New Type Nuclear Reactors (NETNUC) 2008-2011 Final Report

This report summarizes the work done by a consortium consisting of Lappeenranta University of Technology, Aalto University and VTT Technical Research Centre of Finland in the New Type Nuclear Reactors (NETNUC) project during 2008–2011. The project was part of the Sustainable Energy (SusEn) research programme of the Academy of Finland. A wide range of generation IV nuclear technologies were studied during the project and the research consisted of multiple tasks. This report contains short articles summarizing the results of the individual tasks. In addition, the publications produced and the persons involved in the project are listed in the appendices.

Thermal shields for heat loss reduction in Siemens-type CVD reactors

The use of thermal shields to reduce radiation heat loss in Siemens-type CVD reactors is analyzed, both theoretically and experimentally. The potential savings from the use of the thermal shields is first explored using a radiation heat model that takes emissivity variations with wavelength into account, which is important for materials that do not behave as grey bodies. The theoretical calculations confirm that materials with lower surface emissivity lead to higher radiation savings. Assuming that radiation heat loss is responsible for around 50% of the total power consumption, a reduction of 32.9% and 15.5% is obtained if thermal shields with constant emissivities of 0.3 and 0.7 are considered, respectively. Experiments considering different thermal shields are conducted in a laboratory CVD reactor, confirming that the real materials do not behave as grey bodies, and proving that significant energy savings in the polysilicon deposition process are obtained. Using silicon as a thermal shield leads to energy savings of between 26.5-28.5%. For wavelength-dependent emissivities, the model shows that there are significant differences in radiation heat loss, of around 25%, when compared to that of constant emissivity. The results of the model highlight the importance of having reliable data on the emissivities within the relevant range of wavelengths, and at deposition temperatures, which remains a pending issue.

Program Trixy : multi-group analysis of nuclear reactors in three space dimensions : for the IBM type 704 computer /

"Physics and Math. TID-4500 (15th Ed.)"--Title page.

Effect of impeller type and mechanical agitation on the mass transfer and power consumption aspects of ASBR operation treating synthetic wastewater

The effect of flow type and rotor speed was investigated in a round-bottom reactor with 5 L useful volume containing 2.0 L of granular biomass. The reactor treated 2.0 L of synthetic wastewater with a concentration of 800 mgCOD/L in 8-h cycles at 30 degrees C. Five impellers, commonly used in biological processes, have been employed to this end, namely: a turbine and a paddle impeller with six-vertical-flat-blades, a turbine and a paddle impeller with six-45 degrees-inclined-flat-blades and a three-blade-helix impeller. Results showed that altering impeller type and rotor speed did not significantly affect system stability and performance. Average organic matter removal efficiency was about 84% for filtered samples, total volatile acids concentration was below 20 mgHAc/L and bicarbonate alkalinity a little less than 400 mgCaCO(3)/L for most of the investigated conditions. However, analysis of the first-order kinetic model constants showed that alteration in rotor speed resulted in an increase in the values of the kinetic constants (for instance, from 0.57 h(-1) at 50 rpm to 0.84 h(-1) at 75 rpm when the paddle impeller with six-45 degrees-inclined-flat-blades was used) and that axial flow in mechanically stirred reactors is preferable over radial-flow when the vertical-flat-blade impeller is compared to the inclined-flat-blade impeller (for instance at 75 rpm, from 0.52 h(-1) with the six-flat-blade-paddle impeller to 0.84 h(-1) with the six-45 degrees-inclined-flat-blade-paddle impeller), demonstrating that there is a rotor speed and an impeller type that maximize solid-liquid mass transfer in the reaction medium. Furthermore, power consumption studies in this reduced reactor volume showed that no high power transfer is required to improve mass transfer (less than 0.6 kW/10(3) m(3)). (C) 2008 Elsevier Ltd. All rights reserved.

