Dissertations / Theses on the topic 'NUCLEAR REACTOR SYSTEMS'
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Anadani, Mohamed. "Decision support systems for nuclear reactor control." Thesis, University of Sheffield, 2000. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.341828.
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Presby, Andrew L. "Thermophotovoltaic energy conversion in space nuclear reactor power systems." Thesis, Monterey, Calif. : Naval Postgraduate School, 2004. http://edocs.nps.edu/npspubs/scholarly/theses/2004/Dec/04Dec%5FPresby.pdf.
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Thesis (Astronautical Engineer and M. S. in Astronautical Engineering)--Naval Postgraduate School, December 2004.Thesis Advisor(s): Gopinath, Ashok ; Michael, Sherif. "December 2004." Description based on title screen as viewed on March 13, 2009. Includes bibliographical references (p. 123-127). Also available in print.
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CARVALHO, LUIZ S. "Frequencia de danos no nucleo por blecaute em reator nuclear de concepcao avancada." reponame:Repositório Institucional do IPEN, 2004. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11147.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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Kim, Choong Seok. "Reliability assessment of pressurized water reactor auxiliary feedwater systems." Diss., Georgia Institute of Technology, 1985. http://hdl.handle.net/1853/13374.
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Persson, Carl-Magnus. "Reactivity Assessment in Subcritical Systems." Licentiate thesis, Stockholm : Fysiska institutionen, Kungliga Tekniska högskolan, 2007. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-4363.
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Witter, Jonathan Keay. "Modeling for the simulation and control of nuclear reactor rocket systems." Thesis, Massachusetts Institute of Technology, 1993. http://hdl.handle.net/1721.1/12755.
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Wu, Xiao. "Design of a Tritium Mitigation and Control System for Fluoride-salt-cooled High-temperature Reactor Systems." The Ohio State University, 2016. http://rave.ohiolink.edu/etdc/view?acc_num=osu1452249907.
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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 limitations, historically, led to the well founded conclusion that the most appropriate use of nitride fuels was in the fast reactor fuel cycle, where the cost burdens associated with them is substantially less. Indeed, it is within this context that the vast majority of work on nitrides has been and continues to be done. Nevertheless, following the 2011 Fukushima-Daiichi nuclear accident, a concerted governmental-industrial eort was embarked upon to explore the alternatives of so-called \accident tolerant" and \high performance" fuels. These fuels would, at the same time, improve the response of the fuel-clad system to severe accidents and improve the economy of operation for light water reactor systems. Among the various candidates proposed are uranium nitride, uranium silicide, and a third \uranium nitride-silicide" composite featuring a mixture of the former. In this thesis a method has been established for the synthesis, fabrication, and characterization of high purity uranium nitride, and uranium nitride-silicide composites, prepared by the spark plasma sintering (SPS) technique. A specic result has been to isolate the impact of the processing parameters on the microstructure of representative fuel pellets, essentially permitting any conceivable microstructure of interest to be fabricated. This has enabled the development of a highly reproducible technique for the production of pellets with microstructures tailored towards any desired porosity between 88-99.9%TD, any grain size between 6-24 μm, and, in the case of the uranium nitride-silicide composite, a silicide-coated UN matrix. This has permitted the evaluation of these microstructural characteristics on the performance of these materials, specically with respect to their role as accident tolerant fuels. This has generated results which have tightly coupled nitride performance with pellet microstructure, with important implications for the use of nitrides in water-cooled reactors.Under artionden har forskning om nitridbranseln och dess egenskaper bedrivits pa grundval av nitridbransletsatravarda egenskaper avseende dess hoga metall tathet och hog varmeledningsformaga. Dessa egenskaper besitter vasentliga fordelar avseende prestanda, ekonomi och sakerhet for metallkylda som lattvatten reaktorer. Genom forskning har aven centrala begr ansningar identierats for implementering av nitridbranslen for kommersiellt bruk. Begransningar avser den kemiska interaktionen med luft och vattenanga, en uppmarksammad svarighet att sintring av materialet samt hoga kostnader forknippade med den nodvandiga anrikningen av 15-N. Kombinationen av dessa begransningar resulterade, tidigare, i en valgrundad slutsats att nitridbranslet mest andamalsenliga anvandningsomrade var i karnbranslecykeln for snabba reaktorer. Detta da kostnaderna forenade med implementeringen av branslet ar avsevart lagre. Inom detta sammanhang har majoriteten av forskning avseende nitrider bedrivits och fortskrider an idag. Dock, efter karnkraftsolyckan i Fukushima-Daiichi 2011, inleddes en samlad industriell och statlig anstrangning for att undersoka alternativ till sa kallade \olyckstoleranta" och \hogpresterande" branslen. Dessa branslen skulle samtidigt forbattra reaktionstiden for bransleinkapsling systemet mot allvarliga olyckor samt forbattra driftsekonomin av lattvattenreaktorer. Foreslagna kandidater ar urannitrid, uransilicid och en tredje \uran nitrid-silicid", komposit bestaende av en blandning av de foregaende. Genom denna avhandling har en metod faststallts for syntes, tillverkning och karaktarisering av uran nitrid av hog renhet samt uran nitrid-silicid kompositer, forberedda med tekniken SPS (Spark Plasma Sintering). Ett specikt resultat har varit att isolera eekten av processparametrar pa mikrostrukturen pa representativa branslekutsar. Detta mojliggor, i princip, framstallningen av alla tankbara mikrostrukturer utav intresse for tillverkning. Vidare har detta mojliggjort utvecklingen av en hogeligen reproducerbar teknik for framstallningen av branslekutsar med mikrostrukturer skraddarsydda for onskad porositet mellan 88 och 99.9 % TD, och kornstorlek mellan 6 och 24 μm. Dartill har en metod for att belagga en matris av uran nitrid-silicid framarbetats. Detta har mojliggjort utvarderingen av dessa mikrostrukturella parametrars paverkan pa materialens prestanda, sarskilt avseende dess roll som olyckstoleranta branslen. Detta har genererat resultat som ar tatt sammanlankat nitridbranslets prestanda till kutsens mikrostruktur, med viktiga konsekvenser for den potentiella anvandningen av nitrider i lattvatten reaktorer.
QC 20170210
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CONCEICAO, JUNIOR OSMAR. "Aplicacao da tecnica de analise de modos de falha e efeitos ao sistema de resfriamento de emergencia de uma instalacao nuclear experimental." reponame:Repositório Institucional do IPEN, 2009. http://repositorio.ipen.br:8080/xmlui/handle/123456789/9367.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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Morrison, Jonathan J. "Corrosion, transport, and deposition in pressurised water nuclear reactor primary coolant systems." Thesis, University of Birmingham, 2016. http://etheses.bham.ac.uk//id/eprint/6816/.
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Several unscheduled shut downs of the Cruas nuclear power plant in France have been caused by the deposition of corrosion products in flow broaches of the steam generator tube support sheets. The depositions are theorised to be the result of electrokinetically stimulated deposition. In this work, a hot water loop to replicate these depositions in the laboratory was built, along with rigs to characterise supporting phenomena – the corrosion rate of stainless steel and the solubility of the corrosion products. While the data obtained from the hot water loop did not provide conclusive proof of the existence or prevalence of the electrokinetically stimulated deposition mechanism, evidence of deposition caused by cavitation was found. The corrosion rate of stainless steel was measured at high temperatures in solutions of lithium hydroxide at various concentrations. Surface finish was found to have an effect on the corrosion rate, though the difference between mechanically ground surfaces with an order of magnitude difference in roughness was found to be minimal. The solubility of the corrosion products formed was measured and found to be of similar order to that reported in the literature, however the minor alloying elements were found to leach from the surface in substantial quantities.
