Letteratura scientifica selezionata sul tema "PWR TYPE REACTORS"
Cita una fonte nei formati APA, MLA, Chicago, Harvard e in molti altri stili
Consulta la lista di attuali articoli, libri, tesi, atti di convegni e altre fonti scientifiche attinenti al tema "PWR TYPE REACTORS".
Accanto a ogni fonte nell'elenco di riferimenti c'è un pulsante "Aggiungi alla bibliografia". Premilo e genereremo automaticamente la citazione bibliografica dell'opera scelta nello stile citazionale di cui hai bisogno: APA, MLA, Harvard, Chicago, Vancouver ecc.
Puoi anche scaricare il testo completo della pubblicazione scientifica nel formato .pdf e leggere online l'abstract (il sommario) dell'opera se è presente nei metadati.
Articoli di riviste sul tema "PWR TYPE REACTORS":
1
Niearonov, Y. M., T. Y. Baybuzenko, V. Y. Shenderovych, M. I. Vlasenko, О. V. Godun, V. M. Кyrianchuk, G. R. Semenov e L. І. Gromok. "Reactor Technology Rationale for Construction of Substitution and New Power Units in Ukraine after 2035". Nuclear Power and the Environment 18 (2020): 10–22. http://dx.doi.org/10.31717/2311-8253.20.3.2.
Abstract (sommario):
An algorithm of selecting the reactor technology type were constructed. The algorithm is based on a comparative assessment of the respective nuclear power plants. The formation of qualitative and quantitative criteria is performed for the estimation algorithm. Tools of the International Atomic Energy Agency (IAEA) INPRO-KIND project on multi-criteria comparative assessment of nuclear power plants for ranking the obtained results were adapted. The sensitivity analysis of the obtained results to change of numerical values and weight of criteria is carried out. The choice of the type of reactor technology for construction in Ukraine after 2035 is substantiated. It is shown that PWR and SMR reactor technologies in Ukraine are the most promising direction in the development of nuclear energy in Ukraine. Taking into account the factors of uncertainty and sensitivity to the values of the original data and possible risks, results of the analysis shows that there is a trend of advantages of SMR reactors, which generally have higher ratings compared to PWR, BWR and HWR. At the same time, the level of multi-criteria ratings of PWR reactors is close to SMR reactors. Making a further decision on the type of reactor technology for the conditions of Ukraine, it is necessary to take into account the possibility of its maximal total installed capacity deployment. It is necessary to conduct a separate study to determine the optimal ratio of reactor technologies PWR and SMR in the power system of Ukraine, taking into account the prospects for the deployment of renewable energy sources.
2
Kouhia, Virpi, Heikki Purhonen, Vesa Riikonen, Markku Puustinen, Riitta Kyrki-Rajamäki e Juhani Vihavainen. "PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications". Science and Technology of Nuclear Installations 2012 (2012): 1–8. http://dx.doi.org/10.1155/2012/548513.
Abstract (sommario):
This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.
3
Zvezdin, Yu I., D. M. Shur e A. A. Popov. "Service Properties of Reactor Structural Steels and Prediction of NPP Equipment Life With PWR-Type Reactors". Journal of Pressure Vessel Technology 113, n. 2 (1 maggio 1991): 159–62. http://dx.doi.org/10.1115/1.2928741.
Abstract (sommario):
The results obtained from the experimental investigation of Cr-Ni-Mo-V and Mn-Ni-Mo-V steel service properties used in the USSR reactor building, as well as the static crack resistance KIC versus reference temperature dependence and Paris kinetic diagrams of fatigue crack growth rate in air and water of high parameters, are given.
4
Usman, Iyabo, David Vermillion, Howard Hall e Steve Skutnik. "Fingerprinting commercial nuclear reactor forensics using the ORIGEN-S simulation code". International Journal of Modern Physics: Conference Series 48 (gennaio 2018): 1860126. http://dx.doi.org/10.1142/s2010194518601266.
Abstract (sommario):
The ability to determine the origin of a specific spent-fuel sample from a commercial nuclear reactor was studied using the Origen-S simulation code by calculating the plutonium and uranium isotopic concentration data for a range of nuclear power reactors. This range of reactors is based on a typical Westinghouse PWR fuel assembly with a fuel type of W17 X 17, having individual operating histories. Isotopic ratios of plutonium in nuclear reactors during the fuel-cycle period provide information on how the plutonium grows into the fuel as a function of burnup, as well as its attractiveness to proliferators. Using the results from the calculation of uranium and plutonium isotopic ratios, the origin of each spent-fuel assembly for a particular reactor can be predicted and documented for a future nuclear forensics reference database.
5
Bryk, Rafał, Lars Dennhardt e Simon Schollenberger. "Experimental investigation of PWR accident scenarios at the PKL test facility". E3S Web of Conferences 137 (2019): 01016. http://dx.doi.org/10.1051/e3sconf/201913701016.
Abstract (sommario):
PKL is the only test facility in Europe that replicates the entire primary side and the most important parts of the secondary side of western-type Pressurized Water Reactors (PWR) in the scale of 1:1 in heights. It is also worldwide the only test facility with 4 identical reactor coolant loops arranged symmetrically around the Reactor Pressure Vessel (RPV) for simulation of nonsymmetrical boundary conditions between the reactor loops. Thermal-hydraulic phenomena observed in PWRs are simulated in the PKL test facility for over 40 years. The analyses carried out in these years encompass a large spectrum of accident scenario simulations and corresponding cool-down procedures. The overall goal of the PKL experiments is to show that under accident conditions - even for extreme and highly unlikely assumptions as additional loss of safety systems - the core cooling can be maintained or re-established by automatic or operator- performed procedures and that a severe accident e.g. a core melt-down can be avoided under all circumstances. Another goal of the tests performed in the PKL facility is the provision of data for validation of thermal-hydraulic system codes. This paper presents recent modifications of the PKL facility, applied in order to adapt the facility to the latest western-type designs currently built in the world. The paper discusses also important results obtained in the last years.
6
Dmitriev, S. M., D. V. Doronkov, M. A. Legchanov, V. D. Sorokin e A. E. Khrobostov. "REGULARITIES OF FORMATION OF FLOW OF COOLANT BEHIND THE TVS-KVADRAT MIXING SPACING GRID OF THE PWR-TYPE REACTOR". ENERGETIKA. Proceedings of CIS higher education institutions and power engineering associations 61, n. 3 (28 maggio 2018): 258–68. http://dx.doi.org/10.21122/1029-7448-2019-61-3-258-268.
