin is considered: a one-group space-independent (point) kinetics model to describe the neutron kinetics and a one-dimensional model describing the heat transfer (radial in the fuel and axial in the coolant) are used. Mechanical deformations due to temperature gradient, pressure losses, fuel melting, etc., are also calculated. The code is written in FORTRAN-4 language and is running on a IBM-370/135 computer
CHF and flow instability in rod bundles
International Nuclear Information System (INIS)
Data for two very different rod bundles have been analyzed using a new CHF correlation and a crude, but simple, subchannel analysis. The CHF correlation was developed for round uniform tubes and has been shown to accurately predict CHF in nonuniform tubes. The first set of data was for a KWU rod bundle (37 rods) with a heated length of 3.00 m and an O.D. (outside diameter) of 12.9 mm over a range of pressure 70 to 150 bar in upflow. The second set of data was for a 5 x 5 TRIGA rod bundle with a heated length of 0.559 m and 13.75 mm O.D. over a range of pressure of 0.945 to 1.372 bar in downflow. In contrast to the KWU data, the correlation greatly over estimates the CHF values for the TRIGA data. The TRIGA CHF data correlate very well with the variable qsat assuming no mixing, qc,exp = 0.955qsat (stdev = 9.87%). This result strongly suggests that these instabilities, which resulted immediately in CHF, are triggered by the Onset of Flow Instability (OFI) rather than CHF. The wide spread in rod power factors, the low pressure, and the downflow condition all contribute to promoting this type of instability (Ledinegg). The crude subchannel analysis has been compared with calculations of exit conditions of the hot channel using COBRA code. The agreement is fair when the homogeneous equilibrium model is used in the COBRA code. This is expected since the exit of the hot channel is always subcooled. Using Zuber's, along with other, void fraction relations in COBRA yields much lower exit velocities and high positive exit qualities, and, in some cases, convergence difficulties arise. The facts indicate that the bundle has already past the OFI point: which is possible since no CHF calculation was made in these COBRA analyses. (J.P.N)
Interstitial He and Ne in Nanotube Bundles
Stan, G.; Crespi, V. H.; Cole, M. W.; Boninsegni, M.
1998-01-01
We explore the properties of atoms confined to the interstitial regions within a carbon nanotube bundle. We find that He and Ne atoms are of ideal size for physisorption interactions, so that their binding energies are much greater there than on planar surfaces of any known material. Hence high density phases exist at even small vapor pressure. There can result extraordinary anisotropic liquids or crystalline phases, depending on the magnitude of the corrugation within the interstitial channels.
Effective freeness of adjoint line bundles
Heier, Gordon
2001-01-01
In this note we establish a new Fujita-type effective bound for the base point freeness of adjoint line bundles on a compact complex projective manifold of complex dimension $n$. The bound we obtain (approximately) differs from the linear bound conjectured by Fujita only by a factor of the cube root of $n$. As an application, a new effective statement for pluricanonical embeddings is derived.
On Complex Supermanifolds with Trivial Canonical Bundle
Groeger, Josua
2016-01-01
We give an algebraic characterisation for the triviality of the canonical bundle of a complex supermanifold in terms of a certain Batalin-Vilkovisky superalgebra structure. As an application, we study the Calabi-Yau case, in which an explicit formula in terms of the Levi-Civita connection is achieved. Our methods include the use of complex integral forms and the recently developed theory of superholonomy.
Telescope sipping - pinpointing leaking fuel bundles
International Nuclear Information System (INIS)
Given the top priority operators of nuclear power plants assign to safety, even the slightest sign of damage to the fuel assemblies has to be carefully monitored and analyzed. The detection of leaking fuel bundles also plays an important role in ensuring good availability and economy for the plants. ABB Atom has developed a new, highly accurate method, called 'telescope sipping', for identifying defective fuel assemblies. (orig.)
Imperfect Bundling In Public-Private Partnerships
Luciano Greco
2012-01-01
The economic literature on PPPs has generally overlooked agency problems within private consortia. We provide a first contribution in this direction, relying on a simple incomplete contracts framework where a Builder and an Operator set up a Special Purpose Vehicle (SPV) to carry out a contract with the government. Because of incomplete contracts, the bundling of tasks is imperfect, and the SPV ownership structure is the main tool to regulate the power of private incentives. The scope for wel...
Using Advanced Fuel Bundles in CANDU Reactors
International Nuclear Information System (INIS)
Improving the exit fuel burnup in CANDU reactors was a long-time challenge for both bundle designers and performance analysts. Therefore, the 43-element design together with several fuel compositions was studied, in the aim of assessing new reliable, economic and proliferation-resistant solutions. Recovered Uranium (RU) fuel is intended to be used in CANDU reactors, given the important amount of slightly enriched Uranium (~0.96% w/o U235) that might be provided by the spent LWR fuel recovery plants. Though this fuel has a far too small U235 enrichment to be used in LWR's, it can be still used to fuel CANDU reactors. Plutonium based mixtures are also considered, with both natural and depleted Uranium, either for peacefully using the military grade dispositioned Plutonium or for better using Plutonium from LWR reprocessing plants. The proposed Thorium-LEU mixtures are intended to reduce the Uranium consumption per produced MW. The positive void reactivity is a major concern of any CANDU safety assessment, therefore reducing it was also a task for the present analysis. Using the 43-element bundle with a certain amount of burnable poison (e.g. Dysprosium) dissolved in the 8 innermost elements may lead to significantly reducing the void reactivity. The expected outcomes of these design improvements are: higher exit burnup, smooth/uniform radial bundle power distribution and reduced void reactivity. Since the improved fuel bundles are intended to be loaded in existing CANDU reactors, we found interesting to estimate the local reactivity effects of a mechanical control absorber (MCA) on the surrounding fuel cells. Cell parameters and neutron flux distributions, as well as macroscopic cross-sections were estimated using the transport code DRAGON and a 172-group updated nuclear data library. (author)
Uncovering ecosystem service bundles through social preferences.
Directory of Open Access Journals (Sweden)
Berta Martín-López
Full Text Available Ecosystem service assessments have increasingly been used to support environmental management policies, mainly based on biophysical and economic indicators. However, few studies have coped with the social-cultural dimension of ecosystem services, despite being considered a research priority. We examined how ecosystem service bundles and trade-offs emerge from diverging social preferences toward ecosystem services delivered by various types of ecosystems in Spain. We conducted 3,379 direct face-to-face questionnaires in eight different case study sites from 2007 to 2011. Overall, 90.5% of the sampled population recognized the ecosystem's capacity to deliver services. Formal studies, environmental behavior, and gender variables influenced the probability of people recognizing the ecosystem's capacity to provide services. The ecosystem services most frequently perceived by people were regulating services; of those, air purification held the greatest importance. However, statistical analysis showed that socio-cultural factors and the conservation management strategy of ecosystems (i.e., National Park, Natural Park, or a non-protected area have an effect on social preferences toward ecosystem services. Ecosystem service trade-offs and bundles were identified by analyzing social preferences through multivariate analysis (redundancy analysis and hierarchical cluster analysis. We found a clear trade-off among provisioning services (and recreational hunting versus regulating services and almost all cultural services. We identified three ecosystem service bundles associated with the conservation management strategy and the rural-urban gradient. We conclude that socio-cultural preferences toward ecosystem services can serve as a tool to identify relevant services for people, the factors underlying these social preferences, and emerging ecosystem service bundles and trade-offs.
Noncommutative line bundle and Morita equivalence
Jurco, Branislav; Schupp, Peter; Wess, Julius
2001-01-01
Global properties of abelian noncommutative gauge theories based on $\\star$-products which are deformation quantizations of arbitrary Poisson structures are studied. The consistency condition for finite noncommutative gauge transformations and its explicit solution in the abelian case are given. It is shown that the local existence of invertible covariantizing maps (which are closely related to the Seiberg-Witten map) leads naturally to the notion of a noncommutative line bundle with noncommu...
Bundling harvester; Harvennuspuun automaattisen nippukorjausharvesterin kehittaeminen
Energy Technology Data Exchange (ETDEWEB)
Koponen, K. [Eko-Log Oy, Kuopio (Finland)
1997-12-01
The starting point of the project was to design and construct, by taking the silvicultural point of view into account, a harvesting and processing system especially for energy-wood, containing manually driven bundling harvester, automating of the harvester, and automated loading. The equipment forms an ideal method for entrepreneur`s-line harvesting. The target is to apply the system also for owner`s-line harvesting. The profitability of the system promotes the utilisation of the system in both cases. The objectives of the project were: to construct a test equipment and prototypes for all the project stages, to carry out terrain and strain tests in order to examine the usability and durability, as well as the capacity of the machine, to test the applicability of the Eko-Log system in simultaneous harvesting of energy and pulp woods, and to start the marketing and manufacturing of the products. The basic problems of the construction of the bundling harvester have been solved using terrain-tests. The prototype machine has been shown to be operable. Loading of the bundles to form sufficiently economically transportable loads has been studied, and simultaneously, the branch-biomass has been tried to be utilised without loosing the profitability of transportation. The results have been promising, and will promote the profitable utilisation of wood-energy. (orig.)
Nuclear reactor control bundle guide system
International Nuclear Information System (INIS)
Each bundle is formed by several absorbent rods, which are vertically movable and are connected together by a spider to a common axial operating rod, and guide means for the control bundles in their displacement, out of the core; the said means comprise guide boxes containing horizontal plates for discontinuous guiding, at the upper part of the boxes, of absorbent rods positioned in pairs on a radius and individual peripheral absorbent rods of the control bundle. At the lower part of the boxes in a continuous guiding zone, guiding of the absorbent rods positioned in pairs on a radius is effected by association of the horizontal plates for mechanical guiding of the rods, with housings which minimise hydraulic effects by smoothing the coolant flow in the radial direction around the absorbent rods. The hydraulic housings are mounted between the horizontal plates as discontinuous spacers. Pressure differences around each rod are minimised or eliminated and continuous guiding is achieved without affecting the design of the guide boxes, the internal equipment or the pressure vessel. The invention can be applied to PWRs
Energy Technology Data Exchange (ETDEWEB)
Gamez, Abel; Rojas, Leorlen; Rosales, Jesus; Castro, Landy Y.; Gonzalez, Daniel; Garcia, Carlos, E-mail: agamezgmf@gmail.com, E-mail: leored1984@gmail.com, E-mail: jrosales@instec.cu, E-mail: lcastro@instec.cu, E-mail: danielgonro@gmail.com, E-mail: cgr@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba); Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Oliveira, Carlos B. de, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Dominguez, Dany S., E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil)
2015-07-01
The high temperature gas cooled reactor (HTGR) is one of candidates of next generation of nuclear reactor according to IAEA report 2013. Evaluation of thermohydraulic performance and an experimental comparison results were proposed to the international research community. In this article, the tree dimensional CFD thermohydraulic modelation of steady state of HTR-10 modular reactor, using ANSYS CFX v14.0, has been done. Code-to-code and Code-to-experiment benchmark analyses, related to the testing program of the HTR-10 plant including steady state temperature distribution with the reactor at full power, were developed. The 3D real scale representation of reflector zone and fluid path flow inner and outer reflector blocks and cold helium cavity were carried out. The porous medium model was used to simulate the core zone in the reactor. The power distribution of the initial core published by IAEA-TECDOC-1694 obtained by Chief Scientific Investigators (CSIs) from China was used as heat sources in the core zone. (author)
International Nuclear Information System (INIS)
One of the main issues related to the conception of a superconducting magnet cooled by a superfluid helium bath (like the Iseult magnet) is to insure the magnet safety as well as the whole cryogenic facility safety in case of accidental quench. In order to find a solution to this problem, we first have to identify the physical mechanisms which drive the pressure rise during a quench. This is why our study deals with the modeling of the thermohydraulic phenomena taking place during such a magnet quench. First of all, we performed and analyzed local pressure rise experiments in a heated helium channel. A numerical thermohydraulic model was developed for this study. Quench experiments were then performed on an 8-T (Seht) superconducting coil cooled by a superfluid helium bath. These experiments allowed us to make a detailed analysis of the physical mechanisms which drive the global pressure rise in case of quench as well as the strong coupling between this pressure rise and the normal zone propagation. Following this analysis, a complete model of normal zone propagation and pressure rising during a quench was developed. This model is a first step toward predictive modeling of the pressure rise during the quench of a superconducting magnet cooled by a superfluid helium bath. (author)
International Nuclear Information System (INIS)
It has been recognized that nuclear power plants can substitute for diesel engines as the main propulsion engines for merchant ships. It is necessary to steadily accumulate and foster the technologies and knowledges of marine reactors so as to deal with their necessity in future. In this research, the synthetic analysis and evaluation of the operational characteristics under the condition of hull motion and the safety at the time of accidents of one-body type marine reactor are carried out by using the nuclear powered ship engineering simulation system developed by Japan Atomic Energy Research Institute. The research on the safety of the nuclear powered ship, on which one-body type marine reactor is installed, the research on the effect that pulsating flow exerts to the thermo-hydraulic characteristics of marine reactor, and the research on the thermo-hydraulic behavior in containment vessel at the time of accidents in passive safety type marine reactor have been carried out. The outline of these researches and the activities in fiscal year 1995 are reported. The plan for hereafter to advance this research for contributing to the improvement of the marine reactors of next generation is discussed. (K.I.)
