Analysis of the 207Bi, 194Hg/Au and 173Lu distribution in the irradiated MEGAPIE target

Samples obtained from different locations within the prototype liquid metal spallation target MEGAPIE irradiated in 2006 at PSI were analysed using γ-spectrometry. A variety of radionuclides formed by reaction of the target material, lead–bismuth eutectic (LBE), with the proton beam and secondary particles were identified. While nuclides representing the target material itself (207Bi) and nuclides of noble metals were found in LBE samples throughout the target, nuclides of electropositive metals were found to be quantitatively deposited on free surfaces and material interfaces within the target system. This behaviour is analysed in more detail based on results obtained for three nuclides representing groups of elements with distinct chemical behaviour, namely 207Bi, 194Hg/Au and 173Lu. Quantitative analysis results are given and compared with predictions obtained using nuclear physics calculations for those nuclides showing rather homogeneous distribution within the target. Possible reasons for the separation of radionuclides from the liquid metal and their deposition on surfaces are given, and consequences arising for nuclear facilities utilizing liquid metals are discussed.

A comparison between the chemical behaviour of lead-gold and lead-bismuth eutectics towards 316L stainless steel

Lead-gold eutectic (LGE) has been recently proposed as an alternative target material for high power spallation sources. In order to compare the corrosive properties of LGE to the better-studied eutectic of lead-bismuth (LBE), an isothermal twin-loop made of SS 316L was built and operated at the Institute of Physics of the University of Latvia. We have measured the concentration of steel alloying elements dissolved in both alloys at the end of two test campaigns via ICP-OES. In case of LGE, a pronounced concentration increase of Fe, Ni, Mn and Cr is found in the liquid metal, which is significantly higher compared to LBE. Similar results were obtained during complementary investigations on material samples exposed to both alloys in this twin-loop at 400 ◦C and 450 ◦C. These findings indicate that in contact with LGE, SS 316L steel suffers from substantial chemical attack. Detailed investigations using structure materials other than SS 316L have to be undertaken before qualifying LGE as a serious alternative to LBE.

Radiochemical determination of 129I and 36Cl in MEGAPIE, a proton irradiated lead-bismuth eutectic spallation target

The concentrations of the long-lived nuclear reaction products 129I and 36Cl have been measured in samples from the MEGAPIE liquid metal spallation target. Samples from the bulk target material (lead-bismuth eutectic, LBE), from the interface of the metal free surface with the cover gas, from LBE/steel interfaces and from noble metal absorber foils installed in the cover gas system were analysed using Accelerator Mass Spectrometry at the Laboratory of Ion beam Physics at ETH Zürich. The major part of 129I and 36Cl was found accumulated on the interfaces, particularly at the interface of LBE and the steel walls of the target container, while bulk LBE samples contain only a minor fraction of these nuclides. Both nuclides were also detected on the absorber foils to a certain extent (≪ 1% of the total amount). The latter number is negligible concerning the radio-hazard of the irradiated target material; however it indicates a certain affinity of the absorber foils for halogens, thus proving the principle of using noble metal foils for catching these volatile radionuclides. The total amounts of 129I and 36Cl in the target were estimated from the analytical data by averaging within the different groups of samples and summing up these averages over the total target. This estimation could account for about half of the amount of 129I and 36Cl predicted to be produced using nuclear physics modelling codes for both nuclides. The significance of the results and the associated uncertainties are discussed.

Silica final lenses under HiPER laser fusion reactor operation conditions

We have studied the thermo-mechanical response and atomistic degradation of final lenses in HiPER project. Final silica lenses are squares of 75 × 75 cm2 with a thickness of 5 cm. There are two scenarios where lenses are located at 8 m from the centre: •HiPER 4a, bunches of 100 shots (maximum 5 DT shots <48 MJ at ≈0.1 Hz). No blanket in chamber geometry. •HiPER 4b, continuous mode with shots ≈50 MJ at 10 Hz to generate 0.5 GW. Liquid metal blanket in chamber design.

