Assessment of CTF boiling transition and critical heat flux modeling capabilities using the OECD/NRC BFBT and PSBT benchmark databases

The need to refine models for best-estimate calculations, based on good-quality experimental data, has been expressed in many recent meetings in the field of nuclear applications. The modeling needs arising in this respect should not be limited to the currently available macroscopic methods but should be extended to next-generation analysis techniques that focus on more microscopic processes. One of the most valuable databases identified for the thermalhydraulics modeling was developed by the Nuclear Power Engineering Corporation (NUPEC), Japan. From 1987 to 1995, NUPEC performed steady-state and transient critical power and departure from nucleate boiling (DNB) test series based on the equivalent full-size mock-ups. Considering the reliability not only of the measured data, but also other relevant parameters such as the system pressure, inlet sub-cooling and rod surface temperature, these test series supplied the first substantial database for the development of truly mechanistic and consistent models for boiling transition and critical heat flux. Over the last few years the Pennsylvania State University (PSU) under the sponsorship of the U.S. Nuclear Regulatory Commission (NRC) has prepared, organized, conducted and summarized the OECD/NRC Full-size Fine-mesh Bundle Tests (BFBT) Benchmark. The international benchmark activities have been conducted in cooperation with the Nuclear Energy Agency/Organization for Economic Co-operation and Development (NEA/OECD) and Japan Nuclear Energy Safety (JNES) organization, Japan. Consequently, the JNES has made available the Boiling Water Reactor (BWR) NUPEC database for the purposes of the benchmark. Based on the success of the OECD/NRC BFBT benchmark the JNES has decided to release also the data based on the NUPEC Pressurized Water Reactor (PWR) subchannel and bundle tests for another follow-up international benchmark entitled OECD/NRC PWR Subchannel and Bundle Tests (PSBT) benchmark. This paper presents an application of the joint Penn State University/Technical University of Madrid (UPM) version of the well-known subchannel code COBRA-TF, namely CTF, to the critical power and departure from nucleate boiling (DNB) exercises of the OECD/NRC BFBT and PSBT benchmarks

Assessment of CTF boiling transition and critical heat flux modeling capabilities using the OECD/NRC BFBT and PSBT benchmark databases

Over the last few years, the Pennsylvania State University (PSU) under the sponsorship of the US Nuclear Regulatory Commission (NRC) has prepared, organized, conducted, and summarized two international benchmarks based on the NUPEC data—the OECD/NRC Full-Size Fine-Mesh Bundle Test (BFBT) Benchmark and the OECD/NRC PWR Sub-Channel and Bundle Test (PSBT) Benchmark. The benchmarks’ activities have been conducted in cooperation with the Nuclear Energy Agency/Organization for Economic Co-operation and Development (NEA/OECD) and the Japan Nuclear Energy Safety (JNES) Organization. This paper presents an application of the joint Penn State University/Technical University of Madrid (UPM) version of the well-known sub-channel code COBRA-TF (Coolant Boiling in Rod Array-Two Fluid), namely, CTF, to the steady state critical power and departure from nucleate boiling (DNB) exercises of the OECD/NRC BFBT and PSBT benchmarks. The goal is two-fold: firstly, to assess these models and to examine their strengths and weaknesses; and secondly, to identify the areas for improvement.

Nonlocal electron heat flux revisited

A known nonlocal model of electron heat flux, applying for (scale length/thermal ion-electron mean-free path) of order Z)1/2(e*/T)312, ionization number Z, large, and e*~ 6.5 T (the energy of electrons carrying most of the flux), is reconsidered. The large e*/T ratio simplifies the complete formalism. A simple flux formula, exact for both smooth and steep profiles, is given. Thermoelectric effects and other models are discussed.

Self-consistent, nonlocal electron heat flux at arbitrary ion charge number

A single, nonlocal expression for the electron heat flux, which closely reproduces known results at high and low ion charge number 2, and “exact” results for the local limit at all 2, is derived by solving the kinetic equation in a narrow, tail-energy range. The solution involves asymptotic expansions of Bessel functions of large argument, and (Z-dependent)order above or below it, corresponding to the possible parabolic or hyperbolic character of the kinetic equation; velocity space diffusion in self-scattering is treated similarly to isotropic thermalization of tail energies in large Z analyses. The scale length H characterizing nonlocal effects varies with Z, suggesting an equal dependence of any ad hoc flux limiter. The model is valid for all H above the mean-free path for thermal electrons.