International Nuclear Information System (INIS)
LILAC-meltpool has been developed to study thermo-hydraulic behavior of molten pool and thermal behavior of vessel wall during severe accident. To validate LILAC-meltpool code several two and three dimensional thermo-hydraulic problems were selected and solved. The benchmark problems have experimental results or verified numerical results. Through the validation it was found that LILAC-meltpool reproduces very accurate numerical results. Two-layered semicircular pool was solved to study thermal and hydraulic characteristics of pool stratification. The LAVA experiment using alumina/ferrite molten pool was calculated and compared with computed results. Cooling of alumina/ferrite two-layered pool was affected by stratification. In the numerical results temperature of vessel inner was highest at a location below the interface. Crust was developed from upper surface and lower outer surface, but in the area near the interface corium simulant existed as molten state for long time. LAVA-4 experiment was studied using gap-cooling model in LILAC-meltpool code. Temperature increase of LAVA vessel after alumina melt relocation was strongly dependent on gap formation mechanism. Calculated cooling rates of the vessel were very similar to experimental results. For LAVA experiments which do not have heat generation coolant penetrates easily into a gap and it is found that gap-cooling is very effective for cooling of vessel, but it is thought that coolant penetration could be limited near upper part of gap because of decay heat and high temperature of corium crust
International Nuclear Information System (INIS)
This report is a summary of the work performed under a co-ordinated research project (CRP) entitled Harmonization and Validation of Fast Reactor Thermomechanical and Thermo-Hydraulic Codes and Relations using Experimental Data. The project was organized by the IAEA on the recommendation of the IAEA's Technical Working Group on Fast Reactors (TWGFR) and carried out from 1996 to 1999. In certain conditions, temperature fluctuations in the coolant close to a structure caused by thermal striping can lead to thermomechanical damage to structures. Institutes from a number of Member States have an interest in improving engineering tools and prediction techniques concerning the characterization of the thermal striping effects, in which numerical models have a major role. Therefore, the IAEA through its advanced reactor technology development programme supports the activities of Member States in this area. Design analyses applied to thermal striping phenomena need to be firmly established, and the CRP provided a valuable tool in assessing their reliability. Eleven institutes from France, India, Italy, Japan, the Republic of Korea, the Russian Federation and the United Kingdom co-operated in this CRP. This report documents the CRP activities, provides the main results and recommendations and includes the work carried out by the research groups at the participating institutes within the CRP on harmonization and validation of fast reactor thermomechanical and thermohydraulic codes and relations
LMFR core and heat exchanger thermohydraulic design: former USSR and present Russian approaches
International Nuclear Information System (INIS)
The information presented in this report is dealing with liquid metal cooled fast reactors some of which are in operation (France, Japan, Russian federation) or under construction. Comprehensive thermal hydraulic research both experimental and numeric applied to such reactors was carried out in the Institute of Physics and Power Engineering (IPPE), Obninsk, Russian Federation. The IAEA Working Group on fast Reactors (IWGFR) recommended that IPPE should generalize its thermal hydraulic studies as well as results of other countries published previously in the field of liquid metal flow distribution and heat transfer in fuel pin and heat exchanger rod bundles (France, Germany, Japan, India, Russian Federation, United Kingdom and United States). The validity of computer codes and design approaches was proven by comparison of calculated results with measured values of velocity, pressure, temperature distributions in rod bundles cooled/heated by liquid metal, usually sodium. The report includes the methodology and philosophy of the analytical and experimental investigations when applied to core and heat exchanger thermal hydraulic design of Light Water Moderated Fast Reactors (LMFRs)
Thermohydraulic and mechanical analysis of the research reactor Munich II Compact-Core
International Nuclear Information System (INIS)
The new research reactor Munich II (Forschungsreaktor Muenchen II, FRM-II), which is under construction at the Technical University of Munich, Germany, contains a compact reactor core consisting of one single fuel element, assembled by two concentric tubes between which 113 involutely bent fuel plates are located rotationally symmetric. In order to perform the hydraulic and mechanical testing of the FRM-II fuel element, two test facilities have been built at the Department for Nuclear and New Energy Systems of the Ruhr University Bochum. The first mocks up the central region of the reactor coolant system of the FRM-II in a 1:1 scale with emphasis on the fuel element and the inflow and discharge section in order to enable the analysis of the FRM-II core. In the course of the testing the vibration behaviour and the flow resistance of the core were investigated. Likewise start-up and shut down tests of the main pump unit were simulated and the flow profile at the outlet of the element as well as the flow division inside the core were determined. Furthermore an endurance test lasting 60 days (equivalent to 12 operating cycles) was performed, too. Tests including blockages of parts of the reactor cooling system cross section at the core entrance sieve proved the efficiency of the cooling capacity. No major resonances occurred during operation and an endurance test neither showed any incidents nor irregularities. In order to investigate the concept of the decay heat removal in the FRM-II a second test facility was built. This facility simulates the thermohydraulic conditions in one cooling channel of the FRM-II by means of an electrically heated test section, which enables different operating conditions of the decay heat removal system as well as enhanced safety investigations. In the FRM-II the decay heat, which is produced after a shutdown, is removed by means of decay heat removal pumps, which maintain a downward flow in the fuel element for at least three hours
International Nuclear Information System (INIS)
An accurate subchannel database is crucial for modeling the multidimensional two-phase flow in a rod bundle and for validating subchannel analysis codes. Based on available reference, it can be said that a point-measurement sensor for acquiring void fractions and bubble velocity distributions do not infer interactions of the subchannel flow dynamics, such as a cross flow and flow distribution, etc. In order to acquire multidimensional two-phase flow in a 10×10 rod bundle with an o.d. of 10 mm and 3110 mm length, a new sensor consisting of 11-wire by 11-wire and 10-rod by 10-rod electrodes was developed. Electric potential in the proximity region between two wires creates a void fraction in the center subchannel region, like a so-called wire mesh sensor. A unique aspect of the devised sensor is that the void fraction near the rod surface can be estimated from the electric potential in the proximity region between one wire and one rod. The additional 400 points of void fraction and phasic velocity in 10×10 bundle can therefore be acquired. The devised sensor exhibits the quasi three-dimensional flow structures, i.e. void fraction, phasic velocity and bubble chord length distributions. These quasi three-dimensional structures exhibit the complexity of two-phase flow dynamics, such as coalescence and the breakup of bubbles in transient phasic velocity distributions. (author)
[Experience with ablation of the bundle of His using electrical discharge].
Lukl, J; Cernosek, B; Heinc, P
1990-01-01
Ablation of the bundle of His by an electric discharge was made in 10 patients (average age 64 years, range 48-80) unsuccessfully treated with 3-14 antiarrhythmic drugs or their combinations (average 8.7) on account of supraventricular tachycardia occurring repeatedly for 3-44 years (average 12.5 years). By means of a bipolar electrode inserted into the area of the bundle of His a nonsynchronized defibrillation discharge with a mean energy of 323 J (40-380) was administered. On average 2.9 discharges were used (1-9) per patient, in 1-3 sessions. The patients were followed up for a period of 15.3 months (10-19). Permanent complete a-v block was achieved in 6 patients, the remaining 4 patients are also free from complaints with have antiarrhythmic treatment. To all patients a pacemaker was implanted, four times "physiological" stimulation was used. One month after the operation the authors observed once the development of a transient phatic disorder and once the slow development of cardiac tamponade in conjunction with anticoagulant treatment, resolved by pericardial puncture. Ablation of the bundle of His by a defibrillation discharge is thus in carefully selected patients a highly effective method of treatment of stubborn supraventricular tachycardias. PMID:2327082
Adsorption of Argon on Carbon nanotube bundles and its influence on the bundle lattice parameter
International Nuclear Information System (INIS)
We report experimental studies of the adsorption characteristics and structure of both Ar36 and Ar40 on single-wall carbon nanotube bundles. The structural studies make use of the large difference in coherent neutron scattering cross section for the two Ar isotopes to explore the influence of the adsorbate on the nanotube lattice parameter. We observe no dilation of the nanotube lattice with Ar40, and explain the apparent expansion of this lattice upon Ar36 adsorption by the location of the adsorbed Ar atoms on the outer bundle surface
CANDU fuel behaviour under transient conditions
International Nuclear Information System (INIS)
The Canadian R and D program to understand CANDU fuel behaviour under transient conditions is described. Fuel sheath behaviour studies have led to the development of a model of transient plastic strain in inert gas, which integrates the deformation due to several mechanisms. Verification tests demonstrated that on average the model overpredicts strain by 20%. From oxidation kinetics studies a sheath failure embrittlement criterion based on oxygen distribution has been developed. We have also established a rate equation for high-temperature stress-dependent crack formation due to embrittlement of the sheath by beryllium. An electric, simulated fuel element is being used in laboratory tests to characterize the behaviour of fuel in the horizontal. In-reactor, post-dryout tests have been done for several years. There is an axially-segmented, axisymmetric fuel element model in place and a fully two-dimensional code is under development. Laboratory testing of bundles, in its early stages, deals with the effects of geometric distortion and sheath-to-sheath interaction. In-reactor, post-dryout tests of CANDU fuel bundles with extensive central UO2 melting did not result in fuel fragmentation nor damage to the pressure tube. (author)
Modeling the prototype repository project: sensitivity analysis of thermo-hydraulic behavior
International Nuclear Information System (INIS)
general objective of the Prototype Repository Project is to demonstrate that the important processes taking place in the engineered barriers and the host rock are sufficiently well understood. A model for investigating the thermal evolution of the Prototype Repository was developed by Kristensson and Hoekmark (2007). More recently, hydraulic models for the repository have been in preparation as part of an EBS Task Force assignment initiated by SKB. The assignment was divided into three steps: First, the hydraulic evolution was evaluated for the pre-installation model containing only host rock with the excavated deposition tunnel and holes. Second, the buffer, backfill and plug were installed forming the post-installation hydraulic model, which also included relevant hydraulic events observed during the operational stage of the Prototype Repository experiment. The third step in the assignment will be to implement boundary conditions from the second, post-installation, step thermal and hydraulic models, into a single deposition hole model and as a final objective perform a complete THM analysis at this local scale. The work described here concentrates on investigating the hydraulic behavior of both the first step pre-installation model and the second step post-installation model through sensitivity analysis of the parameters and constitutive laws for the rock, buffer and backfill materials. Additionally, sensitivity analysis has been performed regarding the boundary conditions and some physical processes for the post-installation model, and the influence of changes in geometry with respect to nearby tunnels and model boundaries has been examined. A coupled thermo-hydraulic analysis for the post-installation stage has also been considered. The analyses have been performed with the finite element code Code-Bright. An example of the influence of rock intrinsic permeability on the deposition hole inflows and liquid pressure can be seen in Figure 1. Based on these analyses, key
International Nuclear Information System (INIS)
When performing transient analysis in heterogeneous nuclear reactors loaded with different types of fuel bundles is necessary to model the reactor core by a few representative fuel elements with average properties of a region containing a large number of fuel elements. The properties of these representative fuel bundles are obtained by averaging the thermal-hydraulic properties of the fuel elements contained in each region. In this paper we study the different ways to perform the averaging of the thermal-hydraulic properties that can have an influence on the transient results for licence purposes. Also we study the influence of the different averaging methods on the peak clad temperature (PCT) evolution for a LOCA, and on the critical power ratio (CPR) in the hot channels for a turbine trip transient without bypass credit.
International Nuclear Information System (INIS)
The COBRA-IV-I computer code uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV-I as the implicit solution scheme option. In addition to these techniques, a new explicit solution scheme is now available which allows the calculation of severe transients involving flow reversals, recirculations, expulsion and reentry flows, with a pressure or flow boundary condition specified. Significant storage compaction and reduced running times have been achieved to allow the calculation of problems involving hundreds of subchannels
Amplitude death of coupled hair bundles with stochastic channel noise
Kim, Kyung-Joong
2014-01-01
Hair cells conduct auditory transduction in vertebrates. In lower vertebrates such as frogs and turtles, due to the active mechanism in hair cells, hair bundles(stereocilia) can be spontaneously oscillating or quiescent. Recently, the amplitude death phenomenon has been proposed [K.-H. Ahn, J. R. Soc. Interface, {\\bf 10}, 20130525 (2013)] as a mechanism for auditory transduction in frog hair-cell bundles, where sudden cessation of the oscillations arises due to the coupling between non-identical hair bundles. The gating of the ion channel is intrinsically stochastic due to the stochastic nature of the configuration change of the channel. The strength of the noise due to the channel gating can be comparable to the thermal Brownian noise of hair bundles. Thus, we perform stochastic simulations of the elastically coupled hair bundles. In spite of stray noisy fluctuations due to its stochastic dynamics, our simulation shows the transition from collective oscillation to amplitude death as inter-bundle coupling str...
A Tannakian approach to dimensional reduction of principal bundles
Álvarez-Cónsul, Luis; García-Prada, Oscar
2016-01-01
Let $P$ be a parabolic subgroup of a connected simply connected complex semisimple Lie group $G$. Given a compact K\\"ahler manifold $X$, the dimensional reduction of $G$-equivariant holomorphic vector bundles over $X\\times G/P$ was carried out by the first and third authors. This raises the question of dimensional reduction of holomorphic principal bundles over $X\\times G/P$. The method used for equivariant vector bundles does not generalize to principal bundles. In this paper, we adapt to equivariant principal bundles the Tannakian approach of Nori, to describe the dimensional reduction of $G$-equivariant principal bundles over $X\\times G/P$, and to establish a Hitchin--Kobayashi type correspondence. In order to be able to apply the Tannakian theory, we need to assume that $X$ is a complex projective manifold.
Hydrodynamic behavior of a bare rod bundle
International Nuclear Information System (INIS)
The temperature distribution within the rod bundle of a nuclear reactor is of major importance in nuclear reactor design. However temperature information presupposes knowledge of the hydrodynamic behavior of the coolant which is the most difficult part of the problem due to complexity of the turbulence phenomena. In the present work a 2-equation turbulence model--a strong candidate for analyzing actual three dimensional turbulent flows--has been used to predict fully developed flow of infinite bare rod bundle of various aspect ratios (P/D). The model has been modified to take into account anisotropic effects of eddy viscosity. Secondary flow calculations have been also performed although the model seems to be too rough to predict the secondary flow correctly. Heat transfer calculations have been performed to confirm the importance of anisotropic viscosity in temperature predictions. All numerical calculations for flow and heat have been performed by two computer codes based on the TEACH code. Experimental measurements of the distribution of axial velocity, turbulent axial velocity, turbulent kinetic energy and radial Reynolds stresses were performed in the developing and fully developed regions. A 2-channel Laser Doppler Anemometer working on the Reference mode with forward scattering was used to perform the measurements in a simulated interior subchannel of a triangular rod array with P/D = 1.124. Comparisons between the analytical results and the results of this experiment as well as other experimental data in rod bundle array available in literature are presented. The predictions are in good agreement with the results for the high Reynolds numbers
Historical dynamics in ecosystem service bundles.
