Hydrodynamic phenomena in the downcomer during flow rate transients in the primary circuit of a PWR

Mixing phenomena observed when the flow rate in a single loop of the primary circuit is changed can influence the operation of pressurized water reactor (PWR) by inducing local gradients of boron concentration or coolant temperature. Analysis of one-dimensional Laser Doppler Anemometry (LDA) measurements during the start-up and shutdown of pump on a single loop of the ROCOM test facility has been performed. The effect of a step change and a ramped change in the flow rate on the axial and azimuthal velocities was examined. Numerical simulations were also performed for the step change in the flow rate that gave quantitative agreement with the axial velocities. Phenomenological agreement was made on the turbulent kinetic energy; however, observed values were a factor of 2.5 less than the turbulent kinetic energy derived from the measurements. © 2007.

Fracture resistance of weakened roots restored with composite resin and glass fiber post

This study evaluated the fracture resistance of weakened roots restored with glass fiber posts, composite resin cores and complete metal crowns. Thirty maxillary canines were randomly divided into 3 groups of 10 teeth each: teeth without weakened roots (control); teeth with partially weakened roots (PWR) and teeth with and largely weakened roots (LWR). The control group was restored with glass fiber posts and a composite resin core. Teeth in the PWR and LWR groups were flared internally to standardized dimensions in order to simulate root weakness. Thereafter, the roots were partially filled with composite resin and restored in the same way as in the control group. The specimens were exposed to 250,000 cycles in a controlled chewing simulator. All intact specimens were subjected to a static load (N) in a universal testing machine at 45 degrees to the long axis of the tooth until failure. Data were analyzed by one-way ANOVA and Dunnett's test for multiple comparisons (p=0.05). There were statistically significant difference differences (p<0.01) among the groups (control group = 566.73 N; PWR = 409.64 N; and LWR = 410.91 N), with significantly higher fracture strength for the control group. There was no statistically significant difference (p>0.05) between the weakened groups. The results of this study showed that thicker root dentin walls significantly increase the fracture resistance of endodontically treated teeth.

Viabilidade económica da implementação de um reactor nuclear para a produção de energia eléctrica em Portugal

O presente trabalho tem como objectivo avaliar economicamente e determinar a viabilidade da implementação de um reactor nuclear para produção de energia eléctrica. Faz-se uma abordagem a aspectos da energia nuclear no mundo e em particular a energia nuclear na união europeia, faz-se uma análise sobre a estrutura do sector nuclear em Espanha e o futuro da energia no mundo. É realizada uma análise sobre a energia nuclear em Portugal, são abordados aspectos como o planeamento energético, a localização da central nuclear, os recursos nacionais e internacionais, a inspecção e regulação nuclear e o impacto industrial. Por fim, faz-se uma análise sobre o mix energético português. Faz-se uma descrição do ciclo de combustível, isto é, um estudo a todas as fases pela qual o combustível nuclear passa desde a sua extracção, passando pela produção de energia e processamento dos resíduos. São descritos os principais componentes de uma central nuclear e o seu princípio de funcionamento. São analisados em detalhe os principais componentes de um reactor PWR (Objecto de estudo deste trabalho) e faz-se uma breve descrição de alguns modelos de reactores nucleares. É feita uma breve abordagem aos principais acidentes nucleares que ocorreram, e descrita a escala de ocorrências nucleares e as várias fases de desmantelamento de uma central. São apresentados os principais custos da central nuclear. Também é apresentado um estudo de viabilidade económica analisando três cenários diferentes e é apresentada uma análise de sensibilidade do VAL em função de algumas variáveis que têm grande influência na avaliação económica. São apresentadas as principais conclusões.