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Thiele, Roman. "Mechanistic Modeling of Wall-Fluid Thermal Interactions for Innovative Nuclear Systems." Doctoral thesis, KTH, Reaktorteknologi, 2015. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-177370.
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Next generation nuclear power plants (GEN-IV) will be capable of not only producing energy in a reliable, safe and sustainable way, but they will also be capable of reducing the amount of nuclear waste, which has been accumulated over the lifetime of current-generation nuclear power plants, through transmutation. Due to the use of new and different coolants, existing computational tools need to be tested, further developed and improved in order to thermal-hydraulically design these power plants.This work covers two different non-unity Prandtl number fluids which are considered as coolants in GEN-IV reactors, liquid lead/lead-bismuth-eutectic and supercritical water. The study investigates different turbulence modeling strategies, such as Large Eddy Simulation (LES) and Reynolds-Averaged Navier-Stokes (RANS) modeling, and their applicability to these proposed coolants. It is shown that RANS turbulence models are partly capable of predicting wall heat transfer in annular flow configurations. However, improvements in these prediction should be possible through the use of advanced turbulence modeling strategies, such as the use of separate thermal turbulence models. A large blind benchmark study of heat transfer in supercritical water showed that the available turbulence modeling strategies are not capable of predicting deteriorated heat transfer in a 7-rod bundle at supercritical pressures. New models which take into account the strong buoyancy forces and the rapid change of the molecular Prandtl number near the wall occurring during the transition of the fluid through the pseudocritical point need to be developed. One of these strategies to take into account near-wall buoyancy forces is the use of advanced wall functions, which cannot only help in modeling these kind of flows, but also decrease computational time by 1 to 2 orders of magnitude. Different advanced wall function models were implemented in the open-source CFD toolbox OpenFOAM and their performance for different flows in sub- and supercritical conditions were evaluated. Based on those results, the wall function model UMIST-A by Gerasimov is recommended for further investigation and specific modeling tactics are proposed.Near-wall temperature and velocity behavior is important to and influenced by the wall itself. The thermal inertia of the wall influences the temperature in the fluid. However, a more important issue is how temperature fluctuations at the wall can induce thermal fatigue. With the help of LES thermal mixing in a simplified model of a control rod guide tube was investigated, including the temperature field inside the control rod and guide tube walls. The WALE sub-grid turbulence model made it possible to perform LES computations in this complex geometry, because it automatically adapts to near-wall behavior close to the wall, without the use of ad-hoc functions. The results for critical values, such as the amplitude and frequency of the temperature fluctuations at the wall, obtained from the LES computations are in good agreement with experimental results.The knowledge gained from the aforementioned investigations is used to optimize the flow path in a small, passively liquid-metal-cooled pool-type GEN IV reactor, which was designed for training and education purposes, with the help of 3D CFD. The computations were carried out on 1/4 of the full geometry, where the small-detail regions of the heat exchangers and the core were modeled using a porous media approach. It was shown that in order to achieve optimal cooling of the core without changing the global geometry a ratio of close to unity of the pressure drop over the core and the heat exchanger needs to be achieved. This is done by designing a bottom plate which channels enough flow through the core without choking the flow in the core. Improved cooling is also achieved by reducing heat losses from the hot leg through the flow shroud to the cold leg by applying thermal barrier coating similar to methods used in gas turbine design.Nästa generations kärnkraftverk (GEN-IV) kan inte bara producera el på ett pålitligt, säkert och hållbart sätt, utan det kan också reducera mängden kärnavfall, som har producerats under tiden som man använt nuvarande generationen kärnkraftverk, genom att transmutera avfallen. Framtidens kärnkraftverk använder andra kylmedel än nuvarande kraftverk som t.ex. flytande bly, gas eller superkritiskt vatten. Det betyder att många beräkningsverktyg måste testas, utvecklas och förbättras så att man kan genomföra termohydrauliska designberäkningar. Den här avhandlingen omfattar två olika kylmedel, flytande bly och superkritiskt vatten, som har ett Prandtl-tal som skiljer sig från 1 och kommer att användas i GEN-IV reaktorer. Studien undersöker olika strategier för att modellera turbulens som Large Eddy Simulation (LES) och Reynolds-Averaged Navier-Stokes (RANS) och hur man kan använda dessa strategierna i beräkningar av strömning och värmetransfer i den nya kylvätskan. Undersökningen visar att RANS turbulensmodeller delvis kan förutsäga värmeöverföringen vid en vägg i en ringformad strömningsgeometri. Förbättringar av förutsägelsen ska vara möjlig genom användning av avancerade strategier för turbulensmodellering, t.ex. termiska turbulensmodeller. En stor prestandajämförelse för värmeöverföring i superkritiskt vatten visade att ingen av nuvarande strategier för turbulensmodellering kan förutsäga försämrad värmeöverföring i en 7-stavknippet under superkritiskt tryck. Nya modeller, som omfattar de starka flytkrafterna och den snabba förändringen av den molekulära Prandtl-tal vid väggen som uppstår när vätskan går genom pseudokritiska punkten, måste utvecklas. Avancerade väggfunktioner är en av strategierna som kan ta hänsyn till dessa fenomen. Väggfunktioner kan inte bara hjälpa till att modellera de typer av flöden som behövs utan kan också hjälpa till att sänka beräkningstiden med en eller två tiopotenser. Olika avancerade väggfunktioner i open-source beräkningsverktyget OpenFOAM implementerades och deras prestation i sub- och superkritiska vattenflödar värderades. Baserat på detta rekommenderas Gerasimovs modell för ytterligare utredning. Dessutom läggs olika strategier fram för att utöka modellens validitet till flöde med superkritiskt vatten i sammanband med försämrad och förbättrad värmeöverföring. Kunskap om beteendet av temperatur och hastighet i väggens närhet är viktigt för väggens integritet, detta då väggen även påverkar beteendet. Väggens termiska tröghet påverkar flödets temperatur och hastighet. Dock är ett ännu viktigare problem, som kan uppträda, är att temperaturfluktuationer kan framkalla termisk utmattning i en vägg. Med användning av LES utreds termisk blandning av varmt och kallt vatten i en simplifierad modell av ett styrstavsledrör, inklusive temperaturfältet i styrstaven och ledrörsväggen. Användningen av WALE LES-turbulensmodellen gör det möjligt att utföra beräkningar i den komplexa geometrin, detta eftersom modellen anpassar sig automatiskt till fenomenen nära väggen utan användning av ad-hoc funktioner. LES resultaten för alla värden som är viktiga för att bestämma utmattningsbeteende, som amplitud och frekvens av temperaturfluktuationer i väggens närhet och i väggen själv, är i god överensstämmelse med resultaten från experiment från KTH i samma geometri.Kunskapen som vunnits genom ovannämnda utredningar användes för att optimera den termohydrauliska designen av en liten, pool-typ GEN-IV reaktor som är passivt kyld med flytande bly. Reaktorn är designad som en utbildnings- och träningsreaktor och optimeringen genomfördes med hjälp av 3D CFD. Beräkningarna genomfördes på en fjärdedel av reaktorns hela geometrin. Regioner med små detaljer, som de åtta värmeväxlarna och reaktorns kärna, modellerades genom porösa material. Det visar sig att för att ha en optimal kylning av kärnan, utan att förändra reaktorns globala geometri, måste förhållandet mellan tryckförlust i reaktorkärnan och värmeväxlarna vara nära 1. Detta uppnås genom att designa plattan vid ingången till kärnan så att tillräckligt med bly flödar genom kärnan utan att kväva flödet i denna. Ytterligare en förbättring i reaktorkylningen uppnås genom att reducera värmeförlusten genom väggen som skiljer varm och kall vätska. Detta görs med en strategi som förekommer i gasturbinteknologin, genom att man lägger till ett tunt skikt av termiskt isolerande material på väggen, som reducerar värmeöverföring med ungefär 50%.