Renard, Delphine; Rhemtulla, Jeanine M; Bennett, Elena M
2015-10-27
Managing multiple ecosystem services (ES), including addressing trade-offs between services and preventing ecological surprises, is among the most pressing areas for sustainability research. These challenges require ES research to go beyond the currently common approach of snapshot studies limited to one or two services at a single point in time. We used a spatiotemporal approach to examine changes in nine ES and their relationships from 1971 to 2006 across 131 municipalities in a mixed-use landscape in Quebec, Canada. We show how an approach that incorporates time and space can improve our understanding of ES dynamics. We found an increase in the provision of most services through time; however, provision of ES was not uniformly enhanced at all locations. Instead, each municipality specialized in providing a bundle (set of positively correlated ES) dominated by just a few services. The trajectory of bundle formation was related to changes in agricultural policy and global trends; local biophysical and socioeconomic characteristics explained the bundles' increasing spatial clustering. Relationships between services varied through time, with some provisioning and cultural services shifting from a trade-off or no relationship in 1971 to an apparent synergistic relationship by 2006. By implementing a spatiotemporal perspective on multiple services, we provide clear evidence of the dynamic nature of ES interactions and contribute to identifying processes and drivers behind these changing relationships. Our study raises questions about using snapshots of ES provision at a single point in time to build our understanding of ES relationships in complex and dynamic social-ecological systems. PMID:26460005
Numerical simulations of square arrayed rod bundles
International Nuclear Information System (INIS)
Highlights: ► CFD simulations with square arrayed rod bundles. ► Mesh dependency and turbulence model study by comparison with experiments. ► Gibson and Launder Reynolds stress model shows good agreement with experiments. ► Effect of pitch to diameter ratio and Reynolds number is correctly captured. - Abstract: Computational fluid dynamics (CFD) simulations were performed with square arrayed rod bundles featuring pitch to diameter (P/D) ratio of 1.194 and 1.326 in order to find an optimal mesh and turbulence model for simulations with more complex geometries in the future. With the tighter lattice a mesh sensitivity and turbulence model study were accomplished and the post processed turbulence quantities, velocity field and wall shear stress were compared with experimental data ( Developed single phase turbulent flow through a square-pitch rod cluster. Nuclear Engineering and Design 60, 365–379.). The comparisons show that Reynolds-Averaged Navier–Stokes method with the Reynolds stress model of Gibson and Launder in conjunction with an appropriate mesh can provide reasonable agreement with the experiment for this lattice. For pure bundle simulations the body fitted structured meshes are suggested, since slightly better agreement can be captured considering all quantities with the same number of cells. Based on the drawn conclusions the procession was repeated for P/D = 1.326, where, due to lack of experiment, just the correct tendencies of the turbulence quantities and velocity field were established. The results show Reynolds number independency correctly and the increase of P/D issues in more similar flow to axisymmetric pipe flow.
Systematic Bundle Adjustment of HRSC Image Data
Bostelmann, J.; Schmidt, R.; Heipke, C.
2012-07-01
The European Mars Express mission was launched in June 2003 and sent into orbit around Mars. On board the orbiter is the German High Resolution Stereo Camera (HRSC). This multi-line sensor images the Martian surface with a resolution of up to 12m per pixel in three dimensions and provides RGB and infra-red color information. The usage of the stereoscopic image information for the improvement of the observed position and attitude information via bundle adjustment is important to derive high quality 3D surface models, color orthoimages and other data products. In many cases overlapping image strips of different orbits can be used to form photogrammetric blocks, thus allowing the simultaneous adjustment of the exterior orientation data. This reduces not only local, but also regional inconsistencies in the data. With the growing number of HRSC image strips in this ongoing mission, the size and complexity of potential blocks is increasing. Therefore, a workflow has been built up for the systematic improvement of the exterior orientation using single orbit strips and regional blocks. For a successful bundle adjustment of blocks using multiple image strips a sufficient number of tie points in the overlapping area is needed. The number of tie points depends mainly on the geometric and radiometric quality of the images. This is considered by detailed analysis of the tie point accuracy and distribution. The combination of methods for image pre-processing, tie point matching, bundle adjustment and evaluation of the results in an automated workflow allows for all HRSC images a global assessment of the quality and a systematic selection of data for larger blocks.
Current interruption transients calculation
Peelo, David F
2014-01-01
Provides an original, detailed and practical description of current interruption transients, origins, and the circuits involved, and how they can be calculated Current Interruption Transients Calculationis a comprehensive resource for the understanding, calculation and analysis of the transient recovery voltages (TRVs) and related re-ignition or re-striking transients associated with fault current interruption and the switching of inductive and capacitive load currents in circuits. This book provides an original, detailed and practical description of current interruption transients, origins,
Numerical study of the thermo-hydraulic behavior for the Candu type fuel channel
International Nuclear Information System (INIS)
Candu type reactors use fuel channel in a horizontal lattice. The fuel bundles are positioned in two Zircaloy tubes: the pressure tube surrounded by calandria tube. Inside the pressure tube the coolant heavy water flows. The coolant reaches high temperatures and pressures. Due to irregular neutron spatial distribution, the fuel channel stress differs from one channel to other. In one improbable event of severe accident, the fuel channel behaves differently according to its normal function history. Over the years, there have been many research projects trying to analyze thermal hydraulic performance of the design and to add some operational improvements in order to achieve an efficient thermal hydraulic distribution. This paper discusses the thermo hydraulic behavior (influence of the temperature and velocity distribution) of the most solicited channel, simulated with Fluent 6.X. Code. Moreover it will be commented the results obtained using different models and mesh applied. (authors)
Review of thermohydraulic research of fuel assemblies with partial blocking of flow cross-section
International Nuclear Information System (INIS)
A review is presented of the theoretical and experimental investigation of blockage formation and of velocity and temperature fields in fuel rod bundles with partial blockage of the flow section. The temperature and velocity fields in cases of flow blockage are analyzed and the range of the recirculation zone length behind the blockage is shown. Formulas for the evaluation of the coolant flow rate changes and the temperature increments in dependence on the operating parameters and the blocked area are given. Questions of blockage identification and of prevention of emergency situations are discussed. Results of the analysis emphasize the necessity to continue research of blockage formation problems and of velocity and temperature conditions in blocked assemblies. (author)
Tiling spaces are Cantor set fiber bundles
Sadun, Lorenzo; Williams, R F
2001-01-01
We prove that fairly general spaces of tilings of R^d are fiber bundles over the torus T^d, with totally disconnected fiber. This was conjectured (in a weaker form) in [W3], and proved in certain cases. In fact, we show that each such space is homeomorphic to the d-fold suspension of a Z^d subshift (or equivalently, a tiling space whose tiles are marked unit d-cubes). The only restrictions on our tiling spaces are that 1) the tiles are assumed to be polygons (polyhedra if d>2) that meet full-...
Higher order mechanics on graded bundles
International Nuclear Information System (INIS)
In this paper we develop a geometric approach to higher order mechanics on graded bundles in both, the Lagrangian and Hamiltonian formalism, via the recently discovered weighted algebroids. We present the corresponding Tulczyjew triple for this higher order situation and derive in this framework the phase equations from an arbitrary (also singular) Lagrangian or Hamiltonian, as well as the Euler–Lagrange equations. As important examples, we geometrically derive the classical higher order Euler–Lagrange equations and analogous reduced equations for invariant higher order Lagrangians on Lie groupoids. (paper)
Assembly mechanism for nuclear fuel bundles
International Nuclear Information System (INIS)
The invention relates to a nuclear power reactor fuel bundle of the type wherein several rods are mounted in parallel array between two tie plates which secure the fuel rods in place and are maintained in assembled position by means of a number of tie rods secured to both of the end plates. Improved apparatus is provided for attaching the tie rods to the upper tie plate by the use of locking lugs fixed to rotatable sleeves which engage the upper tie plate. (auth)
Compression of a bundle of light rays.
Marcuse, D
1971-03-01
The performance of ray compression devices is discussed on the basis of a phase space treatment using Liouville's theorem. It is concluded that the area in phase space of the input bundle of rays is determined solely by the required compression ratio and possible limitations on the maximum ray angle at the output of the device. The efficiency of tapers and lenses as ray compressors is approximately equal. For linear tapers and lenses the input angle of the useful rays must not exceed the compression ratio. The performance of linear tapers and lenses is compared to a particular ray compressor using a graded refractive index distribution. PMID:20094478
Differential geometry of complex vector bundles
Kobayashi, Shoshichi
2014-01-01
Holomorphic vector bundles have become objects of interest not only to algebraic and differential geometers and complex analysts but also to low dimensional topologists and mathematical physicists working on gauge theory. This book, which grew out of the author's lectures and seminars in Berkeley and Japan, is written for researchers and graduate students in these various fields of mathematics. Originally published in 1987. The Princeton Legacy Library uses the latest print-on-demand technology to again make available previously out-of-print books from the distinguished backlist of Princeto
A Unified Framework for Quasi-Linear Bundle Adjustment
Bartoli, Adrien
2002-01-01
Obtaining 3D models from long image sequences is a major issue in computer vision. One of the main tools used to obtain accurate structure and motion estimates is bundle adjustment. Bundle adjustment is usually performed using nonlinear Newton-type optimizers such as Levenberg-Marquardt which might be quite slow when handling a large number of points or views. We investigate an algorithm for bundle adjustment based on quasi-linear optimization. The method is straightforward to implement and r...
Non-commutative P-1-bundles over commutative schemes
Van den Bergh, Michel
2012-01-01
In this paper we develop the theory of non-commutative P-1-bundles over commutative (smooth) schemes. Such non-commutative P-1-bundles occur in the theory of D-modules but our definition is more general. We can show that every non-commutative deformation of a Hirzebruch surface is given by a non-commutative P-1-bundle over P-1 in our sense.
Contacting single bundles of carbon nanotubes with alternating electric fields
Krupke, R.; Hennrich, F.; Weber, H. B.; Beckmann, D.; Hampe, O.; Malik, S.; Kappes, M. M.; Löhneysen, H. v.
2002-01-01
Single bundles of carbon nanotubes have been selectively deposited from suspensions onto sub-micron electrodes with alternating electric fields. We explore the resulting contacts using several solvents and delineate the differences between Au and Ag as electrode materials. Alignment of the bundles between electrodes occurs at frequencies above 1 kHz. Control over the number of trapped bundles is achieved by choosing an electrode material which interacts strongly with the chemical functional g...
Dark-field illuminated reflectance fiber bundle endoscopic microscope
Liu, Xuan; Huang, Yong; Kang, Jin U.
2011-01-01
We propose a reflectance fiber bundle microscope using a dark-field illumination configuration for applications in endoscopic medical imaging and diagnostics. Our experiment results show that dark-field illumination can effectively suppress strong specular reflection from the proximal end of the fiber bundle. We realized a lateral resolution of 4.4 μm using the dark-field illuminated fiber bundle configuration. To demonstrate the feasibility of using the system to study cell morphology, we ob...
A viscous two-phase model for contractile actomyosin bundles.
Oelz, Dietmar
2014-06-01
A mathematical model in one dimension for a non-sarcomeric actomyosin bundle featuring anti-parallel flows of anti-parallel F-actin is introduced. The model is able to relate these flows to the effect of cross-linking and bundling proteins, to the forces due to myosin-II filaments and to external forces at the extreme tips of the bundle. The modeling is based on a coarse graining approach starting with a microscopic model which includes the description of chemical bonds as elastic springs and the force contribution of myosin filaments. In a second step we consider the asymptotic regime where the filament lengths are small compared to the overall bundle length and restrict to the lowest order contributions. There it becomes apparent that myosin filaments generate forces which are partly compensated by drag forces due to cross-linking proteins. The remaining local contractile forces are then propagated to the tips of the bundle by the viscosity effect of bundling proteins in the filament gel. The model is able to explain how a disordered bundle of comparatively short actin filaments interspersed with myosin filaments can effectively contract the two tips of the actomyosin bundle. It gives a quantitative description of these forces and of the anti-parallel flows of the two phases of anti-parallel F-actin. An asymptotic version of the model with infinite viscosity can be solved explicitly and yields an upper bound to the contractile force of the bundle. PMID:23670678
Robust Mapping of Incoherent Fiber-Optic Bundles
Roberts, Harry E.; Deason, Brent E.; DePlachett, Charles P.; Pilgrim, Robert A.; Sanford, Harold S.
2007-01-01
A method and apparatus for mapping between the positions of fibers at opposite ends of incoherent fiber-optic bundles have been invented to enable the use of such bundles to transmit images in visible or infrared light. The method is robust in the sense that it provides useful mapping even for a bundle that contains thousands of narrow, irregularly packed fibers, some of which may be defective. In a coherent fiber-optic bundle, the input and output ends of each fiber lie at identical positions in the input and output planes; therefore, the bundle can be used to transmit images without further modification. Unfortunately, the fabrication of coherent fiber-optic bundles is too labor-intensive and expensive for many applications. An incoherent fiber-optic bundle can be fabricated more easily and at lower cost, but it produces a scrambled image because the position of the end of each fiber in the input plane is generally different from the end of the same fiber in the output plane. However, the image transmitted by an incoherent fiber-optic bundle can be unscrambled (or, from a different perspective, decoded) by digital processing of the output image if the mapping between the input and output fiber-end positions is known. Thus, the present invention enables the use of relatively inexpensive fiber-optic bundles to transmit images.
Enthalpy and void distributions in subchannels of PHWR fuel bundles
Energy Technology Data Exchange (ETDEWEB)
Park, J. W.; Choi, H.; Rhee, B. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1998-12-31
Two different types of the CANDU fuel bundles have been modeled for the ASSERT-IV code subchannel analysis. From calculated values of mixture enthalpy and void fraction distribution in the fuel bundles, it is found that net buoyancy effect is pronounced in the central region of the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. It is also found that the central region of the DUPIC fuel bundle can be cooled more efficiently than that of the standard fuel bundle. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the mixture enthalpy and void fraction can be highest in the peripheral region of the DUPIC fuel bundle. On the other hand, the enthalpy and the void fraction were found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. This study shows that the subchannel analysis is very useful in assessing thermal behavior of the fuel bundle that could be used in CANDU reactors. 10 refs., 4 figs., 2 tabs. (Author)
Steady-flow characteristics of bundle fluid in drawing
Energy Technology Data Exchange (ETDEWEB)
Huh, You; Kim, Jong Seong [Kyunghee University, Suwon (Korea, Republic of)
2006-07-15
Drawing is a mechanical operation attenuating material thickness to an appropriate level for the next processing or end usage. When the input material has a form of bundle or bundles made of very thin and long shaped wires or fibers, this attenuation operation is called 'bundle drawing' or 'drafting'. Bundle drawing is being used widely in manufacturing micro sized wires or staple yarns. However, the bundle processed by this operation has more or less defects in the evenness of linear density. Such irregularities cause many problems not only for the product quality but also for the efficiency of the next successive processes. In this research a mathematical model for the dynamic behavior of the bundle fluid is to be set up on the basis of general physical laws containing physical variables, i.e. linear density and velocity as the dynamic state variables of the bundle fluid. The governing equations resulting from the modeling show that they appear in a slightly different form from what they do in a continuum fluid. Then, the governing equations system is simplified in a steady state and the bundle dynamics is simulated, showing that the shape of the velocity profiles depends on two model parameters. Experiments confirm that the model parameters are to be well adjusted to show a coincidence with the theoretical analysis. The higher the drawing ratio and drawing speed are, the more sensitive becomes the bundle flow to exogenous disturbances.
