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Relevant bibliographies by topics / Water cooled reactors / Dissertations / Theses

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Dissertations / Theses on the topic 'Water cooled reactors'

Author: Grafiati

Published: 4 June 2021

Last updated: 1 August 2024

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1

Zhao, Jiyun Ph D. Massachusetts Institute of Technology. "Stability analysis of supercritical water cooled reactors." Thesis, Massachusetts Institute of Technology, 2005. http://hdl.handle.net/1721.1/34651.

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Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2005.<br>Includes bibliographical references (p. 269-277).<br>The Supercritical Water-Cooled Reactor (SCWR) is a concept for an advanced reactor that will operate at high pressure (25MPa) and high temperature (500°C average core exit). The high coolant temperature as it leaves the reactor core gives the SCWR the potential for high thermal efficiency (45%). However, near the supercritical thermodynamic point, coolant density is very sensitive to temperature which raises concerns about instabilities in the supe

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2

Chan, Ping-lam. "Development of a simulation model for PWR reactor coolant system /." [Hong Kong] : University of Hong Kong, 1989. http://sunzi.lib.hku.hk/hkuto/record.jsp?B1273617X.

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3

Mowrey, James A. "Control system modeling for a boiling water reactor." Thesis, Georgia Institute of Technology, 1995. http://hdl.handle.net/1853/17083.

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4

陳炳林 and Ping-lam Chan. "Development of a simulation model for PWR reactor coolant system." Thesis, The University of Hong Kong (Pokfulam, Hong Kong), 1989. http://hub.hku.hk/bib/B31209130.

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5

Khamis, Ibrahim Ahmad. "Simulation of nuclear power plant pressurizers with application to an inherently safe reactor." Diss., The University of Arizona, 1988. http://hdl.handle.net/10150/184378.

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Pressurizer modeling for predicting the dynamic pressure of the PIUS system is presented. The transient behavior of this model for the PIUS system was investigated. The validity of this model for the PIUS system is limited to transients that are neither too large nor too long in duration. For example, the model is not capable of describing events following a complete loss of liquid for the pressurizer. However, the model can be used for qualitative prediction of the PIUS system behavior for a wide variety of severe transients. A review of pressurizer modeling indicates that the neglecting of t

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6

CARATIN, REINALDO L. "Estudo da imobilização de rejeitos radioativos em matrizes asfálticas e resíduos elastoméricos utilizando a técnica de microondas." reponame:Repositório Institucional do IPEN, 2007. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11574.

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Made available in DSpace on 2014-10-09T12:53:25Z (GMT). No. of bitstreams: 0<br>Made available in DSpace on 2014-10-09T14:09:32Z (GMT). No. of bitstreams: 1 12213.pdf: 5005382 bytes, checksum: c4bde457760b3a6d6f53b64c21e33010 (MD5)<br>Dissertação (Mestrado)<br>IPEN/D<br>Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

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Khoza, Samukelisiwe Nozipho Purity. "Characteristic behaviour of pebble bed high temperature gas-cooled reactors during water ingress events / Samukelisiwe Nozipho Purity Khoza." Thesis, North-West University, 2012. http://hdl.handle.net/10394/8706.

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The effect of water ingress in two pebble bed high temperature gas-cooled reactors i.e. the PBMR-200 MWthermal and the PBMR-400 MWthermal were simulated and compared using the VSOP 99/05 suite of codes. To investigate the effect of this event on reactivity, power profiles and thermal neutron flux profiles, the addition of partial steam vapour pressures in stages up to 400 bar into the primary circuit for the PBMR-400 and up to 300 bar for the PBMR- 200 was simulated for both reactors. During the simulation, three scenarios were simulated, i.e. water ingress into the core only, water ingress in

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8

UMBEHAUN, PEDRO E. "Desenvolvimento de um elemento combustível instrumentado para o reator de pesquisa IEA-R1." reponame:Repositório Institucional do IPEN, 2016. http://repositorio.ipen.br:8080/xmlui/handle/123456789/26935.

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Submitted by Marco Antonio Oliveira da Silva (maosilva@ipen.br) on 2016-12-21T17:48:50Z No. of bitstreams: 0<br>Made available in DSpace on 2016-12-21T17:48:50Z (GMT). No. of bitstreams: 0<br>Após o aumento de potência do reator IEA-R1 de 2 MW para 5 MW observou-se um aumento da taxa de corrosão nas placas laterais de alguns elementos combustíveis e algumas dúvidas surgiram com relação ao valor de vazão utilizada nas análises termo-hidráulicas. A fim de esclarecer e medir a distribuição de vazão real pelos elementos combustíveis que compõe o núcleo do reator IEA-R1, um elemento combustível pr

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9

MELLO, JOSÉ ROBERTO de. "Regulamentação do sistema elétrico do reator IEA-R1." reponame:Repositório Institucional do IPEN, 2016. http://repositorio.ipen.br:8080/xmlui/handle/123456789/26928.