Abstract (sommario):
Tgratinghis paper presents the results of experimental investigations of the influence of mixing spacer gratings with different types of deflectors on the coolant flow in the TVSKvadrat fuel assembly of the PWR-type reactor. Experimental model of the TVS-Kvadrat of the PWR reactor was made in complete geometric similarity with the full-scale cassettes. Studies were carried out by modeling the flow of coolant in the core with the use of an experimental stand; the latter was an aerodynamic open loop through which air is pumped. To measure the local hydrodynamic characteristics of the coolant flow, special pneumatic sensors were used that were able to measure the full velocity vector at the point by its three components. During the studies of the local fluid dynamics of the coolant, the transverse flow rates were measured; also, the coolant flow rates were measured by cells of the TVS-Kvadrat experimental model. The analysis of the spatial distribution of the projections of the absolute flow velocity made it possible to detail the pattern of the coolant flow behind the mixing spacing gratings with different variants of the deflector design, as well as to choose the deflector of the optimal design. Accumulated data base on the flow of the coolant in the TVS-Kvadrat fuel assembly formed the basis of the engineering justification of the structures of the active zones of PWR reactors. Guidelines for choosing optimal designs mixing spacing grids have been considered by designers of the “Afrikantov OKBM” JSC when they created implementations of the latest TVS-Kvadrat assemblies. The results of experimental studies are used to verify CFD-codes of both foreign and domestic origin, as well as the programs for detailed cell-by-cell calculation of active zones in order to reduce conservatism in the justification of thermal reliability.
7
Ekariansyah, Andi Sofrany, Surip Widodo, Susyadi Susyadi e Hendro Tjahjono. "PRELIMINARY ASSESSMENT OF ENGINEERED SAFETY FEATURES AGAINST STATION BLACKOUT IN SELECTED PWR MODELS". JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA 23, n. 2 (14 giugno 2021): 47. http://dx.doi.org/10.17146/tdm.2021.23.2.6204.
Abstract (sommario):
The 2011 Fukushima accident did not prevent countries to construct new nuclear power plants (NPPs) as part of the electricity generation system. Based on the IAEA database, there are a total of 44 units of PWR type NPPs whose constructions are started after 2011. To assess the technology of engineered safety features (ESFs) of the newly constructed PWRs, a study has been conducted as described in this paper, especially in facing the station blackout (SBO) event. It is expected from this study that there are a number of PWR models that can be considered to be constructed in Indonesia from the year of 2020. The scope of the study is PWRs with a limited capacity from 900 to 1100 MWe constructed and operated after 2011 and small-modular type of reactors (SMRs) with the status of at least under licensing. Based on the ESFs design assessment, the passive core decay heat removal has been applied in the most PWR models, which is typically using steam condensing inside heat exchanger within a water tank or by air cooling. From the selected PWR models, the CPR-1000, HPR-1000, AP-1000, and VVER-1000, 1200, 1300 series have the capability to remove the core decay heat passively. The most innovative passive RHR of AP-1000 and the longest passive RHR time period using air cooling in several VVER models are preferred. From the selected SMR designs, the NuScale design and RITM-200 possess more advantages compared to the ACP-100, CAREM-25, and SMART. NuScale represents the model with full-power natural circulation and RITM-200 with forced circulation. NuScale has the longest time period for passive RHR as claimed by the vendor, however the design is still under licensing process. The RITM-200 reactor has a combination of passive air and water-cooling of the heat exchanger and is already under construction.
8
Rebak, Raul B., Michael Larsen e Young-Jin Kim. "Characterization of oxides formed on iron-chromium-aluminum alloy in simulated light water reactor environments". Corrosion Reviews 35, n. 3 (28 agosto 2017): 177–88. http://dx.doi.org/10.1515/corrrev-2017-0011.
Abstract (sommario):
AbstractTo avoid accidents like that in Fukushima, the US Department of Energy is engaged with a nuclear fuel vendor to evaluate the performance of iron-chromium-aluminum (FeCrAl) alloys such as advanced powder metallurgy tubing (APMT) as accident-tolerant material for uranium dioxide fuel cladding in light water reactors (LWR). It was important to characterize the oxides formed on APMT under both boiling water reactor (BWR) and pressurized water reactor (PWR) environments for a better understanding of its environmental sustainability in LWRs. Coupons of APMT were exposed for 1 year to both hydrogenated and oxygenated high-purity water at 288°C (e.g. simulated BWR water chemistry without Pt injection) and hydrogenated high-purity water at 330°C (e.g. simulated primary PWR water chemistry without Li/B addition). Results show that after 1-year immersion, APMT always developed a chromium-rich protective oxide film on its surface. In oxygen-containing environments, the oxide consisted of a dual layer, an external thicker layer containing mostly iron oxides and a thinner internal layer rich in chromium oxide. In hydrogen environments, only a single oxide layer formed, consisting of chromium oxide. This is a similar finding as for type 304 and 316 stainless steels and for nickel-based alloy 600, which is extensively reported in the literature. General corrosion of APMT alloys under LWR operating conditions would not be a limiting factor for its performance as cladding material.
9
Dmitriev, S. M., O. B. Samoilov, A. E. Khrobostov, A. V. Varentsov, A. A. Dobrov, D. V. Doronkov e V. D. Sorokin. "Combined numerical and experimental investigations of local hydrodynamics and coolant flow mass transfer in Kvadrat-type fuel assemblies of PWR reactors with mixing grids". Thermal Engineering 61, n. 8 (19 luglio 2014): 558–65. http://dx.doi.org/10.1134/s0040601514080059.
10
Kirillov, Pavel L., e Galina P. Bogoslovskaya. "Generation IV supercritical water-cooled nuclear reactors: Realistic prospects and research program". Nuclear Energy and Technology 5, n. 1 (11 aprile 2019): 67–74. http://dx.doi.org/10.3897/nucet.5.34293.