Bundle formation in parallel aligned polymers with competing interactions
Dutta, Sandipan; Benetatos, P.; Jho, Y. S.
2016-04-01
Aggregation of like-charged polymers is widely observed in biological- and soft-matter systems. In many systems, bundles are formed when a short-range attraction of diverse physical origin like charge bridging, hydrogen bonding or hydrophobic interaction, overcomes the longer-range charge repulsion. In this letter, we present a general mechanism of bundle formation in these systems as the breaking of the translational invariance in parallel aligned polymers with competing interactions of this type. We derive a criterion for finite-sized bundle formation as well as for macroscopic phase separation (formation of infinite bundles).
Confinement-dependent friction in peptide bundles.
Erbaş, Aykut; Netz, Roland R
2013-03-19
Friction within globular proteins or between adhering macromolecules crucially determines the kinetics of protein folding, the formation, and the relaxation of self-assembled molecular systems. One fundamental question is how these friction effects depend on the local environment and in particular on the presence of water. In this model study, we use fully atomistic MD simulations with explicit water to obtain friction forces as a single polyglycine peptide chain is pulled out of a bundle of k adhering parallel polyglycine peptide chains. The whole system is periodically replicated along the peptide axes, so a stationary state at prescribed mean sliding velocity V is achieved. The aggregation number is varied between k = 2 (two peptide chains adhering to each other with plenty of water present at the adhesion sites) and k = 7 (one peptide chain pulled out from a close-packed cylindrical array of six neighboring peptide chains with no water inside the bundle). The friction coefficient per hydrogen bond, extrapolated to the viscous limit of vanishing pulling velocity V → 0, exhibits an increase by five orders of magnitude when going from k = 2 to k = 7. This dramatic confinement-induced friction enhancement we argue to be due to a combination of water depletion and increased hydrogen-bond cooperativity. PMID:23528088
International Nuclear Information System (INIS)
A three-dimensional thermo-hydraulic computer code is developed for simulation of incompressible flows in complex geometries. The computer code employs a body-fitted, non orthogonal grid system in order to efficiently handle the complex geometries encountered in many engineering applications. The finite volume method is used to discretize the governing equations and the convection term is treated by higher-order bounded schemes. The cell-centered, non staggered grid arrangement is adopted and the resulting checkerboard pressure oscillation is avoided by use of momentum interpolation practice. The computer code employs the SIMPLE algorithm for pressure and velocity coupling and the K-ε turbulence for turbulent calculation. The computer code has been tested through application to a variety of test problems and some results are presented in this paper
International Nuclear Information System (INIS)
Some studies were done about the effect of the uncertainty in the values of several thermo-hydraulic parameters on the core behaviour of the CAREM-25 reactor.By using the chain codes CITVAP-THERMIT and the perturbation the reference states, it was found that concerning to the total power, the effects were not very important, but were much bigger for the pressure.Furthermore were hardly significant in the presence of any perturbation on the void fraction calculation and the fuel temperature.The reactivity and the power peaking factor had highly important changes in the case of the coolant flow.We conclude that the use of this procedure is adequate and useful to our purpose
International Nuclear Information System (INIS)
PUCHOK BM-DF code is described which is designated for local thermal-hudraulic parameters calculations in rod clusters with large rod numbers (300 and more) using the cell method. In this code iterations are applied for the convective crossflow mixing calculation. For the turbulent heat transfer description several empirical correlations are employed the choice among which can be done by the user. On the basis of the calculated cell thermohydraulic parameters numerous experimental data on the heat transfer crisis in various test facilities have been analyzed including those with high length. A good agreement between the calculated and experimental data allows to recommend with a high degree of reliability the cell method realized in RUCHOK BM-DF code for the fuer rod performance analysis in clusters of high length
International Nuclear Information System (INIS)
An analysis is done of the core behavior for a 1861 MW(th) pressurized water reactor with two coolant loops, during the blowdown phase of a double-ended cold leg rupture, between the main feedwater pump, and the pressure vessel. The analysis is done through a detailed thermohydraulic study of the hot pin channel with RELAP4/MOD 5 code, including the Evaluatin Model options. The problem is solved separately for two values of discharge coefficient (C sub(D)= 1,0 and 0,4). The results show that the maximum clad temperature is lower than the limit value for licensing purposes. Concerning clad material oxidation, the maximum value obtained is also under the limit of acceptance. (author)
International Nuclear Information System (INIS)
Thermo-hydraulic experiment was carried out in order to test performance of the direct cooled liquid hydrogen moderator cell to be installed at the research reactor of the Budapest Neutron Center. Two electric hearers up to 300 W each imitated the nuclear heat release in the liquid hydrogen as well as in construction material. The test moderator cell was also equipped with temperature gauges to measure the hydrogen temperature at different positions as well as the inlet and outlet temperature of cooling he gas. The hydrogen pressure in the connected buffer volume was also controlled. At 140 w expected total heat load the moderator cell was filled with liquid hydrogen within 4 hours. The heat load and hydrogen pressure characteristics of the moderator cell are also presented. (author)
Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition
International Nuclear Information System (INIS)
Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor
Directory of Open Access Journals (Sweden)
Masataka Deie
2015-01-01
Full Text Available Background. Posterior cruciate ligament (PCL injuries are not rare in acute knee injuries, and several recent anatomical studies of the PCL and reconstructive surgical techniques have generated improved patient results. Now, we have evaluated PCL reconstructions performed by either the single-bundle or double-bundle technique in a patient group followed up retrospectively for more than 10 years. Methods. PCL reconstructions were conducted using the single-bundle (27 cases or double-bundle (13 cases method from 1999 to 2002. The mean age at surgery was 34 years in the single-bundle group and 32 years in the double-bundle group. The mean follow-up period was 12.5 years. Patients were evaluated by Lysholm scoring, the gravity sag view, and knee arthrometry. Results. The Lysholm score after surgery was 89.1±5.6 points for the single-bundle group and 91.9±4.5 points for the double-bundle group. There was no significant difference between the methods in the side-to-side differences by gravity sag view or knee arthrometer evaluation, although several cases in both groups showed a side-to-side difference exceeding 5 mm by the latter evaluation method. Conclusions. We found no significant difference between single- and double-bundle PCL reconstructions during more than 10 years of follow-up.
Energy Technology Data Exchange (ETDEWEB)
Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A. [Comision Federal de Electricidad, Central Nucleoelectrica de Laguna Verde, Carretera Cardel-Nautla Km. 42.5, Mpio. Alto Lucero, Veracruz (Mexico); Castlllo D, R. [ININ, Carretera Mexico-Toluca Km 36.5, La Marquesa, Estado de Mexico (Mexico)]. e-mail: gcm9acpp@cfe.gob.mx
2006-07-01
In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the
International Nuclear Information System (INIS)
Plant heat-up is a process which all operating systems such as primary coolant circuit, pressurizer, primary and secondary sides of the steam generators and etc. are transferred from a cold shutdown to a hot standby status. During plant heat-up, some thermohydraulic limitations such as maximum and minimum allowable pressure and maximum rate of increase in pressure and temperature which are recommended by plant commissioning program and NPP safety related documents should be considered. Maximum allowable pressure prevents brittle fracture in reactor vessel, Minimum allowable pressure in the inlet of the reactor coolant pumps (RCPs) prevent pump cavitations and maximum allowable rate of increase in temperature and pressure respectively prevent thermal and mechanical shocks. Thus, tuning pressure and temperature increasing rates during plant heat-up is important from plant safety point of view. The RELAP5 system code was used to model and analysis the behavior of WWER-1000 plants during heat-up. In plant heat-up, at first the primary circuit pressure increases by injection of N2 gas into pressurizer in order to provide minimum required NPSH (net positive suction head) for operation of the RCPs. After short time RCPs are turned on to operate which increases the primary coolant circuit temperature through friction losses. At a time which is specified by heat-up procedure the pressurizer heaters are turning on to increase the primary circuit pressure. Heat transfer from primary to secondary side in the steam generators causes increasing of the secondary side temperature and pressure. Temperature and pressure of primary and secondary circuits increase until plant reaches to hot standby condition. The results show that the thermohydraulic parameters during plant heat-up are in an acceptable range and have a good agreement with available data in technical documents. (authors)
Directory of Open Access Journals (Sweden)
A. Del Nevo
2012-01-01
Full Text Available Accurate prediction of steam volume fraction and of the boiling crisis (either DNB or dryout occurrence is a key safety-relevant issue. Decades of experience have been built up both in experimental investigation and code development and qualification; however, there is still a large margin to improve and refine the modelling approaches. The qualification of the traditional methods (system codes can be further enhanced by validation against high-quality experimental data (e.g., including measurement of local parameters. One of these databases, related to the void fraction measurements, is the pressurized water reactor subchannel and bundle tests (PSBT conducted by the Nuclear Power Engineering Corporation (NUPEC in Japan. Selected experiments belonging to this database are used for the OECD/NRC PSBT benchmark. The activity presented in the paper is connected with the improvement of current approaches by comparing system code predictions with measured data on void production in PWR-type fuel bundles. It is aimed at contributing to the validation of the numerical models of CATHARE 2 code, particularly for the prediction of void fraction distribution both at subchannel and bundle scale, for different test bundle configurations and thermal-hydraulic conditions, both in steady-state and transient conditions.
International Nuclear Information System (INIS)
The need for a multi dimensional analysis of transient thermal hydraulic phenomena in a component of a nuclear reactor is increasing with the advanced design features. Motivated by this, the development of a new thermal hydraulic analysis code, named CUPID, is in progress at KAERI (Korea Atomic Energy Research Institute). The simulation of the passive secondary cooling system, PAFS (Passive Auxiliary Feedwater System) has been considered as one of the practical applications of CUPID. In order to validate the two phase flow models of CUPID, the PASCAL test facility was simulated with a porous media model in our previous study. In the present study, the heat exchanger bundle effect on the natural circulation heat transfer was investigated using CUPID for the passive condensate cooling tank (PCCT) of the PAFS. The calculation results with the porous media model, such as the liquid temperature and velocity, were imposed as boundary conditions of the detailed flow simulation near the single heat exchanger tube of the PASCAL facility. Thereafter, the bundle effect was investigated by comparing the calculation result of the unit cell of the PAFS tube bundle and that of the single tube. This paper presents the two dimensional porous media analysis result for the PCCT of the PASCAL facility, two dimensional open media analysis result for a single heat exchanger tube, and two dimensional open media analysis result for the unit cell of the heat exchanger bundle
Energy Technology Data Exchange (ETDEWEB)
Cho, Hyoung Kyu; Lee, Seung Jun; Yoon, Han Young [KAERI, Daejeon (Korea, Republic of)
2012-10-15
The need for a multi dimensional analysis of transient thermal hydraulic phenomena in a component of a nuclear reactor is increasing with the advanced design features. Motivated by this, the development of a new thermal hydraulic analysis code, named CUPID, is in progress at KAERI (Korea Atomic Energy Research Institute). The simulation of the passive secondary cooling system, PAFS (Passive Auxiliary Feedwater System) has been considered as one of the practical applications of CUPID. In order to validate the two phase flow models of CUPID, the PASCAL test facility was simulated with a porous media model in our previous study. In the present study, the heat exchanger bundle effect on the natural circulation heat transfer was investigated using CUPID for the passive condensate cooling tank (PCCT) of the PAFS. The calculation results with the porous media model, such as the liquid temperature and velocity, were imposed as boundary conditions of the detailed flow simulation near the single heat exchanger tube of the PASCAL facility. Thereafter, the bundle effect was investigated by comparing the calculation result of the unit cell of the PAFS tube bundle and that of the single tube. This paper presents the two dimensional porous media analysis result for the PCCT of the PASCAL facility, two dimensional open media analysis result for a single heat exchanger tube, and two dimensional open media analysis result for the unit cell of the heat exchanger bundle.
