Relevant bibliographies by topics / European Pressurized Reactor

Academic literature on the topic 'European Pressurized Reactor'

Author: Grafiati
Published: 14 March 2023

Create a spot-on reference in APA, MLA, Chicago, Harvard, and other styles

Select a source type:

Contents

Consult the lists of relevant articles, books, theses, conference reports, and other scholarly sources on the topic 'European Pressurized Reactor.'

Next to every source in the list of references, there is an 'Add to bibliography' button. Press on it, and we will generate automatically the bibliographic reference to the chosen work in the citation style you need: APA, MLA, Harvard, Chicago, Vancouver, etc.

You can also download the full text of the academic publication as pdf and read online its abstract whenever available in the metadata.

Journal articles on the topic "European Pressurized Reactor"

1

Leny, J. C. "The European Pressurized Water Reactor / Der Europäische Druckwasserreaktor." Kerntechnik 58, no. 6 (June 1, 1993): 353–57. http://dx.doi.org/10.1515/kern-1993-580615.

Full text
APA, Harvard, Vancouver, ISO, and other styles
2

Czech, J., J. Wirkner, M. Yvon, M. Serret, U. Krugmann, K. E. Schmidt, J. P. Berger, and M. Grenet. "European pressurized water reactor: safety objectives and principles." Nuclear Engineering and Design 187, no. 1 (January 1999): 25–32. http://dx.doi.org/10.1016/s0029-5493(98)00255-6.

Full text
APA, Harvard, Vancouver, ISO, and other styles
3

Baumgartl, B. J., and F. Bouteille. "The European Pressurized Water Reactor (EPR): an advanced PWR." Revue Générale Nucléaire, no. 6 (November 1994): 478–83. http://dx.doi.org/10.1051/rgn/19946478.

Full text
APA, Harvard, Vancouver, ISO, and other styles
4

Bonhomme, Nico. "Systems organization for the European pressurized water reactor (EPR)." Nuclear Engineering and Design 187, no. 1 (January 1999): 71–78. http://dx.doi.org/10.1016/s0029-5493(98)00258-1.

Full text
APA, Harvard, Vancouver, ISO, and other styles
5

Quinot, P., and G. Desfontaines. "The main components of the European pressurized water reactor." Nuclear Engineering and Design 187, no. 1 (January 1999): 121–33. http://dx.doi.org/10.1016/s0029-5493(98)00261-1.

Full text
APA, Harvard, Vancouver, ISO, and other styles
6

Teichel, Holger. "Objectives in developing the European pressurized water reactor (EPR)." Nuclear Engineering and Design 165, no. 1-2 (August 1996): 271–76. http://dx.doi.org/10.1016/0029-5493(95)01158-7.

Full text
APA, Harvard, Vancouver, ISO, and other styles
7

Valenti, Michael. "A Next-Generation Reactor." Mechanical Engineering 120, no. 08 (August 1, 1998): 68–71. http://dx.doi.org/10.1115/1.1998-aug-5.

Full text
Abstract:
This article highlights that Electricité de France’s (EDF) N4 nuclear technology has increased European standards of power, efficiency, and safety beyond previous limits. EDF, the Paris-based French national utility, has developed and is operating the N4 reactor, capable of generating 1450 megawatts, at its power plant in Chooz. The N4 was designed to put public safety concerns to rest while providing more power than the previous generation of 1,300-megawatt EDF reactors. This installation represents the next step in French, European, and possibly the world’s nuclear power. Chooz A began operations in 1967 as the first pressurized water reactor (PWR) in France, originally based on the 185-megawatt synchronized PWR Yankee Rowe plant in Massachusetts. Typical reactor safety systems analyze a problem after it occurs. Such a procedure involves painstaking historical, reconstruction that is time-consuming, often difficult to interpret, and less reliable as time passes.
APA, Harvard, Vancouver, ISO, and other styles
8

Skrzypek, Maciej, and R. Laskowski. "Thermal-hydraulic calculations for a fuel assembly in a European Pressurized Reactor using the RELAP5 code." Nukleonika 60, no. 3 (September 1, 2015): 537–44. http://dx.doi.org/10.1515/nuka-2015-0110.

