Дисертації з теми "Nuclear reactor vessel"
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Everson, Matthew S. "The design of a reduced diameter Pebble Bed Modular Reactor for reactor pressure vessel transport by railcar." Thesis, Massachusetts Institute of Technology, 2009. http://hdl.handle.net/1721.1/53295.
Повний текст джерелаCataloged from PDF version of thesis.
Includes bibliographical references (p. 92).
Many desirable locations for Pebble Bed Modular Reactors are currently out of consideration as construction sites for current designs due to limitations on the mode of transportation for large RPVs. In particular, the PBMR-400 design developed by PBMR Pty of South Africa uses an RPV with an outer diameter of 6.4 meters. Since current SCHNABEL railcars can only haul components up to 4.3 meters wide, the only other possibility for transport is by barge, which limits construction to sites accessible by river, lake or coast. Designing a PBMR with a core able to fit within an RPV able to be transported by railcar would be extremely valuable, especially for potential inland sites only accessible by railway, such as those in the Canadian Oil Sands at which the PBMR would be utilized for oil extraction processes. Therefore, a study was conducted to determine the feasibility of a Pebble Bed Modular Reactor design operating at 250 MWth with a core restricted to fitting inside an RPV with an outer diameter of 4.3 meters. After reviewing the performance of various core configurations satisfying this constraint, an optimized PBMR design operating at this power was found. This new design uses the same fuel management scheme as the PBMR 400, as well as similar inlet and outlet coolant temperatures. This MPBR-250 design includes a pebble bed with an outer diameter of 2.7 meters, an outer reflector 50 cm thick and 12.5% enriched fuel. A mixture of graphite pebbles of 11.7% is also included in the pebble bed to produce an equilibrium core with minimal excess reactivity.
(cont.) This thesis shows that the MPBR-250 can perform up to the standards of the PBMR-400 design with respect to power peaking factors, peak temperatures and RPV fast fluences and can also increase fuel burnup to nearly 110 GWd/T. In addition, the MPBR-250 is a much more agile design, able to be deployed at a wider variety of locations because its RPV can be transported by railcar.
by Matthew S. Everson.
S.M.and S.B.
Tanco, André. "Implementation of 3D-Imaging technique for visual testing in a nuclear reactor pressure vessel." Thesis, KTH, Maskinkonstruktion (Inst.), 2014. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-157475.
Повний текст джерелаDetta examensarbete har utförts på uppdrag av Dekra Industrial AB. Dekra Industrial AB är ett dotterbolag till Dekra. Dekra Industrial AB arbetar främst med kontroller och provningar inom industrin. Kärnkraftindustrin är en industrigren där DEKRA arbetar med sådan kontroll Inspektionerna som utförs består huvudsakligen av oförstörande provning såsom visuell provning. Metoderna som används idag behöver vidareutvecklas och det finns en stark efterfrågan att förbättra den visuella inspektionen. 3D-avbildningsteknik är allt vanligare inom industrin idag och skulle kunna användas som ett mäthjälpmedel för att komplettera den visuella inspektionen. Syftet med examensarbetet är att få en förståelse för hur väl tekniken fungerar samt att föreslå en tillämpning där den kan komma att användas som ett komplement till den visuella inspektionen. Målet med arbetet är att ta fram underlag och föreslå en tillämpning för provning i högstrålande miljö. 3D-avbildningsteknik är ett generellt namn för många olika typer av tekniker som har sina fördelar respektive nackdelar. Arbetet inleds med en litteraturstudie kring 3D-avbildningstekniker, fysik med avseende på avbildningsteknik, den visuella proceduren idag samt hur elektronik påverkas av högstrålande miljö. Information som inte kan fås via studier inhämtas via intervjuer och möten. Tekniken som valdes att analyseras var strukturerat ljus. Tekniken bygger på en trianguleringsprincip som använder en projektor och kamera för att tillförskaffa 3D-koordinater. Genom att projicera mönster på ett objekt kan kameran detektera det reflekterade mönstret och på så vis skapa 3D koordinater. Ett strukturerat ljus system ställdes upp och testades. Testet bestod huvudsakligen av en försöksplanering där de testade faktorerna var trianguleringsvinkel, ljusstyrka och mätavstånd. Testuppställningen som gav bäst resultat var med störst trianguleringsvinkel, högsta ljusstyrka samt kortast mätavstånd. Noggrannheten bestämdes genom att mäta planheten på objektet. Den bästa noggrannheten som uppnåddes med testet var 91.5 μm. Förutom den goda noggrannheten har tekniken visat sin potential genom att avbilda ett svetsprov som genererade ett väldefinierat punktmoln av svetsprofilen. Sammanfattningsvis uppfylldes målen och det uppställda systemet gav en noggrannhet som är jämförbar med en del system ute på marknaden. Detta var möjligt på grund av att en högupplöst stillbildskamera användes. Det finns potential för förbättringar då komponenterna som används i systemet är kommersiella produkter. Nyckelord: Dekra Industrial AB, Visuell inspektion, Avbildningsteknik, Strukturerat ljus
Buongiorno, Jacopo 1971. "Conceptual design of a lead-bismuth cooled fast reactor with in-vessel direct-contact steam generation." Thesis, Massachusetts Institute of Technology, 2001. http://hdl.handle.net/1721.1/32205.