TRACE assessment of the ACHILLES ISP-25 reflood transient
International Nuclear Information System (INIS)
The purpose of this paper is to assess the capability of the best estimate thermal-hydraulic code TRACE Version 5.211 to predict the reflood process in a rod bundle test section using ACHILLES experimental data from the ISP-25 reflood transient. For the purpose of this assessment study, two detailed TRACE models representing the entire ACHILLES test section without the downcomer were developed and applied to simulate the ISP-25 transient. The TRACE models differed only in the hydrodynamic components, VESSEL and PIPE, which were used to represent the rod bundle region of the test section. Code predictions were compared against ISP-25 test measurements for both local- and integral-type quantities. These measurements included rod surface temperatures for individual rods at various axial elevations, sub-channel steam temperatures at different axial elevations, test section exit liquid and steam mass flow rates, quench front location, test section collapsed liquid level, test section overall pressure drop, and differential pressure drops across various axial sections of the test bundle. Considering the involvement of a non-uniform axial power profile combined with radial temperature variations among individual rods in the experimental rod surface temperature data, TRACE exhibited reasonable capability in predicting the ACHILLES ISP-25 reflood transient implementing an average-rod test bundle modeling approach. Consistent with other reflood simulations obtained with recent TRACE code versions, major differences between ACHILLES ISP-25 simulation results and experimental data for rod surface temperatures were observed mainly for the upper part of test section, also caused by lack of spacer grid models in TRACE. (author)
International Nuclear Information System (INIS)
PWR or LMFBR cores or fuel assemblies, PWR steam generators, tubular heat exchangers, are as many components of a nuclear power plant involving single or two-phase flows in tube bundles. The knowledge of the detailed flow patterns (velocity, pressure, temperature, void fraction) on the shell side is necessary to evaluate DNB in reactor cores during incidental transients, singularity effects (grids, wire spacers, support plates, baffles), corrosion on steam generator tube sheet, bypass effects and vibration risks. For that purpose, the Research and Studies Branch of EDF is now developing a new 3D computer code THYC, at first for PWR's and LFMBR's core thermalydraulic calculations and in a second place, for more general applications of flows in tube bundles, thanks to a general porous medium approach. This paper describes the physical model used in THYC (4 partial differential equations, closure relationships specific of tube bundle geometries) then numerical solution method (space and time discretization, solution algorithm). Examples of numerical tests and incidental operation transients in a french PWR 900 MW core, are given. Finally, 3 validation calculations will be presented concerning a tube and shell heat exchanger, a LMFBR fuel assembly in natural, mixed and forced convection, and a lower part of a PWR steam generator
Subanalytic Bundles and Tubular Neighbourhoods of Zero-Loci
Indian Academy of Sciences (India)
Vishwambhar Pati
2003-08-01
We introduce the natural and fairly general notion of a subanalytic bundle (with a finite dimensional vector space of sections) on a subanalytic subset of a real analytic manifold , and prove that when is compact, there is a Baire subset of sections in whose zero-loci in have tubular neighbourhoods, homeomorphic to the restriction of the given bundle to these zero-loci.
Quantum principal bundles as Hopf-Galois extensions
Durdevic, M
1995-01-01
It is shown that every quantum principal bundle with a compact structure group is a Hopf-Galois extension. This property naturally extends to the level of general differential structures, so that every differential calculus over a quantum principal bundle with a compact structure group is a graded-differential variant of the Hopf-Galois extension.
Lexical Bundles in L1 and L2 Academic Writing
Chen, Yu-Hua; Baker, Paul
2010-01-01
This paper adopts an automated frequency-driven approach to identify frequently-used word combinations (i.e., "lexical bundles") in academic writing. Lexical bundles retrieved from one corpus of published academic texts and two corpora of student academic writing (one L1, the other L2), were investigated both quantitatively and qualitatively.…
On the Classification of Complex Vector Bundles of Stable Rank
Indian Academy of Sciences (India)
Constantin Bǎnicǎ; Mihai Putinar
2006-08-01
One describes, using a detailed analysis of Atiyah–Hirzebruch spectral sequence, the tuples of cohomology classes on a compact, complex manifold, corresponding to the Chern classes of a complex vector bundle of stable rank. This classification becomes more effective on generalized flag manifolds, where the Lie algebra formalism and concrete integrability conditions describe in constructive terms the Chern classes of a vector bundle.
On the general elephant conjecture for Mori conic bundles
Prokhorov, Yu G
1996-01-01
Let $f:X\\to S$ be an extremal contraction from a threefolds with terminal singularities onto a surface (so called Mori conic bundle). We study some particular cases of such contractions: quotients of usual conic bundles and index two contractions. Assuming Reid's general elephants conjecture we also obtain a rough classification. We present many examples.
CANDU bundle junction. Misalignment probability and pressure-drop correlation
International Nuclear Information System (INIS)
The pressure drop over the bundle junction is an important component of the pressure drop in a CANDU (Canada Deuterium Uranium) fuel channel. This component can represent from ∼ 15% for aligned bundles to ∼ 26% for rotationally misaligned bundles, and is dependent on the degree of misalignment. The geometry of the junction increases the mixing between subchannels, and hence improves the thermal performance of the bundle immediately downstream. It is therefore important to model the junction's performance adequately. This paper summarizes a study sponsored by COG (CANDU Owners Group) and an NSERC (National Science and Engineering Research Council) Industrial Research Grant, undertaken, at CRL (Chalk River Laboratories) to identify and develop a bundle-junction model for potential implementation in the ASSERT (Advanced Solution of Subchannel Equations in Reactor Thermalhydraulics) subchannel code. The work reported in this paper consists of two components of this project: an examination of the statistics of bundle misalignment, demonstrating that there are no preferred positions for the bundles and therefore all misalignment angles are equally possible; and, an empirical model for the single-phase pressure drop across the junction as a function of the misalignment angle. The second section of this paper includes a brief literature review covering the experimental, analytical and numerical studies concerning the single-phase pressure drop across bundle junctions. 32 refs., 9 figs
Trace extensions, determinant bundles, and gauge group cocycles
Arnlind, J; Arnlind, Joakim; Mickelsson, Jouko
2002-01-01
We study the geometry of determinant line bundles associated to Dirac operators on compact odd dimensional manifolds. Physically, these arise as (local) vacuum line bundles in quantum gauge theory. We give a simplified derivation of the commutator anomaly formula using a construction based on noncyclic trace extensions and associated multiplicative renormalized determinants.
Bundle power history envelope using a theoretical method
International Nuclear Information System (INIS)
This paper gives a simple theoretical method for calculating a bundle power history envelope which envelopes all possible individual bundle power versus burnup histories. The lattice parameters at different burnups were generated by computer code CLUB. (author). 2 refs., 1 tab., 2 figs
Bundled slaty cleavage in laminated argillite, north-central minnesota
Southwick, D.L.
1987-01-01
Exceptional bundled slaty cleavage (defined herein) has been found in drill cores of laminated, folded, weakly metamorphosed argillite at several localities in the early Proterozoic Animikie basin of north-central Minnesota. The cleavage domains are more closely spaced within the cleavage bundles than outside them, the mean tectosilicate grain size of siltstone layers, measured normal to cleavage, is less in the cleavage bundles than outside them, and the cleavage bundles are enriched in opaque phases and phyllosilicates relative to extra-bundle segments. These facts suggest that pressure solution was a major factor in bundle development. If it is assumed that opaque phases have been conserved during pressure solution, the modal differences in composition between intra-bundle and extra-bundle segments of beds provide a means for estimating bulk material shortening normal to cleavage. Argillite samples from the central part of the Animikie basin have been shortened a minimum of about 22%, as estimated by this method. These estimates are similar to the shortening values derived from other strain markers in other rock types interbedded with the argillite, and are also consistent with the regional pattern of deformation. ?? 1987.
The behaviour of Phenix fuel pin bundle under irradiation
International Nuclear Information System (INIS)
An entire Phenix sub-assembly has been mounted and sectioned after irradiation. The examination of cross-sections revealed the effects of mechanical interaction in the bundle (ovalisations and contacts between clads). According to analysis of the sodium channels, cooling of the pin bundle remained uniform. (author)
Restriction Theorem for Principal bundles in Arbitrary Characteristic
DEFF Research Database (Denmark)
Gurjar, Sudarshan
2015-01-01
The aim of this paper is to prove two basic restriction theorem for principal bundles on smooth projective varieties in arbitrary characteristic generalizing the analogues theorems of Mehta-Ramanathan for vector bundles. More precisely, let G be a reductive algebraic group over an algebraically c...
Phase Space Reduction of Star Products on Cotangent Bundles.
N. Kowalzig; N. Neumaier; M. Pflaum
2005-01-01
In this paper we construct star products on Marsden-Weinstein reduced spaces in case both the original phase space and the reduced phase space are (symplectomorphic to) cotangent bundles. Under the assumption that the original cotangent bundle $T^*Q$ carries a symplectic structure of form $\\omega_{B
Transient drainage summary report
International Nuclear Information System (INIS)
This report summarizes the history of transient drainage issues on the Uranium Mill Tailings Remedial Action (UMTRA) Project. It defines and describes the UMTRA Project disposal cell transient drainage process and chronicles UMTRA Project treatment of the transient drainage phenomenon. Section 4.0 includes a conceptual cross section of each UMTRA Project disposal site and summarizes design and construction information, the ground water protection strategy, and the potential for transient drainage
Artificial ciliary bundles with nano fiber tip links
Asadnia, Mohsen; Miao, Jianmin; Triantafyllou, Michael
2015-01-01
Mechanosensory ciliary bundles in fishes are the inspiration for carefully engineered artificial flow sensors. We report the development of a new class of ultrasensitive MEMS flow sensors that mimic the intricate morphology of the ciliary bundles, including the stereocilia, tip links, and the cupula, and thereby achieve threshold detection limits that match the biological example. An artificial ciliary bundle is achieved by fabricating closely-spaced arrays of polymer micro-pillars with gradiating heights. Tip links that form the fundamental sensing elements are realized through electrospinning aligned PVDF piezoelectric nano-fibers that link the distal tips of the polymer cilia. An optimized synthesis of hyaluronic acid-methacrylic anhydride hydrogel that results in properties close to the biological cupula, together with drop-casting method are used to form the artificial cupula that encapsulates the ciliary bundle. In testing, fluid drag force causes the ciliary bundle to slide, stretching the flexible nan...
Introduction to the theory of fiber bundles and connections I
International Nuclear Information System (INIS)
In lectures 1 and 2 we discuss basic concepts of topology and differential geometry: definition of a topological space and of Hausdorff, compact, connected and paracompact spaces; topological groups and actions of groups on spaces; differentiable manifolds, tangent vectors and 1 forms; partitions of unity and Lie groups. In lecture 3 we present the concept of a fiber bundle and discuss vector bundles and principal bundles. The concept of a connection on a smooth vector bundle is defined in lecture 4, together with the associated concepts of curvature and parallel transport; as an illustration we present the Levi-Civita connection on a Riemannian manifold. Finally, in lecture 5 we define connections on principal bundles and present examples with the Lie groups U(1) and SU(2). For reasons of space the present article only includes lectures 1, 2 and 3. Lectures 4 and 5 will be published in a forthcoming paper. (Author)
HORIZONTAL LAPLACE OPERATOR IN REAL FINSLER VECTOR BUNDLES
Institute of Scientific and Technical Information of China (English)
无
2008-01-01
A vector bundle F over the tangent bundle TM of a manifold M is said to be a Finsler vector bundle if it is isomorphic to the pull-back π*E of a vector bundle E over M([1]). In this article the authors study the h-Laplace operator in Finsler vector bundles.An h-Laplace operator is defined, first for functions and then for horizontal Finsler forms on E. Using the h-Laplace operator, the authors define the h-harmonic function and h-harmonic horizontal Finsler vector fields, and furthermore prove some integral formulas for the h-Laplace operator, horizontal Finsler vector fields, and scalar fields on E.
Development of nuclear fuel. Development of CANDU advanced fuel bundle
International Nuclear Information System (INIS)
In order to develop CANDU advanced fuel, the agreement of the joint research between KAERI and AECL was made on February 19, 1991. AECL conceptual design of CANFLEX bundle for Bruce reactors was analyzed and then the reference design and design drawing of the advanced fuel bundle with natural uranium fuel for CANDU-6 reactor were completed. The CANFLEX fuel cladding was preliminarily investigated. The fabricability of the advanced fuel bundle was investigated. The design and purchase of the machinery tools for the bundle fabrication for hydraulic scoping tests were performed. As a result of CANFLEX tube examination, the tubes were found to be meet the criteria proposed in the technical specification. The dummy bundles for hydraulic scoping tests have been fabricated by using the process and tools, where the process parameters and tools have been newly established. (Author)
Composite bundles in Clifford algebras. Gravitation theory. Part I
Sardanashvily, G
2016-01-01
Based on a fact that complex Clifford algebras of even dimension are isomorphic to the matrix ones, we consider bundles in Clifford algebras whose structure group is a general linear group acting on a Clifford algebra by left multiplications, but not a group of its automorphisms. It is essential that such a Clifford algebra bundle contains spinor subbundles, and that it can be associated to a tangent bundle over a smooth manifold. This is just the case of gravitation theory. However, different these bundles need not be isomorphic. To characterize all of them, we follow the technique of composite bundles. In gravitation theory, this technique enables us to describe different types of spinor fields in the presence of general linear connections and under general covariant transformations.
The 2-Hilbert Space of a Prequantum Bundle Gerbe
Bunk, Severin; Szabo, Richard J
2016-01-01
We construct a prequantum 2-Hilbert space for any line bundle gerbe whose Dixmier-Douady class is torsion. Analogously to usual prequantisation, this 2-Hilbert space has the category of sections of the line bundle gerbe as its underlying 2-vector space. These sections are obtained as certain morphism categories in Waldorf's version of the 2-category of line bundle gerbes. We show that these morphism categories carry a monoidal structure under which they are semisimple and abelian. We introduce a dual functor on the sections, which yields a closed structure on the morphisms between bundle gerbes and turns the category of sections into a 2-Hilbert space. We discuss how these 2-Hilbert spaces fit various expectations from higher prequantisation. We then extend the transgression functor to the full 2-category of bundle gerbes and demonstrate its compatibility with the additional structures introduced. We discuss various aspects of Kostant-Souriau prequantisation in this setting, including its dimensional reductio...
Bundles over Quantum RealWeighted Projective Spaces
Directory of Open Access Journals (Sweden)
Tomasz Brzeziński
2012-09-01
Full Text Available The algebraic approach to bundles in non-commutative geometry and the definition of quantum real weighted projective spaces are reviewed. Principal U(1-bundles over quantum real weighted projective spaces are constructed. As the spaces in question fall into two separate classes, the negative or odd class that generalises quantum real projective planes and the positive or even class that generalises the quantum disc, so do the constructed principal bundles. In the negative case the principal bundle is proven to be non-trivial and associated projective modules are described. In the positive case the principal bundles turn out to be trivial, and so all the associated modules are free. It is also shown that the circle (coactions on the quantum Seifert manifold that define quantum real weighted projective spaces are almost free.
Geometries and applications of active fiber bundles
Giglmayr, Josef
2001-10-01
Active fiber bundles (FBs) are aimed to model photonic switching and processing in 3-D without the restrictions of the photonic technology. The 2-D photonic architectures are assumed to be implemented by networks of directional couplers (DCs) and Mach-Zehnder interferometers (MZIs), respectively. For the implementation several crucial problems are expected: (1) proper operation of the spatial couplers/switches (nonblocking interconnections) and (2) coupling in the interstage interconnection section mainly caused by parallel and crossing fibers/waveguides (WGs). For the design of proper operating switches (refinement of couplers) the application of decoupling concepts of modern control theory is proposed. The final goal is to translate the refined couplers into integrated photonic architectures rather than into additional lightwave circuits (LWCs) which simply would increase the coupling. The decoupling concepts are reviewed. The paper is an attempt to prepare for applying well-known system engineering concepts to the upcoming technology of photonics.