QC 20151123
THEMFA
GENIUS
THINS
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Ruan, Xiaoyong. "Structural Integrity Assessment of Nuclear Energy Systems." Kyoto University, 2020. http://hdl.handle.net/2433/253517.
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ABRATE, NICOLO'. "Methods for safety and stability analysis of nuclear systems." Doctoral thesis, Politecnico di Torino, 2022. http://hdl.handle.net/11583/2971611.
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FERREIRA, JUNIOR DECIO B. M. "Desenvolvimento de um sistema computacional para monitoracao dos parametros de reatividade e das oscilacoes axiais de xenonio do reator nuclear de Agra 1." reponame:Repositório Institucional do IPEN, 2001. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10918.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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ROSSI, ROSA H. P. S. "Utilizacao de redes neurais na monitoracao da potencia do reator IEA-R1." reponame:Repositório Institucional do IPEN, 2001. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10895.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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Lundström, Tim. "Radiation chemistry of aqueous solutions related to nuclear reactor systems and spent fuel management /." Linköping : Univ, 2003. http://www.bibl.liu.se/liupubl/disp/disp2003/tek840s.pdf.
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Szakaly, Frank Joseph. "Assessment of uranium-free nitride fuels for spent fuel transmutation in fast reactor systems." Thesis, Texas A&M University, 2003. http://hdl.handle.net/1969.1/31.
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The purpose of this work is to investigate the implementation of nitride fuels containing little or no uranium in a fast-spectrum nuclear reactor to reduce the amount of plutonium and minor actinides in spent nuclear fuel destined for the Yucca Mountain Repository. A two tier recycling strategy is proposed. Thermal spectrum transmutation systems converted from the existing LWR fleet were modeled for the first tier, and the Japanese fast reactor MONJU was used for the fast-spectrum transmutation. The modeling was performed with the Monteburns code. Transmutation performance was investigated as well as delayed neutron fraction, heat generation rates, and radioactivity of the spent material in the short and long term for the different transmutation fuel cycles. A two-tier recycling strategy incorporating fast and thermal transmutation with uranium-free nitride fuel was shown to reduce the long-term heat generation rates and radioactivity of the spent nuclear fuel inventory.
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OLIVEIRA, JOSE R. de. "Programa computacional para estudo da estrategia de controle de um reator nuclear do tipo PWR." reponame:Repositório Institucional do IPEN, 2002. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11060.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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Singh, Mohit S. M. Massachusetts Institute of Technology. "Ssessment methodology for proliferation resistant fast breeder reactor." Thesis, Massachusetts Institute of Technology, 2014. http://hdl.handle.net/1721.1/92085.
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Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014.Thesis: S.M. in Technology and Policy, Massachusetts Institute of Technology, Engineering Systems Division, Technology and Policy Program, 2014.
Cataloged from PDF version of thesis.
Includes bibliographical references (pages 70-72).
Due to perceived proliferation risks, current US fast reactor designs have avoided the use of uranium blankets. While reducing the amount of plutonium produced, this omission also restrains the reactor design space and has several disadvantages over blanketed cores. This study investigated many blanket options that would satisfy the proliferation concern while minimizing negative fuel cycle impact. To do so, a multi-variable metric was developed that combines 6 attributes: proliferation resistance, fuel fabrication, radiotoxicity, breeding gain, reactivity penalty and transportation. The final version of the metric consisted of using a yes or no decision on the proliferation criteria proposed by Bathke (for technologically advanced nations). The remaining 5 attributes are scaled between 0 and 1 with assigned weights for each. For our analysis, a 2400MWth sodium cooled core was considered. One row of blanket was added radially. Metal fuel composed of depleted uranium, zirconium and Np/Pu from light water reactor used fuel was used for the driver. It was determined that to meet the prescribed proliferation resistance criteria, a minimum of 4% MA (by volume) was needed in the blanket assemblies. However, increasing the amount of MA past 4% became detrimental to the combination of the other 5 attributes, mainly impacting the radiotoxicity, fuel fabrication and transportation. The addition of moderation by itself did not provide any means of dissipating proliferation issues. In the cases studied, it was determined that ZrH1.6 and BeO were the most promising moderating materials. They both provided some reduction in required MA concentration but at the expense of the radiotoxicity of the end product. Using our defined metric, it was determined that moderation provided no immediate benefit. It should also be noted that the homogeneous or heterogeneous addition of moderators has minimal impact on such scoping studies. Separation of the Cm/Bk/Cf vector from the Am was also studied. The blankets were composed of Am while the remaining Cm/Bk/Cf was left to decay in storage. The metric was then applied to the combined streams for all attributes except proliferation. The separated case performed worst in all cases examined. Also, as expected, varying the uranium composition vector from natural (NU), depleted (DU) and recycled (RU) had very little impact on our metric, thus the choice of uranium vector would be mostly left to cost and initial fabrication considerations. It should however be noted that the k-infinity at beginning-of-life was obviously higher for the recycled and natural cases. Looking at the reactivity over the first cycle indicates that NU provides an additional -40pcm over DU while RU provides -60pcm, which could provide 30 and 45 extra days of operation, respectively, or a reduction in driver core enrichment for a given cycle length.
by Mohit Singh.
S.M.
S.M. in Technology and Policy
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Berglöf, Carl. "On measurement and monitoring of reactivity in subcritical reactor systems." Doctoral thesis, KTH, Reaktorfysik, 2010. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-12483.
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Accelerator-driven systems have been proposed for incineration of transuranic elements from spent nuclear fuel. For safe operation of such facilities, a robust method for reactivity monitoring is required. Experience has shown that the performance of reactivity measurement methods in terms of accuracy and applicability is highly system dependent. Further investigations are needed to increase the knowledge data bank before applying the methods to an industrial facility and to achieve license to operate such a facility. In this thesis, two systems have been subject to investigation of various reactivity measurement methods. Conditions for successful utilization of the methods are presented, based on the experimental experience. In contrast to previous studies in this field, the reactivity has not only been determined, but also monitored based on the so called beam trip methodology which is applicable also to non-zero power systems. The results of this work constitute a part of the knowledge base for the definition of a validated online reactivity monitoring methodology for facilities currently being under development in Europe (XT-ADS and EFIT).QC 20100621
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Ablay, Gunyaz. "Sliding Mode Approaches for Robust Control, State Estimation, Secure Communication, and Fault Diagnosis in Nuclear Systems." The Ohio State University, 2012. http://rave.ohiolink.edu/etdc/view?acc_num=osu1354551858.
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Zamxaka, Lwandiso Lindani. "The impact of quality management systems during a pebble bed modular reactor project. A case study." Thesis, Cape Peninsula University of Technology, 2010. http://hdl.handle.net/20.500.11838/1226.
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Thesis(Mtech (Industrial Engineering)--Cape Peninsula University of Technology, 2010In the nuclear industry, Quality Management Systems are extremely important, especially if one wishes to improve public acceptance of radioactive solutions. There is normally minimum communication between the public and scientists, especially in nuclear science. People are not comfortable with nuclear technology, based on the past history of the Chernobyl catastrophe. Consequently, it is difficult to discuss important and sensitive issues like disposing of nuclear waste. Quality Management Systems can improve public confidence and communication. Integrated Management Systems in the project planning stage of the project can be a proactive step towards preventing unnecessary delays and costs. There is a perception that quality is implemented or executed at the implementation stage of the Project Life cycle. Most people believe that a Quality Management System is quality control only and forget the aspect of Quality assurance. The project managers are more concerned with finishing the project and saving costs. Quality holds together the three pillars of project management, which are schedule, costs and scope. There are a plethora of things that can go wrong if the Quality Management System is not implemented on time, like scope changes that are not captured, monitored and controlled. This can lead to scope creep, unnecessary costs and schedule overruns. If there is no cost control, the project can also overrun its budget and consequently be stopped. PBMR is the only company that is active in new nuclear projects in South Africa, except Koeberg, which was commissioned about thirty years ago.