Full text
Abstract:
Abstract The main object of interest was a typical fuel assembly, which constitutes a core of the nuclear reactor. The aim of the paper is to describe the phenomena and calculate thermal-hydraulic characteristic parameters in the fuel assembly for a European Pressurized Reactor (EPR). To perform thermal-hydraulic calculations, the RELAP5 code was used. This code allows to simulate steady and transient states for reactor applications. It is also an appropriate calculation tool in the event of a loss-of-coolant accident in light water reactors. The fuel assembly model with nodalization in the RELAP5 (Reactor Excursion and Leak Analysis Program) code was presented. The calculations of two steady states for the fuel assembly were performed: the nominal steady-state conditions and the coolant flow rate decreased to 60% of the nominal EPR flow rate. The calculation for one transient state for a linearly decreasing flow rate of coolant was simulated until a new level was stabilized and SCRAM occurred. To check the correctness of the obtained results, the authors compared them against the reactor technical documentation available in the bibliography. The obtained results concerning steady states nearly match the design data. The hypothetical transient showed the importance of the need for correct cooling in the reactor during occurrences exceeding normal operation. The performed analysis indicated consequences of the coolant flow rate limitations during the reactor operation.
APA, Harvard, Vancouver, ISO, and other styles
9

Yan, Jia-chuan, Xiao-fei Jin, Feng Qin, Zheng Li, Feng Fan, and Jin-ping Ou. "Modular construction mechanics of a European pressurized reactor steel containment liner." Journal of Zhejiang University-SCIENCE A 18, no. 6 (June 2017): 467–86. http://dx.doi.org/10.1631/jzus.a1600136.

Full text
APA, Harvard, Vancouver, ISO, and other styles
10

Bury, Tomasz. "Evaluation of passive autocatalytic recombiners operation efficiency by means of the lumped parameter approach." Nukleonika 60, no. 2 (June 1, 2015): 339–45. http://dx.doi.org/10.1515/nuka-2015-0042.

Full text
Abstract:
Abstract The problem of hydrogen behavior in containment buildings of nuclear reactors belongs to thermal-hydraulic area. Taking into account the size of systems under consideration and, first of all, safety issues, such type of analyses cannot be done by means of full-scale experiments. Therefore, mathematical modeling and numerical simulations are widely used for these purposes. A lumped parameter approach based code HEPCAL has been elaborated in the Institute of Thermal Technology of the Silesian University of Technology for simulations of pressurized water reactor containment transient response. The VVER-440/213 and European pressurised water reactor (EPR) reactors containments are the subjects of analysis within the framework of this paper. Simulations have been realized for the loss-of-coolant accident scenarios with emergency core cooling system failure. These scenarios include core overheating and hydrogen generation. Passive autocatalytic recombiners installed for removal of hydrogen has been taken into account. The operational efficiency of the hydrogen removal system has been evaluated by comparing with an actual hydrogen concentration and flammability limit. This limit has been determined for the three-component mixture of air, steam and hydrogen. Some problems related to the lumped parameter approach application have been also identified.
APA, Harvard, Vancouver, ISO, and other styles
More sources

Dissertations / Theses on the topic "European Pressurized Reactor"

1

Nie, Markus. "Temporary melt retention in the reactor pit of the European pressurized water reactor (EPR)." [S.l. : s.n.], 2005. http://www.bsz-bw.de/cgi-bin/xvms.cgi?SWB11759373.

Full text
APA, Harvard, Vancouver, ISO, and other styles
2

Lebas, Elise. "Analysis of fire risk aspects and development of the fire protection strategy for the European Pressurized Reactor at Hinkley Point C, United Kingdom." Thesis, KTH, Fysik, 2015. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-172411.