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Submitted by Marco Antonio Oliveira da Silva (maosilva@ipen.br) on 2016-12-21T12:55:48Z No. of bitstreams: 0<br>Made available in DSpace on 2016-12-21T12:55:48Z (GMT). No. of bitstreams: 0<br>O reator IEA-R1 do Instituto de Pesquisas Energéticas e Nucleares (IPENCNEN/ SP) é um reator de pesquisa tipo piscina aberta, projetado e construído pela empresa norte-americana \"Babcock & Wilcox\", tendo, como refrigerante e moderador, água leve deionizada e berílio e grafite como refletores. Até cerca de 1988, os sistemas de segurança do reator recebiam alimentação de uma única fonte de energia. Nos a

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10

SANTOS, DIOGO F. dos. "Caracterização dos campos neutrônicos obtidos por meio de armadilhas de nêutrons a partir da utilização de água pesada (D2O) no interior do núcleo do reator nuclear IPEN/MB-01." reponame:Repositório Institucional do IPEN, 2015. http://repositorio.ipen.br:8080/xmlui/handle/123456789/23825.

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Submitted by Claudinei Pracidelli (cpracide@ipen.br) on 2015-07-23T11:04:17Z No. of bitstreams: 0<br>Made available in DSpace on 2015-07-23T11:04:17Z (GMT). No. of bitstreams: 0<br>Dissertação (Mestrado em Tecnologia Nuclear)<br>IPEN/D<br>Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

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11

Johnson, Kyle D. "High Performance Fuels for Water-Cooled Reactor Systems." Doctoral thesis, KTH, Reaktorfysik, 2016. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-201604.

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Investigation of nitride fuels and their properties has, for decades, been propelled on the basis of their desirable high metal densities and high thermal conductivities, both of which oer intrinsic advantages to performance, economy, and safety in fast and light water reactor systems. In this time several key obstacles have been identied as impeding the implementation of these fuels for commercial applications; namely chemical interactions with air and steam, the noted diculty in sintering of the material, and the high costs associated with the enrichment of 15N. The combination of these limi

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12

Fischer, Kai. "Design of a supercritical water cooled reactor pressure vessel and internals /." Karlsruhe : Forschungszentrum Karlsruhe, 2008. http://d-nb.info/991370759/34.

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13

Fischer, Kai [Verfasser]. "Design of a supercritical water-cooled reactor : pressure vessel and internals / Kai Fischer." Karlsruhe : FZKA, 2008. http://d-nb.info/996911936/34.

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14

Fischer, Kai [Verfasser]. "Design of a supercritical water cooled reactor : pressure vessel and internals / Kai Fischer." Karlsruhe : Forschungszentrum Karlsruhe, 2008. http://nbn-resolving.de/urn:nbn:de:0005-074298.

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15

Guillemin, Perrine. "Recherche de la haute conversion en cycle thorium dans les réacteurs CANDU et REP : Développement des méthodes de simulation associées et étude de scénarios symbiotiques." Grenoble INPG, 2009. http://www.theses.fr/2009INPG0176.

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Ce travail explore la compétitivité des réacteurs actuels (CANDU et REP) en cycle thorium afin de pallier au problème d'utilisation des ressources naturelles en uranium et aux problèmes de disponibilité et de coût des réacteurs de quatrième génération dans le cadre d'un nucléaire durable. Ce travail s'est focalisé sur l'analyse neutronique de deux types de gestion du combustible thorium : la production d'233U à partir de combustible Th/Pu et la conversion d'233U dans ces réacteurs. En particulier, la régénération en CANDU Th/U multirecyclé a été établie. En amont de ces études, les outils et l

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16

Saltanov, Eugene. "Steam-reheat option for supercritical-water-cooled reactors." Thesis, 2010. http://hdl.handle.net/10155/139.

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SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. Main objectives of the development are to increase thermal efficiency of a Nuclear Power Plant (NPP) and to decrease capital and operational costs. The first objective can be achieved by introducing nuclear steam reheat inside a reactor and utilizing regenerative feedwater heaters. The second objective can be achieved by designing a steam cycle that closely matches that of the mature supercritical fossil-fuelled power plants. The feasibility of these objectives is discussed. As

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17

Mokry, Sarah. "Development of a heat-transfer correlation for supercritical water in supercritical water-cooled reactor applications." Thesis, 2009. http://hdl.handle.net/10155/84.