Abstract (sommario):
Existing conditions make possible obtaining information that being discussed openly by wide scientific community could help outlining or even establishing the expediency of a particular area of present and future research. Use link http://www.sciencedirect.com to learn about the topics or areas that most attract researchers from different countries. The Generation IV International Forum (GIF-IV) established in January 2000 has set a goal to improve the new generation of nuclear technologies in the following areas: stability, safety and reliability, economic competitiveness, proliferation resistance and physical protection. The purpose of the present publication is to prepare a discussion of one of the directions of development of fourth-generation NPPs, the groundwork for which has already been laid in thermal power engineering in various countries. The number of papers published annually on this topic is the largest among other similar topics dedicated to nuclear power plants of the fourth generation. Judging from the operating experience of existing nuclear power plants using water as a coolant, it can be ascertained that the tendency of building water-cooled nuclear power plants will remain during the next 30 to 50 years. During the present stage the task in the development of alternative types of reactors will be limited to demonstration of their performance and acceptability for future power engineering and the society. The project of supercritical water-cooled reactor is based on the operating experience of VVER, PWR, BWR reactors (more than 14,000 reactor-years); many years of experience accumulated in operating fossil thermal power plants (more than 400 power units; 20,000 years of operation of power units) using supercritical (25 MPa, 540°C) and super-supercritical (35–37 MPa, 620–700°C) water steam. In Russia more than 140 supercritical pressure units are currently in operation. Numerical calculation and design of supercritical water-cooled reactor (similarly to BR-10 reactor) will allow not only training personnel for future development of this technology, but will also help revealing the most difficult points requiring experimental confirmation with application of independent test facilities, as well as formulating the plan of first priority experimental studies. Knowledge accumulated over the last 10 years in the world allows the following: further specifying the already developed concept; developing a plan of specific priority studies; compiling task order for designing small-power pilot VVER SKP-30 reactor (30 MW-th). The scope of problems that are to be solved to substantiate VVER-SCP reactor and commence designing an experimental reactor with thermal capacity of 30 MW is the same as that in developing any type of nuclear reactor: physics of the reactor core; material related matters (primarily concerned with the reactor pressure vessel, fuel, and fuel rod cladding); thermal hydraulics of rod bundles in the near- and supercritical areas; water chemistry at supercritical pressure; corrosion of materials, development of safety systems. Research must be carried out both in static conditions and under irradiation. The absence in Russia during the extended time period of approved program with allocation of appropriate funding and preservation of the existing status during the coming two or three years will lead to the situation when Russia will be hopelessly lagging behind in the development of SCWR technology.
Tesi sul tema "PWR TYPE REACTORS":
1
MAI, LUIZ A. "Analise tecnico-economico do ciclo de combustivel 'Tandem'. Um estudo do caso Brasil-Argentina". reponame:Repositório Institucional do IPEN, 1997. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10684.
Abstract (sommario):
Made available in DSpace on 2014-10-09T12:43:01Z (GMT). No. of bitstreams: 0Made available in DSpace on 2014-10-09T14:07:41Z (GMT). No. of bitstreams: 1 06024.pdf: 8432008 bytes, checksum: aedffb47b1226a13b5e8a41e796ee3c2 (MD5)
Tese (Doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
2
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.
Abstract (sommario):
Made available in DSpace on 2014-10-09T12:26:23Z (GMT). No. of bitstreams: 0Made available in DSpace on 2014-10-09T14:04:18Z (GMT). No. of bitstreams: 0
Tese (Doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
3
PORFIRIO, ROGILSON N. da S. "Modelagem e simulacao do termo-fonte radioativo de produtos de fissao em reatores nucleares do tipo PWR". reponame:Repositório Institucional do IPEN, 1996. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10450.
Abstract (sommario):
Made available in DSpace on 2014-10-09T12:38:52Z (GMT). No. of bitstreams: 0Made available in DSpace on 2014-10-09T14:02:08Z (GMT). No. of bitstreams: 1 02806.pdf: 2555475 bytes, checksum: dd5ac6dfe2c65a240c1aa4feafd10dbc (MD5)
Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
4
ROSSINI, MARCOS R. "Sistema de calculo para gerenciamento de combustivel em reatores tipo PWR atraves da teoria de perturbacao de primeira ordem". reponame:Repositório Institucional do IPEN, 1992. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10311.
Abstract (sommario):
Made available in DSpace on 2014-10-09T12:37:11Z (GMT). No. of bitstreams: 0Made available in DSpace on 2014-10-09T14:08:54Z (GMT). No. of bitstreams: 1 04607.pdf: 2576039 bytes, checksum: 603b16c8476069d7b49cc9e583ab15bc (MD5)
Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
5
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.
Abstract (sommario):
Made available in DSpace on 2014-10-09T12:49:42Z (GMT). No. of bitstreams: 0Made available in DSpace on 2014-10-09T14:03:16Z (GMT). No. of bitstreams: 1 09834.pdf: 11774838 bytes, checksum: 940cf578592bd4491d6495c63535f0a7 (MD5)
Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
6
TRINDADE, CARLOS E. "Determinacao das propriedades modais de elementos combustiveis utilizados em reatores do tipo PWR". reponame:Repositório Institucional do IPEN, 1992. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10299.
Abstract (sommario):
Made available in DSpace on 2014-10-09T12:37:03Z (GMT). No. of bitstreams: 0Made available in DSpace on 2014-10-09T13:56:36Z (GMT). No. of bitstreams: 1 04490.pdf: 6751452 bytes, checksum: b98c1dc1b829918d01b97249140225c3 (MD5)
Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
7
ABE, ALFREDO Y. "Analise de transmutacao considerando o tratamento explicito dos produtos de fissao num sistema acoplado, composto pelos codigos Hammer-Technion e". reponame:Repositório Institucional do IPEN, 1990. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10231.
Abstract (sommario):
Made available in DSpace on 2014-10-09T12:36:18Z (GMT). No. of bitstreams: 0Made available in DSpace on 2014-10-09T13:59:17Z (GMT). No. of bitstreams: 1 03980.pdf: 2316245 bytes, checksum: 417b388f4f1b84fef2568b5807e84716 (MD5)
Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
8
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.
Abstract (sommario):
Made available in DSpace on 2014-10-09T12:26:16Z (GMT). No. of bitstreams: 0Made available in DSpace on 2014-10-09T14:09:46Z (GMT). No. of bitstreams: 0
Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
9
LOPEZ, LUIZ A. N. M. "Concepcao e simulacao estatica do circuito secundario de usinas nucleares de pequena potencia". reponame:Repositório Institucional do IPEN, 1989. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10275.