Extendability of parallel sections in vector bundles
Kirschner, Tim
2016-01-01
I address the following question: Given a differentiable manifold M, what are the open subsets U of M such that, for all vector bundles E over M and all linear connections ∇ on E, any ∇-parallel section in E defined on U extends to a ∇-parallel section in E defined on M? For simply connected manifolds M (among others) I describe the entirety of all such sets U which are, in addition, the complement of a C1 submanifold, boundary allowed, of M. This delivers a partial positive answer to a problem posed by Antonio J. Di Scala and Gianni Manno (2014). Furthermore, in case M is an open submanifold of Rn, n ≥ 2, I prove that the complement of U in M, not required to be a submanifold now, can have arbitrarily large n-dimensional Lebesgue measure.
TRIGA modified bundle thermal hydraulic analysis
Energy Technology Data Exchange (ETDEWEB)
Negut, G.; Mladin, M.; Preda, M. [Inst. for Nuclear Research, Pitesti (Romania)
2001-07-01
TRIGA 14 MW steady state reactor (SSR) has more than 20 years of operation experience. It was used as a material test reactor to accomplish full range of experiments of CANDU type fuel, tests on structure material as Zircaloy and stainless steel. We did, also, isotope production for industrial and medical use, neutronography, gamma prompt, neutron diffractometry and activation analysis. In order to optimize the core for a more homogenous burnup we did some experiments on a modified fuel bundle. The paper is dedicated to the computations done in order to validate the optimized core configuration. The analysis has shown no significant impact on the central fuel temperatures, to affect the core safety. (orig.)
TRIGA modified bundle thermal hydraulic analysis
International Nuclear Information System (INIS)
TRIGA 14 MW steady state reactor (SSR) has more than 20 years of operation experience. It was used as a material test reactor to accomplish full range of experiments of CANDU type fuel, tests on structure material as Zircaloy and stainless steel. We did, also, isotope production for industrial and medical use, neutronography, gamma prompt, neutron diffractometry and activation analysis. In order to optimize the core for a more homogenous burnup we did some experiments on a modified fuel bundle. The paper is dedicated to the computations done in order to validate the optimized core configuration. The analysis has shown no significant impact on the central fuel temperatures, to affect the core safety. (orig.)
Effect of Reynolds number and bundle geometry on the turbulent flow in tight lattice bundle
International Nuclear Information System (INIS)
The flow structure in tight lattice is still of great interest to nuclear industry. The accurate prediction of flow parameter in subchannels of tight lattice is likable. Unsteady Reynolds Averaged Navier Stokes (URANS) is a promising approach to achieve this goal. The implementation of URANS (Unsteady Reynolds Averaged Navier Stokes) approach will be validated by comparing computational results with the experimental data of Krauss (1998). In this paper, the turbulent flow with different Reynolds number (5000~215000) and different P/D(1.005~1.2) are simulated with CFD code CFX12.The effects of the Reynolds number and the bundle geometry(P/D) on wall shear stress, turbulent kinetic energy, turbulent mixing and large scale coherent structure in tight lattice are analyzed in details. It is hoped that the present work will contribute to the understanding of these important flow phenomena and facilitate the prediction and design of rod bundles. (author)
Fiber bundle model under fluid pressure
Amitrano, David; Girard, Lucas
2016-03-01
Internal fluid pressure often plays an important role in the rupture of brittle materials. This is a major concern for many engineering applications and for natural hazards. More specifically, the mechanisms through which fluid pressure, applied at a microscale, can enhance the failure at a macroscale and accelerate damage dynamics leading to failure remains unclear. Here we revisit the fiber bundle model by accounting for the effect of fluid under pressure that contributes to the global load supported by the fiber bundle. Fluid pressure is applied on the broken fibers, following Biot's theory. The statistical properties of damage avalanches and their evolution toward macrofailure are analyzed for a wide range of fluid pressures. The macroscopic strength of the new model appears to be strongly controlled by the action of the fluid, particularly when the fluid pressure becomes comparable with the fiber strength. The behavior remains consistent with continuous transition, i.e., second order, including for large pressure. The main change concerns the damage acceleration toward the failure that is well modeled by the concept of sweeping of an instability. When pressure is increased, the exponent β characterizing the power-law distribution avalanche sizes significantly decreases and the exponent γ characterizing the cutoff divergence when failure is approached significantly increases. This proves that fluid pressure plays a key role in failure process acting as destabilization factor. This indicates that macrofailure occurs more readily under fluid pressure, with a behavior that becomes progressively unstable as fluid pressure increases. This may have considerable consequences on our ability to forecast failure when fluid pressure is acting.
Fiber bundle model under fluid pressure.
Amitrano, David; Girard, Lucas
2016-03-01
Internal fluid pressure often plays an important role in the rupture of brittle materials. This is a major concern for many engineering applications and for natural hazards. More specifically, the mechanisms through which fluid pressure, applied at a microscale, can enhance the failure at a macroscale and accelerate damage dynamics leading to failure remains unclear. Here we revisit the fiber bundle model by accounting for the effect of fluid under pressure that contributes to the global load supported by the fiber bundle. Fluid pressure is applied on the broken fibers, following Biot's theory. The statistical properties of damage avalanches and their evolution toward macrofailure are analyzed for a wide range of fluid pressures. The macroscopic strength of the new model appears to be strongly controlled by the action of the fluid, particularly when the fluid pressure becomes comparable with the fiber strength. The behavior remains consistent with continuous transition, i.e., second order, including for large pressure. The main change concerns the damage acceleration toward the failure that is well modeled by the concept of sweeping of an instability. When pressure is increased, the exponent β characterizing the power-law distribution avalanche sizes significantly decreases and the exponent γ characterizing the cutoff divergence when failure is approached significantly increases. This proves that fluid pressure plays a key role in failure process acting as destabilization factor. This indicates that macrofailure occurs more readily under fluid pressure, with a behavior that becomes progressively unstable as fluid pressure increases. This may have considerable consequences on our ability to forecast failure when fluid pressure is acting. PMID:27078437
Feasibility evaluation of x-ray imaging for measurement of fuel rod bowing in CFTL test bundles
International Nuclear Information System (INIS)
The Core Flow Test Loop (CFTL) is a high temperature, high pressure, out-of-reactor helium-circulating system. It is designed for detailed study of the thermomechanical performance, at prototypic steady-state and transient operating conditions, of electrically heated rods that simulate segments of core assemblies in the Gas-Cooled Fast Breeder reactor demonstration plant. Results are presented of a feasibility evaluation of x-ray imaging for making measurements of the displacement (bowing) of fuel rods in CFTL test bundles containing electrically heated rods. A mock-up of a representative CFTL test section consisting of a test bundle and associated piping was fabricated to assist in this evaluation
Anatomic double-bundle anterior crucial ligament reconstruction with G-ST
Kuroda, Ryosuke; Matsushita, Takehiko
2011-01-01
The anterior cruciate ligament (ACL) consists of two primal functional bundles, anteromedial bundle and posterolateral bundles. Those two bundles play different functional roles and contribute differently to knee stability throughout the range of motion. Recent advancement in studies of anatomy and biomechanics of ACL has led surgeons to perform double-bundle ACL reconstruction to obtain better stability and kinematics. Consequently, variable surgical techniques of double-bundle ACL reconstru...
The Failure Effect of Primary Coolant Pump to Thermo-Hydraulic Characteristic of TRIGA 2000 Reactor
International Nuclear Information System (INIS)
Has been done analysis of transient, when TRIGA 2000 reactor loss of primary coolant flow because primary pump loss of electric power, so fail in function.The calculation using RELAP5/MOD32 computer code with reactor core is modeled in the form of different seven channels as representation of different seven areas in core with 116 fuels. This reactor model also considers position of tip of primary pipe of input tank which is below of core, form of lower part core geometry influencing direction and coolant flow rate into core, and existence of diffuser system. The result of calculation in condition of steady state is obtained initiation condition of steady state is reached after 2500 seconds from reactor starts operation on 2000 kW power. On steady state, the channel-3 cladding temperature (hottest) is 149.63℃, the coolant temperature outlet from the channel-3 (hottest) is 105.66℃ , reactor inlet temperature is 32.2℃, and reactor outlet temperature is 46.79℃. The primary coolant entering reactor with flow rate 59.64 kg/s, distributed to core 31.44 kg/s and to by-pass of core or by-pass of chimney 28.20 kg/s. The result of calculation transient is obtained, before scram occur the channel-3 cladding temperature (hottest) is 161.03℃ and the coolant temperature outlet from the channel-3 (hottest) is 117.66℃. In the reactor core is a natural circulation as well (from reactor core, to chimney, to by-pass of chimney, to by-pass of core and back to platform) which is cooling reactor core. Scram occur on 250 seconds after failure of the primary pump. Based on result of this study is known that, when transient condition is happened because primary pump failure, reactor is predicted to stays in safety margin. (author)
Thermohydraulic analysis of the IAEA standard problem test on the PMK-NHV facility
International Nuclear Information System (INIS)
International Atomic Energy Agency (IAEA) has supported a standard test problem simulating small break loss of coolant accident on the test facility PMH-NHV in Budapest. The present pretest analysis of that transient was done using the computer code RELAP4/MOD6. The results were compared to the measurements data and to data of 19 other laboratories around the world that have performed the same analysis. The correspondence of the results to the measured data is reasonable. There are bigger discrepancies, which in turn influence other variables. (author)
An experimental assessment of cooling of a 54-rod bundle by in-bundle injection
International Nuclear Information System (INIS)
Highlights: ► Rewetting of a 54-rod bundle assembled with a central coolant tube is investigated. ► The coolant tube injects the coolant radially outwards at different axial levels. ► Above a minimum flow rate, coolant quenches all the rods throughout their length. ► Rapid cooling of rods occurs up to around 100 °C of the rod surface temperature. ► Counter current flow of steam–water gets generated which affects cooling adversely. - Abstract: The performance of an in-bundle coolant injection system for the quenching of dry heated rods has been experimentally investigated. The rod bundle contains 54 fuel rods of 11.2 mm diameter, 3700 mm long, arranged in three concentric rings with a central coolant supply tube. The coolant tube supplies the coolant in the form of jets through a series of circumferential holes at different axial levels inside the rod bundle. Visualization during cold state injection tests ensures that the liquid spray can reach different levels of all the rods above a certain flow rate of water through the coolant tube. Extensive cooling experiments were done to assess the suitability of the proposed scheme of in-bundle coolant injection. Time–temperature curves have been derived from rods at different locations, from different heights of the rods, over a range of coolant flow rate as well as for different rod temperatures. The effect of the presence of the spacers on local cooling has also been investigated. The cooling curves follow a general trend of a rapid temperature drop up to almost 100 °C of the rod surface temperature irrespective of the operating parameters and the location of the rod. Thereafter, the temperature falls slowly reaching the coolant temperature almost asymptotically. Moreover, the second phase of cooling is often marked by temperature fluctuations of random nature. It was also observed that a large volume of steam generates during cooling and comes out through the top of the test section expelling a
Energy Technology Data Exchange (ETDEWEB)
Caro, R.
1976-07-01
In this report the analysis of neutronics thermohydraulics and shielding of the 10 HWt swimming pool reactor C.E.N.E is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs.
Uren, Kenneth Richard; Schoor, George van
2013-01-01
This second paper in a two part series presents the application of a developed state space model extraction methodology applied to a Brayton cycle-based PCU (power conversion unit) of a PBMR (pebble bed modular reactor). The goal is to investigate if the state space extraction methodology can cope with larger and more complex thermohydraulic systems. In Part I the state space model extraction methodology for the purpose of control was described in detail and a state space represen...
Babitz Philip; Choe Dongok; Jevremovic Tatjana
2013-01-01
The thermodynamic conditions of the University of Utah's TRIGA Reactor were simulated using SolidWorks Flow Simulation, Ansys, Fluent and PARET-ANL. The models are developed for the reactor's currently maximum operating power of 90 kW, and a few higher power levels to analyze thermohydraulics and heat transfer aspects in determining a design basis for higher power including the cost estimate. It was found that the natural convection current becomes much mor...
International Nuclear Information System (INIS)
The computer code NAIADQ is designed to simulate the course and consequences of non-destructive reactivity accidents in low power, experimental, water-cooled reactor cores fuelled with metal plate elements. It is a coupled neutron kinetics-hydrodynamics-heat transfer code which uses point kinetics and one-dimensional thermohydraulic equations. Nucleate boiling, which occurs at the fuel surface during transients, is modelled by the growth of a superheated layer of water in which vapour is generated at a non-equilibrium rate. It is assumed that this vapour is formed at its saturation temperature and that it mixes homogeneously with the water in this layer. The code is written in FORTRAN IV and has been programmed to run as a catalogued procedure on an IBM operating system such as MVT or MVS, with facility for the inclusion of user routines
Two-categorical bundles and their classifying spaces
DEFF Research Database (Denmark)
Baas, Nils A.; Bökstedt, M.; Kro, T.A.
2012-01-01
For a 2-category 2C we associate a notion of a principal 2C-bundle. In case of the 2-category of 2-vector spaces in the sense of M.M. Kapranov and V.A. Voevodsky this gives the the 2-vector bundles of N.A. Baas, B.I. Dundas and J. Rognes. Our main result says that the geometric nerve of a good 2......-category is a classifying space for the associated principal 2-bundles. In the process of proving this we develop a lot of powerful machinery which may be useful in further studies of 2-categorical topology. As a corollary we get a new proof of the classification of principal bundles. A calculation based...... on the main theorem shows that the principal 2-bundles associated to the 2-category of 2-vector spaces in the sense of J.C. Baez and A.S. Crans split, up to concordance, as two copies of ordinary vector bundles. When 2C is a cobordism type 2-category we get a new notion of cobordism-bundles which turns out...