Variation of bulk properties of anaerobic granules with wastewater type

Development of a granular sludge with high strength, high biological activity and a narrow settling distribution is necessary for optimal operation of high-rate upflow anaerobic treatment systems. Several studies have compared granules produced from different wastewaters but these have largely been from laboratory-fed reactors or compared granules from full-scale reactors fed similar wastewater types. Though two authors have commented on the inferiority of granules produced by a protein-based feed, the properties of these granules have not been characterised. In this paper, granules from full-scale reactors treating fruit and vegetable cannery effluent, two brewery effluents and a pig abattoir (slaughterhouse) were compared in terms of basic composition, size distribution, density, settling velocity, shear strength, and EPS content. The results supported previous qualitative observations by other researchers that indicate granule properties depend more on wastewater type rather than reactor design or operating conditions such as pre-acidification level. The cannery-fed granules bad excellent shear strength, settling distribution and density. Granules from the two brewery-fed reactors had statistically the same bulk properties, which were still acceptable for upflow applications. The protein-grown granule had poor strength and settling velocity. (C) 2001 Elsevier Science Ltd. All rights reserved.

Producció de biopolímers amb cultius bacterians mixtes. Comparació de tres reactors discontinus seqüencials (SBR) amb diferents condicions operacionals

La producció de biopolímers (polihidroxialcanoats (PHA) i substàncies polimèriques extracel·lulars (EPS)) a nivell industrial, resulta una nova àrea d’investigació que recull diverses disciplines, entre elles les Ciències Ambientals. Aquest projecte final de carrera amb el títol: “Producció de biopolímers amb cultius bacterians mixtes”, s’ha desenvolupat sota la supervisió de la directora de projecte Dra. María Eugenia Suárez Ojeda del Departament d’Enginyeria Química de la Universitat Autònoma de Barcelona (UAB) i s’ha dut a terme per l’estudiant Jordi Pérez i Forner de la Llicenciatura de Ciències Ambientals, Facultat de Ciències de la UAB, en el Departament d’Enginyeria Química de la mateixa universitat. L’objectiu d’aquest projecte ha estat produir biopolímers simultàniament amb l’eliminació de fòsfor i matèria orgànica en aigües residuals per obtenir un residu final amb un alt valor afegit. Aquests biopolímers reuneixen les característiques necessàries per a poder competir amb els plàstics convencionals i així, reduir l’elevat consum del petroli i la generació de residus no biodegradables. En aquest projecte s’ha dut a terme la posta en marxa d’un reactor discontinu seqüencial (SBR) per a l’acumulació de biopolímers amb cultius bacterians mixtes. Diferents investigadors han estudiat que aquests tipus de cultius bacterians arriben a nivells de fins el 53-97% [Pijuan et al., 2009] de contingut de biopolímers a la biomassa, sometent als microorganismes a diferents situacions d’estrés ja sigui per dèficit de nutrients o per variacions en les fases de feast-famine (festí-fam). Durant el projecte, s’ha realitzat el monitoratge del reactor alimentat amb una aigua sintètica, elaborada en el laboratori, amb les característiques d’un aigua residual provinent de la industria làctica. S’ha sotmès als microorganismes a diferents condicions operacionals, una d’elles amb limitació de fòsfor com a nutrient i una tercera condició amb una variació a les fases feast-famine. D’altra banda, com a segon objectiu, s’ha analitzat el contingut de biopolímers a la biomassa de dos SBRs més, del grup de recerca Bio-GLS del Departament d’Enginyeria Química de la UAB, alimentats amb diferents fonts de carboni, glicerol i àcids grassos de cadena llarga (AGCLL), per observar les influències que té el tipus de substrat en l’acumulació de biopolímers. Els resultats obtinguts en la primera part d’aquest projecte han estat similars als resultats d’altres investigadors [Pijuan et al., 2009; Guerrero et al., 2012]. S’ha determinat que sotmetre als microorganismes a situacions d’estrés té un efecte directe pel que fa a l’acumulació de biopolímers. També s’ha observat com al mateix temps que acumulaven aquests compostos, els microorganismes desenvolupaven la seva tasca de depurar l’aigua residual, obtenint al final del cicle una aigua amb un baix contingut en matèria orgànica i altres contaminants com amoni i fòsfor, en aquest cas. En la segona part del projecte, s’ha observat com el tipus de substrat té un efecte directe pel que fa a l’acumulació de biopolímers i també a l’activitat metabòlica dels microorganismes. Per tant, s’ha conclòs que la producció de biopolímers mitjançant la depuració d’aigües residuals es una via d’investigació molt prometedora pel que fa als resultats obtinguts. Alhora que es tracta un residu, s’obté una producte residual amb un alt valor afegit que pot ser utilitzat per la producció de bioplàstics 100% biodegradables.

