On track for solar grade silicon through a siemens process-type laboratory reactor: operating conditions and energy savings

Polysilicon cost impacts significantly on the photovoltaics (PV) cost and on the energy payback time. Nowadays, the besetting production process is the so called Siemens process, polysilicon deposition by chemical vapor deposition (CVD) from Trichlorosilane. Polysilicon purification level for PV is to a certain extent less demanding that for microelectronics. At the Instituto de Energía Solar (IES) research on this subject is performed through a Siemens process-type laboratory reactor. Through the laboratory CVD prototype at the IES laboratories, valuable information about the phenomena involved in the polysilicon deposition process and the operating conditions is obtained. Polysilicon deposition by CVD is a complex process due to the big number of parameters involved. A study on the influence of temperature and inlet gas mixture composition on the polysilicon deposition growth rate, based on experimental experience, is shown. Moreover, CVD process accounts for the largest contribution to the energy consumption of the polysilicon production. In addition, radiation phenomenon is the major responsible for low energetic efficiency of the whole process. This work presents a model of radiation heat loss, and the theoretical calculations are confirmed experimentally through a prototype reactor at our disposal, yielding a valuable know-how for energy consumption reduction at industrial Siemens reactors.

A comparative study of Monte Carlo-coupled depletion codes applied to a Sodium Fast Reactor design loaded with minor actinides

inor actinides (MAs) transmutation is a main design objective of advanced nuclear systems such as generation IV Sodium Fast Reactors (SFRs). In advanced fuel cycles, MA contents in final high level waste packages are main contributors to short term heat production as well as to long-term radiotoxicity. Therefore, MA transmutation would have an impact on repository designs and would reduce the environment burden of nuclear energy. In order to predict such consequences Monte Carlo (MC) transport codes are used in reactor design tasks and they are important complements and references for routinely used deterministic computational tools. In this paper two promising Monte Carlo transport-coupled depletion codes, EVOLCODE and SERPENT, are used to examine the impact of MA burning strategies in a SFR core, 3600 MWth. The core concept proposal for MA loading in two configurations is the result of an optimization effort upon a preliminary reference design to reduce the reactivity insertion as a consequence of sodium voiding, one of the main concerns of this technology. The objective of this paper is double. Firstly, efficiencies of the two core configurations for MA transmutation are addressed and evaluated in terms of actinides mass changes and reactivity coefficients. Results are compared with those without MA loading. Secondly, a comparison of the two codes is provided. The discrepancies in the results are quantified and discussed.

The european project NURISP for nuclear reactor simulation

The NURISP project aims at developing the European NURESIM reference simulation platform [1] for nuclear reactor. A first version of NURESIM was delivered in 2008. 22 organizations from 14 European countries contribute to the further development of this platform. NURISP also includes a User’s Group (UG) whose members are not NURISP partners and come from the industrial nuclear sector or European and non-European R&D labs. Users can benefit from the use of the NURESIM platform, methods, results and modules and they provide concrete input and feedback on the use of these elements.

Design and analysis of helium Brayton power cycles for HiPER reactor

Helium Brayton cycles have been studied as power cycles for both fission and fusion reactors obtaining high thermal efficiency. This paper studies several technological schemes of helium Brayton cycles applied for the HiPER reactor proposal. Since HiPER integrates technologies available at short term, its working conditions results in a very low maximum temperature of the energy sources, something that limits the thermal performance of the cycle. The aim of this work is to analyze the potential of the helium Brayton cycles as power cycles for HiPER. Several helium Brayton cycle configurations have been investigated with the purpose of raising the cycle thermal efficiency under the working conditions of HiPER. The effects of inter-cooling and reheating have specifically been studied. Sensitivity analyses of the key cycle parameters and component performances on the maximum thermal efficiency have also been carried out. The addition of several inter-cooling stages in a helium Brayton cycle has allowed obtaining a maximum thermal efficiency of over 36%, and the inclusion of a reheating process may also yield an added increase of nearly 1 percentage point to reach 37%. These results confirm that helium Brayton cycles are to be considered among the power cycle candidates for HiPER.

Review of the C-nat(n,gamma) cross section and criticality calculations of the graphite moderated reactor BR1

A review of the experimental data for natC(n,c) and 12C(n,c) was made to identify the origin of the natC capture cross sections included in evaluated data libraries and to clarify differences observed in neutronic calculations for graphite moderated reactors using different libraries. The performance of the JEFF-3.1.2 and ENDF/B-VII.1 libraries was verified by comparing results of criticality calculations with experimental results obtained for the BR1 reactor. This reactor is an air-cooled reactor with graphite as moderator and is located at the Belgian Nuclear Research Centre SCK-CEN in Mol (Belgium). The results of this study confirm conclusions drawn from neutronic calculations of the High Temperature Engineering Test Reactor (HTTR) in Japan. Furthermore, both BR1 and HTTR calculations support the capture cross section of 12C at thermal energy which is recommended by Firestone and Révay. Additional criticality calculations were carried out in order to illustrate that the natC thermal capture cross section is important for systems with a large amount of graphite. The present study shows that only the evaluation carried out for JENDL-4.0 reflects the current status of the experimental data.