Повний текст джерелаIncludes bibliographical references (p. 357-366).
The feasibility of a lead-bismuth (Pb-Bi) cooled fast reactor that eliminates the need for steam generators and coolant pumps was explored. The working steam is generated by direct contact vaporization of water and liquid metal in the chimney above the core and then is sent to the turbine. The presence of a lighter fluid in the chimney drives the natural circulation of the Pb- Bi within the reactor pool. Three key technical issues were addressed: 1) the maximum thermal power removable by direct contact heat transfer without violating the fuel, clad and vessel temperature limits, 2) the consequences of Pb-Bi aerosol transport on the design and operation of the turbine and 3) the release of radioactive polonium (a product of coolant activation) to the steam. Modeling of the multi-phase phenomena occurring in the chimney confirmed the effectiveness of the direct contact heat transfer mode within a well-defined design envelope for the reactor power, chimney height and steam superheat. A 1260MWth power is found possible for 10m chimney height and 25°C superheat. The temperature of the low-nickel steel clad is maintained below 600°C, which results in limited corrosion if tight control of the coolant oxygen concentration is adopted.
Generation, transport and deposition of Pb-Bi aerosols were also modeled. It was found that the design of a chevron steam separator reduces the heavy liquid metal in the steam lines by about three orders of magnitude. Nevertheless, the residual Pb-Bi is predicted to cause embrittlement of the turbine blades. Four solutions to this problem were assessed: blade coating, employment of alternative materials, electrostatic precipitation and oxidation of the Pb-Bi droplets. An experimental campaign was conducted to investigate the polonium release from a hot Pb- Bi bath to a gas-streamn. Th thermodynamics of the polonium hydride formation reaction (free- energy vs. temperature). as welQ as the vapor pressure of the lead-polonide were measured and then utilized to model the polonium transport in the reactor. It was found that the polonium concentration in the steam and on the surface of the power cycle components is significantly above the acceptable limits, which makes the very concept of a direct contact reactor open to question.
by Jacopo Buongiorno.
Ph.D.
Viehrig, Hans-Werner, Eberhard Altstadt, Mario Houska, Gudrun Mueller, Andreas Ulbricht, Joerg Konheiser, and Matti Valo. "Investigation of decommissioned reactor pressure vessels of the nuclear power plant Greifswald." Helmholtz-Zentrum Dresden - Rossendorf, 2018. http://nbn-resolving.de/urn:nbn:de:bsz:d120-qucosa-235681.
Повний текст джерелаMaples, Allen B. "Design of a robust acoustic positioning system for an underwater nuclear reactor vessel inspection robot." Thesis, This resource online, 1993. http://scholar.lib.vt.edu/theses/available/etd-06232009-063217/.
Повний текст джерелаWells, Peter Benjamin. "The Character, Stability and Consequences of Mn-Ni-Si Precipitates in Irradiated Reactor Pressure Vessel Steels." Thesis, University of California, Santa Barbara, 2016. http://pqdtopen.proquest.com/#viewpdf?dispub=10103547.
Повний текст джерелаFormation of a high density of Mn-Ni-Si nanoscale precipitates in irradiated reactor pressure vessel steels could lead to severe, unexpected embrittlement, which may limit the lifetimes of our nation’s light water reactors. While the existence of these precipitates was hypothesized over 20 years ago, they are currently not included in embrittlement prediction models used by the Nuclear Regulatory Commission. This work aims to investigate the mechanisms and variables that control Mn-Ni-Si precipitate (MNSP) formation as well as correlate their formation with hardening and embrittlement.