Full text
APA, Harvard, Vancouver, ISO, and other styles
3

Toppila, Timo, Ulrich Rohde, Bengt Hemström, Yuri Bezrukov, and Sören Kliem. "The European project FLOMIX-R: Description of the slug mixing and buoyancy related experiments at the different test facilities(Final report on WP 2)." Forschungszentrum Dresden, 2010. http://nbn-resolving.de/urn:nbn:de:bsz:d120-qucosa-28632.

Full text
Abstract:
The goal of the work described in this report was the experimental investigation of the mixing of coolant with different quality (temperature, boron concentration) in nuclear reactors on the way from the cold leg through the downcomer and lower plenum to the core inlet in a systematic way. The obtained data were used for the clarification of the mixing mechanisms and form a data basis for the validation of computational fluid dynamics (CFD) codes. For these purposes, experiments on slug mixing have been performed at two test facilities, modelling different reactor types in scale 1:5, the Rossendorf and Vattenfall test facilities. The corresponding accident scenario is the start-up of first main coolant pump (MCP) after formation of a slug of lower borated water during the reflux-condenser mode phase of a small break loss of coolant accident (LOCA). The matrices for the experiments were elaborated on the basis of the key phenomena, being responsible for the coolant mixing during pump start-up. Slug mixing tests have also been performed at the VVER-1000 facility of EDO Gidropress to meet the specifics of this reactor type. The mixing of slugs of water of different quality is also very important for pre-stressed thermal shock (PTS) situations. In emergency core cooling (ECC) situations after a LOCA, cold ECC water is injected into the hot water in the cold leg and downcomer. Due to the large temperature differences, thermal shocks are induced at the reactor pressure vessel (RPV) wall. Temperature distributions near the wall and temperature gradients in time are important to be known for the assessment of thermal stresses. One of the important phenomena in connection with PTS is thermal stratification, a flow condition with a vertical temperature profile in a horizontal pipe. Due to the fluctuating character of the flow, this may cause thermal fatigue in the pipe. Besides of thermal fatigue, a single thermal shock can also be relevant for structural integrity, if it is large enough, especially in the case, that the brittle fracture temperature of the RPV material is reduced due to radiation embrittlement. Therefore, additional to the investigations of slug mixing during re-start of coolant circulation, the mixing of slugs or streams of water with higher density with the ambient fluid in the RPV was investigated. The aim of these investigations was to study the process of turbulent mixing under the influence of buoyancy forces caused by the temperature differences. Heat transfer to the wall and thermal conductivity in the wall material have not been considered. Experiments on density driven mixing were carried out at the Rossendorf and the Fortum PTS facilities.
APA, Harvard, Vancouver, ISO, and other styles
4

Toppila, Timo, Ulrich Rohde, Bengt Hemström, Yuri Bezrukov, and Sören Kliem. "The European project FLOMIX-R: Description of the slug mixing and buoyancy related experiments at the different test facilities(Final report on WP 2)." Forschungszentrum Rossendorf, 2005. https://hzdr.qucosa.de/id/qucosa%3A21690.