The Geometry of Tangent Bundles: Canonical Vector Fields
Directory of Open Access Journals (Sweden)
Tongzhu Li
2013-01-01
Full Text Available A canonical vector field on the tangent bundle is a vector field defined by an invariant coordinate construction. In this paper, a complete classification of canonical vector fields on tangent bundles, depending on vector fields defined on their bases, is obtained. It is shown that every canonical vector field is a linear combination with constant coefficients of three vector fields: the variational vector field (canonical lift, the Liouville vector field, and the vertical lift of a vector field on the base of the tangent bundle.
On the existence of n-dimensional indecomposable vector bundles
International Nuclear Information System (INIS)
Let X be an arbitrary smooth irreducible complex projective curve of genus g with g ≥ 4. In this paper we extend the existence theorem of special divisors to high dimensional indecomposable vector bundles. We give a necessary and sufficient condition on the existence of n-dimensional indecomposable vector bundles E with deg(E) = d, dimH0(X,E) ≥ h. We also determine under what condition the set of all such vector bundles will be finite and how many elements it contains. (author). 9 refs
Line bundles and the Thom construction in noncommutative geometry
Beggs, E. J.; Brzezinski, T.
2010-01-01
The idea of a line bundle in classical geometry is transferred to noncommutative geometry by the idea of a Morita context. From this we can construct Z and N graded algebras, the Z graded algebra being a Hopf-Galois extension. A non-degenerate Hermitian metric gives a star structure on this algebra, and an additional star operation on the line bundle gives a star operation on the N graded algebra. In this case, we can carry out the associated circle bundle and Thom constructions. Starting wit...
On separation axioms of uniform bundles and sheaves
Clara M. Neira U.; Januario Varela
2004-01-01
In the context of the theory of uniform bundles in the sense of J. Dauns and K. H. Hofmann, the topology of the fiber space of a uniform bundle depends on the assumption of upper semicontinuity of its defining set of pseudometrics when composed with local sections. In this paper we show that the additional hypothesis of lower semicontinuity of these functions secures that the fiber space of the uniform bundle is Hausdorff, regular or completely regular provided that the base space has the cor...
Energy Technology Data Exchange (ETDEWEB)
Nam, Seung Hyun; Lee, Jeong Ik; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)
2014-05-15
{sub th} power and electricity generation with 100 kW{sub th} idle power. Consequently, KANUTER has the characteristics of a compact and lightweight system, excellent propellant efficiency, bimodal capability, and mission versatility as indicated in the reference design parameters. This thermo-hydraulic design analysis was carried out to estimate the optimum FWT of the unique SLHC fuel design in the core and thereby the maximum rocket performance. The FWT affects the mechanical strength of the SLHC fuel assembly as well as the thermo-hydraulic capability mainly depending on the heat transfer area of fuel. The thicker fuel wafer is mechanically strong with low pressure drop, while the thinner fuel wafer is thermally robust with less mechanical strength and higher shear stress in the core.
International Nuclear Information System (INIS)
electricity generation with 100 kWth idle power. Consequently, KANUTER has the characteristics of a compact and lightweight system, excellent propellant efficiency, bimodal capability, and mission versatility as indicated in the reference design parameters. This thermo-hydraulic design analysis was carried out to estimate the optimum FWT of the unique SLHC fuel design in the core and thereby the maximum rocket performance. The FWT affects the mechanical strength of the SLHC fuel assembly as well as the thermo-hydraulic capability mainly depending on the heat transfer area of fuel. The thicker fuel wafer is mechanically strong with low pressure drop, while the thinner fuel wafer is thermally robust with less mechanical strength and higher shear stress in the core
International Nuclear Information System (INIS)
In consequence of activities on uncovering the reasons for through faults formation in cooling tubes of reactor control and protection system (CPS) channels of Bilibino-2 reactor the conclusion is made that corrosion failure development takes place against the backdrop of periodic increase of total moisture in reactor space at transient and standby modes at top of cooling tubes of CPS channels. Formation of corrosion defects in cooling tubes of four CPS channels of unit 2 in 2011-2012 is specific factor for this plant unit and do not effect on operation of other plant units. It is pointed out that ingress of moisture into gas system of the reactor is the critical factor providing integrity of structure elements of EhPG-6 reactor core cooling system. This fact agrees nicely with the results obtained during operation AM reactor of the First NPP
Analysis of integral experiments at ISB-VVER-1000 with the thermohydraulic code ATHLET
International Nuclear Information System (INIS)
Post test calculations of LOCA experiments on the integral test facility ISB-VVER with the thermalhydraulic code ATHLET have been implemented. Fulfillment of the work has shown not only capabilities of the code to simulate VVER specifics but revealed peculiarities of the test facility itself. The ISB-VVER test facility models VVER-1000 reactor in the power-volumetric scale l:3000 with full pressure and nominal power. The thermalhydraulic code ATHLET has been developed by the German 'Gesellschaft fuer Anlagen- und Reaktorsicherheit' for PWR LOCA and transients analysis. Three experiments have been analyzed: 2.4% break in the upper plenum, 11% break in the upper plenum, 2.4% break in the downcomer with operation of HPIS. The calculations demonstrated rather good agreement of experimental and numerical data (pressures, temperatures, pressure differences, flow rates, void fractions). (author)
Use of sensitivity-information for the adaptive simulation of thermo-hydraulic system codes
International Nuclear Information System (INIS)
Within the scope of this thesis the development of methods for online-adaptation of dynamical plant simulations of a thermal-hydraulic system code to measurement data is depicted. The described approaches are mainly based on the use of sensitivity-information in different areas: statistical sensitivity measures are used for the identification of the parameters to be adapted and online-sensitivities for the parameter adjustment itself. For the parameter adjustment the method of a ''system-adapted heuristic adaptation with partial separation'' (SAHAT) was developed, which combines certain variants of parameter estimation and control with supporting procedures to solve the basic problems. The applicability of the methods is shown by adaptive simulations of a PKL-III experiment and by selected transients in a nuclear power plant. Finally the main perspectives for the application of a tracking simulator on a system code are identified.
Large-break LOCA studies. Computational analysis of clad ballooning and thermohydraulics in a PWR
International Nuclear Information System (INIS)
A new multi-pin model of the re-flood phase of a large break loss of coolant accident has been created through the dynamic coupling between the thermal-hydraulic code RELAP5 and multiple instances of the single-pin thermal-mechanics code MABEL. After a brief description of the codes and their linkage, a series of tests to assess the capabilities of the linked codes is described, and their results analysed. It is shown that the current coupled multi-pin code is a stable and reliable tool for ballooning transient analysis. A complete validation process with the simulation of the MT-3 test in the NRU reactor at Chalk River is in progress.(author)
Development of multidimensional two-phase flow measurement sensor in rod bundle
International Nuclear Information System (INIS)
In order to acquire multidimensional two-phase flow in 10x10 bundle, SubChannel Void Sensor (SCVC) consisting of 11-wire by 11-wire and 10-rod by 10-rod electrodes is developed. A conductance value in a proximity region of one wire and another gives void fraction in the center of subchannel region. A phasic velocity can be estimated by using two layers of wire meshes, like as so-called wire mesh sensor. 121 points (=11x11) of void fraction as well as those of phasic velocity are acquired. It is peculiarity of the devised sensor that void fraction near rod surface can be estimated by a conductance value in a proximity region of one wire and one rod. 400 additional points of void fraction in 10x10 bundle can be, therefore, acquired. The time resolution of measurement is up to 1250 frames (cross sections) per second. We capability in a 10x10 bundle with o.d. 10 mm and 3110 mm long is demonstrated. The devised sensor is installed in 8 height levels to acquire the two-phase flow dynamics along axial direction. A pair of sensor layers is mounted in each level and is placed by 30 mm apart with each other to estimate a phasic velocity distribution on the basis of cross-correlation function of the two layers. Air bubbles are injected through sintered metal nozzles from the bottom end of 10x10 rods. Air flow rate distribution can vary with a controlled valves connected to each nozzle. The devised sensor exhibited the quasi three-dimensional flow structures, i.e. void fraction, phasic velocity and bubble chord length distributions. These quasi three-dimensional structures explorer complexity of two-phase flow dynamics such as coalescence and breakup of bubbles in the transient phasic velocity distributions. (author)
Assembly mechanism for nuclear fuel bundles
International Nuclear Information System (INIS)
A method of securing a fuel bundle to permit easy remote disassembly is described. Fuel rods are held loosely between end plates, each end of the rods fitting into holes in the end plates. At the upper end of each fuel rod there is a spring pressing against the end plate. Tie rods are used to hold the end plates together securely. The lower end of each tie rod is screwed into the lower end plate; the upper end of each tie rod is attached to the upper end plate by means of a locking assembly described in the patent. In order to remove the upper tie plate during the disassembly process, it is necessary only to depress the tie plate against the pressure of the springs surrounding the fuel rods and then to rotate each locking sleeve on the tie rods from its locked to its unlocked position. It is then possible to remove the tie plate without disassembling the locking assembly. (LL)
Gehrels, Neil
2012-01-01
We present an overview of high energy transients in astrophysics, highlighting important advances over the past 50 years. We begin with early discoveries of gamma-ray transients, and then delve into physical details associated with a variety of phenomena. We discuss some of the unexpected transients found by Fermi and Swift, many of which are not easily classifiable or in some way challenge conventional wisdom. These objects are important insofar as they underscore the necessity of future, more detailed studies.
International Nuclear Information System (INIS)
Transient osteoporosis or transient bone marrow oedema is a rare cause of acute hip pain that predominantly affects adults of middle and younger age. We report on the MR image in 8 patients with transient bone marrow oedema of the hip and in one patient with affection of the knee joint. In three of these, sympathetic nerve blockade has been performed. The MR image after sympathicolysis is discussed. (orig.)
Heat transfer to water from a vertical tube bundle under natural-circulation conditions
International Nuclear Information System (INIS)
The natural circulation heat transfer data for longitudinal flow of water outside a vertical rod bundle are needed for developing correlations which can be used in best estimate computer codes to model thermal-hydraulic behavior of nuclear reactor cores under accident or shutdown conditions. The heat transfer coefficient between the fuel rod surface and the coolant is the key parameter required to predict the fuel temperature. Because of the absence of the required heat transfer coefficient data base under natural circulation conditions, experiments have been performed in a natural circulation loop. A seven-tube bundle having a pitch-to-diameter ratio of 1.25 was used as a test heat exchanger. A circulating flow was established in the loop, because of buoyancy differences between its two vertical legs. Steady-state and transient heat transfer measurements have been made over as wide a range of thermal conditions as possible with the system. Steady state heat transfer data were correlated in terms of relevant dimensionless parameters. Empirical correlations for the average Nusselt number, in terms of Reynolds number, Rayleigh number and the ratio of Grashof to Reynolds number are given
The THYC three-dimensional thermal-hydraulic codes for rod bundles: Recent developments and tests
International Nuclear Information System (INIS)
Pressurized water reactor (PWR) or liquid-metal fast breeder reactor cores or fuel assemblies, PWR steam generators, condensers, and tubular heat exchangers are basic components of a nuclear power plant that involve two-phase flows in tube or rod bundles. A deep knowledge of the detailed flow patterns on the shell side is necessary to evaluate departure from nucleate boiling (DNB) margins in reactor cores, singularity effects (grids, wire spacers, support plates, and baffles), corrosion on the steam generator tube sheet, bypass effects, and vibration risks. For that purpose, Electricite de France has developed since 1986 a general purpose thermal-HYdraulic Code (THYC) to study three-dimensional single- and two-phase flows in rod or tube bundles (PWR codes, steam generators, condensers, and heat exchangers). It considers the three-dimensional domain to contain two kinds of components: fluid and solids. The THYC model is obtained by space-time averaging of the instantaneous equations (mass, momentum, and energy) of each phase over control volumes including fluid and solids. The physical model of THYC is validated under several French and international experiments for single- and two-phase flows. The THYC is used for the calculation of transients such as steam-line break (coupled with a three-dimensional neutronics code), for DNB predictions, and for various steam generator or condenser studies
Analysis of CHF experiment data for finned fuel bundle
International Nuclear Information System (INIS)
The HANARO uses finned-element fuel bundles. For thermal-hydraulic safety analysis, used is the MATRA-h code which is a modified version of KAERI's MATRA-α. The subchannel analysis model was determined by using the in-core irradiation test results and hydraulic experiment results for fuel bundle. The validity of the analysis model was investigated by comparing the MATRA-h predictions with the experimental results from several bundle CHF tests. The comparison showed that the code predictions for the CHF power were very close to or less than the experimental results. Thus, it was confirmed that the subchannel analysis using MATRA-h is to be applicable to the prediction of CHF phenomenon in HANARO fuel bundle
Bundles of Norms About Teen Sex and Pregnancy.
Mollborn, Stefanie; Sennott, Christie
2015-09-01
Teen pregnancy is a cultural battleground in struggles over morality, education, and family. At its heart are norms about teen sex, contraception, pregnancy, and abortion. Analyzing 57 interviews with college students, we found that "bundles" of related norms shaped the messages teens hear. Teens did not think their communities encouraged teen sex or pregnancy, but normative messages differed greatly, with either moral or practical rationalizations. Teens readily identified multiple norms intended to regulate teen sex, contraception, abortion, childbearing, and the sanctioning of teen parents. Beyond influencing teens' behavior, norms shaped teenagers' public portrayals and post hoc justifications of their behavior. Although norm bundles are complex to measure, participants could summarize them succinctly. These bundles and their conflicting behavioral prescriptions create space for human agency in negotiating normative pressures. The norm bundles concept has implications for teen pregnancy prevention policies and can help revitalize social norms for understanding health behaviors. PMID:25387911
Zeta Functions for Elliptic Curves I. Counting Bundles
Weng, Lin
2012-01-01
To count bundles on curves, we study zetas of elliptic curves and their zeros. There are two types, i.e., the pure non-abelian zetas defined using moduli spaces of semi-stable bundles, and the group zetas defined for special linear groups. In lower ranks, we show that these two types of zetas coincide and satisfy the Riemann Hypothesis. For general cases, exposed is an intrinsic relation on automorphism groups of semi-stable bundles over elliptic curves, the so-called counting miracle. All this, together with Harder-Narasimhan, Desale-Ramanan and Zagier's result, gives an effective way to count semi-stable bundles on elliptic curves not only in terms of automorphism groups but more essentially in terms of their $h^0$'s. Distributions of zeros of high rank zetas are also discussed.