Intercomparison of the response of different photon and neutron detectors around a spent fuel cask

An intercomparison of the response of different photon and neutron detectors was performed in several measurement positions around a spent fuel cask (type TN 12/2B) filled with 4 MOX and 8 UO2 15 x 15 PWR fuel assemblies at the nuclear power plant Gosgen (KKG) in Switzerland. The instruments used in the study were both active and passive, photon and neutron detectors calibrated either for ambient or personal dose equivalent. The aim of the measurement campaign was to compare the responses of the radiation instruments to routinely used detectors. It has been shown that especially the indications of the neutron detectors are strongly dependent on the neutron spectra around the cask due to their different energy responses. However, routinely used active photon and neutron detectors were shown to be reliable instruments. (C) 2012 Elsevier Ltd. All rights reserved.

Modeling of Pressurizer Using APROS and TRACE Thermal Hydraulic Codes

Diplomityön tavoitteena on paineistimen yksityiskohtainen mallintaminen APROS- ja TRACE- termohydrauliikkaohjelmistoja käyttäen. Rakennetut paineistinmallit testattiin vertaamalla laskentatuloksia paineistimen täyttymistä, tyhjentymistä ja ruiskutusta käsittelevistä erilliskokeista saatuun mittausdataan. Tutkimuksen päätavoitteena on APROSin paineistinmallin validoiminen käyttäen vertailuaineistona PACTEL ATWS-koesarjan sopivia paineistinkokeita sekä MIT Pressurizer- ja Neptunus- erilliskokeita. Lisäksi rakennettiin malli Loviisan ydinvoimalaitoksen paineistimesta, jota käytettiin turbiinitrippitransientin simulointiin tarkoituksena selvittää mahdolliset voimalaitoksen ja koelaitteistojen mittakaavaerosta johtuvat vaikutukset APROSin paineistinlaskentaan. Kokeiden simuloinnissa testattiin erilaisia noodituksia ja mallinnusvaihtoehtoja, kuten entalpian ensimmäisen ja toisen kertaluvun diskretisointia, ja APROSin sekä TRACEn antamia tuloksia vertailtiin kattavasti toisiinsa. APROSin paineistinmallin lämmönsiirtokorrelaatioissa havaittiin merkittävä puute ja laskentatuloksiin saatiin huomattava parannus ottamalla käyttöön uusi seinämälauhtumismalli. Työssä tehdyt TRACE-simulaatiot ovat osa United States Nuclear Regulatory Commissionin kansainvälistä CAMP-koodinkehitys-ja validointiohjelmaa.

Validation Studies of Thermal-hydraulic Code for Safety Analysis of Nuclear Power Plants

This thesis gives an overview of the validation process for thermal hydraulic system codes and it presents in more detail the assessment and validation of the French code CATHARE for VVER calculations. Three assessment cases are presented: loop seal clearing, core reflooding and flow in a horizontal steam generator. The experience gained during these assessment and validation calculations has been used to analyze the behavior of the horizontal steam generator and the natural circulation in the geometry of the Loviisa nuclear power plant. The cases presented are not exhaustive, but they give a good overview of the work performed by the personnel of Lappeenranta University of Technology (LUT). Large part of the work has been performed in co-operation with the CATHARE-team in Grenoble, France. The design of a Russian type pressurized water reactor, VVER, differs from that of a Western-type PWR. Most of thermal-hydraulic system codes are validated only for the Western-type PWRs. Thus, the codes should be assessed and validated also for VVER design in order to establish any weaknesses in the models. This information is needed before codes can be used for the safety analysis. Theresults of the assessment and validation calculations presented here show that the CATHARE code can be used also for the thermal-hydraulic safety studies for VVER type plants. However, some areas have been indicated which need to be reassessed after further experimental data become available. These areas are mostly connected to the horizontal stem generators, like condensation and phase separation in primary side tubes. The work presented in this thesis covers a large numberof the phenomena included in the CSNI code validation matrices for small and intermediate leaks and for transients. Also some of the phenomena included in the matrix for large break LOCAs are covered. The matrices for code validation for VVER applications should be used when future experimental programs are planned for code validation.