A series of RPV model steels with systematic variations in Cu and Ni contents, two variables that have been shown to have a dominant effect on hardening, were irradiated in a series of test reactor and power reactor surveillance irradiations. Atom probe tomography (APT) measurements show that large volume fractions (fv) of MNSPs form in all the steels irradiated at high fluence, even those containing no added Cu, which were previously believed to have low sensitivity to embrittlement. It is demonstrated that while Cu enhances the rate of MNSP formation, it does not appear to significantly alter their saturation fv or composition. The high fluence MNSPs have compositions consistent with known intermetallic phases in the Mn-Ni-Si system and have fv very near those predicted by equilibrium thermodynamic models. In addition, X-ray diffraction experiments by collaborators shows that these precipitates also have the expected crystal structure of the predicted Mn-Ni-Si phases.
Post irradiation annealing experiments are used to measure the hardness recovery at various temperatures as well as to determine if the large f v of MNSPs that form under high fluence neutron irradiation are thermodynamically stable phases or non-equilibrium solute clusters, enhanced or induced by irradiation, respectively. Notably, while post irradiation annealing of a Cu-free, high Ni steel at 425°C results in dissolution of most precipitates, a few larger MNSPs appear to remain stable and may begin to coarsen after long times. A cluster dynamics model rationalizes the dissolution and reduction in precipitate number density, since most are less than the critical radius at the annealing temperature and decomposed matrix composition. The stability of larger precipitates suggests that they are an equilibrium phase, consistent with thermodynamic models.
Charged particle irradiations using Fe3+ ions are also used to investigate the precipitates which form under irradiation. Two steels irradiated to a dose of 0.2 dpa using both neutrons and ions show precipitates with very similar compositions. The ion irradiation shows a smaller f v, likely due to the much higher dose rate, which has been previously shown to delay precipitation to higher fluences. While the precipitates in the ion irradiated condition are slightly deficient in Mn and enriched in Ni and Si compared to neutron irradiated condition, the overall similarities between the two conditions suggest that ion irradiations can be a very useful tool to study the susceptibility of a given steel to irradiation embrittlement.
Finally, the large fv of MNSPs that are shown to form in all steels, including those low in Cu, at high fluence, even those without added Cu, result in large amounts of hardening and embrittlement. A preliminary embrittlement prediction model, which incorporates MNSPs at high fluence, is presented, along with results from a recent test reactor irradiation to fluences representative of extended lifetimes. This model shows very good agreement with the data.
Petersson, Jens. "CFD-analysis of buoyancy-driven flow inside a cooling pipe system attached to a reactor pressure vessel." Thesis, Linköpings universitet, Mekanisk värmeteori och strömningslära, 2014. http://urn.kb.se/resolve?urn=urn:nbn:se:liu:diva-112796.
Повний текст джерелаLongmire, Pamela. "Nonparametric statistical methods applied to the final status decommissioning survey of Fort St. Vrains prestressed concrete reactor vessel." The Ohio State University, 1998. http://rave.ohiolink.edu/etdc/view?acc_num=osu1407398430.
Повний текст джерелаLim, Joven Jun Hua. "Electron microscopy studies of precipitation in nuclear reactor pressure vessel steels under neutron irradiation and thermally ageing." Thesis, University of Oxford, 2014. http://ora.ox.ac.uk/objects/uuid:33daab2c-5c3f-466b-bdd6-0cc022169a6b.
Повний текст джерелаRuan, Xiaoyong. "Structural Integrity Assessment of Nuclear Energy Systems." Kyoto University, 2020. http://hdl.handle.net/2433/253517.
Повний текст джерелаReza, S. M. Mohsin. "Design modification for the modular helium reactor for higher temperature operation and reliability studies for nuclear hydrogen production processes." [College Station, Tex. : Texas A&M University, 2007. http://hdl.handle.net/1969.1/ETD-TAMU-1354.
Повний текст джерелаMatlack, Kathryn H. "Nonlinear ultrasound for radiation damage detection." Diss., Georgia Institute of Technology, 2014. http://hdl.handle.net/1853/51965.
Повний текст джерелаLi, Xiaohua. "Etude des processus de formation des microcavités dans les alliages ferritiques des cuves de réacteurs nucléaires." Université Joseph Fourier (Grenoble), 1996. http://www.theses.fr/1996GRE10010.
