Validation of the sub-channel code F-COBRA-TF: Part II. Recalculation of void measurements

The present paper is a continuation of Part I: “Recalculation of single-phase and two-phase pressure loss measurements”. It deals with recalculations of void distribution measurements with the advanced two-phase, three-field sub-channel code F-COBRA-TF. Again, experimental data of both the OECD/NRC BFBT benchmark and in-house tests in AREVA NP's KATHY loop are used.The results of the recalculations of the measurements especially demonstrate the capability of a three-field code to predict void fractions with good accuracy, whereas the code is not based on a conventional void correlation which derives void fraction from quality according to an empirical function. In fact, the code relies on interfacial friction correlations for each flow regime. The quantities volume fraction of continuous liquid, volume fraction of entrained liquid and volume fraction of vapor are variables in the basic transport equations, which it is directly solved for. Thereby, the F-COBRA-TF standard models - which are usually applied for all sorts of calculations (pressure loss, void distribution, lateral mixing, critical heat flux, etc.) - were used. As already in Part I of the present paper, it was not necessary to do special code tuning with respect to certain experiments.     

Data requirements for neutron activation: Part I: Cross sections

Existing neutron cross section libraries such as EAF-2003 contain a large number of reactions. Of these only a small proportion are significant in the production of the nuclides that dominate activation quantities such as activity or γ-dose rate at various decay times. These reactions were identified in a recent extensive set of calculations on all the elements and reported in the ‘Activation Handbook’. These 1340 reactions are considered here and the set of reactions with no differential or integral data are listed and prioritised. A set of 339 reactions with very limited possibility of measurement are given; these require further study by theoretical calculations. A set of 100 reactions with either discrepant or no data, that might be measurable are listed and are recommended to be the subject of future experimental work.     

Heat transfer in two-component internal mist cooling: Part I. Experimental investigation

An experimental investigation was conducted to examine steady, internal, nozzle-generated, gas/liquid mist cooling in vertical channels. The ideal primary cooling mechanism in this situation is surface evaporation of an ultra-thin, subcooled liquid film that forms on the heated surface. The aim was to quantify the effects of various operating and design parameters on the cooling effectiveness. Parameters tested included the liquid atomization nozzle design, inlet flow condition (liquid mass fraction; carrier gas velocity, temperature and humidity; liquid temperature; liquid droplet size distribution; and gas/liquid combination), channel characteristics (cross-section geometry, length and surface wettability), and flow direction. Interest in this research has been motivated by the need for a highly efficient cooling mechanism in high-power lasers for inertial fusion reactor applications. A fully instrumented experimental test facility that included three cylindrical and two rectangular electrically heated test sections with different cross-sections and unheated entry lengths was used. The channel hydraulic diameters covered the range 16-26.7 mm, and the heated length-to-hydraulic-diameter ratio varied in the range from 23.3 to 51. Water was used as the mist liquid, with air or helium as the carrier gas. Three types of mist generating nozzles with significantly different spray characteristics were used. Local heat transfer coefficients, defined based on the temperature difference between the heated surface and the bulk gas, were obtained along the channels for a wide range of operating conditions. The data indicate that mist cooling can increase the heat transfer coefficient by more than an order of magnitude compared to forced convection using only the carrier gas.     

Data requirements for neutron activation: Part II: Decay data

Existing data libraries such as EAF-2003 contain information for a large number of nuclides. Of these only a proportion makes a significant contribution to quantities such as activity or γ-dose rate at various decay times. These nuclides were identified in a recent extensive set of calculations on all the elements and reported in the ‘Activation Handbook’. These 754 nuclides are considered here and the set of nuclides judged to have poorly known decay data are listed and prioritised. A set of 338 prioritised nuclides are given and the most important of these are recommended for measurement and evaluation work in the foreseeable future.     

Optimized thick-wall cylinders by virtue of Poisson's ratio selection: Part 1: Isochoric pressure application

The principal stress distributions in thick-wall cylinders due to variation in the Poisson's ratio are predicted using analytical and finite element methods. Analyses of appropriate brittle and ductile failure criteria show that under the isochoric pressure conditions investigated that auextic (i.e. those possessing a negative Poisson's ratio) materials act as stress concentrators; hence they are predicted to fail before their conventional (i.e. possessing a positive Poisson's ratio) material counterparts. The key finding of the work presented shows that for constrained thick-wall cylinders the maximum tensile principal stress can vanish at a particular Poisson's ratio and aspect ratio. This phenomenon is exploited in order to present an optimized design criterion for thick-wall cylinders. Moreover, via the use of a cogent finite element model, this criterion is also shown to be applicable for the design of micro-porous materials.     