Monte Carlo Neutronics and Thermal-hydraulics coupling applied to Fast Reactors

Un escenario habitualmente considerado para el uso sostenible y prolongado de la energía nuclear contempla un parque de reactores rápidos refrigerados por metales líquidos (LMFR) dedicados al reciclado de Pu y la transmutación de actínidos minoritarios (MA). Otra opción es combinar dichos reactores con algunos sistemas subcríticos asistidos por acelerador (ADS), exclusivamente destinados a la eliminación de MA. El diseño y licenciamiento de estos reactores innovadores requiere herramientas computacionales prácticas y precisas, que incorporen el conocimiento obtenido en la investigación experimental de nuevas configuraciones de reactores, materiales y sistemas. A pesar de que se han construido y operado un cierto número de reactores rápidos a nivel mundial, la experiencia operacional es todavía reducida y no todos los transitorios se han podido entender completamente. Por tanto, los análisis de seguridad de nuevos LMFR están basados fundamentalmente en métodos deterministas, al contrario que las aproximaciones modernas para reactores de agua ligera (LWR), que se benefician también de los métodos probabilistas. La aproximación más usada en los estudios de seguridad de LMFR es utilizar una variedad de códigos, desarrollados a base de distintas teorías, en busca de soluciones integrales para los transitorios e incluyendo incertidumbres. En este marco, los nuevos códigos para cálculos de mejor estimación ("best estimate") que no incluyen aproximaciones conservadoras, son de una importancia primordial para analizar estacionarios y transitorios en reactores rápidos. Esta tesis se centra en el desarrollo de un código acoplado para realizar análisis realistas en reactores rápidos críticos aplicando el método de Monte Carlo. Hoy en día, dado el mayor potencial de recursos computacionales, los códigos de transporte neutrónico por Monte Carlo se pueden usar de manera práctica para realizar cálculos detallados de núcleos completos, incluso de elevada heterogeneidad material. Además, los códigos de Monte Carlo se toman normalmente como referencia para los códigos deterministas de difusión en multigrupos en aplicaciones con reactores rápidos, porque usan secciones eficaces punto a punto, un modelo geométrico exacto y tienen en cuenta intrínsecamente la dependencia angular de flujo. En esta tesis se presenta una metodología de acoplamiento entre el conocido código MCNP, que calcula la generación de potencia en el reactor, y el código de termohidráulica de subcanal COBRA-IV, que obtiene las distribuciones de temperatura y densidad en el sistema. COBRA-IV es un código apropiado para aplicaciones en reactores rápidos ya que ha sido validado con resultados experimentales en haces de barras con sodio, incluyendo las correlaciones más apropiadas para metales líquidos. En una primera fase de la tesis, ambos códigos se han acoplado en estado estacionario utilizando un método iterativo con intercambio de archivos externos. El principal problema en el acoplamiento neutrónico y termohidráulico en estacionario con códigos de Monte Carlo es la manipulación de las secciones eficaces para tener en cuenta el ensanchamiento Doppler cuando la temperatura del combustible aumenta. Entre todas las opciones disponibles, en esta tesis se ha escogido la aproximación de pseudo materiales, y se ha comprobado que proporciona resultados aceptables en su aplicación con reactores rápidos. Por otro lado, los cambios geométricos originados por grandes gradientes de temperatura en el núcleo de reactores rápidos resultan importantes para la neutrónica como consecuencia del elevado recorrido libre medio del neutrón en estos sistemas. Por tanto, se ha desarrollado un módulo adicional que simula la geometría del reactor en caliente y permite estimar la reactividad debido a la expansión del núcleo en un transitorio. éste módulo calcula automáticamente la longitud del combustible, el radio de la vaina, la separación de los elementos de combustible y el radio de la placa soporte en función de la temperatura. éste efecto es muy relevante en transitorios sin inserción de bancos de parada. También relacionado con los cambios geométricos, se ha implementado una herramienta que, automatiza el movimiento de las barras de control en busca d la criticidad del reactor, o bien calcula el valor de inserción axial las barras de control. Una segunda fase en la plataforma de cálculo que se ha desarrollado es la simulació dinámica. Puesto que MCNP sólo realiza cálculos estacionarios para sistemas críticos o supercríticos, la solución más directa que se propone sin modificar el código fuente de MCNP es usar la aproximación de factorización de flujo, que resuelve por separado la forma del flujo y la amplitud. En este caso se han estudiado en profundidad dos aproximaciones: adiabática y quasiestática. El método adiabático usa un esquema de acoplamiento que alterna en el tiempo los cálculos neutrónicos y termohidráulicos. MCNP calcula el modo fundamental de la distribución de neutrones y la reactividad al final de cada paso de tiempo, y COBRA-IV calcula las propiedades térmicas en el punto intermedio de los pasos de tiempo. La evolución de la amplitud de flujo se calcula resolviendo las ecuaciones de cinética puntual. Este método calcula la reactividad estática en cada paso de tiempo que, en general, difiere de la reactividad dinámica que se obtendría con la distribución de flujo exacta y dependiente de tiempo. No obstante, para entornos no excesivamente alejados de la criticidad ambas reactividades son similares y el método conduce a resultados prácticos aceptables. Siguiendo esta línea, se ha desarrollado después un método mejorado para intentar tener en cuenta el efecto de la fuente de neutrones retardados en la evolución de la forma del flujo durante el transitorio. El esquema consiste en realizar un