Methodologies for an improved prediction of the isotopic content in high burnup samples. Application to Vandellos-II reactor core

Fuel cycles are designed with the aim of obtaining the highest amount of energy possible. Since higher burnup values are reached, it is necessary to improve our disposal designs, traditionally based on the conservative assumption that they contain fresh fuel. The criticality calculations involved must consider burnup by making the most of the experimental and computational capabilities developed, respectively, to measure and predict the isotopic content of the spent nuclear fuel. These high burnup scenarios encourage a review of the computational tools to find out possible weaknesses in the nuclear data libraries, in the methodologies applied and their applicability range. Experimental measurements of the spent nuclear fuel provide the perfect framework to benchmark the most well-known and established codes, both in the industry and academic research activity. For the present paper, SCALE 6.0/TRITON and MONTEBURNS 2.0 have been chosen to follow the isotopic content of four samples irradiated in the Spanish Vandellós-II pressurized water reactor up to burnup values ranging from 40 GWd/MTU to 75 GWd/MTU. By comparison with the experimental data reported for these samples, we can probe the applicability of these codes to deal with high burnup problems. We have developed new computational tools within MONTENBURNS 2.0. They make possible to handle an irradiation history that includes geometrical and positional changes of the samples within the reactor core. This paper describes the irradiation scenario against which the mentioned codes and our capabilities are to be benchmarked.

Neutronics and activation of the preliminary reaction chamber of HiPER reactor based in a SCLL blanket

The HiPER reactor design is exploring different reaction chambers. In this study, we tackle the neutronicsand activation studies of a preliminary reaction chamber based in the following technologies: unpro-tected dry wall for the First Wall, self-cooled lead lithium blanket, and independent low activation steelVacuum Vessel. The most critical free parameter in this stage is the blanket thickness, as a function ofthe6Li enrichment. After a parametric study, we select for study both a ?thin? and ?thick? blanket, with?high? and ?low?6Li enrichment respectively, to reach a TBR = 1.1. To help to make a choice, we com-pute, for both blanket options, in addition to the TBR, the energy amplification factor, the tritium partialpressure, the203Hg and210Po total activity in the LiPb loop, and the Vacuum Vessel thickness requiredto guarantee the reweldability during its lifetime. The thin blanket shows a superior performance in thesafety related issues and structural viability, but it operates at higher6Li enrichment. It is selected forfurther improvements. The Vacuum Vessel shows to be unviable in both cases, with the thickness varyingbetween 39 and 52 cm. Further chamber modifications, such as the introduction of a neutron reflector,are required to exploit the benefits of the thin blanket with a reasonable Vacuum Vessel.

Nuclear Data Uncertainty Propagation to Reactivity Coefficients of a Sodium Fast Reactor

The assessment of the uncertainty levels on the design and safety parameters for the innovative European Sodium Fast Reactor (ESFR) is mandatory. Some of these relevant safety quantities are the Doppler and void reactivity coefficients, whose uncertainties are quantified. Besides, the nuclear reaction data where an improvement will certainly benefit the design accuracy are identified. This work has been performed with the SCALE 6.1 codes suite and its multigroups cross sections library based on ENDF/B-VII.0 evaluation.

Heat losses in a CVD reactor for polysilicon: comprehensive model and experimental validation

This work addresses heat losses in a CVD reactor for polysilicon production. Contributions to the energy consumption of the so-called Siemens process are evaluated, and a comprehensive model for heat loss is presented. A previously-developed model for radiative heat loss is combined with conductive heat loss theory and a new model for convective heat loss. Theoretical calculations are developed and theoretical energy consumption of the polysilicon deposition process is obtained. The model is validated by comparison with experimental results obtained using a laboratory-scale CVD reactor. Finally, the model is used to calculate heat consumption in a 36-rod industrial reactor; the energy consumption due to convective heat loss per kilogram of polysilicon produced is calculated to be 22-30 kWh/kg along a deposition process.

Assessment of damage domains of the High-Temperature Engineering Test Reactor (HTTR)

This paper presents an assessment analysis of damage domains of the 30 MWth prototype High-Temperature Engineering Test Reactor (HTTR) operated by the Japan Atomic Energy Agency (JAEA). For this purpose, an in-house deterministic risk assessment computational tool was developed based on the Theory of Stimulated Dynamics (TSD). To illustrate the methodology and applicability of the developed modelling approach, assessment results of a control rod (CR) withdrawal accident during subcritical conditions are presented and compared with those obtained by the JAEA.

Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking

The new reactor concepts proposed in the Generation IV International Forum (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs sodium fast reactors (SFRs) stand out due to their technological feasibility as demonstrated in several countries during the last decades. As part of the contribution of EURATOM to GIF the CP-ESFR is a collaborative project with the objective, among others, to perform extensive analysis on safety issues involving renewed SFR demonstrator designs. The verification of computational tools able to simulate the plant behaviour under postulated accidental conditions by code-to-code comparison was identified as a key point to ensure reactor safety. In this line, several organizations employed coupled neutronic and thermal-hydraulic system codes able to simulate complex and specific phenomena involving multi-physics studies adapted to this particular fast reactor technology. In the “Introduction” of this paper the framework of this study is discussed, the second section describes the envisaged plant design and the commonly agreed upon modelling guidelines. The third section presents a comparative analysis of the calculations performed by each organisation applying their models and codes to a common agreed transient with the objective to harmonize the models as well as validating the implementation of all relevant physical phenomena in the different system codes.

Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs.

Estudio de la influencia de la relación C/N en la desnitrificación y metanogénesis de aguas residuales en un reactor anaeróbico de lecho fluidizado inverso

La nitrificación-desnitrificación es el proceso biológico tradicional para la remoción de nitrógeno de las aguas residuales (Ruiz G. et al., 2006a), siendo fundamental ya que contribuye a controlar la eutroficación de los cuerpos receptores. Debido al deterioro que sobre la disponibilidad de los recursos han ejercido las actividades antropogénicas, es necesario orientar el tratamiento de las aguas residuales hacia tecnologías que ofrezcan el mayor grado de sustentabilidad, planteando innovaciones en el tratamiento. El presente proyecto de tesis doctoral versa sobre el estudio de la influencia de la relación C/N en la desnitrificación y metanogénesis de aguas residuales urbanas en un reactor anaeróbico de lecho fluidizado inverso (RLFI). Previamente a la realización de las pruebas experimentales de variación de la relación C/N, se llevó a cabo la etapa de arranque del RLFI la cual se inició en modo batch, favoreciendo la formación y adhesión de biopelícula al medio de soporte utilizado (Extendosphere). Después, sobrevino la operación en modo continuo desde una carga volumétrica aplicada (CVA) de 0.5 g DQOs/L⋅d hasta alcanzar 4 g DQOs/L⋅d, carga volumétrica a la cual se logró la plena estabilización del reactor, siendo la alta variabilidad de la concentración de DQOs en el agua residual urbana de alimentación, la principal problemática que ocasionó retrasos en la estabilidad del reactor. A una CVA de 4 g DQOs/L⋅d en estado estacionario, el valor mínimo de eficiencia de remoción de DQOs fue del 32.36% y el máximo de 66.99%. En estas condiciones el porcentaje de metano presente en el biogás producido tuvo un valor medio de 85.57 ± 2.93%, siendo un valor alto comparado con otros porcentajes de metano encontrados en la digestión anaerobia de aguas residuales urbanas. El YCH4 tuvo un valor medio de 0.316 ± 0.110 LCH4/g DQOrem⋅día. Los porcentajes de metanización variaron en el rango de 20.50 a 100%, registrándose un valor medio de 73.42 ± 25.63%. La considerable variabilidad en el porcentaje de metanización se debió principalmente a que se presentaron eventos de lavado de soporte colonizado, lo cual propició que las actividades metabólicas fueran orientadas hacia formación de biopelícula (anabolismo) en vez de estar dirigidas hacia producción de metano (catabolismo). En relación a los ensayos con variación de la relación C/N, se manejaron relaciones DQOs/N-NO3 en el rango de 1.65 a 21.1 g DQOs/g N-NO3. La tasa de remoción anaerobia de DQOs se incrementó con la concentración de sustrato en una relación casi lineal, ajustándose a una cinética de primer orden, lo que regularmente se presenta a concentraciones bajas de sustrato. La eficiencia del proceso de desnitrificación fue por lo regular alta, incrementándose ligeramente con la concentración de DQOs en el influente, con valores en el rango de 73.8 a 99.1%. Por otra parte, la tasa de remoción por metanogénesis se incrementó con la concentración relativa de sustrato (es decir, a mayores relaciones DQOs/N-NO3), siendo más sensitiva la metanogénesis a la concentración relativa de sustrato que la desnitrificación. Conforme aumentó la relación DQOs/N-NO3, la desnitrificación, de ser la ruta metabólica principal de utilización de la materia orgánica (comparada con la metanización), empezó a combinarse con la metanización. De manera evidente, a las relaciones DQOs/N-NO3 probadas, se manifestaron más las actividades desnitrificantes, quedando reflejadas por el alto porcentaje de utilización de la DQOs removida hacia la desnitrificación. La