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VALERIO, DOMENICO. "Multi-physics modelling of liquid metals in Advanced Nuclear Systems." Doctoral thesis, Politecnico di Torino, 2022. http://hdl.handle.net/11583/2970997.
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UGGENTI, ANNA CHIARA. "Safety assessment of next generation nuclear systems: methodology development and case studies on fission and fusion devices." Doctoral thesis, Politecnico di Torino, 2019. http://hdl.handle.net/11583/2762332.
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Moisseytsev, Anton. "Passive load follow analysis of the STAR-LM and STAR-H2 systems." Texas A&M University, 2003. http://hdl.handle.net/1969.1/390.
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A steady-state model for the calculation of temperature and pressure distributions, and heat and work balance for the STAR-LM and the STAR-H2 systems was developed. The STAR-LM system is designed for electricity production and consists of the lead cooled reactor on natural circulation and the supercritical carbon dioxide Brayton cycle. The STAR-H2 system uses the same reactor which is coupled to the hydrogen production plant, the Brayton cycle, and the water desalination plant. The Brayton cycle produces electricity for the on-site needs. Realistic modules for each system component were developed. The model also performs design calculations for the turbine and compressors for the CO2 Brayton cycle. The model was used to optimize the performance of the entire system as well as every system component. The size of each component was calculated.
For the 400 MWt reactor power the STAR-LM produces 174.4 MWe (44% efficiency) and the STAR-H2 system produces 7450 kg H2/hr. The steady state model was used to conduct quasi-static passive load follow analysis. The control strategy was developed for each system; no control action on the reactor is required. As a main safety criterion, the peak cladding temperature is used. It was demonstrated that this temperature remains below the safety limit during both normal operation and load follow.
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Skwarcan-Bidakowski, Alexander. "Nuclear reactor core model for the advancednuclear fuel cycle simulator FANCSEE. Advanceduse of Monte Carlo methods in nuclear reactorcalculations." Thesis, Institutionen för Reaktorfysik, 2017. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-324260.
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A detailed reactor core modeling of the LOVIISA-2 PWR and FORSMARK-3BWR was performed in the Serpent 2 Continuous Energy Monte-Carlocode.Both models of the reactors were completed but the approximations ofthe atomic densities of nuclides present in the core differedsignificantly.In the LOVIISA-2 PWR, the predicted atomic density for the nuclidesapproximated by Chebyshev Rational Approximation method (CRAM)coincided with the corrected atomic density simulated by the Serpent2 program. In the case of FORSMARK-3 BWR, the atomic density fromCRAM poorly approximated the data returned by the simulation inSerpent 2. Due to boiling of the moderator in the core of FORSMARK-3,the model seemed to encounter problems of fission density, whichyielded unusable results.The results based on the models of the reactor cores are significantto the FANCSEE Nuclear fuel cycle simulator, which will be used as adataset for the nuclear fuel cycle burnup in the reactors.FANCSEE
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Popescu, George. "Digital Signal Processing Methods for Safety Systems Employed in Nuclear Power Industry." University of Cincinnati / OhioLINK, 2016. http://rave.ohiolink.edu/etdc/view?acc_num=ucin1479815935917872.
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Laubscher, Ryno. "Development aspects of a high temperature heat pipe heat exchanger for high temperature gas-cooled nuclear reactor systems." Thesis, Stellenbosch : Stellenbosch University, 2013. http://hdl.handle.net/10019.1/80096.
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Thesis (MScEng)--Stellenbosch University, 2013.ENGLISH ABSTRACT: High temperature heat sources are becoming an ever-increasing imperative in the process industry for the production of plastics, ammonia and fertilisers, hydrogen, coal-toliquid fuel and process heat. Currently, high temperature reactor (HTR) technology is capable of producing helium temperatures in excess of 950°C; however, at these temperatures, tritium, which is a radioactive contaminant found in the helium coolant stream, is able to diffuse though the steel retaining wall of the helium-to-steam heat exchanger. To circumvent this radioactivity problem, regulations require an intermediate heat exchange loop between the helium and the process heat streams. In this paper, the use of a uniquely designed sodium-charged heat pipe heat exchanger is considered, and has the distinct advantage of having almost zero exergy loss as it eliminates the intermediate heat exchange circuit. In order to investigate this novel heat pipe heat exchanger concept, a special intermediate-temperature (± 240°C) experimental heat pipe heat exchanger (HPHE) was designed. This experimental HPHE uses Dowtherm A as working fluid and has two glass windows to enable visual observation of the boiling and condensation two-phase flow processes. A high temperature air-burner supply simulates the high temperature stream, and the cold stream is provided by water from a constant-heat supply tank. This experimental apparatus can be used to evaluate the validity of steady-state and start-up transient theoretical models that have been developed. This paper will highlight the special design aspects of this HPHE, the theoretical model and the solution algorithm described. Experimental results will be compared with the theoretically calculated results. The theoretical model will then be used to predict the performance of a high temperature (sodium working fluid at 850°C) HPHE will be undertaken and conclusions and recommendation made.
AFRIKAANSE OPSOMMING: Hoë temperatuur hitte bronne is besig om ‘n toenemende noodsaaklikheid te raak in die proses industrie vir die vervaardiging van plastieke, ammoniak, kunsmis, waterstof, steenkool-tot-vloeibare brandstof en proses hitte. Huidige hoë temperatuur reaktor tegnologie is in staat om helium te verhit tot temperature hoër as 950°C, maar by sulke hoë temperature is die vorming van tritium, wat ‘n radioaktiewe produk is, in die helium verkoeling stroom wat deur die reaktor vloei, ‘n probleem. Die tritium is in staat om deur die staal wand van ‘n enkel fase warmte uitruiler te diffundeer. Om hierdie radioaktiewe probleem te uitoorlê, stel huidige regulasies voor dat ‘n oorgangs hitte uitruil lus gebruik raak tussen die helium en proses strome van die reaktor stelsel. In hierdie tesis word ‘n unieke natrium gevulde hitte pyp warmte uitruiler nagevors, hierdie ontwerp het die voordeel dat dit geen “exergy” verlies het omdat dit nie ‘n oorgangs hitte uitruil lus benodig nie. Hierdie unieke konsep was nagevors deur ‘n spesiale oorgangs temperatuur (± 230°C) eksperimentiële hitte pyp warmte uitruiler te ontwerp. Hierdie eksperimentiële hitte pyp warmte uitruiler gebruik Dowtherm A as oordrags medium tussen die warm en koue strome en het twee glas venters waardeur die kook en kondensasie van die oorgangs medium dop gehou kan word. ‘n Hoë temperatuur verbrander simuleer die warm stroom deur die reaktor en die koue stroom word gesimuleer deur koue water. Die eksperimentiële opstelling sal gebruik word om die tyd afhangklike en tyd onafhangklike teoretiese wiskundige modele te valideer. Hierdie tesis sal die spesiale ontwerp aspekte van die hitte pyp warmte uitruiler, teoretiese modelle en oplos algoritme te bespreek. Eksperimentiele resultate sal met die teoretiese resultate vergelyk word en dan sal die teoretiese modelle gebruik word om ‘n natrium gevulde warmte uitruiler te simuleer. Gevolgtrekkings en aanbevelings sal in die lig van die resultate verskaf word.
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PONTES, EDUARDO W. "Analise de sistemas de medicao de fluxo de neutrons utilizando funcoes estatisticas." reponame:Repositório Institucional do IPEN, 1997. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10648.