cálculo cuasiestacionario por cada paso de tiempo con MCNP. La simulación cuasiestacionaria se basa EN la aproximación de fuente constante de neutrones retardados, y consiste en dar un determinado peso o importancia a cada ciclo computacial del cálculo de criticidad con MCNP para la estimación del flujo final. Ambos métodos se han verificado tomando como referencia los resultados del código de difusión COBAYA3 frente a un ejercicio común y suficientemente significativo. Finalmente, con objeto de demostrar la posibilidad de uso práctico del código, se ha simulado un transitorio en el concepto de reactor crítico en fase de diseño MYRRHA/FASTEF, de 100 MW de potencia térmica y refrigerado por plomo-bismuto. ABSTRACT Long term sustainable nuclear energy scenarios envisage a fleet of Liquid Metal Fast Reactors (LMFR) for the Pu recycling and minor actinides (MAs) transmutation or combined with some accelerator driven systems (ADS) just for MAs elimination. Design and licensing of these innovative reactor concepts require accurate computational tools, implementing the knowledge obtained in experimental research for new reactor configurations, materials and associated systems. Although a number of fast reactor systems have already been built, the operational experience is still reduced, especially for lead reactors, and not all the transients are fully understood. The safety analysis approach for LMFR is therefore based only on deterministic methods, different from modern approach for Light Water Reactors (LWR) which also benefit from probabilistic methods. Usually, the approach adopted in LMFR safety assessments is to employ a variety of codes, somewhat different for the each other, to analyze transients looking for a comprehensive solution and including uncertainties. In this frame, new best estimate simulation codes are of prime importance in order to analyze fast reactors steady state and transients. This thesis is focused on the development of a coupled code system for best estimate analysis in fast critical reactor. Currently due to the increase in the computational resources, Monte Carlo methods for neutrons transport can be used for detailed full core calculations. Furthermore, Monte Carlo codes are usually taken as reference for deterministic diffusion multigroups codes in fast reactors applications because they employ point-wise cross sections in an exact geometry model and intrinsically account for directional dependence of the ux. The coupling methodology presented here uses MCNP to calculate the power deposition within the reactor. The subchannel code COBRA-IV calculates the temperature and density distribution within the reactor. COBRA-IV is suitable for fast reactors applications because it has been validated against experimental results in sodium rod bundles. The proper correlations for liquid metal applications have been added to the thermal-hydraulics program. Both codes are coupled at steady state using an iterative method and external files exchange. The main issue in the Monte Carlo/thermal-hydraulics steady state coupling is the cross section handling to take into account Doppler broadening when temperature rises. Among every available options, the pseudo materials approach has been chosen in this thesis. This approach obtains reasonable results in fast reactor applications. Furthermore, geometrical changes caused by large temperature gradients in the core, are of major importance in fast reactor due to the large neutron mean free path. An additional module has therefore been included in order to simulate the reactor geometry in hot state or to estimate the reactivity due to core expansion in a transient. The module automatically calculates the fuel length, cladding radius, fuel assembly pitch and diagrid radius with the temperature. This effect will be crucial in some unprotected transients. Also related to geometrical changes, an automatic control rod movement feature has been implemented in order to achieve a just critical reactor or to calculate control rod worth. A step forward in the coupling platform is the dynamic simulation. Since MCNP performs only steady state calculations for critical systems, the more straight forward option without modifying MCNP source code, is to use the flux factorization approach solving separately the flux shape and amplitude. In this thesis two options have been studied to tackle time dependent neutronic simulations using a Monte Carlo code: adiabatic and quasistatic methods. The adiabatic methods uses a staggered time coupling scheme for the time advance of neutronics and the thermal-hydraulics calculations. MCNP computes the fundamental mode of the neutron flux distribution and the reactivity at the end of each time step and COBRA-IV the thermal properties at half of the the time steps. To calculate the flux amplitude evolution a solver of the point kinetics equations is used. This method calculates the static reactivity in each time step that in general is different from the dynamic reactivity calculated with the exact flux distribution. Nevertheless, for close to critical situations, both reactivities are similar and the method leads to acceptable practical results. In this line, an improved method as an attempt to take into account the effect of delayed neutron source in the transient flux shape evolutions is developed. The scheme performs a quasistationary calculation per time step with MCNP. This quasistationary simulations is based con the constant delayed source approach, taking into account the importance of each criticality cycle in the final flux estimation. Both adiabatic and quasistatic methods have been verified against the diffusion code COBAYA3, using a theoretical kinetic exercise. Finally, a transient in a critical 100 MWth lead-bismuth-eutectic reactor concept is analyzed using the adiabatic method as an application example in a real system.