Abstract (sommario):
Made available in DSpace on 2014-10-09T12:36:47Z (GMT). No. of bitstreams: 0Made available in DSpace on 2014-10-09T13:57:28Z (GMT). No. of bitstreams: 1 04288.pdf: 4625254 bytes, checksum: 9001b53adf99cafca24341f1052507a2 (MD5)
Dissertacao (Mestrado)
IPEN/D
Escola Politecnica, Universidade de Sao Paulo - POLI/USP
10
RODRIGUES, LUIZ A. H. "Modelagem teorica-experimental da equacao da quantidade de movimento para geradores de vapor de reatores PWR". reponame:Repositório Institucional do IPEN, 1994. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10391.
Abstract (sommario):
Made available in DSpace on 2014-10-09T12:38:09Z (GMT). No. of bitstreams: 0Made available in DSpace on 2014-10-09T14:04:38Z (GMT). No. of bitstreams: 1 05652.pdf: 4753146 bytes, checksum: 955c8142395925630bbfa71f9f9ba9ba (MD5)
Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
Libri sul tema "PWR TYPE REACTORS":
1
OECD/NEA-CSNI international standard problem ISP36: CORA-W2 experiment on severe fuel damage for a Russsian type PWR : comparison report. Issy-les-Moulineaux, France: Committee on the Safety of Nuclear Installations, OECD Nuclear Energy Agency, 1996.
Capitoli di libri sul tema "PWR TYPE REACTORS":
1
Solomon, H. D., C. Amzallag, A. J. Vallee, A. J. Vallee e R. E. Delair. "Fatigue Limit and Hysteresis Behavior of Type 304L Stainless Steel in Air and PWR Water, at 150°C and 300°C". In 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 583–603. Hoboken, New Jersey, Canada: John Wiley & Sons, Inc., 2012. http://dx.doi.org/10.1002/9781118456835.ch60.
2
Solomon, H. D., C. Amzallag, A. J. Vallee e R. E. DeLair. "Fatigue limit and Hysteresis Behavior of Type 304L Stainless Steel in Air and PWR Water, at 150°C and 300°C". In Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors, 583–603. Cham: Springer International Publishing, 2011. http://dx.doi.org/10.1007/978-3-319-48760-1_35.
3
Pandey, Sanjay Kumar, S. Jalaldeen, K. Velusamy e P. Puthiyavinayagam. "Creep–Fatigue Assessment for Interaction Between IHX Seal Holder and Inner Vessel Stand Pipe in a Pool-Type Fast Reactor as Per RCC-MR". In Lecture Notes in Mechanical Engineering, 299–306. Singapore: Springer Singapore, 2017. http://dx.doi.org/10.1007/978-981-10-6002-1_23.
4
Fox, Michael H. "About Those Accidents". In Why We Need Nuclear Power. Oxford University Press, 2014. http://dx.doi.org/10.1093/oso/9780199344574.003.0017.
Abstract (sommario):
A nuclear power plant is undergoing an emergency shutdown procedure known as a “scram” when there is an unusual vibration and the coolant level drops precipitously. Subsequent investigation by a shift supervisor reveals that X-rays of welds have been falsified and other problems exist with the plant that could potentially cause a core meltdown that would breach the containment building and cause an explosion. However, the results of the investigation are squelched and the plant is brought up to full power. The shift supervisor takes the control room hostage but is then shot by a SWAT team as the reactor is scrammed. A meltdown does not actually occur. No, this did not really happen, but these events—portrayed in the movie The China Syndrome —evoked a scenario in which a nuclear core meltdown could melt its way to China and contaminate an area the size of Pennsylvania. It also exposed a nuclear power culture that covered up safety issues rather than fixing them. It made for a compelling anti-nuclear story that scared a lot of people. And then a real core meltdown happened, 12 days later. The worst commercial nuclear power reactor accident in US history began on Three Mile Island, an island in the Susquehanna River three miles downstream from Middletown, Pennsylvania (hence its name). Two nuclear reactors were built on this island, but one of them (TMI-1) was shut down for refueling while the other one (TMI-2) was running at full power, rated at 786 MWe. At 4:00 a.m., what should have been a minor glitch in the secondary cooling loop began a series of events that led to a true core meltdown, but no China syndrome occurred and there was little contamination outside the plant. Nevertheless, it caused panic, roused anti-nuclear sentiment in the country, and shut down the construction of new nuclear power plants in the United States for decades. The nuclear reactors at Three Mile Island were pressurized water reactors (PWR), the type of reactor that Admiral Rickover had designed for power plants in US Navy nuclear submarines.
5
Doraiswamy, L. K. "Reactions and Reactors Basic Concepts". In Organic Synthesis Engineering. Oxford University Press, 2001. http://dx.doi.org/10.1093/oso/9780195096897.003.0009.
Abstract (sommario):
Organic synthesis is replete with countless classes of reactions, including several that are named after their discoverers (the name reactions), but fortunately they can all be conducted in less than a half-dozen broad types of reactors. Choosing a reactor for a given reaction is based on several considerations and combines reaction analysis with reactor analysis. Thus in this chapter we consider the following aspects of reactions and reactors, much of which should serve as an introduction to chemists and a refresher to chemical engineers: reaction rates, stoichiometry, and rate equations; the basic reactor types, as a prelude to a more rigorous treatment of these in Parts III and IV; transport of mass (represented by reactant and product molecules) and heat across phase boundaries for heterogeneous reactions; and types of laboratory reactors used by chemists and chemical engineers for their specific objectives. The first step in any consideration of reaction rates is the definition of reaction time. This depends on the mode of reactor operation, batch or continuous. For the batch reactor, the reaction time is the elapsed time; whereas for the continuous reactor, it is given by the time the reactant spends in the reactor, called the residence time, that is measured by the ratio of reactor volume to flow rate (volume/volume per unit time with units of time). An equally important consideration is the concept of reaction space (which can have units of volume, surface, or weight), leading to different definitions of the reaction rate. We begin this section by considering different ways of defining the reaction rate based on different definitions of reaction time and space. The basis of all reactor design is an equation for the reaction rate.
6
Meier, Paul F. "Nuclear". In The Changing Energy Mix, 120–51. Oxford University Press, 2020. http://dx.doi.org/10.1093/oso/9780190098391.003.0005.
Abstract (sommario):
With the exception of nuclear submarines and some military applications, nuclear energy is only used to generate electricity. In the United States, uranium and plutonium are the fuels of choice, while some other countries, notably India, are developing thorium as the nuclear fuel. There are two main types of nuclear reactors—the pressurized water reactor (PWR) and the boiling water reactor (BWR). The PWR is the more common design, where the water used to generate steam and drive the turbine is isolated from the reactor core. In contrast, the water that moderates reactor heat in the BWR is also used to generate the steam, so this water must be contained to prevent radioactive contamination. In the United States, nuclear energy accounts for about 20% of electricity generation. Worldwide uranium reserves are about 6 million tonnes based on a price of $130/kg, but if this price constraint is relaxed, the supply of uranium is virtually unlimited since it is present in seawater at parts per billion levels.