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Tese (Doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
30
Bhattacharyya, Sampriti. "Reliability Analysis and Controls for Accelerator Driven Systems Based On Project X." The Ohio State University, 2012. http://rave.ohiolink.edu/etdc/view?acc_num=osu1343340152.
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BONFIETTI, GERSON. "Analise da confiabilidade do sistema de suprimento de energia eletrica de emergencia de um reator nuclear de pequeno porte." reponame:Repositório Institucional do IPEN, 2003. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11129.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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BOCANEGRA, CIFUENTES JOHAN AUGUSTO. "Lattice Boltzmann Method: applications to thermal fluid dynamics and energy systems." Doctoral thesis, Università degli studi di Genova, 2021. http://hdl.handle.net/11567/1060259.
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In many energy systems fluids play a fundamental role, and computational simulations are a valuable tool to
study their complex dynamics. The Lattice Boltzmann Method (LBM) is a relatively new numerical method
for computational fluid dynamics, but its applications can be extended to physical phenomena beyond fluid
flows. This thesis presents applications of the LBM to thermal fluid dynamics and energy systems. Specific
applications considered are: application to nuclear reactor engineering problems; thermal fluid dynamic
behavior of a Natural Circulation Loop; nanoparticles gravitational sedimentation; acoustical problems.
The main original contributions derived from this work are: first, the systematic description of the current
status of LBM applications to nuclear reactors problems, including test cases and benchmark simulations;
second, the development and validation of a LBM model for a single-phase natural circulation loop; third,
the development and validation of a LBM model for gravitational sedimentation of nanoparticles, and fourth,
the systematic description of the current status of LBM applications to acoustics, including simulations of
test cases. The development of this thesis was not limited to simulations; experimental studies in parallel
connected natural circulation loops of small inner diameter were conducted, showing the wide applicability
of the one-dimensional theoretical models used to validate the LBM results. Additional contributions derived
from this work: 1. the applicability of the method to study neutron transport and nuclear waste disposal using
porous materials was shown. 2. changes in the thermophysical performance of the natural circulation loop
when the loop reached a non-laminar (transition) regime were found at a Reynolds number lower than the
typical range. 3. variable diffusion and sedimentation parameters were effective to model the experimental
sedimentation curves. In conclusion, this work shows that the LBM is a versatile and powerful computational
tool that can be used beyond the common Computational Fluid Dynamics applications.In many energy systems fluids play a fundamental role, and computational simulations are a valuable tool to study their complex dynamics. The Lattice Boltzmann Method (LBM) is a relatively new numerical method for computational fluid dynamics, but its applications can be extended to physical phenomena beyond fluid flows. This thesis presents applications of the LBM to thermal fluid dynamics and energy systems. Specific applications considered are: application to nuclear reactor engineering problems; thermal fluid dynamic behavior of a Natural Circulation Loop; nanoparticles gravitational sedimentation; acoustical problems. The main original contributions derived from this work are: first, the systematic description of the current status of LBM applications to nuclear reactors problems, including test cases and benchmark simulations; second, the development and validation of a LBM model for a single-phase natural circulation loop; third, the development and validation of a LBM model for gravitational sedimentation of nanoparticles, and fourth, the systematic description of the current status of LBM applications to acoustics, including simulations of test cases. The development of this thesis was not limited to simulations; experimental studies in parallel connected natural circulation loops of small inner diameter were conducted, showing the wide applicability of the one-dimensional theoretical models used to validate the LBM results. Additional contributions derived from this work: 1. the applicability of the method to study neutron transport and nuclear waste disposal using porous materials was shown. 2. changes in the thermophysical performance of the natural circulation loop when the loop reached a non-laminar (transition) regime were found at a Reynolds number lower than the typical range. 3. variable diffusion and sedimentation parameters were effective to model the experimental sedimentation curves. In conclusion, this work shows that the LBM is a versatile and powerful computational tool that can be used beyond the common Computational Fluid Dynamics applications.
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D'AMICO, Salvatore. "Integral approach to the safety design of the EU-DEMO Helium-Cooled Pebble Beds with reference to the associated relevant systems." Doctoral thesis, Università degli Studi di Palermo, 2020. http://hdl.handle.net/10447/395442.
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SANTOS, GEAN R. dos. "Algoritmo de colônia de formigas e redes neurais artificiais aplicados na monitoração e detecção de falhas em centrais nucleares." reponame:Repositório Institucional do IPEN, 2016. http://repositorio.ipen.br:8080/xmlui/handle/123456789/26798.
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Um desafio recorrente em processos produtivos é o desenvolvimento de sistemas de monitoração e diagnóstico. Esses sistemas ajudam na detecção de mudanças inesperadas e interrupções, prevenindo perdas e mitigando riscos. Redes Neurais Artificiais (RNA) têm sido largamente utilizadas na criação de sistemas de monitoração. Normalmente as RNA utilizadas para resolver este tipo de problema são criadas levando-se em conta apenas parâmetros como o número de entradas, saídas e quantidade de neurônios nas camadas escondidas. Assim, as redes resultantes geralmente possuem uma configuração onde há uma total conexão entre os neurônios de uma camada e os da camada seguinte, sem que haja melhorias em sua topologia. Este trabalho utiliza o algoritmo de Otimização por Colônia de Formigas (OCF) para criar redes neurais otimizadas. O algoritmo de busca OCF utiliza a técnica de retropropagação de erros para otimizar a topologia da rede neural sugerindo as melhores conexões entre os neurônios. A RNA resultante foi aplicada para monitorar variáveis do reator de pesquisas IEA-R1 do IPEN. Os resultados obtidos mostram que o algoritmo desenvolvido é capaz de melhorar o desempenho do modelo que estima o valor de variáveis do reator. Em testes com diferentes números de neurônios na camada escondida, utilizando como comparativos o erro quadrático médio, o erro absoluto médio e o coeficiente de correlação, o desempenho da RNA otimizada foi igual ou superior ao da tradicional.
Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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BAPTISTA, FILHO BENEDITO D. "Redes neurais para controle de sistemas de reatores nucleares." reponame:Repositório Institucional do IPEN, 1998. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10723.
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Tese (Doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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CARNEIRO, ALVARO L. G. "Desenvolvimento de sistema de monitoracao e diagnostico aplicado a valvulas moto-operadas utilizadas em centrais nucleares." reponame:Repositório Institucional do IPEN, 2003. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11109.
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Tese (Doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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DeWitte, Jacob D. (Jacob Dominic). "Reactor protection system design alternatives for sodium fast reactors." Thesis, Massachusetts Institute of Technology, 2011. http://hdl.handle.net/1721.1/76523.
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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011."January 2011." Cataloged from PDF version of thesis.
Includes bibliographical references (p. 110-112).