Design summary report of LCRE primary coolant pumps and sump /

"Contract AT(30-1)-2789."

Design summary report of LCRE reflector coolant pumps and sump /

"Contract AT(30-1)-2789."

Design summary report of LCRE secondary coolant pump and sump /

"Contract AT(30-1)-2789."

Final report : LCRE valve development /

"Contract AT(30-1)-2789."

Failure data handbook for nuclear power facilities : a guide for the design, construction, and maintenance of nuclear power plants from a reliability improvement standpoint /

Photocopy.

Experimental results of forced convection boiling potassium heat transfer and pressure drop tests /

"Contract AT(30-1)-2789."

Closeout costs, Clinch River breeder reactor project [microform] : hearing before the Committee on Science and Technology, U.S. House of Representatives, Ninety-seventh Congress, first session, May 11, 1981.

"No. 30."

A three region steam drum model for a nuclear power plant simulator (BRENDA) /

"Date published: August 1981."

The automation of resonant thermometer probe measurements

An ultrasonic thermometer has been developed for high temperature measurement over a wide temperature range. It is particularly suitable for use in measuring nuclear fuel rod centerline temperatures in advanced liquid metal and high flux nuclear reactors. The thermometer which was designed to determine fuel temperature up to the fuel melting point, utilizes the temperature dependence of the ultrasonic propagation velocity (related to the elastic modulus} in a thin rod sensor as the temperature transducing mechanism. A pulse excitation technique has been used, where the mechanical resonator at the remote end of the acoustic·line is madto vibrate. Its natural frequency is proportional to the ultrasonic velocity in the material. This is measured by the electronic instrumentation and enables a frequency­ temperature or period-temperature calibration to be obtained. A completely digital automatic instrument has been designed, constructed and tested to track the resonance frequency of the temperature sensors. It operates smoothly over a frequency range of about 30%, more than the maximum working range of most probe materials. The control uses the basic property of a resonator that the stored energy decays exponentially at the natural frequency of the resonator.The operation of the electronic system is based on a digital multichannel transmitter that is capable of operating with a predefined number of cycles in the burst. this overcomes a basic defect in the previous deslgn where the analogue time-delayed circuits failed to hold synchronization and hence automatic control could be lost. Development of a particular type of temperature probe, that is small enough to fit into a standard 2 mm reactor tube has made the ultrasonic thermometer a practicable device for measuring fuel temperature. The bulkiness of previous probes has been overcome, the new design consists of a tuning fork, integral with a 1mm line, while maintaining a frequency of no more than 100 kHz. A magnetostrictive rod, acoustically matched to the probe is used to launch and receive the acoustic oscillations. This requires a magnetic bias and the previously used bulky magnets have been replaced by a direct current coil. The probe is supported by terminating the launcher with a short heavy isolating rod which can be secured to the reactor structure. This support, the bias and launching coil and the launcher are made up into a single compact unit. On the material side an extensive study of a wide range of refractory materials identified molybdenum, iridium, rhenium and tungsten as satisfactory for a number of applications but mostly exhibiting to some degree a calibration drift with thermal cycling. When attention was directed to ceramic materials, Sapphire (single crystal alumina) was found to have numerous advantages, particularly in respect of stability of calibration which remained with ±2°C after many cycles to 1800oC. Tungsten and thoriated tungsten (W - 2% Tho2) were also found to be quite satisfactory to 1600oC, the specification for a Euratom application.

Multiphase modelling of the characteristics of close coupled gas atomization

Not withstanding the high demand of metal powder for automotive and High Tech applications, there are still many unclear aspects of the production process. Only recentlyhas supercomputer performance made possible numerical investigation of such phenomena. This thesis focuses on the modelling aspects of primary and secondary atomization. Initially two-dimensional analysis is carried out to investigate the influence of flow parameters (reservoir pressure and gas temperature principally) and nozzle geometry on final powder yielding. Among the different types, close coupled atomizers have the best performance in terms of cost and narrow size distribution. An isentropic contoured nozzle is introduced to minimize the gas flow losses through shock cells: the results demonstrate that it outperformed the standard converging-diverging slit nozzle. Furthermore the utilization of hot gas gave a promising outcome: the powder size distribution is narrowed and the gas consumption reduced. In the second part of the thesis, the interaction of liquid metal and high speed gas near the feeding tube exit was studied. Both axisymmetric andnon-axisymmetric geometries were simulated using a 3D approach. The filming mechanism was detected only for very small metal flow rates (typically obtained in laboratory scale atomizers). When the melt flow increased, the liquid core overtook the adverse gas flow and entered in the high speed wake directly: in this case the disruption isdriven by sinusoidal surface waves. The process is characterized by fluctuating values of liquid volumes entering the domain that are monitored only as a time average rate: it is far from industrial robustness and capability concept. The non-axisymmetric geometry promoted the splitting of the initial stream into four cores, smaller in diameter and easier to atomize. Finally a new atomization design based on the lesson learned from previous cases simulation is presented.