7
Verma, Shashi Kant, S. L. Sinha e D. K. Chandraker. "Selection of Appropriate Turbulance Model in Fuel Bundle of Nuclear Energy". In Soft Computing Techniques and Applications in Mechanical Engineering, 249–66. IGI Global, 2018. http://dx.doi.org/10.4018/978-1-5225-3035-0.ch012.
Abstract (sommario):
This chapter presents an overview of various types of turbulence model and their effect on thermal-hydraulic characteristics of nuclear fuel bundle, both past and present using Computational Fluid Dynamic (CFD) approach. It includes the mathematical definition related to fuel bundle in nuclear energy. The various types of geometrical arrangement like Pressurized Water Reactor (PWR), Boiling Water Reactor (BWR), etc., are stressed. The solution procedures that are applicable to the various reactor types are introduced here and presented in detail for different types of turbulence models. Study of these characteristics enables the student to appreciate the effect of the different types of turbulence models on turbulent mixing and related phenomena. Finally, recommendations of turbulence model for rod bundle are finalized. The inclusion of related references provides a starting point for the interested reader / researchers /industrialists.
8
Oriakhi, Christopher O. "Volumetric Analysis". In Chemistry in Quantitative Language. Oxford University Press, 2009. http://dx.doi.org/10.1093/oso/9780195367997.003.0018.
Abstract (sommario):
Volumetric analysis is a chemical analytical procedure based on measurement of volumes of reaction in solutions. It uses titration to determine the concentration of a solution by carefully measuring the volume of one solution needed to react with another. In this process, a measured volume of a standard solution, the titrant, is added from a burette to the solution of unknown concentration. When the two substances are present in exact stoichiometric ratio, the reaction is said to have reached the equivalence or stoichiometric point. In order to determine when this occurs, another substance, the indicator, is also added to the reaction mixture. This is an organic dye which changes color when the reaction is complete. This color change is known as the end point; ideally, it will coincide with the equivalence point. For various reasons, there is usually some difference between the two, though if the indicator is carefully chosen, the difference will be negligible. A typical titration is based on a reaction of the general type aA+bB → products where A is the titrant, B the substance titrated, and a:b is the stoichiometric ratio between the two. Some indicators include Litmus, Methyl Orange, Methyl Red, Phenolphthalein, and Thymol Blue. Titration can be applied to any of the following chemical reactions: • Acid–base • Complexation • Oxidation–reduction • Precipitation Only acid–base and oxidation–reduction titration will be treated here, though the fundamental principles are the same in all cases. Acid–base titration involves measuring the volume of a solution of the acid (or base) that is required to completely react with a known volume of a solution of a base (or acid). The relative amounts of acid and base required to reach the equivalence point depend on their stoichiometric coefficients. It is therefore critical to have a balanced equation before attempting calculations based on acid–base reactions. Below we define some of the common terms associated with acid–base reactions. A molar solution is one that contains one mole of the substance per liter of solution. For example, a molar solution of sodium hydroxide contains 40 g (NaOH=40 g/mol) of the solute per liter of solution. As described in chapter 13, the concentration of a solution expressed in moles per liter of solution is known as the molarity of the solution.
9
Ostresh, J. M., e R. A. Houghten. "Manual multiple synthesis methods". In Fmoc Solid Phase Peptide Synthesis. Oxford University Press, 1999. http://dx.doi.org/10.1093/oso/9780199637256.003.0018.
Abstract (sommario):
Simultaneous multiple peptide synthesis enables the parallel synthesis of large numbers of peptides. The T-bag (tea-bag) method was developed along with other methods, e.g. pin synthesis, synthesis on paper plates, synthesis on parallel columns, and synthesis on cellulose, as technology to facilitate simultaneous multiple synthesis. Large numbers of peptides, peptidomimetics, and small organic molecules have been prepared using the T-bag method to address different research fields, such as conformational analysis, structure activity analysis, synthesis methodologies, and antibody-antigen interaction studies. Using the T-bag method, more than 150 peptides can be prepared in parallel in flexible amounts, with easily enough material for biological tests and analytical studies. The synthesis of peptides of length of up to 26 amino acid residues has been reported. Moreover, the T-bag technology is easy to apply in practice and requires very little special equipment. T-bags are prepared by containing solid phase resins within polypropylene mesh material. Polypropylene is rather chemically inert as well as fairly thermally stable (to 150°C), allowing a wide range of chemical reactions to be used for solid phase synthesis without affecting the bag material. Polystyrene cross-linked with 1% divinylbenzene, 100-200 mesh, is mainly used as the solid support, but other types of base resin can be used as well, e.g. TentaGel. The size of the resin beads must exceed the size of the pores of the polypropylene mesh material of the T-bags to avoid resin loss during synthesis. Syntheses are carried out manually, using semi automation, or within a multiple peptide synthesizer. The preparation of T-bags for solid phase synthesis, starting with 100 mg resin per bag, is described in Protocol 1. Synthesis using the T-bag method can be performed using either Boc or Fmoc synthetic strategies. For all manipulations, enough solvent should be used to cover the T-bags (about 3-4 ml per bag containing 100 mg of resin). To enable efficient washings and reactions, the reaction vessels (polyethylene bottles) should be shaken vigorously, preferably through the use of a reciprocating shaker.
10
Herz, Norman, e Ervan G. Garrison. "Radiation-Damage, Cosmogenic, and Atom-Counting Methods". In Geological Methods for Archaeology. Oxford University Press, 1998. http://dx.doi.org/10.1093/oso/9780195090246.003.0010.