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A large set of experimental data, obtained in Russia, was analyzed and a new heat-transfer correlation for supercritical water was developed. This experimental dataset was obtained within conditions similar to those for proposed SuperCritical Water-cooled nuclear Reactor (SCWR) concepts. Thus, this new correlation, for forced convective heat transfer in the normal heat-transfer regime, can be used for preliminary heat-transfer calculations in SCWR fuel channels. It has demonstrated a good fit for Heat Transfer Coefficient (HTC) values (±25%) and for wall temperature calculations (±15) for the

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18

Grande, Lisa Christine. "Thermal aspects of using alternative nuclear fuels in supercritical water-cooled reactors." Thesis, 2010. http://hdl.handle.net/10155/140.

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A SuperCritical Water-cooled Nuclear Reactor (SCWR) is a Generation IV concept currently being developed worldwide. Unique to this reactor type is the use of light-water coolant above its critical point. The current research presents a thermal-hydraulic analysis of a single fuel channel within a Pressure Tube (PT) - type SCWR with a single-reheat cycle. Since this reactor is in its early design phase many fuel-channel components are being investigated in various combinations. Analysis inputs are: steam cycle, Axial Heat Flux Profile (AHFP), fuel-bundle geometry, and thermophysical properties o

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19

Abdalla, Ayman. "Sensitivity analysis of fuel centerline temperatures in SuperCritical water-cooled reactors (SCWRs)." Thesis, 2012. http://hdl.handle.net/10155/292.

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SuperCritical Water-cooled Reactors (SCWRs) are one of the six nuclear-reactor concepts currently being developed under the Generation-IV International Forum (GIF). A main advantage of SCW Nuclear Power Plants (NPPs) is that they offer higher thermal efficiencies compared to those of current conventional NPPs. Unlike today’s conventional NPPs, which have thermal efficiencies between 30 ‒ 35%, SCW NPPs will have thermal efficiencies within a range of 45 ‒ 50%, owing to high operating temperatures and pressures (i.e., coolant temperatures as high as 625°C at 25 MPa pressure). The use of current

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20

Samuel, Jeffrey. "Conceptual design for a re-entrant type fuel channel for supercritical water-cooled nuclear reactors." Thesis, 2011. http://hdl.handle.net/10155/144.

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Current CANDU-type nuclear reactors use a once-through fuel-channel with an annulus gas insulating it from the moderator. The current reference design for a CANDU-type SuperCritical Water-Cooled Reactor (SCWR) is to eliminate the annulus gap and use a ceramic insert to insulate the coolant from the moderator. While such a design may work, alternative fuel-channel design concepts are under development to explore the optimum efficiency of SCWRs. One such alternative approach is called the Re-Entrant fuel-channel. The Re-Entrant fuel-channel consists of three tubes, the inner tube (flow tube), pr

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21

TASSONE, ALESSANDRO. "Study on liquid metal magnetohydrodynamic flows and numerical application to a water-cooled blanket for fusion reactors." Doctoral thesis, 2019. http://hdl.handle.net/11573/1243658.

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Il breeding blanket \`{e} uno dei componenti chiave per il funzionamento di un reattore a fusione termonucleare, in quanto responsabile dell'estrazione della potenza termica generata dalle reazioni nucleari, della surgenerazione del trizio, e della schermatura per i componenti sensibili alle radiazioni e il personale. I metalli liquidi, come la lega eutettica di piombo e litio (PbLi), sono considerati come attraenti fluidi tecnici da impiegare in questo componente a causa della loro combinazione di eccellenti propriet\`{a} termiche, alta temperatura di ebollizione e capacit\`{a} di generare tr

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22

Merritt, Patrick A. "MCNP benchmarking and analysis of Oregon State University TRIGA Reactor." Thesis, 1999. http://hdl.handle.net/1957/33101.

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23

Chinudomsub, Kittisak. "Analysis and proposed revision of the radiation protection and waste management programs as described in the Oregon State University TRIGA research reactor safety analysis report." Thesis, 1999. http://hdl.handle.net/1957/33385.