relación experimental DQOs/N-NO3 a la cual se pudiera haber cumplido con el requerimiento de materia orgánica (en términos de DQOs) para la desnitrificación de nitratos en las aguas residuales urbanas tratadas resultó aproximadamente ser igual a 7.1 g DQOs/g N-NO3. A una CVA de 4 g DQOs/L⋅d, se obtuvo un diámetro promedio máximo de soporte colonizado igual a 266.106 ± 69.279 μm aunque, hay que indicarlo, se presentaron fluctuaciones, las cuales se reflejaron también en el espesor de la biopelícula, el cual tuvo un valor máximo de 50.099 μm y un valor promedio de 37.294 ± 11.199 μm. Estas fluctuaciones pudieron deberse a la existencia de corrientes preferenciales dentro del reactor, las cuales no permitieron un acceso equitativo del sustrato a todo el lecho. Nitrification-denitrification is the traditional biological process for nitrogen removal from wastewaters (Ruiz G. et al., 2006a), being fundamental since it contributes to control the eutrophication of the receiving waters. Due to the deterioration that on the availability of the aquatic resources the anthropogenic activities have exerted, it is necessary to orient the treatment of wastewaters towards technologies that offer the greater degree of sustainability, raising innovations in the treatment. This work studied the influence of C/N ratio on denitrification and methanogenesis of urban wastewaters in an inverse fluidized bed reactor (IFBR). Previously to the accomplishment of the experimental tests with variation of C/N ratio, the start up of the IFBR was carried out in batch way, encouraging the formation and adhesion of biofilm to Extendosphere, which it was used as support. The operation in continuous way carried out from an organic loading rate (OLR) of 0.5 g CODs/L ∙ d to 4 g CODs/L ∙ d, when the steady-state was reached. The high variability of the CODs of the urban wastewaters caused delays in the stability of the reactor. Once stationary state was reached, the removal efficiency of CODs ranged from 32.36 to 66.99% to 4 g CODs/L ∙ d. In these conditions the percentage of methane in produced biogas had an average value of 85.57 ± 2.93%, being a high value compared with other studies treating anaerobically urban wastewaters. The YCH4 had an average value of 0.316 ± 0.110 LCH4/g CODrem ∙ d. The percentage of methanisation ranged from 20.50 to 100%, with an average value of 73.42 ± 25.63%. The considerable variability in the methanisation percentage occurred mainly due events of wash-out of colonized support, which caused that the metabolic activities were oriented towards formation of biofilm (anabolism) instead of methane production (catabolism). Concerning the tests with variation of C/N ratio, CODs/NO3-N ratios from 1.65 to 21.1 g CODs/g NO3-N were proved. The CODs anaerobic removal rate increased with the substrate concentration in an almost linear relation, adjusting to a kinetic of first order, which regularly appears to low concentrations of substrate. Efficiency of the denitrification process was regularly high, and it increased slightly with the CODs concentration in the influent, ranging from 73.8 to 99.1%. On the other hand, the CODs removal rate by methanogenesis increased with the substrate relative concentration (e.g., to greater CODs/NO3-N ratios), being more sensitive the methanogenesis to the substrate relative concentration that the denitrification. When the CODs/NO3-N ratio increased, the denitrification, of being the main metabolic route of use of the organic matter (compared with the methanogenesis), began to be combined with the methanogenesis. Definitively, to the proven CODs/NO3-N ratios the denitrification processes were more pronounced, being reflected by the high percentage of use of the removed CODs towards denitrification. The experimental CODs/NO3-N ratio to which it was possible to have been fulfilled the requirement of organic matter (in terms of CODs) for the denitrification of nitrates in urban wastewaters turned out to be approximately 7.1 g CODs/g NO3-N. It was obtained a maximum average diameter of colonized support of 266.106 ± 69.279 μm to 4 g CODs/L ∙ d, although it is necessary to indicate that appeared fluctuations in the thickness of biofilm, which had a maximum value of 50.099 μm and an average value of 37.294 ± 11.199 μm. These fluctuations could be due to the existence of preferential currents within the reactor, which did not allow an equitable access of the substrate to all the bed.

Simulating the remote handling of the blanket segments in DEMO fusion reactor with thermo-mechanical meshfree multibody dynamics

The present study shows a first approach to the simulation of the remote handling oper- ation which takes into account the thermal and flexible behavior of the blanket segments and its implications on the remote handling equipment, in order to validate and improve its design.

Biomass Gasification in a Fluidized Bed Reactor: Effect of Temperature, Stoichiometric Ratio and Biomass Type

This paper investigates the gasification of two biomass types (pine wood and olive stones) in a laboratory scale bubbling fluidized bed reactor, in order to evaluate comparatively their potential in the production of syngas.