Lämpötilakerrostumien syntyminen ydinvoimalaitoksen ensiöpiirissä luonnonkierron olosuhteissa

Tässä diplomityössä on tutkittu lämpötilakerrostumien syntymistä RENATA-koelaitteistolla, joka muistutti geometrialtaan painevesireaktorin paineastian ylätilaa. Kokeet tehtiin siten, että aluksi RENATA täytettiin lämpimällä vedellä, jonka jälkeen koelaitteistoon juoksutettiin pohjasta käsin kylmää vettä. Kokeiden tuloksia verrattiin kirjallisuudessa esitettyyn korrelaatioon. Koetilanne mallinnettiin myös Fluent-virtauslaskentaohjelmalla, jolloin saatiin tietoa ohjelman kyvystä käsitellä lämpötilakerrostumia. Kokeiden tuloksissa havaittiin olevan selvää yhteyttä korrelaatioon. Korrelaation kriittistä rajaa suuremmilla arvoilla kylmä vesi kerrostui lämpimän veden alapuolelle. Lämpimän ja kylmän veden väliin muodostui muutaman senttimetrin paksuinen rajakerros, lämpötilakerrostuma, jossa lämpötilan muutos oli suurimmillaan parinkymmenen asteen luokkaa. Tämä lämpötilakerrostuma nousi hitaasti ylöspäin kokeen edetessä. Vastaavasti korrelaation kriittistä rajaa pienemmillä arvoilla lämmin ja kylmä vesi sekoittuivat keskenään. Myös Fluentilla lasketuissa simuloinneissa kylmä vesi kerrostui lämpimän veden alapuolelle. Lämpötilakerrostuma ei kuitenkaan noussut ylöspäin niin kuin kokeessa tapahtui, vaan se seisahtui koelaitteiston yläosaan.

Painevesilaitosten höyrystinkonstruktioiden vertailu

Tässä kirjallisuustyössä tutustutaan kolmeen kaupallisissa painevesilaitoksissa käytettyyn höyrystinkonstruktioon: U-putkipystyhöyrystimeen, U-putkivaakahöyrystimeen ja läpivirtaushöyrystimeen. Työssä tutustutaan näiden ratkaisujen erityispiirteisiin ja vertaillaan niitä keskenään.

Experimental analysis of pressure drop and flow redistribution in axial flows in rod bundles

Fuel elements of PWR type nuclear reactors consist of rod bundles, arranged in a square array, and held by spacer grids. The coolant flows, mainly, axially along the rods. Although such elements are laterally open, experiments are performed in closed type test sections, originating the appearance of subchannels with different geometries. In the present work, utilizing a test section of two bundles of 4x4 pins each, experiments were performed to determine the friction and the grid drag coefficients for the different subchannels and to observe the effect of the grids in the crossflow, in cases of inlet flow maldistribution.

Luonnonkiertovirtauksen kääntyminen pystyhöyrystimen lämmönvaihtoputkissa

Diplomityössä tutkitaan virtauksen kääntymistä Lappeenrannan teknillisen yliopiston PWR PACTEL –koelaitteiston pystyhöyrystimen lämmönvaihtoputkissa käyttäen APROS–prosessisimulointiohjelmaa. Työn teoriaosassa esitellään pystyhöyrystimillä varustettuja koelaitteistoja, erityisesti PWR PACTEL ja sen höyrystin. Lisäksi esitellään virtauksen kääntymisestä tehtyjä havaintoja ja käsitellään kääntymistä teoreettisesta näkökulmasta. Simulointiosan alussa esitellään työssä käytetty APROS –prosessisimulointiohjelma, sekä sen avulla höyrystimestä luodut mallit. Työssä on tutkittu virtauksen käännöstapahtumaa simuloimalla useita eri transienttitilanteita pienillä primäärimassavirroilla. Simulaatiotapauksissa havaittiin virtauksen kääntyvän höyrystimen eripituisissa lämmönvaihtoputkissa, tilanteesta riippuen pääosin lyhimmissä tai toisiksi lyhimmissä lämmönvaihtoputkissa. Transienttien eri vaiheiden, ts. primäärimassavirran muutos- ja tasaantumisvaiheiden pituuden havaittiin vaikuttavan siihen, minkä pituisissa putkissa kääntyminen tapahtuu ja missä järjestyksessä.