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The Safety Analysis Report (SAR) of the Oregon State University TRIGA Research Reactor (OSTR) was prepared and used as a safety baseline for the reactor's operation since 1968. Although, in general, revision of the Safety Analysis Report of a research reactor is not a regulation requirement, it should be revised from time to time to include changes to the facility or procedures or update to current regulatory standards. The ANS 15.21 workgroup developed a draft standard format and content for safety analysis reports for research reactors. An area of this guidance, which was selected for this w

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24

Hallee, Brian Todd. "Feed-and-bleed transient analysis of OSU APEX facility using the modern Code Scaling, Applicability, and Uncertainty method." Thesis, 2013. http://hdl.handle.net/1957/37872.

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The nuclear industry has long relied upon bounding parametric analyses in predicting the safety margins of reactor designs undergoing design-basis accidents. These methods have been known to return highly-conservative results, limiting the operating conditions of the reactor. The Best-Estimate Plus Uncertainty (BEPU) method using a modernized version of the Code-Scaling, Applicability, and Uncertainty (CSAU) methodology has been applied to more accurately predict the safety margins of the Oregon State University Advanced Plant Experiment (APEX) facility experiencing a Loss-of-Feedwater Acciden

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25

Lukomski, Andrew John. "Study on linking a SuperCritical water-cooled nuclear reactor to a hydrogen production facility." Thesis, 2011. http://hdl.handle.net/10155/179.

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The SuperCritical Water-cooled nuclear Reactor (SCWR) is one of six Generation-IV nuclear-reactor concepts currently being designed. It will operate at pressures of 25 MPa and temperatures up to 625°C. These operating conditions make a SuperCritical Water (SCW) Nuclear Power Plant (NPP) suitable to support thermochemical-based hydrogen production via co-generation. The Copper-Chlorine (Cu‒Cl) cycle is a prospective thermochemical cycle with a maximum temperature requirement of ~530°C and could be linked to an SCW NPP through a piping network. An intermediate Heat eXchanger (HX) is considered a

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26

Thind, Harwinder. "Heat-transfer analysis of double-pipe heat exchangers for indirect-cycle SCW NPP." Thesis, 2012. http://hdl.handle.net/10155/227.

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SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. SuperCritical Water (SCW) Nuclear Power Plants (NPPs) are expected to have much higher operating parameters compared to current NPPs, i.e., pressure of about 25 MPa and outlet temperature up to 625 oC. This study presents the heat transfer analysis of an intermediate Heat exchanger (HX) design for indirect-cycle concepts of Pressure-Tube (PT) and Pressure-Vessel (PV) SCWRs. Thermodynamic configurations with an intermediate HX gives a possibility to have a single-reheat option f

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27

Richards, Graham. "Study of heat transfer in a 7-element bundle cooled with the upward flow of supercritical Freon-12." Thesis, 2012. http://hdl.handle.net/10155/228.

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Experimental data on SuperCritical-Water (SCW) cooled bundles are very limited. Major problems with performing such experiments are: 1) small number of operating SCW experimental setups and 2) difficulties in testing and experimental costs at very high pressures, temperatures and heat fluxes. However, SuperCritical Water-cooled nuclear Reactor (SCWRs) designs cannot be finalized without such data. Therefore, as a preliminary approach experiments in SCW-cooled bare tubes and in bundles cooled with SC modeling fluids can be used. One of the SC modeling fluids typically used is Freon-12 (R-12) wh

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28

Peiman, Wargha. "Thermal aspects of high efficiency channel with conventional and alternative fuels in SuperCritical water-cooled reactor (SCWR) applications." Thesis, 2011. http://hdl.handle.net/10155/158.

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Chosen as one of six Generation‒IV nuclear-reactor concepts, SuperCritical Water-cooled Reactors (SCWRs) are expected to have high thermal efficiencies within the range of 45 ‒ 50% owing to reactor‘s high outlet temperatures. A generic pressure-channel (or pressure-tube)SCWR operates at a pressure of 25 MPa with inlet- and outlet-coolant temperatures of 350°C and 625°C. Consequently, the sheath and fuel centerline temperatures are higher in SCWRs than those of the current nuclear reactors. Previous studies have shown that the sheath and fuel centerline temperatures could exceed the design and

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29

DOBEŠ, Petr. "Porovnání výpustí českých a světových JE." Master's thesis, 2007. http://www.nusl.cz/ntk/nusl-47418.

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In this work, which deals with problematics of releases from nuclear power plants, I tried to make an overview of various types of nuclear power plants and radioizotopes which are released through liquid and gasseous effluents. As a part of this comparison evaluation of czech and world nuclear power plants gaseous and liquid releases was made. Introductory part of this work contains information about different types of nuclear power plants and radioizotopes, which are produced in their reactors. It continues with today{\crq}s legislative and information about releasing levels and methods and s

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