Self-similar expansion of laser plasmas with nonlocal heat flux

A previous hydrodynamic model of the expansion of a laser-produced plasma, using classical (Spitzer) heat flux, is reconsidered with a nonlocal heat flux model. The nonlocal law is shown to be valid beyond the range of validity of the classical law, breaking down ultimately, however, in agreement with recent predictions.

Nonlocal electron heat-flux

Electron thermal conduction in a not quite collisional unmagnetlzed plasma is analysed. The failure of classical results for temperature scale-length up to 100 times larger than thermal mean-free-path for electron scattering, and large ion-charge number Z , is discussed. Recent results from a nonlocal model of conduction at large Z are reviewed. Closed form expressions for Braginskii's coefficients a ,/3 , y for Z =0(1) are derived. An extension of the nonlocal model for Z =0(1) is discussed.

High power beam dump project for the accelerator prototype LIPAc: Cooling design and analysis

En el campo de la fusión nuclear y desarrollándose en paralelo a ITER (International Thermonuclear Experimental Reactor), el proyecto IFMIF (International Fusion Material Irradiation Facility) se enmarca dentro de las actividades complementarias encaminadas a solucionar las barreras tecnológicas que aún plantea la fusión. En concreto IFMIF es una instalación de irradiación cuya misión es caracterizar materiales resistentes a condiciones extremas como las esperadas en los futuros reactores de fusión como DEMO (DEMOnstration power plant). Consiste de dos aceleradores de deuterones que proporcionan un haz de 125 mA y 40 MeV cada uno, que al colisionar con un blanco de litio producen un flujo neutrónico intenso (1017 neutrones/s) con un espectro similar al de los neutrones de fusión [1], [2]. Dicho flujo neutrónico es empleado para irradiar los diferentes materiales candidatos a ser empleados en reactores de fusión, y las muestras son posteriormente examinadas en la llamada instalación de post-irradiación. Como primer paso en tan ambicioso proyecto, una fase de validación y diseño llamada IFMIFEVEDA (Engineering Validation and Engineering Design Activities) se encuentra actualmente en desarrollo. Una de las actividades contempladas en esta fase es la construcción y operación de una acelarador prototipo llamado LIPAc (Linear IFMIF Prototype Accelerator). Se trata de un acelerador de deuterones de alta intensidad idéntico a la parte de baja energía de los aceleradores de IFMIF. Los componentes del LIPAc, que será instalado en Japón, son suministrados por diferentes países europeos. El acelerador proporcionará un haz continuo de deuterones de 9 MeV con una potencia de 1.125 MW que tras ser caracterizado con diversos instrumentos deberá pararse de forma segura. Para ello se requiere un sistema denominado bloque de parada (Beam Dump en inglés) que absorba la energía del haz y la transfiera a un sumidero de calor. España tiene el compromiso de suministrar este componente y CIEMAT (Centro de Investigaciones Energéticas Medioambientales y Tecnológicas) es responsable de dicha tarea. La pieza central del bloque de parada, donde se para el haz de iones, es un cono de cobre con un ángulo de 3.5o, 2.5 m de longitud y 5 mm de espesor. Dicha pieza está refrigerada por agua que fluye en su superficie externa por el canal que se forma entre el cono de cobre y otra pieza concéntrica con éste. Este es el marco en que se desarrolla la presente tesis, cuyo objeto es el diseño del sistema de refrigeración del bloque de parada del LIPAc. El diseño se ha realizado utilizando un modelo simplificado unidimensional. Se han obtenido los parámetros del agua (presión, caudal, pérdida de carga) y la geometría requerida en el canal de refrigeración (anchura, rugosidad) para garantizar la correcta refrigeración del bloque de parada. Se ha comprobado que el diseño permite variaciones del haz respecto a la situación nominal siendo el flujo crítico calorífico al menos 2 veces superior al nominal. Se han realizado asimismo simulaciones fluidodinámicas 3D con ANSYS-CFX en aquellas zonas del canal de refrigeración que lo requieren. El bloque de parada se activará como consecuencia de la interacción del haz de partículas lo que impide cualquier cambio o reparación una vez comenzada la operación del acelerador. Por ello el diseño ha de ser muy robusto y todas las hipótesis utilizadas en la realización de éste deben ser cuidadosamente