Full text
Abstract:
The goal of the work described in this report was the experimental investigation of the mixing of coolant with different quality (temperature, boron concentration) in nuclear reactors on the way from the cold leg through the downcomer and lower plenum to the core inlet in a systematic way. The obtained data were used for the clarification of the mixing mechanisms and form a data basis for the validation of computational fluid dynamics (CFD) codes. For these purposes, experiments on slug mixing have been performed at two test facilities, modelling different reactor types in scale 1:5, the Rossendorf and Vattenfall test facilities. The corresponding accident scenario is the start-up of first main coolant pump (MCP) after formation of a slug of lower borated water during the reflux-condenser mode phase of a small break loss of coolant accident (LOCA). The matrices for the experiments were elaborated on the basis of the key phenomena, being responsible for the coolant mixing during pump start-up. Slug mixing tests have also been performed at the VVER-1000 facility of EDO Gidropress to meet the specifics of this reactor type. The mixing of slugs of water of different quality is also very important for pre-stressed thermal shock (PTS) situations. In emergency core cooling (ECC) situations after a LOCA, cold ECC water is injected into the hot water in the cold leg and downcomer. Due to the large temperature differences, thermal shocks are induced at the reactor pressure vessel (RPV) wall. Temperature distributions near the wall and temperature gradients in time are important to be known for the assessment of thermal stresses. One of the important phenomena in connection with PTS is thermal stratification, a flow condition with a vertical temperature profile in a horizontal pipe. Due to the fluctuating character of the flow, this may cause thermal fatigue in the pipe. Besides of thermal fatigue, a single thermal shock can also be relevant for structural integrity, if it is large enough, especially in the case, that the brittle fracture temperature of the RPV material is reduced due to radiation embrittlement. Therefore, additional to the investigations of slug mixing during re-start of coolant circulation, the mixing of slugs or streams of water with higher density with the ambient fluid in the RPV was investigated. The aim of these investigations was to study the process of turbulent mixing under the influence of buoyancy forces caused by the temperature differences. Heat transfer to the wall and thermal conductivity in the wall material have not been considered. Experiments on density driven mixing were carried out at the Rossendorf and the Fortum PTS facilities.
APA, Harvard, Vancouver, ISO, and other styles
5

Nie, Markus [Verfasser]. "Temporary melt retention in the reactor pit of the European pressurized water reactor (EPR) / vorgelegt von Markus Nie." 2005. http://d-nb.info/974884073/34.

Full text
APA, Harvard, Vancouver, ISO, and other styles
6

Montwedi, Ontlametse Emmanuel. "Neutronic simulation of a European Pressurised Reactor / Ontlametse Emmanuel Montwedi." Thesis, 2014. http://hdl.handle.net/10394/15437.

Full text
Abstract:
The South African government’s integrated resource plan for electricity IRP2010 states that the country plans to have an additional 9.6 GW of nuclear power on the national electricity grid by 2030. In support of this, the NRF-funded SARChI Research Chair in Nuclear Engineering within the School of Mechanical and Nuclear Engineering at the North-West University recently initiated research studies focused on Light Water Reactor (LWR) systems. These studies inter alia involve coupled neutronic and thermal hydraulic analyses of selected LWR systems. This study focuses on the steady state neutronic analysis of the European Pressurised Reactor (EPR) using Monte-Carlo N-Particle (MCNP5). The neutronic model will in due course be coupled to a thermal hydraulic model forming part of a broader study of the system. The Monte Carlo neutron transport code MCNP5 has been widely used since the 1950s for analysis of existing and future reactor systems due to its ability to simulate complex fuel assemblies without making any significant approximations. The primary aim of the study was to develop an input model for a representative fresh fuel assembly of the US EPR reactor core from which the fluxes and fission power of the reactor can be obtained. There after a 3D model of full EPR core developed by the school of mechanical and nuclear engineering based on findings of this work is also tested. The results are compared to those in the US EPR Final Safety Analysis Report. Agreement in major core operational parameters including the keff eigenvalue, axial and radial power profiles and control rod worth are evaluated, from which consistency of the model and results will be confirmed. Further convergence of the model within a reasonable time is assessed.
MSc (Engineering Sciences in Nuclear Engineering), North-West University, Potchefstroom Campus, 2014
APA, Harvard, Vancouver, ISO, and other styles

Books on the topic "European Pressurized Reactor"

1

Germany), Kerntechnische Gesellschaft (Bonn, and Société française d'énergie nucléaire, eds. The European pressurized water reactor, EPR: Proceedings, KTG/SFEN conference, Maritim Hotel Cologne, 19-21 October 1997. Bonn: INFORUM Verlag, 1997.