Deformation Quantization of Principal Fibre Bundles and Classical Gauge Theories
Wei\\ss, Stefan
2010-01-01
In this dissertation the notion of deformation quantization of principal fibre bundles is established and investigated in order to find a geometric formulation of classical gauge theories on noncommutative space-times. As a generalization, the notion of deformation quantization of surjective submersions is also discussed. It is shown that deformation quantizations of surjective submersions and principal fibre bundles always exist and are unique up to equivalence. These statements concerning complex-valued functions are moreover formulated and proved for sections of arbitrary vector bundles over the total space, in particular equivariant vector bundles. The commutants of the deformed right module structures within the differential operators, playing an inportant role with regard to the infinitesimal gauge transformations, are computed explicitly in each case. Depending on the choice of specific covariant derivatives and connections the commutants are isomorphic to the formal power series of the respective vert...
T-duality for circle bundles via noncommutative geometry
Mathai, Varghese
2013-01-01
Recently Baraglia showed how topological T-duality can be extended to apply not only to principal circle bundles, but also to non-principal circle bundles. We show that his results can also be recovered via two other methods: the homotopy-theoretic approach of Bunke and Schick, and the noncommutative geometry approach which we previously used for principal torus bundles. This work has several interesting byproducts, including a study of the K-theory of crossed products by Isom(R), the universal cover of O(2), and some interesting facts about equivariant K-theory for Z/2. In the final section of this paper, these results are extended to the case of bundles with singular fibers, or in other words, non-free O(2)-actions.
Steric effects induce geometric remodeling of actin bundles in filopodia
Dobramysl, Ulrich; Erban, Radek
2016-01-01
Filopodia are ubiquitous fingerlike protrusions, spawned by many eukaryotic cells, to probe and interact with their environments. Polymerization dynamics of actin filaments, comprising the structural core of filopodia, largely determine their instantaneous lengths and overall lifetimes. The polymerization reactions at the filopodial tip require transport of G-actin, which enter the filopodial tube from the filopodial base and diffuse toward the filament barbed ends near the tip. Actin filaments are mechanically coupled into a tight bundle by cross-linker proteins. Interestingly, many of these proteins are relatively short, restricting the free diffusion of cytosolic G-actin throughout the bundle and, in particular, its penetration into the bundle core. To investigate the effect of steric restrictions on G-actin diffusion by the porous structure of filopodial actin filament bundle, we used a particle-based stochastic simulation approach. We discovered that excluded volume interactions result in partial and the...
Introductory lectures on fibre bundles and topology for physicists
International Nuclear Information System (INIS)
These lectures may provide useful background material for understanding gauge theories, particularly the nonperturbative effects such as instantons and monopoles. The mathematical language of topology and fibre bundles is introduced
Mechanical Models of Microtubule Bundle Collapse in Alzheimer's Disease
Sendek, Austin; Singh, Rajiv; Cox, Daniel
2013-03-01
Amyloid-beta aggregates initiate Alzheimer's disease, and downstream trigger degradation of tau proteins that act as microtubule bundle stabilizers and mechanical spacers. Currently it is unclear which of tau cutting by proteases, tau phosphorylation, or tau aggregation are responsible for cytoskeleton degradation., We construct a percolation simulation of the microtubule bundle using a molecular spring model for the taus and including depletion force attraction between microtubules and membrane/actin cytoskeletal surface tension. The simulation uses a fictive molecular dynamics to model the motion of the individual microtubules within the bundle as a result of random tau removal, and calculates the elastic modulus of the bundle as the tau concentration falls. We link the tau removal steps to kinetic tau steps in various models of tau degradation. Supported by US NSF Grant DMR 1207624
National Partnership for Maternal Safety Consensus Bundle on Obstetric Hemorrhage.
Main, Elliott K; Goffman, Dena; Scavone, Barbara M; Low, Lisa Kane; Bingham, Debra; Fontaine, Patricia L; Gorlin, Jed B; Lagrew, David C; Levy, Barbara S
2015-01-01
Hemorrhage is the most frequent cause of severe maternal morbidity and preventable maternal mortality and therefore is an ideal topic for the initial national maternity patient safety bundle. These safety bundles outline critical clinical practices that should be implemented in every maternity unit. They are developed by multidisciplinary work groups of the National Partnership for Maternal Safety under the guidance of the Council on Patient Safety in Women's Health Care. The safety bundle is organized into 4 domains: Readiness, Recognition and Prevention, Response, and Reporting and Systems Learning. Although the bundle components may be adapted to meet the resources available in individual facilities, standardization within an institution is strongly encouraged. References contain sample resources and "Potential Best Practices" to assist with implementation. PMID:26059199
National Partnership for Maternal Safety: consensus bundle on obstetric hemorrhage.
Main, Elliott K; Goffman, Dena; Scavone, Barbara M; Low, Lisa Kane; Bingham, Debra; Fontaine, Patricia L; Gorlin, Jed B; Lagrew, David C; Levy, Barbara S
2015-07-01
Hemorrhage is the most frequent cause of severe maternal morbidity and preventable maternal mortality and therefore is an ideal topic for the initial national maternity patient safety bundle. These safety bundles outline critical clinical practices that should be implemented in every maternity unit. They are developed by multidisciplinary work groups of the National Partnership for Maternal Safety under the guidance of the Council on Patient Safety in Women's Health Care. The safety bundle is organized into four domains: Readiness, Recognition and Prevention, Response, and Reporting and System Learning. Although the bundle components may be adapted to meet the resources available in individual facilities, standardization within an institution is strongly encouraged. References contain sample resources and "Potential Best Practices" to assist with implementation. PMID:26091046
Interactive hypermedia training manual for spent-fuel bundle counters
International Nuclear Information System (INIS)
Spent-fuel bundle counters, developed by the Canadian Safeguards Support Program for the International Atomic Energy Agency, provide a secure and independent means of counting the number of irradiated fuel bundles discharged into the fuel storage bays at CANDU nuclear power stations. Paper manuals have been traditionally used to familiarize IAEA inspectors with the operation, maintenance and extensive reporting capabilities of the bundle counters. To further assist inspectors, an interactive training manual has been developed on an Apple Macintosh computer using hypermedia software. The manual uses interactive animation and sound, in conjunction with the traditional text and graphics, to simulate the underlying operation and logic of the bundle counters. This paper presents the key features of the interactive manual and highlights the advantages of this new technology for training
Topological T-duality for torus bundles with monodromy
Baraglia, David
2015-05-01
We give a simplified definition of topological T-duality that applies to arbitrary torus bundles. The new definition does not involve Chern classes or spectral sequences, only gerbes and morphisms between them. All the familiar topological conditions for T-duals are shown to follow. We determine necessary and sufficient conditions for existence of a T-dual in the case of affine torus bundles. This is general enough to include all principal torus bundles as well as torus bundles with arbitrary monodromy representations. We show that isomorphisms in twisted cohomology, twisted K-theory and of Courant algebroids persist in this general setting. We also give an example where twisted K-theory groups can be computed by iterating T-duality.
Infinitely stably extendable vector bundles on projective spaces
Coanda, Iustin
2009-01-01
According to Horrocks (1966), a vector bundle E on the projective n-space extends stably to the projective N-space, N>n, if there exists a vector bundle on the larger space whose restriction to the smaller one is isomorphic to E plus a direct sum of line bundles. We show that E extends stably to the projective N-space for every N>n if and only if E is the cohomology of a free monad (with three terms). The proof uses the method of Coanda and Trautmann (2006). Combining this result with a theorem of Mohan Kumar, Peterson and Rao (2003), we get a new effective version of the Babylonian tower theorem for vector bundles on projective spaces.
SEU43 fuel bundle shielding analysis during spent fuel transport
Energy Technology Data Exchange (ETDEWEB)
Margeanu, C. A.; Ilie, P.; Olteanu, G. [Inst. for Nuclear Research Pitesti, No. 1 Campului Street, Mioveni 115400, Arges County (Romania)
2006-07-01
The basic task accomplished by the shielding calculations in a nuclear safety analysis consist in radiation doses calculation, in order to prevent any risks both for personnel protection and impact on the environment during the spent fuel manipulation, transport and storage. The paper investigates the effects induced by fuel bundle geometry modifications on the CANDU SEU spent fuel shielding analysis during transport. For this study, different CANDU-SEU43 fuel bundle projects, developed in INR Pitesti, have been considered. The spent fuel characteristics will be obtained by means of ORIGEN-S code. In order to estimate the corresponding radiation doses for different measuring points the Monte Carlo MORSE-SGC code will be used. Both codes are included in ORNL's SCALE 5 programs package. A comparison between the considered SEU43 fuel bundle projects will be also provided, with CANDU standard fuel bundle taken as reference. (authors)
Fuel rod bundles proposed for advanced pressure tube nuclear reactors
International Nuclear Information System (INIS)
The paper aims to be a general presentation for fuel bundles to be used in Advanced Pressure Tube Nuclear Reactors (APTNR). The characteristics of such a nuclear reactor resemble those of known advanced pressure tube nuclear reactors like: Advanced CANDU Reactor (ACRTM-1000, pertaining to AECL) and Indian Advanced Heavy Water Reactor (AHWR). We have also developed a fuel bundle proposal which will be referred as ASEU-43 (Advanced Slightly Enriched Uranium with 43 rods). The ASEU-43 main design along with a few neutronic and thermalhydraulic characteristics are presented in the paper versus similar ones from INR Pitesti SEU-43 and CANDU-37 standard fuel bundles. General remarks regarding the advantages of each fuel bundle and their suitability to be burned in an APTNR reactor are also revealed. (authors)
Design and synthesis of DNA four-helix bundles
Energy Technology Data Exchange (ETDEWEB)
Rangnekar, Abhijit; Gothelf, Kurt V [Department of Chemistry, Centre for DNA Nanotechnology (CDNA) and Interdisciplinary Nanoscience Center (iNANO), Aarhus University, DK-8000 Aarhus C (Denmark); LaBean, Thomas H, E-mail: kvg@chem.au.dk, E-mail: thl@cs.duke.edu [Department of Chemistry, Duke University, Durham, NC 27708 (United States)
2011-06-10
The field of DNA nanotechnology has evolved significantly in the past decade. Researchers have succeeded in synthesizing tile-based structures and using them to form periodic lattices in one, two and three dimensions. Origami-based structures have also been used to create nanoscale structures in two and three dimensions. Design and construction of DNA bundles with fixed circumference has added a new dimension to the field. Here we report the design and synthesis of a DNA four-helix bundle. It was found to be extremely rigid and stable. When several such bundles were assembled using appropriate sticky-ends, they formed micrometre-long filaments. However, when creation of two-dimensional sheet-like arrays of the four-helix bundles was attempted, nanoscale rings were observed instead. The exact reason behind the nanoring formation is yet to be ascertained, but it provides an exciting prospect for making programmable circular nanostructures using DNA.
Wire-wrapped rod-bundle heat-transfer analysis for LMFBR
International Nuclear Information System (INIS)
Helical wire wraps are widely used in the LMFBR fuel and blanket assemblies to provide coolant mixing and maintain proper spacing between fuel pins. The presence of the helical wire, however, may possibly induce heat transfer problems, such as the uncertainty of the maximum clad temperature as a result of the contact between the wires and the pins. In this study, the detailed transient three dimensional velocity and temperature distributions for the coolant around the pin will be determined by solving the governing momentum and energy equation numerically. A computer code HEATRAN has been developed to perform this calculation. Before the computer code HEATRAN is applied to the wire wrapped rod bundle problem, it is used to analyze a wide range of fluid and heat transfer problem to verify its capabilities
International Nuclear Information System (INIS)
The Tunnel Sealing Experiment (TSX) was a two-phase international project funded by Canada, Japan, France, and the United States. The first phase was pressurizing the TSX chamber to 4 MPa to investigate the ability of clay and concrete bulkheads to reduce hydraulic flows. The second phase involved circulating heated water through the chamber to evaluate the influence of elevated temperature on the performance of the bulkheads and adjacent rock. A numerical analysis to simulate thermohydraulic evolution of the bulkheads and surrounding rock of the TSX was conducted to help in understanding the physical test process and the interaction between heat and pore pressure evolutions. The simulated rock temperature matched the measured data quite well; however the simulated bulkhead temperatures were greater than the measured temperatures. The difference may have been caused by entrapped air or formation of microchannels in the chamber sand, which would decrease the amount of heat reaching the bulkheads. The simulated thermally induced pore pressure increase in the clay bulkhead reasonably matched the measured data for the saturated portion. The difference in magnitude between simulated and measured rock pore pressures indicates that thermo hydraulic simulation should be coupled with a mechanical component when the stiffness of the media is large and hydraulic conductivity is low. (author)
Energy Technology Data Exchange (ETDEWEB)
Arndt, S.A.
1997-07-01
The real-time reactor simulation field is currently at a crossroads in terms of the capability to perform real-time analysis using the most sophisticated computer codes. Current generation safety analysis codes are being modified to replace simplified codes that were specifically designed to meet the competing requirement for real-time applications. The next generation of thermo-hydraulic codes will need to have included in their specifications the specific requirement for use in a real-time environment. Use of the codes in real-time applications imposes much stricter requirements on robustness, reliability and repeatability than do design and analysis applications. In addition, the need for code use by a variety of users is a critical issue for real-time users, trainers and emergency planners who currently use real-time simulation, and PRA practitioners who will increasingly use real-time simulation for evaluating PRA success criteria in near real-time to validate PRA results for specific configurations and plant system unavailabilities.
International Nuclear Information System (INIS)
The real-time reactor simulation field is currently at a crossroads in terms of the capability to perform real-time analysis using the most sophisticated computer codes. Current generation safety analysis codes are being modified to replace simplified codes that were specifically designed to meet the competing requirement for real-time applications. The next generation of thermo-hydraulic codes will need to have included in their specifications the specific requirement for use in a real-time environment. Use of the codes in real-time applications imposes much stricter requirements on robustness, reliability and repeatability than do design and analysis applications. In addition, the need for code use by a variety of users is a critical issue for real-time users, trainers and emergency planners who currently use real-time simulation, and PRA practitioners who will increasingly use real-time simulation for evaluating PRA success criteria in near real-time to validate PRA results for specific configurations and plant system unavailabilities