Historically, unprotected transients have been viewed as design basis events that can significantly challenge sodium-cooled fast reactors. The perceived potential consequences of a severe unprotected transient in a sodium-cooled fast reactor include an energetic core disruptive accident, vessel failure, and a large early release. These consequences can be avoided if unprotected transients are properly defended against, potentially improving the economics of sodium fast reactors. One way to defend against such accidents is to include a highly reliable reactor protection system. The perceived undesirability of the consequences arising from an unprotected transient has led some sodium fast reactor designers to consider incorporating several design modifications to the reactor protection system, including: self-actuated shutdown systems, articulated control rods, and seismic anticipatory scram systems. This study investigates the performance of these systems in sodium fast reactors. To analyze the impact of these proposed design alternatives, a model to analyze plant performance that incorporates uncertainty analysis is developed using RELAP5-3D and the ABR-1000 as the reference design. The performance of the proposed alternatives is analyzed during unprotected loss of flow and unprotected transient overpower scenarios, each exacerbated by a loss of heat sink. The recently developed Technology Neutral Framework is used to contextually rate performance of the proposed alternatives. Ultimately, this thesis offers a methodology for a designer to analyze reactor protection system design efficacy. The principle results of this thesis suggest that when using the Technology Neutral Framework as a licensing framework for a sodium-cooled fast reactor, the two independent scram systems of the ABR- 1000's reactor protection system perform well enough to screen unprotected transients from the design basis. While a regulator may still require consideration of accidents involving the failure of the reactor protection system, these events will not drive the design of the system. However, self-actuated shutdown systems may be called for to diversify the reactor protection system. Of these, the Curie point latch marginally reduces the conditional cladding damage probability for metal cores because of their rapid inherent feedback effects, but is more effective for the more sluggish oxide cores given reasonably long pump coastdown times. Flow levitated absorbers are highly effective at mitigating unprotected loss of flow events for both fuel types, but are limited in response during unprotected transient overpower events. When considered from a risk-informed perspective, a clear rationale and objective is needed to justify the inclusion of an additional feature such as self-actuated shutdown systems. The use of articulated safety rods as one of the diverse means of reactivity insertion and the implementation of an anticipatory seismic scram system may be the most cost-effective alternatives to provide defense in depth in light of the sodium fast reactor's susceptibility to seismic events.
by Jacob D. DeWitte.
S.M.
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Kingdon, David Ross. "Safety characteristics of a suspended-pellet fission reactor system." Thesis, National Library of Canada = Bibliothèque nationale du Canada, 1998. http://www.collectionscanada.ca/obj/s4/f2/dsk1/tape11/PQDD_0001/NQ42856.pdf.
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Singo, Thifhelimbilu Daphney. "Development of a high flux neutron radiation detection system for in-core temperature monitoring." Thesis, Stellenbosch : Stellenbosch University, 2012. http://hdl.handle.net/10019.1/19999.
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Thesis (PhD)--Stellenbosch University, 2012.ENGLISH ABSTRACT: The objective of this research was to develop a neutron detection system that incorporates a mass spectrometer to measure high neutron flux in a nuclear reactor environment. This system consists of slow and fast neutron detector elements for measuring fluxes in those energy regions respectively. The detector should further be capable of withstanding the harsh conditions associated with a high temperature reactor. This novel detector which was initially intended for use in the PBMR reactor has possible applications as an in-core neutron and indirect temperature-monitoring device in any of the HTGR. Simulations of a generic HTGR core model were performed in order to obtain the neutron energy spectrum with emphasis on the behavior of three energy regions, slow, intermediate and fast neutrons within the core at different temperatures. The slow neutron flux which has the characteristic of a Maxwell- Boltzmann distribution were found to shift to larger values of neutron flux at higher energies as the fuel temperature increased, while fast neutron flux spectra remained relatively constant. In addition, the results of the fit of the slow neutron flux with a modified Maxwell-Boltzmann equation confirmed that in the presence of the neutron source, leakage and absorption, the effective neutron temperatures is above the medium temperatures. From these results, it was clear that the detection system will need to monitor both slow and fast neutron flux. Placing neutron detectors inside the reactor core, that are sensitive to a particular energy range of slow and fast neutrons, would thus provide information about the change of temperature in the fuel and hence act as an in-core temperature monitor. A detection mechanism was developed that employs the neutron-induced break-up reaction of 6Li and 12C into α-particles. These materials make excellent neutron converters without interference due to γ-rays, as the contributions from 6Li(γ,np)4He and 12C(γ,3α) reactions are negligible. The mass spectrometer measures the 4He partial pressure as a function of time under high vacuum with the help of pressure gradient provided by a high-vacuum turbomolecular pump and a positive-displacement fore-vacuum pump connected in series. A cryogenic trap, which contains a molecular sieve made of pellets 1.6 mm in diameter, was also designed and manufactured to remove impurities which cause a background in the lighter mass region of the spectrum. The development and testing of the high flux neutron detection system were performed at the iThemba Laboratory for Accelerator Based Sciences (LABS), South Africa. These tests were carried out with a high energy proton beam at the D-line neutron facility, and with a fast neutron beam at the neutron radiation therapy facility. To test the principle and capability of the detection system in measuring high fluxes, a high intensity 66 MeV proton beam was used to produce a large yield of α-particles. This was done because the proton inelastic scattering cross-section with 12C nuclei is similar to that of neutrons, with a threshold energy of about 8 MeV for both reactions. Secondly, the secondary fast neutrons produced from the 9Be(p,n)9B reaction were also measured with the fast neutron detector. The response of this detection system during irradiation was found to be relatively fast, with a rise time of a few seconds. This is seen as a sharp increase in the partial pressure of 4He gas as the proton or neutron beam bombards the 12C material. It was found that the production of 4He with the proton beam was directly proportional to the beam intensity. The number of 4He atoms produced per second was deduced from the partial pressure observed during the irradiation period. With a neutron beam of 1010 s−1 irradiating the detector, the deduced number of 4He atoms was 109 s−1. When irradiation stops, the partial pressure drops exponentially. This response is attributed to a small quantity of 4He trapped in the present design. Overall, the measurements of 4He partial pressure produced during the tests with proton and fast neutron beams were successful and demonstrated proof of principle of the new detection technique. It was also found that this system has no upper neutron flux detection limit; it can be even higher than 1014 n·cm−2·s−1. The lifetime of this detection system in nuclear reactor environment is practically unlimited, as determined by the known ability of stainless steel to keeps its integrity under the high radiation levels. Hence, it is concluded that this high flux neutron detection system is excellent for neutron detection in the presence of high γ-radiation level and provides real-time flux measurements.