Thermal-Hydraulic Studies on the Safety Of VVER-440 Type Nuclear Power Plants

This thesis includes several thermal hydraulic analyses related to the Loviisa WER 440 nuclear power plant units. The work consists of experimental studies, analysis of the experiments, analysis of some plant transits and development of a calculational model for calculation of boric acid concentrations in the reactor. In the first part of the thesis, in the case of won of boric acid solution behaviour during long term cooling period of LOCAs, experiments were performed in scaled down test facilities. The experimental data together with the results of RELAPS/MOD3 simulations were used to develop a model for calculations of boric acid concentrations in the reactor during LOCAs. The results of calculations showed that margins to critical concentrations that would lead to boric acid crystallization were large, both in the reactor core and in the lower plenum. This was mainly caused by the fact that water in the primary cooling circuit includes borax (Na)BsO,.IOHZO), which enters the reactor when ECC water is taken from the sump and greatly increases boric acid solubility in water. In the second part, in the case of simulation of horizontal steam generators, experiments were performed with PACTEL integral test loop to simulate loss of feedwater transients. The PACTEL experiments, as well as earlier REWET III natural circulation tests, were analyzed with RELAPS/MOD3 Version Sm5 code. The analysis showed that the code was capable of simulating the main events during the experiments. However, in the case of loss of secondary side feedwater the code was not completely capable to simulate steam superheating in the secondary side of the steam generators. The third part of the work consists of simulations of Loviisa VVER reactor pump trip transients with RELAPSlMODI Eur, RELAPS/MOD3 and CATHARE codes. All three codes were capable to simulate the two selected pump trip transients and no significant differences were found between the results of different codes. Comparison of the calculated results with the data measured in the Loviisa plant also showed good agreement.

Radiocarbon impact in a nearby tree of a light-water VVER type nuclear power plant, Paks, Hungary

Tree-ring series were collected for radiocarbon analyses from the vicinity of Paks nuclear power plant (NPP) and a background area (Dunaföldvár) for a 10-yr period (2000–2009). Samples of holocellulose were prepared from the wood and converted to graphite for accelerator mass spectrometry (AMS) 14C measurement using the MICADAS at ETH Zürich. The 14C concentration data from these tree rings was compared to the background tree rings for each year. The global decreasing trend of atmospheric 14C activity concentration was observed in the annual tree rings both in the background area and in the area of the NPP. As an average of the past 10 yr, the excess 14C emitted by the pressurized-water reactor (PWR) NPP to the atmosphere shows only a slight systematic excess (~6‰) 14C in the annual rings. The highest 14C excess was 13‰ (in 2006); however, years with the same 14C level as the background were quite frequent in the tree-ring series.

A proposed parameterization of interface discontinuity factors depending on neighborhood for pin-by-pin diffusion computations for LWR

There exists an interest in performing full core pin-by-pin computations for present nuclear reactors. In such type of problems the use of a transport approximation like the diffusion equation requires the introduction of correction parameters. Interface discontinuity factors can improve the diffusion solution to nearly reproduce a transport solution. Nevertheless, calculating accurate pin-by-pin IDF requires the knowledge of the heterogeneous neutron flux distribution, which depends on the boundary conditions of the pin-cell as well as the local variables along the nuclear reactor operation. As a consequence, it is impractical to compute them for each possible configuration. An alternative to generate accurate pin-by-pin interface discontinuity factors is to calculate reference values using zero-net-current boundary conditions and to synthesize afterwards their dependencies on the main neighborhood variables. In such way the factors can be accurately computed during fine-mesh diffusion calculations by correcting the reference values as a function of the actual environment of the pin-cell in the core. In this paper we propose a parameterization of the pin-by-pin interface discontinuity factors allowing the implementation of a cross sections library able to treat the neighborhood effect. First results are presented for typical PWR configurations.