El Outsourcing en la dinámica de sistemas hacia mercados extranjeros

Este trabajo recopila literatura académica relevante sobre estrategias de entrada y metodologías para la toma de decisión sobre la contratación de servicios de Outsourcing para el caso de empresas que planean expandirse hacia mercados extranjeros. La manera en que una empresa planifica su entrada a un mercado extranjero, y realiza la consideración y evaluación de información relevante y el diseño de la estrategia, determina el éxito o no de la misma. De otro lado, las metodologías consideradas se concentran en el nivel estratégico de la pirámide organizacional. Se parte de métodos simples para llegar a aquellos basados en la Teoría de Decisión Multicriterio, tanto individuales como híbridos. Finalmente, se presenta la Dinámica de Sistemas como herramienta valiosa en el proceso, por cuanto puede combinarse con métodos multicriterio.

Zeta potential and doping in polyaniline dispersions

Polyaniline (PAni) has been classified as an intractable polymer, particularly in its conducting form, the emeraldine salt (ES). Therefore one can consider the mixture of water and PAni as a suspension. The conducting form of PAni can be obtained by a doping process known as acid doping, in which a strong acid turns PAni from its insulating form, the emeraldine base (EB), into the conducting form, the emeraldine salt. With the objective of establishing a correlation between the doping level and the zeta potential of polyaniline dispersions, polyaniline + HCl aqueous suspensions were prepared. Positive zeta potential values for the various suspensions of PAni showed that it acquired positive charges after the doping process. It was also observed an increase in zeta potential values as HCl concentration increased, which could be correlated to UV-visible spectra of PAni suspensions.

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.

Accident Management Actions In An Upper-Head Small-Break Loss-Of-Coolant Accident With High-Pressure Safety Injection Failed

Since the Three Mile Island accident, an important focus of pressurized water reactor (PWR) transient analyses has been a small-break loss-of-coolant accident (SBLOCA). In 2002, the discovery of thinning of the vessel head wall at the Davis Besse nuclear power plant reactor indicated the possibility of an SBLOCA in the upper head of the reactor vessel as a result of circumferential cracking of a control rod drive mechanism penetration nozzle - which has cast even greater importance on the study of SBLOCAs. Several experimental tests have been performed at the Large Scale Test Facility to simulate the behavior of a PWR during an upper-head SBLOCA. The last of these tests, Organisation for Economic Co-operation and Development Nuclear Energy Agency Rig of Safety Assessment (OECD/NEA ROSA) Test 6.1, was performed in 2005. This test was simulated with the TRACE 5.0 code, and good agreement with the experimental results was obtained. Additionally, a broad analysis of an upper-head SBLOCA with high-pressure safety injection failed in a Westinghouse PWR was performed taking into account different accident management actions and conditions in order to check their suitability. This issue has been analyzed also in the framework of the OECD/NEA ROSA project and the Code Applications and Maintenance Program (CAMP). The main conclusion is that the current emergency operating procedures for Westinghouse reactor design are adequate for these kinds of sequences, and they do not need to be modified.

Dynamic chanel model LMS updating for RSS-based localization

Received signal strength-based localization systems usually rely on a calibration process that aims at characterizing the propagation channel. However, due to the changing environmental dynamics, the behavior of the channel may change after some time, thus, recalibration processes are necessary to maintain the positioning accuracy. This paper proposes a dynamic calibration method to initially calibrate and subsequently update the parameters of the propagation channel model using a Least Mean Squares approach. The method assumes that each anchor node in the localization infrastructure is characterized by its own propagation channel model. In practice, a set of sniffers is used to collect RSS samples, which will be used to automatically calibrate each channel model by iteratively minimizing the positioning error. The proposed method is validated through numerical simulation, showing that the positioning error of the mobile nodes is effectively reduced. Furthermore, the method has a very low computational cost; therefore it can be used in real-time operation for wireless resource-constrained nodes.