Abstract (sommario):
Fission-track dating, one of the more recent techniques involving the use of radioactivity, has developed one of the widest ranges of applications. Dates of objects have been obtained ranging from 6 months to 109 years BP. Volcanic tephra, obsidian, man-made and basaltic glass, meteorites, and mica have been dated. A more apt term is nuclear-track dating because fissionable elements do not have to be present in the material. Fission, which produces one form of nuclear track, is a rare mode of radioactive decay. A more common decay is alpha decay, which produces a different type of track. Uranium 238 fissions spontaneously and has a well-defined half-life. It also fissions in the presence of neutrons such as are produced by reactors, accelerators, or neutron "howitzers." About 99.27% of all uranium is uranium 238. Robert L. Fleischer, Paul B. Price, and Robert M. Walker, who have done most of the original work in this field, have determined that most minerals contain this isotope in amounts from a few parts per billion (ppb) to many parts per million (ppm). These researchers devised a chart which characterizes the ease of use of this technique as a function of the uranium concentration. A high uranium concentration allows an "easily measured" age where the observer spends an hour at the microscope counting chemically etched fission tracks. For "considerable labor," 40 hours of such work is assumed. Ancient synthetic glass typically contains 1-2 ppm of uranium, so most glasses older than 8,000 years are datable. Most pottery clay contains about 5 ppm of uranium in either the clay itself or other minerals that occur as inclusions. It is very probable that some pottery clays or the mineral inclusions, such as zircon, might contain higher concentrations than this, which would make the age measurement lie between "easily" and "with considerable labor." It is important to point out that mineral inclusions such as zircons or micas act as solid-state detectors in that they register fissions as a track on the surface in contact with the pottery clay. Both fission and alpha events can do this.
Atti di convegni sul tema "PWR TYPE REACTORS":
1
Rivai, Abu Khalid, Ferhat Aziz e Minoru Takahashi. "Design Study of 300 MWt PWR Fueled With UO2 Coated Fuel Particle". In 14th International Conference on Nuclear Engineering. ASMEDC, 2006. http://dx.doi.org/10.1115/icone14-89287.
Abstract (sommario):
A neutronic design was performed for 300 MWt Pressurized Water Reactor (PWR) with UO2 compacts made of coated fuel particles (CFP) comparing that with sintered pellets made of UO2 powder as ordinary fuel type. UO2 CFP type was enriched 4.8% of 235U and UO2 ordinary type was enriched 5% of 235U. Both reactors were operated with single batch refueling system with a cycle period of 3 years. The purpose of the design was to investigate the applicability of UO2 CFP type to PWR comparing with UO2 ordinary type that commonly used for PWR. The calculation was done with SRAC (Standard Reactor Analysis Code) computer code and nuclear library of JENDL-33. The results of calculation showed that k-effective for both type of fuel could be maintained at critical condition for 3 years operation without refueling. The k-effective and the Doppler coefficients have been calculated for all types of fuel at 600 K and 900 K degrees. The results of calculation showed that for all types of fuel Doppler coefficient was negative, which was good for inherent safety characteristic. The size optimization design showed that the active core dimensions of UO2 CFP type reactor was about 2 times larger than the UO2 ordinary type reactor.
2
Guo, Rui, e Akifumi Yamaji. "Conceptual Core Design of Breeding BWR". In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-66829.
Abstract (sommario):
High breeding with light water cooling is not easy to be achieved. The main obstacle is the moderating effect of light water, which softens the neutron spectrum. Decreasing the volume ratio of coolant to fuel is normally introduced as a way to harden the neutron spectrum and achieve breeding with light water cooling. Therefore, the tight-lattice assembly was proposed to design reactors cooled by light water with hard neutron spectrum. However, most of them were High Conversion LWRs and none achieved high breeding to meet the growth rate of energy demand in advanced countries. Tightly packed fuel assembly is designed for the purpose of high breeding. The number ratio of hydrogen atoms to heavy metal atoms (H/HM) in this assembly is significantly reduced to less than 0.1 which is about 1/6 of that of Reduced-Moderation Water Reactor (RMWR). Super Fast Breeding Reactor (Super FBR) is one kind of Supercritical Light Water Cooled Reactors (Super LWRs), which adopts these assemblies, obtaining high breeding of CSDT (less than 50 years). The high breeding performance of Super FBR indicates that, application of the tightly packed fuel assembly on conventional LWR-type reactors, such as BWR-type or PWR-type reactor, may also be effective in achieving high breeding. Compared with Super FBR, the conventional LWR-type reactors with technologies which are currently in use are expected to be easier to implement. When comparing the two main LWR types, BWR-type and PWR-type, BWR-type gains more advantages on breeding, since the coolant is boiling water that generates larger amount of void in the reactor core, leading to a harder neutron spectrum. Meanwhile, from the viewpoint of safety, the negative void reactivity should be satisfied, which is consistent with conventional LWRs. From the viewpoint of neutron economy, high enrichment should be avoided as well. This study aims to design the BWR-type reactor with the tightly packed fuel assemblies, which attains both high breeding and negative reactivity.
3
Mascari, Fulvio, Giuseppe Vella e Brian G. Woods. "TRACE Code Analyses for the IAEA ICSP on “Integral PWR Design Natural Circulation Flow Stability and Thermo-Hydraulic Coupling of Containment and Primary System During Accidents”". In ASME 2011 Small Modular Reactors Symposium. ASMEDC, 2011. http://dx.doi.org/10.1115/smr2011-6594.
Abstract (sommario):
Considering the world energy demand increase in order to fulfill an environmental and economic sustainability, the energy policy of each country has to diversify the sources of energy and use stable, safe energy production option able of producing electricity in a clean way contributing in cutting the CO2 emission. In the framework of the sustainable development, today the use of advanced nuclear power plant, have an important role in the environmental and economic sustainability of country energy strategy. In the last 20 years, in fact, the international community, taking into account the operational experience of the nuclear reactors, starts the development of new advanced reactor designs considering also the use of natural circulation for the cooling of the core in normal and transient conditions. In this framework, Oregon State University (OSU) has constructed, under a U.S. Department of Energy grant, a system level test facility to examine natural circulation phenomena characterizing the Multi-Application Small Light Water Reactor (MASLWR) design, a small modular integral pressurized light water reactor relying on natural circulation during both steady state and transient operation. It includes an integrated helical coil steam generator as well. Starting from an experimental campaign in support of the MASLWR concept design verification, the planned work, will be not only to specifically investigate the concept design further but also advance the broad understanding of integral natural circulation reactor plants and accompanying passive safety features as well. An IAEA International Collaborative Standard Problem (ICSP) on “Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System During Accidents” is hosting at OSU and the experimental data will be developed at the OSU-MASLWR facility. The purpose of this IAEA ICSP is to provide experimental data on single/two-phase flow instability phenomena under natural circulation conditions and coupled containment/reactor vessel behavior in integral-type light water reactors. These data can be used to assess thermal hydraulic codes for reactor system design and analysis as well. The first planned test investigates a stepwise reduction in the primary mass inventory of the facility while operating at reduced power (decay power). The second planned test, investigates a loss of feed water transient with subsequent primary blowdown due to automatic depressurization system actuation and long term cooling phase. The target of this paper is to contribute to the thermal hydraulic analysis of the expected phenomena of these transients on the basis of the TRACE V5 Patch 01 calculated data developed during the double-blind phase of the ICSP.