Find full text
APA, Harvard, Vancouver, ISO, and other styles
2

Engineers, Institution of Mechanical, ed. Developments in industrial compressors and their systems: European conference, 12-13 April 1994, Institution of Mechanical Engineers, One Great George Street, London. Bury St. Edmunds: Mechanical Engineering Publications for IMechE, 1994.

Find full text
APA, Harvard, Vancouver, ISO, and other styles
3

(Editor), Damien Feron, and J. M. Olive (Editor), eds. Corrosion issues in light water reactors: Stress corrosion cracking (EFC 51) (European Federation of Corrosion Publications). CRC, 2007.

Find full text
APA, Harvard, Vancouver, ISO, and other styles
4

Proceedings of the Institution of Mechanical Engineers: Developments in Industrial Compressors and Their Systems, European Conference (Imeche Event Publications). Wiley, 1994.

Find full text
APA, Harvard, Vancouver, ISO, and other styles

Book chapters on the topic "European Pressurized Reactor"

1

Bouteille, François. "The French German Project of the European Pressurized Water Reactor (EPR)." In Preparing the Ground for Renewal of Nuclear Power, 49–54. Boston, MA: Springer US, 1999. http://dx.doi.org/10.1007/978-1-4615-4679-5_5.

Full text
APA, Harvard, Vancouver, ISO, and other styles
2

Keefer, Robert F. "Fertilizers." In Handbook of Soils for Landscape Architects. Oxford University Press, 1999. http://dx.doi.org/10.1093/oso/9780195121025.003.0017.

Full text
Abstract:
Fertilizers for soil on which plants grow come in a variety of forms, such as organic, inorganic, single nutrient, double nutrient, complete fertilizer (contains N, P, and K in that order), speciality fertilizers, composts, and manures. Information about each of these forms follows. Most of the N used in fertilizers is derived from a synthetic process developed by Europeans called the “Claude-Haber process.” This process uses nitrogen gas (N2) from the atmosphere along with hydrogen gas (H2) from natural gas in a device where pressure can be increased and temperature can be raised. The reaction is accelerated using an iron catalyst and removing the product (NH3) as it is formed. The Fe catalyst is subject to poisoning from impurities, such as As, Co, P, or S. Anhydrous ammonia has the highest percentage of N and the cheapest per unit of N since no processing is involved. Anhydrous (without water) ammonia is a gas but when compressed changes to a liquid. For application to soils a pressurized tank is required with a device to inject the liquid ammonia into the soil. Upon release of pressure, the liquid changes back to a gas; however, the ammonia gas reacts with the moisture in the soil to form NH4+ that is available for plants. One problem with ammonia is that NH3 gas is toxic to seedlings and growing plants, so must be applied prior to planting. This limits its use for landscape projects. Salt solutions of aqua ammonia are obtained by dissolving ammonia gas, ammonium nitrate, or urea in water. The amount dissolved will vary the concentration of N in the final product. This can be used in landscape projects, but care must be used as this material can salt out and plug up orifices when sprayed onto a soil. There is no real difference between liquid or solid fertilizers, provided the percentage of N is the same. Ammonia Nitrate [NH4NO3] (33.5% N) Ammonium nitrate is formed by ammonia gas reacting with nitric acid: . . . NH3 + HNO3 → NH4NO3 . . . This material is hygroscopic (absorbs water from the air) and requires moisture-proof bags for storage.
APA, Harvard, Vancouver, ISO, and other styles

Conference papers on the topic "European Pressurized Reactor"

1

Nene, Chandan, B. Bandyopadhyay, and A. P. Tiwari. "Modeling of a large pressurized heavy water reactor by nodal approximation." In 2001 European Control Conference (ECC). IEEE, 2001. http://dx.doi.org/10.23919/ecc.2001.7076273.