comprobadas. Gran parte del esfuerzo de la tesis se centra en la estimación del coeficiente de transferencia de calor que es determinante en los resultados obtenidos, y que se emplea además como condición de contorno en los cálculos mecánicos. Para ello por un lado se han buscado correlaciones cuyo rango de aplicabilidad sea adecuado para las condiciones del bloque de parada (canal anular, diferencias de temperatura agua-pared de decenas de grados). En un segundo paso se han comparado los coeficientes de película obtenidos a partir de la correlación seleccionada (Petukhov-Gnielinski) con los que se deducen de simulaciones fluidodinámicas, obteniendo resultados satisfactorios. Por último se ha realizado una validación experimental utilizando un prototipo y un circuito hidráulico que proporciona un flujo de agua con los parámetros requeridos en el bloque de parada. Tras varios intentos y mejoras en el experimento se han obtenido los coeficientes de película para distintos caudales y potencias de calentamiento. Teniendo en cuenta la incertidumbre de las medidas, los valores experimentales concuerdan razonablemente bien (en el rango de 15%) con los deducidos de las correlaciones. Por motivos radiológicos es necesario controlar la calidad del agua de refrigeración y minimizar la corrosión del cobre. Tras un estudio bibliográfico se identificaron los parámetros del agua más adecuados (conductividad, pH y concentración de oxígeno disuelto). Como parte de la tesis se ha realizado asimismo un estudio de la corrosión del circuito de refrigeración del bloque de parada con el doble fin de determinar si puede poner en riesgo la integridad del componente, y de obtener una estimación de la velocidad de corrosión para dimensionar el sistema de purificación del agua. Se ha utilizado el código TRACT (TRansport and ACTivation code) adaptándalo al caso del bloque de parada, para lo cual se trabajó con el responsable (Panos Karditsas) del código en Culham (UKAEA). Los resultados confirman que la corrosión del cobre en las condiciones seleccionadas no supone un problema. La Tesis se encuentra estructurada de la siguiente manera: En el primer capítulo se realiza una introducción de los proyectos IFMIF y LIPAc dentro de los cuales se enmarca esta Tesis. Además se describe el bloque de parada, siendo el diseño del sistema de rerigeración de éste el principal objetivo de la Tesis. En el segundo y tercer capítulo se realiza un resumen de la base teórica así como de las diferentes herramientas empleadas en el diseño del sistema de refrigeración. El capítulo cuarto presenta los resultados del relativos al sistema de refrigeración. Tanto los obtenidos del estudio unidimensional, como los obtenidos de las simulaciones fluidodinámicas 3D mediante el empleo del código ANSYS-CFX. En el quinto capítulo se presentan los resultados referentes al análisis de corrosión del circuito de refrigeración del bloque de parada. El capítulo seis se centra en la descripción del montaje experimental para la obtención de los valores de pérdida de carga y coeficiente de transferencia del calor. Asimismo se presentan los resultados obtenidos en dichos experimentos. Finalmente encontramos un capítulo de apéndices en el que se describen una serie de experimentos llevados a cabo como pasos intermedios en la obtención del resultado experimental del coeficiente de película. También se presenta el código informático empleado para el análisis unidimensional del sistema de refrigeración del bloque de parada llamado CHICA (Cooling and Heating Interaction and Corrosion Analysis). ABSTRACT In the nuclear fusion field running in parallel to ITER (International Thermonuclear Experimental Reactor) as one of the complementary activities headed towards solving the technological barriers, IFMIF (International Fusion Material Irradiation Facility) project aims to provide an irradiation facility to qualify advanced materials resistant to extreme conditions like the ones expected in future fusion reactors like DEMO (DEMOnstration Power Plant). IFMIF consists of two constant wave deuteron accelerators delivering a 125 mA and 40 MeV beam each that will collide on a lithium target producing an intense neutron fluence (1017 neutrons/s) with a similar spectra to that of fusion neutrons [1], [2]. This neutron flux is employed to irradiate the different material candidates to be employed in the future fusion reactors, and the samples examined after