Повний текст джерелаŘíha, Tomáš. "Studium radiačního poškození nádoby reaktoru VVER-440 jaderné elektrárny Dukovany." Master's thesis, Vysoké učení technické v Brně. Fakulta strojního inženýrství, 2011. http://www.nusl.cz/ntk/nusl-229835.
Повний текст джерелаAQUINO, CARLOS T. E. de. "Uma Nova abordagem ao fenomeno da varia‡ao da tenacidade a fratura na transi‡ao ductil-fragil de a‡os para vasos de pressao nucleares." reponame:Repositório Institucional do IPEN, 1997. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10663.
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Tese(doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
Hannink, Ryan Christopher. "Investigation of the use of nanofluids to enhance the In-Vessel Retention capabilities of Advanced Light Water Reactors." Thesis, Massachusetts Institute of Technology, 2007. http://hdl.handle.net/1721.1/41314.
Повний текст джерелаIncludes bibliographical references (p. 126-130).
Nanofluids at very low concentrations experimentally exhibit a substantial increase in Critical Heat Flux (CHF) compared to water. The use of a nanofluid in the In-Vessel Retention (IVR) severe accident management strategy, employed in Advanced Light Water Reactors, was investigated. A model simulating the two-phase flow and heat transfer on the reactor vessel outer surface quantified the increase in decay power that can be removed using a nanofluid, predicting that the use of a nanofluid will allow a stable operating power ~40% greater than the power allowable using water to be achieved, while holding the Minimum Departure from Nucleate Boiling Ratio (MDNBR) constant. A nanofluid injection system that would take advantage of the enhanced CHF properties of the nanofluid in order to provide a higher safety margin than the current IVR strategy or, for given margin, enable IVR at higher core power, is proposed. A risk-informed analysis has revealed that this injection system has a reasonably high success probability of 0.99, comparable to the success probability without the injection system. Potential regulatory, environmental, and health risk issues were analyzed, and it was concluded that the current regulatory regimes are adequate for ensuring that the implementation of nanofluids in IVR will not endanger public health and safety. However, experimental verification of nanofluid CHF enhancement at prototypical IVR conditions and periodic nanofluid property testing as a surveillance requirement are needed to reduce the key uncertainties related to nanofluid performance. Finally, a periodic review of the health and environmental risks of nanofluids and, if necessary,follow-up research are ecommended to ensure the health of the public and environment.
by Ryan Christopher Hannink.
S.M.
Tigges, Domini. "Nocivité des défauts sous revêtement des cuves de réacteurs à eau sous pressions." Paris, ENMP, 1995. http://www.theses.fr/1995ENMP0588.
Повний текст джерелаALBUQUERQUE, LEVI B. de. "Categorizacao de tensoes em modelos de elementos finitos de conexoes bocal-vaso de pressao." reponame:Repositório Institucional do IPEN, 1999. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10761.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
Hartnick, Angelo. "Effects of thermal stresses on Pressurised Water Reactor nuclear containment vessels following a Loss of Coolant Accident with assimilated containment filtered venting system." Master's thesis, Faculty of Engineering and the Built Environment, 2021. http://hdl.handle.net/11427/32718.
Повний текст джерелаSantos, Wilton Fogaça da Silva. "Uma nova técnica para contenção de acidentes em reatores nucleares de água pressurizada." Universidade de São Paulo, 2018. http://www.teses.usp.br/teses/disponiveis/3/3139/tde-09042018-144934/.
Повний текст джерелаDuring a severe nuclear power plant accident, the integrity of the reactor pressure vessel must be assured. In response to a possible fuel meltdown, operators of the current generation of nuclear power plants are likely to inject water into the reactor pressure vessel to cool down the reactor vessel wall, preserving its integrity and avoiding leakage of radioactive material. This study considers the use of seawater to flood a reactor pressure vessel combined with the attachment of a honeycomb porous plate (HPP) on the vessel outer wall as a way to improve the safety margins for in-vessel retention of fuel. In long-duration experiments, saturated pool boiling of artificial seawater was performed with an upward-facing plain copper heated surface 30 mm in diameter. The resulting value for critical heat flux (CHF) was 1; 6 MW/m2 at atmospheric pressure, a value significantly higher than the CHF obtained when the working fluid was distilled water (1; 0 MW/m2). It was verified that sea-salt deposits could greatly improve surface wettability and capillarity, enhancing the CHF. The combination of artificial seawater and an HPP attached to the heated surface improved the boiling heat transfer coefficient and increased the CHF up to 110% (2; 1 MW/m2) as compared to distilled water on a bare surface. After the artificial seawater experiments, most of the wall micropores of the HPP were clogged because of sea-salt aggregation on the HPP top and bottom surfaces. Thus, the CHF enhancement observed in this case was attributed mainly to the separation of liquid and vapor phases provided by the HPP channel structure and improvement of surface wettability and capillarity by sea-salt deposition.