Full text
APA, Harvard, Vancouver, ISO, and other styles
2

Bellet, Serge, Nicolas Goreaud, and Norbert Nicaise. "Evolution of the Thermal Hydraulics Design Methodologies for the European Pressurized Reactor (EPR) Vessel." In ASME/JSME 2004 Pressure Vessels and Piping Conference. ASME, 2004. http://dx.doi.org/10.1115/pvp2004-2889.

Full text
APA, Harvard, Vancouver, ISO, and other styles
3

Tiwari, A. P., and B. Bandyopadhyay. "Design of fault-tolerant spatial control system for a large pressurized heavy water reactor." In 1999 European Control Conference (ECC). IEEE, 1999. http://dx.doi.org/10.23919/ecc.1999.7099887.

Full text
APA, Harvard, Vancouver, ISO, and other styles
4

Niffenegger, M., O. Costa Garrido, D. F. Mora, G. Qian, R. Mukin, M. Sharabi, N. Lafferty, and B. Niceno. "Uncertainties in Pressurized Thermal Shock Analyses." In ASME 2019 Pressure Vessels & Piping Conference. American Society of Mechanical Engineers, 2019. http://dx.doi.org/10.1115/pvp2019-94076.

Full text
Abstract:
Abstract Integrity assessment of reactor pressure vessels (RPVs) can be performed either by deterministic fracture mechanics (DFM) or/and by probabilistic fracture mechanics (PFM) analyses. In European countries and Switzerland, only DFM analyses are required. However, in order to establish the probabilistic approach in Switzerland, the advantages and shortcomings of the PFM are investigated in the frame of a national research project. Both, the results from DFM and PFM depend strongly on the previous calculated thermal-hydraulic boundary conditions. Therefore, complete integrity analyses involving several integrated numerical codes and methods were performed for a reference pressurized water reactor (PWR) RPV subjected to pressurized thermal shock (PTS) loads. System analyses were performed with the numerical codes RELAP5 and TRACE, whereas for structural and fracture mechanics calculations, the FAVOR and ABAQUS codes were applied. Additional computational fluid dynamics analyses were carried out with ANSYS/FLUENT, and the plume cooling effect was alternatively considered with GRS-MIX. The results from the different analyses tools are compared, to judge the expected overall uncertainty and reliability of PTS safety assessments. It is shown that the scatter band of the stress intensities for a fixed crack configuration is rather significant, meaning that corresponding safety margins should be foreseen. The conditional probabilities of crack initiation and RPV failure might also differ, depending on the considered random parameters and applied rules.
APA, Harvard, Vancouver, ISO, and other styles
5

Scheuerer, Martina. "Simulation of OECD/NEA International Standard Problem No. 43 on Boron Mixing Transients in a Pressurized Water Reactor." In ASME 2002 Joint U.S.-European Fluids Engineering Division Conference. ASMEDC, 2002. http://dx.doi.org/10.1115/fedsm2002-31407.

Full text
Abstract:
The International Standard Problem, ISP 43, was defined by the OECD/NEA and the US NRC for the validation of three-dimensional Computational Fluid Dynamics (CFD) software. The underlying experiment was performed in the 2×4 Loop Facility of the University of Maryland, College Park, U.S.A (UMCP). The test facility is a scaled-down model of the Three Mile Island TMI-2 reactor with detailed reconstruction of the reactor pressure vessel (RPV). The ISP 43 experiments focussed on rapid boron dilution transients in the RPV cold leg and downcomer. The simulations of the ISP 43 were performed with the CFX-TASCflow software. Numerical errors were monitored by comparing results obtained with different higher order discretisation schemes. Uncertainties related to physical modelling, like buoyancy effects and reactor core models, were also investigated. The simulation results show good agreement with data and prove that CFD methods can be usefully applied to this class of nuclear reactor problems.
APA, Harvard, Vancouver, ISO, and other styles
6

He, Wei, Jing Jiang, Chen Xu, Qiang Lei, Chunyan Xu, Xinhua Liu, Yu Wang, and Feng Xie. "Research on Setting Alarm Thresholds of Gaseous Effluent Radiation Monitoring From Nuclear Power Plants in China." In 2021 28th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2021. http://dx.doi.org/10.1115/icone28-62558.