Heat transfer in two-component internal mist cooling: Part II. Mechanistic modeling

The prediction of a mechanistic, three-dimensional, two-phase flow model is compared with experimental heat transfer data presented in the experimental part of this study for steady, internal, nozzle-generated, gas/liquid mist flow in vertical channels. The mechanistic model is based on the modification of the KIVA-3V computer code. The KIVA-3V code has been modified to solve the heat conduction equation in the surrounding structure with either steady or pulsed heat generation simultaneously with the fluid transport equations, and allow modeling of the various channel geometries and droplet injection methods. Among the numerically examined operating and design parameters are: the liquid atomization nozzle design, heat flux, carrier gas velocity and inlet temperature, liquid mass fraction at inlet, and flow direction. Comparison is made between the experimental data for wall and fluid bulk temperatures and heat transfer coefficients, and the predictions of the numerical model. Overall, reasonable agreement is obtained for downward mist flow, in particular at moderate heat fluxes; at high heat fluxes, the model slightly underpredicts the local heat transfer coefficients. For upward mist flow, the model underpredicts the local heat transfer coefficients typically by about 20%, and appears to predict dryout at the test section exit earlier than experiment. Some parametric and sensitivity calculation results are also presented and discussed.     

Two-group interfacial area transport in vertical air-water flow: I. Mechanistic model

The prediction of the dynamical evolution of interfacial area concentration is one of the most challenging tasks in two-fluid model application. This paper is focused on developing theoretical models for interfacial area source and sink terms for a two-group interfacial area transport equation. Mechanistic models of major fluid particle interaction phenomena involving two bubble groups are proposed, including the shearing-off of small bubbles from slug/cap bubbles, the wake entrainment of spherical/distorted bubble group into slug/cap bubble group, the wake acceleration and coalescence between slug/cap bubbles, and the breakup of slug/cap bubbles due to turbulent eddy impacts. The existing one-group interaction terms are extended in considering the generation of cap bubbles, as well as different parametric dependences when these terms are applied to the slug flow regime. The complete set of modeling equations is closed and continuously covers the bubbly flow, slug flow, and churn-turbulent flow regimes. Prediction of the interfacial area concentration evolution using a one-dimensional two-group transport equation and evaluation with experimental results are described in a companion paper.     

Neutron irradiation of sapphire for compressive strengthening.: I. Processing conditions and compressive strength

Sapphire suffers a dramatic loss of c-axis compression strength at elevated temperatures. Irradiation of sapphire with fission-spectrum neutrons to an exposure of ∼10²² neutrons/m² in the core of a 1 MW fission reactor increased the c-axis compression strength by a factor of ∼3 at 600 °C. Strength was similarly improved when 99% of slow neutrons (?0.1 eV) were removed by 10B and Cd shields during irradiation. Annealing at 600 °C for 10 min changed the yellow-brown color of irradiated sapphire to pale yellow, but had no effect on compressive strength. Annealing irradiated sapphire at 1200 °C for 24 h reduced the compressive strength to its baseline value. Transmission electron microscopy suggests that fast-neutron-induced displacement damage inhibits the propagation of r-plane twins which are responsible for the low compressive strength. When irradiated with 10B and Cd shielding, sapphire that was not grown in iridium crucibles is safe for unrestricted handling after 1 month.     

The stability boundary for supercritical flow in natural convection loops: Part I: H2O studies

Non-dimensional parameters governing supercritical flow instability boundary, derived in an earlier study, are now examined through a numerical experiment comprising 94 simulated cases. Part I of this study reports on the stability of supercritical light water in a natural-convection loop and confirms the validity of these non-dimensional parameters for stability predictions. Part II reports on supercritical CO₂ and H₂.     

On the dissemination of scientific information: Part I The role of the European community in disseminating knowledge

The European Community has clearly defined responsibilities for disseminating knowledge originating with its activities in the field of nuclear research and development. The extent and the significance of the dissemination of knowledge as exerted by the services of the Community are described with special reagrd to the emphasis given to improving the access to nuclear information by creating the Euratom Nuclear Documentation System, which is encompassing more than 1.2 million data on nuclear literature and, on the other hand, to sponsoring of and participating in conferences on selected topics and publishing their proceedings.     