AFRIKAANSE OPSOMMING: Die doel van hierdie navorsing was om ’n neutrondetektorstelsel te ontwikkel wat hoë neutronvloed binne in ’n kernreaktor kan meet. Die stelsel bevat twee aparte detektorelemente sodat die termiese sowel as snelneutronvloed gemeet kan word. Die detektor moet verder in staat wees om die strawwe toestande, kenmerkend aan ’n hoë temperatuur reaktor, te kan weerstaan. Die innoverende detektorstelsel, oorspronklik geoormerk vir gebruik in die PBMR reaktor, het toepassingsmoontlikhede as in-kern neutron- sowel as indirekte temperatuurmonitor. Simulasies van ’n generiese model van ’n HTGR reaktorkern is uitgevoer ten einde die neutronenergiespektrum in die kern by verskillende temperature te bekom met klem op die gedrag van neutrone in drie energiegroepe: stadig (termies), intermediêr en snel (vinnig). Daar is bevind dat die stadige neutrone, wat ’n Maxwell-Boltzman verdeling toon, in intensiteit toeneem en dat die piek na hoër energie verskuif met toename in temperatuur, terwyl die vinnige neutronspektrum relatief onveranderd bly. ’n Passing van die stadige spektrum op ’n gemodifiseerde Maxwell-Boltzmann verdeling het bevestig dat die effektiewe neutrontemperatuur weens die teenwoordigheid van bronterme, verliese en absorpsie, hoër as die temperatuur van die medium is. Hierdie resultate maak dit duidelik dat die detektorstelsel beide die stadige sowel as die vinnige neutronvloed moet kan waarneem. Deur detektorelemente wat sensitief is vir die onderskeie spekrale gebiede in die reaktorhart te plaas, kan informasie bekom word wat tot in-kern temperatuur herleibaar is sodat die stelsel inderdaad as indirekte temperatuurmonitor kan dien. Die feit dat alfa-deeltjies geproduseer word in neutron-geïnduseerde opbreekreaksies van 6Li en 12C is as die basis van die nuwe opsporingsmeganisme aangewend. Hierdie materiale funksioneer uitstekend as neutron-selektiewe omsetters in die teenwoordigheid van gamma-strale aangesien laasgenoemde se bydraes tot helium produksie via die 6Li(γ,np)4He en 12C(γ,3α) reaksies, weglaatbaar is. Die massaspektrometer meet die tydgedrag van die 4He parsiële druk binne ’n hoogvakuum wat met behulp van ’n seriegeskakelde kombinasie van ’n turbomolekulêre en positiewe-verplasingsvoorpomp verkry word. ’n Koueval met ’n molekulêre sif, bestaande uit 1.6 mm diameter korrels, is ontwerp en vervaardig om onsuiwerhede te verwyder wat andersins as agtergrond by die ligter gedeelte van die massaspektrum sou wys. Die ontwikkeling en toetsing van die hoëvloed detektorstelsel is te iThembaLABS (iThemba Laboratories for Accelerator Based Sciences) gedoen. Dit is uitgevoer deur gebruik te maak van die hoë energie protonbundel van die D-lyn neutronfasiliteit asook van die bundel vinnige neutrone by die neutronterapiefasiliteit. Om die beginsel en vermoë te toets om by ’n hoë neutronvloed te kan meet, is van die intense 66 MeV protonbudel gebruik gemaak om ’n hoë opbrengs alfa-deeltjies te verkry. Dit is gedoen omdat die reaksiedeursnit vir onelastiese verstrooiing van protone vanaf 12C kerne soortgelyk is aan die van neutrone, met ’n drumpelenergie van 8 MeV vir beide reaksies. Tweedens is die sekondêre vinnige neutrone afkomstig van die 9Be(p,n)9B reaksie ook met die neutrondetektor gemeet. Daar is bevind dat die reaksietyd van die deteksiestelsel tydens bestraling relatief vinnig is, soos gekenmerk deur ’n stygtyd van etlike sekondes. Laasgenoemde manifesteer as ’n toename in die parsiële druk van die 4He sodra die proton- of neutronbundel op die 12C teiken inval. Daar is verder bevind dat die 4He produksie direk eweredig aan die bundelintensiteit is. Vir ’n neutronbundel van nagenoeg 1010 s−1, invallend op die neutrondetektor, is vanaf die gemete parsiële druk afgelei dat die produksie van 4He atome sowat 109 s−1 beloop. In die geheel beoordeel, was die meting van die 4He parsiële druk tydens die toetse met vinnige protone en neutrone suksesvol en het dit die nuwe meetbeginsel bevestig. Dit is verder bevind dat die meetstelsel nie ’n beperking op die boonste neutronvloed plaas nie, maar dat dit vloede van selfs hoër as 1014 s−1 kan hanteer. Die leeftyd van die detektorstelsel in die reaktor is prakties onbeperk en onderhewig aan die bevestigde integriteit van vlekvrystaal onder hoë bestraling. Die gevolgtrekking is dus dat die nuwe detektorstelsel uitstekend geskik is vir die in-tyd meting van ’n baie hoë vloed van neutrone ook in die teenwoordigheid van intense gammabestraling.
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ARONNE, IVAN D. "Desenvolvimento de um sistema de identificacao e classificacao de transientes para um reator nuclear a agua pressurizada integral." reponame:Repositório Institucional do IPEN, 2009. http://repositorio.ipen.br:8080/xmlui/handle/123456789/9380.
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Tese (Doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
41
SILVEIRA, RENATO C. da. "Avaliacao da estabilidade estrutural de contencoes metalicas de centrais nucleares." reponame:Repositório Institucional do IPEN, 2000. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10795.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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JONG, RUDOLF P. de. "Avaliacao de tubulacoes trincadas em sistemas primarios de reatores nucleares PWR." reponame:Repositório Institucional do IPEN, 2004. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11228.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
43
Elshahat, Ayah Elsayed. "Enhancing nuclear energy sustainability using advanced nuclear reactors." Thesis, University of Manchester, 2015. https://www.research.manchester.ac.uk/portal/en/theses/enhancing-nuclear-energy-sustainability-using-advanced-nuclear-reactors(2c39b9ca-86a9-446f-8832-ae9469485a2d).html.
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The safety performance of nuclear power reactors is a very important factor in evaluating nuclear energy sustainability. Improving the safety performance of nuclear reactors can enhance nuclear energy sustainability as it will improve the environmental indicator used to evaluate the overall sustainability of nuclear energy. Great interest is given now to advanced nuclear reactors especially those using passive safety components. Investigation of the improvement in nuclear safety using advanced reactors was done by comparing the safety performance of a conventional reactor which uses active safety systems, such as Pressurized Water Reactor (PWR), with an advanced reactor which uses passive safety systems, such as AP1000, during a design basis accident, such as Loss of Coolant Accident (LOCA), using the PCTran as a simulation code. To assess the safety performance of PWR and AP1000, the “Global Safety Index” GSI model was developed by introducing three indicators: probability of accident occurrence, performance of safety system in case of an accident occurrence, and the consequences of the accident. Only the second indicator was considered in this work. A more detailed model for studying the performance of passive safety systems in AP1000 was developed. That was done using SCDAPSIM/RELAP5 code as it is capable of modelling design basis accidents (DBAs) in advanced nuclear reactors.
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Sunnevik, Klas. "Comparison of MAAP and MELCOR : and evaluation of MELCOR as a deterministic tool within RASTEP." Thesis, Uppsala universitet, Tillämpad kärnfysik, 2014. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-233768.
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This master's thesis is an investigation and evaluation of MELCOR (a software tool for severe accident analyses regarding nuclear power plants), or more correctly of the (ASEA-Atom BWR 75) reactor model developed for version 1.8.6 of MELCOR. The main objective was to determine if MELCOR, with the reactor model in question, is able to produce satisfactory results in severe accident analyses compared to results made by MAAP, which is currently the only official software tool for this application in Sweden. The thesis work is related to the RASTEP project. This project has been carried out in several stages on behalf of SSM since 2009, with a number of specific issues explored within an NKS funded R&D project carried out 2011-2013. This investigation is related to the NKS part of the project. The purpose with the RASTEP project is to develop a method for rapid source term prediction that could aid the authorities in decision making during a severe accident in a nuclear power plant. A software tool, which also gave the project its name, i.e. RASTEP (RApid Source TErm Prediction), is therefore currently under development at Lloyd's Register Consulting. A software tool for severe accident analyses is needed to calculate the source terms which are the end result from the predictions made by RASTEP. A set of issues have been outlined in an earlier comparison between MAAP and MELCOR. The first objective was therefore to resolve these pre-discovered issues, but also to address new issues, should they occur. The existing MELCOR reactor model also had to be further developed through the inclusion of various safety systems, since these systems are required for certain types of scenarios. Subsequently, a set of scenarios was simulated to draw conclusions from the additions made to the reactor model. Most of the issues (pre-discovered as well as new ones) could be resolved. However the work also rendered a set of issues which are in need of further attention and investigation. The overall conclusion is that MELCOR is indeed a promising alternative for severe accident analyses in the Swedish work with nuclear safety. Several potential benefits from making use of MELCOR besides MAAP have been identified. In conclusion, they would be valuable assets to each other, e.g. since deviations in the results (between the two codes) would highlight possible weaknesses of the simulations. Finally it is recommended that the work on improving the MELCOR reactor model should continue.RASTEP
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Adams, Imani Noel. "Comprehensive Analysis of a Scaled-Down Low-Temperature Direct Reactor Auxiliary Cooling System for Fluoride Salt-Cooled High-Temperature Reactors." The Ohio State University, 2013. http://rave.ohiolink.edu/etdc/view?acc_num=osu1366464705.