Full text
Abstract:
Abstract During the past 30 years, several types of nuclear power plants have gained importance in China; these primarily include CPR1000, Russian advanced pressurized water reactor (WWER), passive advanced pressurized water reactor (AP1000) and European advanced pressurized water reactor (EPR). The treatment technologies for the radioactive solid, liquid, and gaseous wastes adopted by various types of nuclear power plants are different; consequently, emission control requirements and specifications for gaseous and liquid effluents also vary accordingly. Especially, the alarm thresholds of gaseous effluent radiation monitoring from nuclear power plants are highly concerned by the regulatory agencies since they determine the upper limit of the emission amount of gaseous radioactivity under normal operation and provide the alarm timely in case of accidents. In this paper, the setting method and basis of the radiation monitoring alarm thresholds for gaseous effluents from various nuclear power plants in China are examined, and the calculation method for the alarm thresholds is compared and analyzed. The unified setting principles and methods for pressurized water reactors (PWRs) are proposed, which can serve as a reference for setting and reviewing the radiation monitoring alarm thresholds of the gaseous effluents from the new nuclear power plants in China, and supply technical support for the improvement of the related regulations and standards for the nuclear power plants in future.
APA, Harvard, Vancouver, ISO, and other styles
7

de Jong, W., J. Andries, and K. R. G. Hein. "Coal-Biomass Gasification in a Pressurized Fluidized Bed Gasifier." In ASME 1998 International Gas Turbine and Aeroengine Congress and Exhibition. American Society of Mechanical Engineers, 1998. http://dx.doi.org/10.1115/98-gt-159.

Full text
Abstract:
In the framework of a multi-national European Joule project, experimental research and modeling concerning co-gasification of biomass and coal in a bubbling pressurized fluidized bed reactor is performed. The impact of fuel characteristics (biomass type, mixing ratio) and process conditions (pressure, temperature, gas residence time, air-fuel ratio and air-steam ratio) on the performance of the gasifier (carbon conversion, fuel gas composition, non-steady state behaviour) was studied experimentally and theoretically. Pelletized straw and miscanthus were used as biomass fuels. The process development unit has a maximum thermal capacity of 1.5 MW and was operated at pressures up to 10 bar and bed temperatures in the range of 650 °C–900 °C. The bed zone of the reactor is 2 m high with a diameter of 0.4 m and is followed by an adiabatic freeboard, approximately 4 m high with a diameter of 0.5 m. Time-averaged as well as time-dependent characteristics of the fuel gas were determined experimentally. The results will be compared with the gas turbine requirements provided by a gas turbine manufacturer, one of the partners in the project. The evaluation of the results will ultimately be used to implement and test an adequate control strategy for the pressurized fluidized bed gasifier integrated with a gas turbine combustion chamber.
APA, Harvard, Vancouver, ISO, and other styles
8

He, Cuizhu, and Yinhui Lan. "Research on Martensitic Stainless Steel Used on the Latch Housing of European Pressurized Reactor (EPR) Control Rod Drive Mechanism (CRDM)." In 2022 29th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2022. http://dx.doi.org/10.1115/icone29-91786.