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 1: Test model and pressure test

In Japan, the Nuclear Power Engineering Corporation (NUPEC), sponsored by the Ministry of Economy, Trade and Industry (METI), has been conducting a series of seismic reliability proving tests using full-scale or close to full-scale models to simulate actual important equipment that is critical for seismic safety of nuclear power plants. The tests are intended to validate the seismic design and reliability with a sufficient margin even under destructive earthquakes. A series of tests was carried out on a reinforced concrete containment vessel (RCCV) for advanced boiling water reactor (ABWR) from 1992 to 1999. A large-scale high-performance shaking table at Tadotsu Engineering Laboratory, the largest in the world, was used for this test. Part 1 reports the test model and the results of pressure and leak tests. Part 2 describes test procedures, input waves and the results of verification tests such as changes of stiffness, characteristic frequency and damping ratio, the failure of the model and the load deflection. Part 3 shows the seismic safety margin that was evaluated from the energy input during the failure test to a design basis earthquake. Part 4 reports simulation analysis results by a stick model with lumped masses.     

A study of the effects of the location of flashing inception on maximum and minimum critical two-phase flow rates:: Part II: analysis and modelling

A general unified model is developed to predict one-component critical two-phase pipe flow. An extension of the Henry [Henry, R.E., 1970. The Two-Phase Critical Discharge of Initially Saturated or Subcooled Liquid. Nucl. Sci. Eng. 41, 336-342.] and Henry and Fauske [Henry, R.E., Fauske, H.K., 1970. The two-Phase critical Flow of One-Component Mixtures in Nozzles; Orifices and Short Tubes, ASME J. Heat Transfer, May 1970.] models to incorporate the effects of wall friction and the location of flashing inception is proposed. Modelling of the two-phase flow is accomplished by describing the evolution of the flow between the location of flashing inception and the exit (critical) plane. The model approximates the nonequilibrium phase change process via thermodynamic equilibrium paths. Included are the relative effects of varying the location of flashing inception, pipe geometry, fluid properties and length to diameter ratio. The model predicts that a range of critical mass fluxes exist and is bound by a maximum and minimum value for a given thermodynamic state. This range is more pronounced at lower subcooled stagnation states and can be attributed to the variation in the location of flashing inception. The model is based on the experimental study of critical two-phase flow rates of saturated and subcooled water through long tubes given in Part I of this work. In that study, the location of flashing inception was accurately controlled and adjusted through the use of a new device. The data obtained revealed that for fixed stagnation conditions, the maximum critical mass fluxes occurred with flashing inception located near the pipe exit; while minimum critical mass fluxes occurred with the flashing front located further upstream. The results of the present study, as well as available data since 1970 are compared with the model predictions. These data cover a wide range of conditions and include test section L/D ratios from 25 to 302 and a temperature and pressure range of 110-280°C and 0.16-6.9 Mpa, respectively. The predicted maximum and minimum critical mass fluxes show an excellent agreement with the range observed in the experimental data.     

Pressurized water reactor (PWR) hot-leg streaming: Part 1: Computational fluid dynamics (CFD) simulations

A thermo-hydraulic analysis model was developed to analyze thermal stratification phenomena observed in the hot-legs of pressurized water reactors (PWR). The model uses VIPREW code to determine the flow field and temperature distribution in the reactor fuel region. The temperature readings from the thermal couples located at the exit of the reactor core were used to compare with the VIPREW computed results. The predicted values agree well with the measurements. The VIPREW results are then used as the boundary conditions for the CFD analysis. The CFD computational domain includes the upper plenum and hot-legs and the fifty two (52) control rod guiding tubes to properly include the additional obstructions imposed to the fluid. Different fuel loading patterns were studied to investigate the effects of different power distribution and fuel channel exit water temperature on hot-leg thermal stratification magnitude. The analysis results show that the 52 control rod guide tubes have major contribution to the mixing effect in the upper plenum. The sudden expansion of the cross sectional area in the upper plenum leads to the formation of recirculation vortex that prolongs the duration of coolant in the reactor vessel. The hotter coolant from the center portion tends to flow upwards to the top before exiting at the upper portion of the hot-leg pipes. It leads to higher temperature in the upper portion of the hot-legs. Water from the cooler outer fuel channels tends to trap in the recirculation region before exiting from the lower portion of the hot-legs.     