4
Pignatel, Jean-Franc¸ois. "The Integral PWR SIR Transients: Comparisons Between CATHARE and RELAP Codes". In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22460.
Abstract (sommario):
Within the framework of the research program on innovative light water reactors, the SERI (Service of Studies on Innovative Reactors) of the French Atomic Energy Commission (CEA), is presenting a predictive study on the modeling of a low-power integral Pressurized Water Reactor, using the CATHARE thermalhydraulic code. The concept selected for this study is that of the SIR reactor project, developed by AEA-T and ABB consortium. This very interesting concept is no doubt that which is the most complete to this date, and on which most information in the literature can be obtained. Many safety calculations made with the RELAP code are also available and represent a highly interesting base for comparison purposes, in order to improve the approach on the results obtained with CATHARE. A comparison of the behavior of the two codes is thus presented in this article. This study therefore shows that CATHARE finely models this type of new PWR concept. The transients studied cover a large area, ranging from natural circulation to loss of primary coolant accidents. The ATWS and a power transient have also been calculated. The comparison made between the CATHARE and RELAP results shows a very good agreement between the two codes, and leads to a very positive conclusion on the pertinence of simulating an integral PWR. Moreover, even though this study is a thorough investigation on the subject, it confirms the potentially safe nature of the SIR reactor.
5
Xu, Tingting, Jiesheng Min, Guofei Chen, Samuel Delepine, Serge Bellet, Jian Ge e Wenxi Tian. "Numerical Investigation of Flow Diffuser Optimization for a PWR Reactor With Code_Saturne: Analysis on EPR Type Reactor". In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60934.
Abstract (sommario):
In order to provide scientific basis for new reactor design in terms of optimal Flow Diffuser, a study was launched on the analysis and investigation on flow distribution of existing flow diffuser design of five different reactors including EPR, VVER, Konvoi, APR+ and Westinghouse. The strategy is to change each time only the flow diffuser within the 1/5 scale BORA mock-up1. The authors consider that the optimal design needs to reach a homogeneous inlet core flow rate, which is defined as figures of merit. This study combines PIRT (Phenomena Identification and Ranking Table) methodology and Computational Fluid Dynamics (CFD) calculations to identify the optimal flow diffuser design. This paper introduces main physical phenomena analysis with PIRT methodology to list the most important phenomena and parameters from cold legs to lower core plate which have a high level of influence on the flow distribution at reactor core inlet for EPR reactor type. CFD calculations are performed under the injection condition of 0.1 m3/s per cold leg at 1/5 scale BORA mock-up for EPR flow diffuser configuration. EDF in-house open-source codes are applied to perform CFD calculations including Salomé2 for pre-processing and postprocessing and Code_Saturne3 for solver.
6
Yan, Yikuan, Shanbin Shi e Mamoru Ishii. "Scaling Analysis and Facility Design for Stability Investigation During Accidents in a PWR-Type SMR". In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60476.
Abstract (sommario):
Small modular reactor (SMR) concept has been developed as one of the key solutions for the growing demand of safe and clean energy. SMR designs can be applied extensively in areas such as sea water desalination and small-scale power generation etc. Unlike conventional light water reactors, most SMRs greatly simplify the structure of reactor pressure vessel, usually eliminate pumps and use natural circulation to cool down the core and transfer energy. However, flow instability may easily occur and affect the entire two phase natural circulation, which is of great importance for the start-up and normal operation process of BWR-type SMRs. For PWR-type SMRs, two-phase natural circulation could exist during accidents such as small break loss of coolant accident (SBLOCA) and loss of heat sink. Current research aims to experimentally investigate potential flow instabilities related to natural circulation for a PWR-type SMR during the accidents. For current research, the NuScale reactor design is selected as the research prototype. In this paper, the design and scaling analysis of a scaled PWR-type experimental facility are provided. In order to experimentally study the natural circulation behavior of PWR-type SMR during accidental scenarios, detailed scaling analyses are necessary to ensure that the scaled phenomena could be obtained in a laboratory test facility. A three-level scaling method is used to get the scaling ratios derived from various non-dimensional numbers. An ideally scaled facility is first accomplished based on derived scaling ratios. RELAP5 simulations of both steady state and transient cases for the ideally scaled facility are performed and compared to the prototype to ensure the accuracy of the scaling analysis. Then the ideally scaled facility is modified under engineering considerations and an engineering scaled facility is designed. Similar RELAP5 analyses are performed on the engineering scaled facility and the results match well with those in the prototype and ideally scaled facility.
7
Awan, Muhammad Qasim, Liangzhi Cao e Hongchun Wu. "Burnable Poisons Alternative Configurations in AT-FCM PWR Fuel Assembly". In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-66444.
Abstract (sommario):
The Fukushima Daiichi Nuclear Power Plant accident in Japan was one of the severest accidents in history of nuclear power plants. This accident changed the prospect of the Nuclear Engineers all around the globe, regarding safety enhancement of nuclear reactors. Since then many aspects of nuclear reactors regarding the improvements of safety features are under investigation and R&D efforts are underway around the world. Use of innovative fuels in present as well as future reactor designs is one of the major potential areas of these ongoing efforts. Fully Ceramic Micro-Encapsulated (FCM) fuel originally developed for use in high temperature gas cooled nuclear reactors, has proven worth for operating in high temperature environment with high burn-up. Due to its additional fission product barrier in the form of strong SiC layer, it is worth using for application as PWR fuel, thus providing potential benefits related to safety and operational aspects of power plant. However, use of FCM fuel in a PWR also has some operational constraints such as the moderator temperature coefficient (MTC) of reactivity has less negative value and even becomes positive when higher concentration of soluble boron is used for the reactivity control. Thus, use of burnable poison material becomes more important to control the access reactivity throughout the cycle length in such a way the quantity of soluble boron to be used is much lower to prevent the positive MTC value or even soluble boron free operation is possible. In present studies a new candidate designs of PWR fuel assembly of 12×12 square array configuration has been used to study the BP material impact on cycle length. Monte Carlo code MVP-BURN is utilized for the analysis to accurately model the double heterogeneity arising due to TRISO type fuel. Two standard materials i.e. Erbia and Gadolinia are used and different configuration including mixing of BP in matrix material, fuel kernel and in QUADRISO form are analyzed and compared with each other. Impact of the residual poison are also analyzed and additional enrichment required to overcome the impact of residual BP material reactivity are calculated. Different configurations support different BP materials. However, it has been found that with an appropriate combination of both materials and configuration, it is possible to minimize the use of soluble boron. Finally, the recommended assembly configuration is analyzed for MTC value during the entire cycle length, showing sustainability of negative values of MTC for the region of interest. With this kind of arrangements, it is possible to use FCM type fuel for present as well as future generations of the PWRs.