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HIROMOTO, MARIA Y. K. "PSINCO-um programa para calculo da distribuicao de potencia e supervisao do nucleo de reatores nucleares, utilizando sinais de detetores tipo 'SPD'." reponame:Repositório Institucional do IPEN, 1998. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10706.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
47
REIS, JUNIOR JOSE S. B. "Métodos e softwares para análise da produção científica e detecção de frentes emergentes de pesquisa." reponame:Repositório Institucional do IPEN, 2015. http://repositorio.ipen.br:8080/xmlui/handle/123456789/26929.
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O progresso de projetos anteriores salientou a necessidade de tratar o problema dos softwares para detecção, a partir de bases de dados de publicações científicas, de tendências emergentes de pesquisa e desenvolvimento. Evidenciou-se a carência de aplicações computacionais eficientes dedicadas a este propósito, que são artigos de grande utilidade para um melhor planejamento de programas de pesquisa e desenvolvimento em instituições. Foi realizada, então, uma revisão dos softwares atualmente disponíveis, para poder-se delinear claramente a oportunidade de desenvolver novas ferramentas. Como resultado, implementou-se um aplicativo chamado Citesnake, projetado especialmente para auxiliar a detecção e o estudo de tendências emergentes a partir da análise de redes de vários tipos, extraídas das bases de dados científicas. Através desta ferramenta computacional robusta e eficaz, foram conduzidas análises de frentes emergentes de pesquisa e desenvolvimento na área de Sistemas Geradores de Energia Nuclear de Geração IV, de forma que se pudesse evidenciar, dentre os tipos de reatores selecionados como os mais promissores pelo GIF - Generation IV International Forum, aqueles que mais se desenvolveram nos últimos dez anos e que se apresentam, atualmente, como os mais capazes de cumprir as promessas realizadas sobre os seus conceitos inovadores.
Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP
48
CARAMELLO, MARCO. "Thermal-hydraulics of passive safety systems for advanced Nuclear Reactors." Doctoral thesis, Politecnico di Torino, 2017. http://hdl.handle.net/11583/2681218.
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The present thesis addresses several topics of thermal-hydraulics applied to the heat transfer in the nuclear field, both in normal operating and accidental conditions. The problem is to demonstrate that innovative heat exchanger geometries and safety systems are able to effectively remove thermal power from a primary system of a nuclear reactor. Normal operating condition was addressed by looking at several heat exchanger and steam generator geometries for heat removal in advanced nuclear reactors, such as Small Modular Reactors and generation IV reactors. Helical coil, microchannel and bayonet tube geometries have been studied. Concerning helical coil heat exchangers, a numerical model has been developed to study the thermal performance of a unit on Matlab Environment. The model requires input data such as the geometry, the type of fluids and boundary conditions, and is able to compute the spatial behaviour of the most important thermal-hydraulic parameters of the flow such as temperature, pressure, void fraction, quality, heat transfer coefficients, etc. The model adopts an iterative solving scheme of solution for the different control volumes of the component and semi-empirical correlations to compute heat transfer coefficients and pressure drops. The model has been validated against reference data of the steam generator of IRIS reactor and BREST fast reactor, and demonstrated to be able to reproduce the reference data with a very high accuracy. The microchannel configuration has been studied for the large modular reactor I2S by means of the system code RELAP5-3D. The bayonet configuration has been widely studied for ALFRED reactor in LEADER configuration by means of the system code RELAP5-3D. Sensitivity studies have revealed the impact of some important geometrical parameters and the role of regenerative heat transfer on heat exchanger performance. Results obtained during the PhD period have been used for optimization purposes of ALFRED steam generator. Some heat transfer devices connecting safety systems to the environment have been studied in the light of accidental sequences. In particular, air heat exchangers and pool heat exchangers have been investigated through several RELAP5-3D simulations in the light of a Station Black Out (SBO) event for the large integral reactor I2S. The most important topic of the thesis is the decay heat removal system of ALFRED reactor. The safety system is both able to remove power due to decay heat and to delay primary coolant freezing in the long term of operation by means of noncondensable gases. Both safety functions are performed in a passive manner. The safety system makes use of the bayonet steam generator for normal operation and connects it to a ultimate heat sink, which is an isolation condenser. A tank is connected to the bottom of the isolation condenser and allows the housing of noncondensable gases. Following an accident, noncondensable gases are firstly collected in the tank and then released during time to reduce the heat transfer in the isolation condenser, so to avoid primary system overcooling. In this framework, the present PhD thesis presents a set of sensitivity studies performed on the safety system by means of the system code RELAP5-3D. Initial gas pressure and tank volume have been addressed. The study have highlighted the most important parameters for design optimization and revealed some important physical phenomena. In a second part, the safety system has been used as reference to design an experimental facility able to reproduce the most important physical phenomena on which the safety system relies. The scaled facility will be built in SIET laboratories in Piacenza thanks to the SIRIO project, partially funded by the Italian Ministry of Economic Development. The present thesis reports the development of the conceptual design of the facility and some preliminary simulations to ensure that the facility can reproduce the most important phenomena of the safety system.
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Törnblom, Nils. "Underwater 3D Surface Scanning using Structured Light." Thesis, Uppsala universitet, Centrum för bildanalys, 2010. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-138205.
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In this thesis project, an underwater 3D scanner based on structured light has been constructed and developed. Two other scanners, based on stereoscopy and a line-swept laser, were also tested. The target application is to examine objects inside the water filled reactor vessel of nuclear power plants. Structured light systems (SLS) use a projector to illuminate the surface of the scanned object, and a camera to capture the surfaces' reflection. By projecting a series of specific line-patterns, the pixel columns of the digital projector can be identified off the scanned surface. 3D points can then be triangulated using ray-plane intersection. These points form the basis the final 3D model. To construct an accurate 3D model of the scanned surface, both the projector and the camera need to be calibrated. In the implemented 3D scanner, this was done using the Camera Calibration Toolbox for Matlab. The codebase of this scanner comes from the Matlab implementation by Lanman & Taubin at Brown University. The code has been modified and extended to meet the needs of this project. An examination of the effects of the underwater environment has been performed, both theoretically and experimentally. The performance of the scanner has been analyzed, and different 3D model visualization methods have been tested. In the constructed scanner, a small pico projector was used together with a high pixel count DSLR camera. Because these are both consumer level products, the cost of this system is just a fraction of commercial counterparts, which uses professional components. Yet, thanks to the use of a high pixel count camera, the measurement resolution of the scanner is comparable to the high-end of industrial structured light scanners.
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Freas, Rosemarv M. "Analysis of required supporting systems for the Supercritical CO2 power conversion system." Thesis, Cambridge Massachusetts Institute of Technology, 2007. http://hdl.handle.net/10945/2992.
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Recently, attention has been drawn to the viability of using S-CO(2) as a working fluid in modern reactor designs. Near the critical point, CO2 has a rapid rise in density allowing a significant reduction in the compressor work of a closed Brayton Cycle. Therefore, 45% efficiency can be achieved at much more moderate temperatures than is optimal for the helium Brayton cycles. An additional benefit of the S-CO2 system is its universal applicability as an indirect secondary Power Conversion System (PCS) coupled to most GEN-IV concept reactors, as well as fusion reactors. The United States DOE's GNEP is now focusing on the liquid Na cooled primary as an alternative to conventional Rankine steam cycles. This primary would also benefit from being coupled to an S-CO2 PCS. Despite current progress on designing the S-CO2 PCS, little work has focused on the principal supporting systems required. Many of the required auxiliary systems are similar to those used in other nuclear or fossil-fired units; others have specialized requirements when CO2 is used as the working fluid, and are therefore given attention in this thesis. Auxiliary systems analyzed within this thesis are restricted to those specific to using CO2 as the working fluid. Particular systems discussed include Coolant Make-up and Storage, Coolant Purification, and Coolant Leak Detection.Contract number: N62271-97-G-0026.
US Navy (USN) author