Heat transfer to liquid metals flowing turbulently and longitudinally through closely spaced rod bundles : Part I

This paper presents results of a theoretical study of heat transfer to liquid metals in fully developed turbulent, in-line flow through unbaffled, spacer-free rod bundles. The bundles have equilateral triangular arrangement; and the rod spacings, rod design, and ranges of independent variables covered were chosen with reference to liquid-metal-cooled nuclear reactor applications. Three different sets of thermal boundary conditions are considered: (A) uniform heat flux in the axial direction with uniform temperature in the circumferential direction, on the outer surface of the cladding; (B) uniform heat flux in both directions, on the outer surface of the cladding; and (C) uniform heat flux in both directions on the inner surface of the cladding. The results of the third set are presented in Part II.     

Theoretical transition probabilities and lifetimes in Kr I and Xe I spectra

Transition probabilities and lifetimes are computed for Kr I and Xe I spectra. The values are obtained in a central field model which takes into account intermediate coupling and configuration mixing. Theoretical values are compared with previous experimental and theoretical data. Bibliographies, made in January 1978, on previous works are given.     

New genetic algorithms (GA) to optimize PWR reactors: Part I: Loading pattern and burnable poison placement optimization techniques for PWRs

The objective of this study was to develop a unique scientific methodology as well as a practical tool for designing the loading pattern (LP) and burnable poison (BP) pattern for a given Pressurized Water Reactor (PWR) core. Because of the large number of possible combinations for the fuel assembly (FA) loading in the core, the design of the core configuration is a complex optimization problem. It requires finding an optimal FA arrangement and corresponding BP placement design that will achieve maximum cycle length while satisfying the safety constraints. To solve this optimization problem, a core reload optimization package, GARCO (Genetic Algorithm Reactor Code Optimization) code is developed. This code is applicable for all types of PWR cores having different geometries and designs with an unlimited number of FA types in the inventory. GARCO has three modes: the user can optimize the core configuration (LP pattern) with or without BPs in the first mode; the second mode is the optimization of BP placement in the core and the last mode is the user can optimize LP and BP placements simultaneously in mode 3. In this study, the first mode finds the optimal LPs using the Haling Power Depletion Method (HPD) for placing BPs in the core. The second mode, which depletes the core accurately, places BPs in the selected optimum LP pattern. This methodology is applied only to the TMI-1 PWR. However, the improved Mode 1 GA option was applied to both the VVER-1000 and the TMI-1 to demonstrate and verify the advantages of the new enhancements in optimizing the LP pattern only. The “Moby-Dick” code is used as reactor physics code for VVER-1000 analysis in this research. The SIMULATE-3 code, which is an advanced two-group nodal code, is used to analyze the TMI-1. The libraries of the BP designs used in SIMULATE-3 in this study were produced by Yilmaz (2005) [Yilmaz, S., 2005. Multilevel optimization of burnable poison utilization for advanced PWR fuel management. Ph.D. Thesis in Nuclear Engineering. the Pennsylvania State University].     

The stability boundary for supercritical flow in natural-convection loops: Part II: CO2 and H2

The universality of non-dimensional parameters that govern the stability boundary of supercritical flow in natural-convection loops is examined through 146 numerical experiments. Part I reported studies of supercritical H₂O. This part extends that study to CO₂ and H₂. The finding is that the derived non-dimensional parameters appear to be valid regardless of the fluid, the geometric dimensions or the flow conditions.     

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 2: Results of shaking table tests

A 1/8-scale model was constructed of a reinforced concrete containment vessel (RCCV) used in the latest advanced boiling water reactors (ABWR). Shaking table tests were conducted on it with input motions corresponding to or exceeding a design earthquake assumed for a real Nuclear Power Plant.The objectives of the tests were to verify the structural integrity and the leak-proof functional soundness of the RCCV subjected to design earthquakes, and to determine the ultimate strength and seismic margin by an excitation that led to the model's collapse. The model, the test sequence and the pressure and leak test results were addressed in Part 1. The shaking table test method, the input motions and the test results, including the transition of the model's stiffness, natural frequencies and damping factors and the effects of vertical input motions and internal pressure on the model's characteristics and behavior, the load-deformation, the ultimate strength, the failure mode of the reinforced concrete portion and the liner plate are described here. The seismic safety margin that was evaluated by the energy input during the failure test to a design basis earthquake will be described in Part 3. The analytical results of simulation using the multi-lumped mass model will be described in Part 4.