On the Extension of the Analytic Nodal Diffusion Solver ANDES to Sodium Fast Reactors

Within the framework of the Collaborative Project for a European Sodium Fast Reactor, the reactor physics group at UPM is working on the extension of its in-house multi-scale advanced deterministic code COBAYA3 to Sodium Fast Reactors (SFR). COBAYA3 is a 3D multigroup neutron kinetics diffusion code that can be used either as a pin-by-pin code or as a stand-alone nodal code by using the analytic nodal diffusion solver ANDES. It is coupled with thermalhydraulics codes such as COBRA-TF and FLICA, allowing transient analysis of LWR at both fine-mesh and coarse-mesh scales. In order to enable also 3D pin-by-pin and nodal coupled NK-TH simulations of SFR, different developments are in progress. This paper presents the first steps towards the application of COBAYA3 to this type of reactors. ANDES solver, already extended to triangular-Z geometry, has been applied to fast reactor steady-state calculations. The required cross section libraries were generated with ERANOS code for several configurations. The limitations encountered in the application of the Analytic Coarse Mesh Finite Difference (ACMFD) method –implemented inside ANDES– to fast reactors are presented and the sensitivity of the method when using a high number of energy groups is studied. ANDES performance is assessed by comparison with the results provided by ERANOS, using a mini-core model in 33 energy groups. Furthermore, a benchmark from the NEA for a small 3D FBR in hexagonal-Z geometry and 4 energy groups is also employed to verify the behavior of the code with few energy groups.

Análisis Integrado de Seguridad de un accidente de SGTR en un reactor nuclear tipo PWR