8
Jiang, Baihui, Zhiwei Zhou, Zhaoyang Xia e Qian Sun. "Comparative Thermal Analyses Between Theoretical Mode and RELAP5 Code Simulation for OTSG of a Small PWR". In 2020 International Conference on Nuclear Engineering collocated with the ASME 2020 Power Conference. American Society of Mechanical Engineers, 2020. http://dx.doi.org/10.1115/icone2020-16281.
Abstract (sommario):
Abstract Due to the low nuclear safety risk, low initial investment cost and short construction period, integrated small nuclear reactors have received wide attention from all over the world. As advanced new type of nuclear reactors, the technologies of integrated small nuclear reactors are in the process of exploration and development. Steam generators are used as the heat transfer system for energy exchange between the primary and secondary circuit in reactors, and their heat transfer analysis is very important for reactor design and development. Due to the simple structure, strong heat exchange capacity and timely load following, Once-Through Steam Generators (OTSGs) are the mainly used steam generators in the design of integrated small nuclear reactors. RELAP5/MOD4.0 is a commercial software developed by Innovative System Software, LLC for transient analysis of light water reactors (LWR). After years of development and improvement, RELAP5 has been a basic tool for analysis and calculation of various simulators of nuclear power plants. However, when RELAP5 models steam generators, only structural models related to straight pipes can be established, which is very inconvenient for the heat transfer research of Once-Through Steam Generators. In this paper, Once-Through Steam Generators with specified structural parameters are taken as the research object. The heat transfer calculation is performed on the simplified inclined tube models by RELAP5 code and the theoretical calculation of the spiral tube heat transfer models is also carried out. Comparing the steam outlet temperature on the primary and secondary sides, the heat exchange power, the average heat transfer coefficient and the tube length of different heat exchange zones under given primary and secondary side inlet fluid conditions, it is confirmed that the RELAP5 heat transfer calculation is verified for simplifying Once-Through Steam Generators with inclined tube models.
9
Mogami, Yuichi, Toru Matsubara, Seiji Yaguchi, Tomohiro Tsuda e Koji Fujimoto. "Swelling Characteristics of a Type 304SS Baffle Plate Irradiated up to 50 DPA in PWR and Validation of a Swelling Equation". In ASME 2017 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/pvp2017-66248.
Abstract (sommario):
When austenitic stainless steel is highly irradiated in a reactor under high temperature, voids will be created in the material, leading to volumetric expansion of structures. This phenomenon is known as void swelling. The deformation caused by the swelling possibly deteriorates the functionality of reactor internals in Pressurized Water Reactors (PWRs), especially baffle former assemblies. To evaluate the functionality of the internals against the swelling and to assure the structural integrity, simulation technologies play a key role, enabling the estimation of the swelling behavior of the internals through plant life. The simulation results strongly depend on inputs, especially on a swelling equation; however, it includes uncertainty to some extent because quite limited swelling data in PWR environment have been available, which are necessary for the validation of the equation. To enable the validation of swelling equations and improve the reliability on the simulations, the authors investigated the swelling characteristics of a type 304SS baffle plate removed from a decommissioned PWR plant. A total of nine swelling data were obtained with the variety in neutron dose (33 to 47 dpa) and irradiation temperature (299 to 327°C). The swelling ratios obtained are ranging from 0.02 to 0.08%, which corresponds well with the swelling equation, showing the similar temperature dependency with the equation. Since the irradiation temperature range of the obtained data, up to 327°C, covers major part of baffle former assemblies, swelling ratios of most part of them are expected to be small, which is probably too small to harm the functionality of the assemblies. The results contribute to the better confidence on swelling simulations and to assure the integrity of PWR reactor internals.
10
Hourcade, E. "Physics of Plutonium and Americium Recycling in PWR Using Advanced Fuel Concepts". In 12th International Conference on Nuclear Engineering. ASMEDC, 2004. http://dx.doi.org/10.1115/icone12-49604.
Abstract (sommario):
PWR waste inventory management is considered in many countries including Frances as one of the main current issues. On this subject, the French 1991 Bataille’s law set up a 15 years research program on three main axes: sub-surface storage, deep geological storage, transmutation using critical or subcritical burners. Amongst the output Actinides, Pu and Am are the 2 main contents both in term of volume and long term radio-toxicity. Waiting for the Generation IV systems implementation (2035–2050), one of the mid-term solutions for their transmutation involves the use of advanced fuels in Pressurized Water Reactors (PWR). These have to require as little modification as possible of the core internals, the cooling system and fuel cycle facilities (fabrication and reprocessing). The present paper is focussed on the reactor physics characteristics, as a preliminary step in the evaluation of options, knowing that others homogeneous and heterogeneous assemblies have been studied by the CEA ([1] to [5]). The main neutronic parameters to be considered for Pu and Am recycling in PWR are void coefficient (αvoid), Doppler coefficient (αDopp), fraction of delayed neutrons (β) and power distribution (especially for heterogeneous configurations). The modification of the moderation ratio, the opportunity to use inert matrices (targets), the optimisation of Uranium, Plutonium and Americium contents are the key parameters to play with. One of the solutions presented here is a heterogeneous assembly with regular moderation ratio composed with both target fuel rods (Pu and Am embedded in an inert matrix) and standard UO2 fuel rods. An EPR (European Pressurised Reactor) type reactor, loaded only with assemblies containing 84 peripheral targets, can reach an Americium consumption rate of [4.4; 23 kg/TWhe] depending on the assembly concept. For Pu and Am inventories stabilisation, the theoretical fraction of reactors loaded with Pu + Am or Pu assemblies is about 60%. For Americium inventory stabilisation, the fraction decreases down to 16%, but Pu is produced at a rate of 18.5 Kg/Twhe (−25% compared to one through UOX cycle).