irradiation at the so called post-irradiative facilities. As a first step in such an ambitious project, an engineering validation and engineering design activity phase called IFMIF-EVEDA (Engineering Validation and Engineering Design Activities) is presently going on. One of the activities consists on the construction and operation of an accelerator prototype named LIPAc (Linear IFMIF Prototype Accelerator). It is a high intensity deuteron accelerator identical to the low energy part of the IFMIF accelerators. The LIPAc components, which will be installed in Japan, are delivered by different european countries. The accelerator supplies a 9 MeV constant wave beam of deuterons with a power of 1.125 MW, which after being characterized by different instruments has to be stopped safely. For such task a beam dump to absorb the beam energy and take it to a heat sink is needed. Spain has the compromise of delivering such device and CIEMAT (Centro de Investigaciones Energéticas Medioambientales y Tecnológicas) is responsible for such task. The central piece of the beam dump, where the ion beam is stopped, is a copper cone with an angle of 3.5o, 2.5 m long and 5 mm width. This part is cooled by water flowing on its external surface through the channel formed between the copper cone and a concentric piece with the latter. The thesis is developed in this realm, and its objective is designing the LIPAc beam dump cooling system. The design has been performed employing a simplified one dimensional model. The water parameters (pressure, flow, pressure loss) and the required annular channel geometry (width, rugoisty) have been obtained guaranteeing the correct cooling of the beam dump. It has been checked that the cooling design allows variations of the the beam with respect to the nominal position, being the CHF (Critical Heat Flux) at least twice times higher than the nominal deposited heat flux. 3D fluid dynamic simulations employing ANSYS-CFX code in the beam dump cooling channel sections which require a more thorough study have also been performed. The beam dump will activateasaconsequenceofthe deuteron beam interaction, making impossible any change or maintenance task once the accelerator operation has started. Hence the design has to be very robust and all the hypotheses employed in the design mustbecarefully checked. Most of the work in the thesis is concentrated in estimating the heat transfer coefficient which is decisive in the obtained results, and is also employed as boundary condition in the mechanical analysis. For such task, correlations which applicability range is the adequate for the beam dump conditions (annular channel, water-surface temperature differences of tens of degrees) have been compiled. In a second step the heat transfer coefficients obtained from the selected correlation (Petukhov- Gnielinski) have been compared with the ones deduced from the 3D fluid dynamic simulations, obtaining satisfactory results. Finally an experimental validation has been performed employing a prototype and a hydraulic circuit that supplies a flow with the requested parameters in the beam dump. After several tries and improvements in the experiment, the heat transfer coefficients for different flows and heating powers have been obtained. Considering the uncertainty in the measurements the experimental values agree reasonably well (in the order of 15%) with the ones obtained from the correlations. Due to radiological reasons the quality of the cooling water must be controlled, hence minimizing the copper corrosion. After performing a bibligraphic study the most adequate water parameters were identified (conductivity, pH and dissolved oxygen concentration). As part of this thesis a corrosion study of the beam dump cooling circuit has been performed with the double aim of determining if corrosion can pose a risk for the copper beam dump , and obtaining an estimation of the corrosion velocitytodimension the water purification system. TRACT code(TRansport and ACTivation) has been employed for such study adapting the code for the beam dump case. For such study a collaboration with the code responsible (Panos Karditsas) at Culham (UKAEA) was established. The work developed in this thesis has supposed the publication of three articles in JCR journals (”Journal of Nuclear Materials” y ”Fusion Engineering and Design”), as well as presentations in more than four conferences and relevant meetings.