El accidente de rotura de tubos de un generador de vapor (Steam Generator Tube Rupture, SGTR) en los reactores de agua a presión es uno de los transitorios más exigentes desde el punto de vista de operación. Los transitorios de SGTR son especiales, ya que podría dar lugar a emisiones radiológicas al exterior sin necesidad de daño en el núcleo previo o sin que falle la contención, ya que los SG pueden constituir una vía directa desde el reactor al medio ambiente en este transitorio. En los análisis de seguridad, el SGTR se analiza desde un punto determinista y probabilista, con distintos enfoques con respecto a las acciones del operador y las consecuencias analizadas. Cuando comenzaron los Análisis Deterministas de Seguridad (DSA), la forma de analizar el SGTR fue sin dar crédito a la acción del operador durante los primeros 30 min del transitorio, lo que suponía que el grupo de operación era capaz de detener la fuga por el tubo roto dentro de ese tiempo. Sin embargo, los diferentes casos reales de accidentes de SGTR sucedidos en los EE.UU. y alrededor del mundo demostraron que los operadores pueden emplear más de 30 minutos para detener la fuga en la vida real. Algunas metodologías fueron desarrolladas en los EEUU y en Europa para abordar esa cuestión. En el Análisis Probabilista de Seguridad (PSA), las acciones del operador se tienen en cuenta para diseñar los cabeceros en el árbol de sucesos. Los tiempos disponibles se utilizan para establecer los criterios de éxito para dichos cabeceros. Sin embargo, en una secuencia dinámica como el SGTR, las acciones de un operador son muy dependientes del tiempo disponible por las acciones humanas anteriores. Además, algunas de las secuencias de SGTR puede conducir a la liberación de actividad radiológica al exterior sin daño previo en el núcleo y que no se tienen en cuenta en el APS, ya que desde el punto de vista de la integridad de núcleo son de éxito. Para ello, para analizar todos estos factores, la forma adecuada de analizar este tipo de secuencias pueden ser a través de una metodología que contemple Árboles de Sucesos Dinámicos (Dynamic Event Trees, DET). En esta Tesis Doctoral se compara el impacto en la evolución temporal y la dosis al exterior de la hipótesis más relevantes encontradas en los Análisis Deterministas a nivel mundial. La comparación se realiza con un modelo PWR Westinghouse de tres lazos (CN Almaraz) con el código termohidráulico TRACE, con hipótesis de estimación óptima, pero con hipótesis deterministas como criterio de fallo único o pérdida de energía eléctrica exterior. Las dosis al exterior se calculan con RADTRAD, ya que es uno de los códigos utilizados normalmente para los cálculos de dosis del SGTR. El comportamiento del reactor y las dosis al exterior son muy diversas, según las diferentes hipótesis en cada metodología. Por otra parte, los resultados están bastante lejos de los límites de regulación, pese a los conservadurismos introducidos. En el siguiente paso de la Tesis Doctoral, se ha realizado un análisis de seguridad integrado del SGTR según la metodología ISA, desarrollada por el Consejo de Seguridad Nuclear español (CSN). Para ello, se ha realizado un análisis termo-hidráulico con un modelo de PWR Westinghouse de 3 lazos con el código MAAP. La metodología ISA permite la obtención del árbol de eventos dinámico del SGTR, teniendo en cuenta las incertidumbres en los tiempos de actuación del operador. Las simulaciones se realizaron con SCAIS (sistema de simulación de códigos para la evaluación de la seguridad integrada), que incluye un acoplamiento dinámico con MAAP. Las dosis al exterior se calcularon también con RADTRAD. En los resultados, se han tenido en cuenta, por primera vez en la literatura, las consecuencias de las secuencias en términos no sólo de daños en el núcleo sino de dosis al exterior. Esta tesis doctoral demuestra la necesidad de analizar todas las consecuencias que contribuyen al riesgo en un accidente como el SGTR. Para ello se ha hecho uso de una metodología integrada como ISA-CSN. Con este enfoque, la visión del DSA del SGTR (consecuencias radiológicas) se une con la visión del PSA del SGTR (consecuencias de daño al núcleo) para evaluar el riesgo total del accidente. Abstract Steam Generator Tube Rupture accidents in Pressurized Water Reactors are known to be one of the most demanding transients for the operating crew. SGTR are special transient as they could lead to radiological releases without core damage or containment failure, as they can constitute a direct path to the environment. The SGTR is analyzed from a Deterministic and Probabilistic point of view in the Safety Analysis, although the assumptions of the different approaches regarding the operator actions are quite different. In the beginning of Deterministic Safety Analysis, the way of analyzing the SGTR was not crediting the operator action for the first 30 min of the transient, assuming that the operating crew was able to stop the primary to secondary leakage within that time. However, the different real SGTR accident cases happened in the USA and over the world demonstrated that operators can took more than 30 min to stop the leakage in actual sequences. Some methodologies were raised in the USA and in Europe to cover that issue. In the Probabilistic Safety Analysis, the operator actions are taken into account to set the headers in the event tree. The available times are used to establish the success criteria for the headers. However, in such a dynamic sequence as SGTR, the operator actions are very dependent on the time available left by the other human actions. Moreover, some of the SGTR sequences can lead to offsite doses without previous core damage and they are not taken into account in PSA as from the point of view of core integrity are successful. Therefore, to analyze all this factors, the appropriate way of analyzing that kind of sequences could be through a Dynamic Event Tree methodology. This Thesis compares the impact on transient evolution and the offsite dose of the most relevant hypothesis of the different SGTR analysis included in the Deterministic Safety Analysis. The comparison is done with a PWR Westinghouse three loop model in TRACE code (Almaraz NPP), with best estimate assumptions but including deterministic hypothesis such as single failure criteria or loss of offsite power. The offsite doses are calculated with RADTRAD code, as it is one of the codes normally used for SGTR offsite dose calculations. The behaviour of the reactor and the offsite doses are quite diverse depending on the different assumptions made in each methodology. On the other hand, although the high conservatism, such as the single failure criteria, the results are quite far from the regulatory limits. In the next stage of the Thesis, the Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermohydraulical analysis of a Westinghouse 3-loop PWR plant with the MAAP code. The ISA methodology allows obtaining the SGTR Dynamic Event Tree taking into account the uncertainties on the operator actuation times. Simulations are performed with SCAIS (Simulation Code system for Integrated Safety Assessment), which includes a dynamic coupling with MAAP thermal hydraulic code. The offsite doses are calculated also with RADTRAD. The results shows the consequences of the sequences in terms not only of core damage but of offsite doses. This Thesis shows the need of analyzing all the consequences in an accident such as SGTR. For that, an it has been used an integral methodology like ISA-CSN. With this approach, the DSA vision of the SGTR (radiological consequences) is joined with the PSA vision of the SGTR (core damage consequences) to measure the total risk of the accident.