Edgar, Christopher Austin. "Improvements to the pool critical assembly benchmark using 3-D discrete ordinate transport with adaptive difference." Thesis, Georgia Institute of Technology, 2013. http://hdl.handle.net/1853/49087.
Повний текст джерелаGraves, Joshua D. "Top-down scaling analysis of the integral reactor vessel test facility." Thesis, 2012. http://hdl.handle.net/1957/36207.
Повний текст джерелаGraduation date: 2013
Hicks, Peter David. "Corrosion fatigue studies in a nuclear pressure vessel steel in simulated pressurized water reactor environments." Thesis, 2015. http://hdl.handle.net/10539/16779.
Повний текст джерелаCHEN, PO-YI, and 陳柏沂. "Investigation on the Design of Reactor Pressure Vessel Water Level Display in the Lungmen Nuclear Power Plant." Thesis, 2009. http://ndltd.ncl.edu.tw/handle/45q24w.
Повний текст джерела國立臺北科技大學
工業工程與管理研究所
97
The RPV(Reactor Pressure Vessel) water level is one of the most critical monitoring parameters in nuclear plant operation, however, the two RPV water level displays in the Lungmen Nuclear Power Plant: Wide Display Panel (WDP) and Video Display Unit (VDU) SDPS C92、C93 displays both are unable to effectively provide operators the RPV water level information; it may also cause the misread of the water level. Therefore, a new RPV water level displays which can assist operators to effectively monitor the water level is essential. In the concerns of feasibility and contribution, this research aim to redesign the VDU SDPS C92、C93 displays on the prerequisite of maintaining the data content. In addition, based on an interface display design theory of EID (Ecological Interface Design), this research combines document research and interview approach in order to elicit the operators’ actual needs of the RPV water level information. The newly designed RPV water level display was evaluated by 17 operators in the Lungmen Nuclear Power Plant by two kinds of questionnaires. Use revised SA (SWORD) questionnaire to evaluate workloads and situation awareness and revised SUS questionnaires to evaluate usability. The research has shown that new RPV water level displays are able to reduce human errors, enhance operator’s performance of and further improve the safety of nuclear plant.
Wei, Hong-Lin, and 魏宏霖. "Effect of Clad Thickness on Reliability of Reactor Pressure Vessels in Nuclear Power Plants." Thesis, 2011. http://ndltd.ncl.edu.tw/handle/80415965040954219618.
Повний текст джерела國立臺灣大學
機械工程學研究所
99
Nowadays, we are facing problems of global energy shortage as well as the need of environmental protection. The advantage of low cost and small amount of CO2 discharge makes nuclear power an important choice for energy. However, the safety of structures, systems and mechanical components employed in a nuclear power plant has to be assured before a plant can be constructed. One of the most important pressure boundary components in the steam supply system of a nuclear power plant is the reactor pressure vessel (RPV). It is welded together by several steel plates. Cracks occur more frequently in welds rather than in base plates of a RPV. When a predominant crack grows along with operating time to a certain size, it may result in brittle fracture in the weld of a RPV. It has been pointed out that clad thickness and crack size affect the embrittlement and fracture of the weld. The present study employs a probabilistic fracture mechanics approach by taking into account radiation embrittlement to find fracture-failure probabilities of RPV welds. The result shows that, when the clad is thicker than 0.35 inch, the failure probability at axial weld should be paid more attention to. As for the effect of inspection and repair, it is found that adopting a more advanced inspection instrument reduces failure probability more than increasing inspection cycles or covering more inspection areas. It is also found the probability of failure at circumferential welds is smaller than that at axial welds. The finding reassures the proposition made by the United States Nuclear Regulatory Commission (USNRC) that inspection of circumferential welds of a RPV can be exempted.