Ion charge number and flux saturation effects in the corona of a laser-irradiated pellet

The quasisteady structure of the corona of a laser-irradiated pellet is completely determined for arbitrary Z, (ion charge number} and re/ra (ratio of critical and ablation radii), and for heat-flux saturation factor/above approximately 0.04. The ion-to-electron temperature ratio at rc grows sensibly with Z,; all other quantities depend weakly and nonmonotonically on Z,. For rc /ra close to unity, and all Z, of interest (Z, < 47}, the flow is subsonic at rc. For a given laser power W, flux saturation may decrease (low/) or increase (high/) the ablation pressure Pa relative to the value obtained when saturation is not considered; in some cases a decrease in/with W fixed increases Pa. For intermediate^ ~0.1), Pa cc (W/r* )2/3 p\n\pc = critical density), independently of rc/ra; for/~0.6, Pa «s larger by a factor of about [rc/raf13. For rjra > 1.2 roughly, the mass ablation rate is C{Z,) [{m/kZ.f^Kr^Pl) l,\ independent of pc and/, and barely dependent on Z,(m, is ion mass; k, Boltzmann's constant; K, conductivity coefficient; and C, a tabulated function).

Electron temperature versus laser intensity times wavelength squared: a comparison of theory and experiments

The peak temperature in the corona of plasma ejected by a laser-irradiated slab is discussed in terms of a one-electron-temperature model. Both heat-flux saturation and pulse rise-time effects are considered;the intensity in the rising half of the pulse is approximated by a linear function of time, I(t) = Iot/r. The temperature is found to be proportional to (IQX2)273 and a function of I0X4/r. Above a certain value of I0X4/T, the plasma presents two characteristic temperatures (at saturation and at the critical surface) which can be identified with experimentally observed cold- and hot-electron temperatures. The results are compared with extensive experimental data available for both nd and CO2 lasers, I0(W'cnf2) X2 (/um) starting around 1012. The agreement is good if substantial flux inhibition is assumed (flux-limit factor f = 0.03), and fails for I0X2 above 1O1S. Results for both ablation pressure and mass ablation rate are also given.

Nonlocal electron heat relaxation in a plasma shock at arbitrary ionization number

A recently obtained nonlocal expression for the electron heat flux valid for arbitrary ionization numbers Z is used to study the structure of a plane shock wave in a fully ionized plasma. Nonlocal effects are only important in the foot of the electronic preheating region, where the electron temperature gradient is the steepest. The results are quantified as a function of a characteristic Knudsen number of that region. This work also generalizes to arbitrary values of Z previous results on plasma shock wave structure.

Two-temperature model of the coronal irradiated pellets

A two electron-temperature, quasi-steady model of the corona of a laser-ablated pellet is considered. Ablation pressure, critical radius and mass flow rate are determined. Results are close to those obtained with heat flux saturation well below the free-streaming limit.

Bubble growth by injection of gas into viscous liquids in cylindrical and conical tubes.

The laminar low Mach number flow of a gas in a tube is analyzed for very small and very large values of the inlet-to-wall temperature ratio. When this ratio tends to zero, pressure forces confine the cold gas to a thin core around the axis of the tube. This core is neatly bounded by an ablation front that consumes it at a finite distance from the tube inlet. When the temperature ratio tends to infinity, the temperature of the gas increases smoothly from the wall to the axis of the tube and the shear stress and heat flux are positive at the wall despite the fact that the viscosity and thermal conductivity of the gas scaled with their inlet values tend to zero at the wall.

Laminar flow of a gas in a tube with large temperature differences.

The laminar low Mach number flow of a gas in a tube is analyzed for very small and very large values of the inlet-to-wall temperature ratio. When this ratio tends to zero, pressure forces confine the cold gas to a thin core around the axis of the tube. This core is neatly bounded by an ablation front that consumes it at a finite distance from the tube inlet. When the temperature ratio tends to infinity, the temperature of the gas increases smoothly from the wall to the axis of the tube and the shear stress and heat flux are positive at the wall despite the fact that the viscosity and thermal conductivity of the gas scaled with their inlet values tend to zero at the wall

A technical note on application of internally finned tubes in solar parabolic trough absorber pipes

The heterogeneous incoming heat flux in solar parabolic trough absorber tubes generates huge temperature difference in each pipe section. Helical internal fins can reduce this effect, homogenising the temperature profile and reducing thermal stress with the drawback of increasing pressure drop. Another effect is the decreasing of the outer surface temperature and thermal losses, improving the thermal efficiency of the collector. The application of internal finned tubes for the design of parabolic trough collectors is analysed with computational fluid dynamics tools. Our numerical approach has been qualified with the computational estimation of reported experimental data regarding phenomena involved in finned tube applications and solar irradiation of parabolic trough collector. The application of finned tubes to the design of parabolic trough collectors must take into account issues as the pressure losses, thermal losses and thermo-mechanical stress, and thermal fatigue. Our analysis shows an improvement potential in parabolic trough solar plants efficiency by the application of internal finned tubes.