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.

Potential issues related to emergency core cooling system strainers performance at boiling water reactors: application to Cofrentes NPP (Spain)

From the 60s to the 90s, a great number of events related to the Emergency Core Cooling Systems Strainers have been happened in all kind of reactors all over the world. Thus, the Nuclear Regulatory Commission of the USA emitted some Bulletins to address the concerns about the adequacy of Emergency Core Cooling Systems (ECCS) strainer performance at boiling water reactors (BWR). In Spain the regulatory body (Consejo de Seguridad Nuclear, CSN) adopted the USA regulation and Cofrentes NPP installed new strainers with a considerable bigger size than the old strainers. The nuclear industry conducted significant and extensive research, guidance development, testing, reviews, and hardware and procedure changes during the 90s to resolve the issues related to debris blockage of BWR strainers. In 2001 the NRC and CSN closed the Bulletins. Thereafter, the strainers issues were moved to the PWR reactors. In 2004 the NRC issued a Generic Letter (GL). It requested the resolution of several effects which were not noted in the past. The GL regarded to be resolved by the PWR reactors but the NRC in USA and the CSN in Spain have requested that the BWR reactors investigate differences between the methodologies used by the BWRs and PWRs. The developments and improvements done for Cofrentes NPP are detailed. Studies for this plant show that the head loss due to the considered debris is at most half of the limited head loss for the ECCS strainer and the NPSH (Net Positive Suction Head) required for the ECCS pumps is at least three times lower than the NPSH available.

Análisis de Ruido en Reactores PWR

Esta tesis se ha llevado a cabo persiguiendo dos objetivos principales: uno de ellos es el desarrollo y la aplicación de modelos para el mantenimiento predictivo de sensores en centrales nucleares, y el otro es profundizar en el entendimiento de los fenómenos que tienen influencia en el ruido de la señal de los detectores de neutrones de los reactores de agua a presión con ayuda de herramientas de simulación 3D. Para el desarrollo de los trabajos se ha contado con medidas de ruido de reactores PWR actualmente en operación registradas en el curso de la tesis. El análisis de estas medidas ha permitido desarrollar los modelos de los sensores a partir de sus señales reales y comparar lo obtenido en las simulaciones con la realidad. El estudio de los sensores y la elaboración de los modelos se han llevado a cabo mediante la aplicación de técnicas autorregresivas a las señales tomadas en planta. Para la reproducción de los fenómenos que tienen lugar en el núcleo del reactor y que pueden influir en el ruido neutrónico se ha contado con códigos neutrónicos ampliamente utilizados en la industria y con modelos actualizados y validados de las plantas. ABSTRACT There are two goals in this thesis. The first one is the development of models and its application for predictive maintenance of sensors in nuclear power plants. The second one is to improve the understanding of the phenomena that influence the neutron noise in pressurized water reactors by using 3D simulators. Real plant measurements recorded during this thesis have been used to achieve such goals. The information provided by the data led the development of the models and the comparison of the results provided by the computational simulations. Sensor models were obtained by applying autorregresive techniques to the signals recorded in the plant. Wide known codes in the nuclear industry as well as updated and validated models have been used for the reproduction of the phenomena that take place in the core an may influence the neutron noise.

PWR fuel element specifications.

"Reactors - Power (TID-4500, 13th Edition)."