Full text
Abstract:
Abstract As one of the most important equipment for reactivity control, Control Rod Drive Mechanism (CRDM), which is widely used in Pressurized Water Reactor (PWR) nuclear power plant, has a series of important security functions. As part of the reactor pressure boundary, materials for European Pressurized Reactor (EPR) CRDM pressure housing needs to have relatively high strength and sufficient toughness (i.e., appropriate yield ratio). As an important part of the magnetic circuit, martensitic stainless steel used for latch housing (a part of pressure housing) needs to have good magnetic properties. It is difficult for the yield ratio and the magnetic properties of the materials to meet the specification requirements simultaneously. Based on the design specification, this research refers to the relevant international manufacturing experience and combines the current manufacture with engineering practice. The manufacturing process and key parameters of the material for latching housing of CRDM are established. The martensitic stainless steel which meets the design requirements has been successfully prepared. The comprehensive properties (e.g. yield ratio and magnetic properties) of the material are improved as well. The performance indicators have reached international advanced level. The successful development of the martensitic steel for EPR CRDM latch housing provides strong technical support for the construction of nuclear power projects at home and abroad.
APA, Harvard, Vancouver, ISO, and other styles
9

Moinereau, Dominique, Stéphane Chapuliot, Stéphane Marie, and Philippe Gilles. "NESC VII: A European Project for Application of WPS in RPV Assessment Including Biaxial Loading." In ASME 2012 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2012. http://dx.doi.org/10.1115/pvp2012-78044.

Full text
Abstract:
The Reactor Pressure Vessel (RPV) is an essential component liable to limit the lifetime duration of pressurized water reactor (PWR) plants. The assessment of defects in RPV subjected to PTS (pressurized thermal shock) transients made at an European level does not necessarily take into account the beneficial effect of load history (warm pre-stress WPS) on the resistance of RPV material regarding the risk of brittle failure. Numerous experimental, analytical and numerical results are available, which confirm the beneficial effect of warm pre-stress on RPV steels, with an effective increase of the material resistance regarding the risk of brittle failure. NESC VII (Network for Evaluating Steel Components), a new project dealing with WPS, has been launched in 2008 (in link with the European Network of Excellence NULIFE) with the participation of international organizations (involving R&D and Technical support, Utilities, Manufacturers, Safety & Regularory organizations). Based on experimental, analytical and numerical tasks, the project is focussed on topics generally non covered by past experience on WPS : biaxiality of loading on small and large-scale specimens, influence and effect of irradiation, applicability to intergranular fracture, modeling (including analytical and numerical models) … Among these tasks, new original WPS experiments are conducted on small, medium and large scale specimens to study the influence of biaxial loading on WPS effect, using a fully representative RPV steel (18MND5 steel similar to A533B steel). A full description of NESC VII project, including the present status, is presented in this paper.
APA, Harvard, Vancouver, ISO, and other styles
10

Chen, Yujia, Weifeng Lv, Zhenyu Jiang, and Yonghai Zhou. "Study on Methodology for Quantification of Radioactive Discharges and Limits for Pressurized Water Reactor HPR1000 Based on Operating Experience." In 2022 29th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2022. http://dx.doi.org/10.1115/icone29-92385.

Full text
Abstract:
Abstract For a nuclear power plant, the radioactive gaseous and liquid discharges are the main contributor to radiation exposure to the member of public and non-human biota during normal operation, which hence need to be quantified and to support the environmental impact assessment. When applying the traditional theoretical methodology, due to various and complex mechanisms involved in radioactive gaseous and liquid effluent streams, a number of assumptions need to be made to support the theoretical modeling. The combination of these assumptions can easily lead to overestimate or underestimate of the radioactive discharges and limits and may not represent the actual performance of the plants. As such, to obtain predicted discharges and limits closer to the future actual performance of the plant, it is meaningful and necessary to develop a methodology based on operating experience. This paper has studied and developed a systematic methodology based on operating experience for quantification of radioactive discharges and limits for the 3rd generation pressurized water reactor HPR1000 during normal operation, taking into account the differences on design features and operation management between the HPR1000 and the operating units, the fluctuations due to the variations of plant and system operation parameters and the potential influences from expected events within the normal operation range. This methodology has been successfully applied to HPR1000 and the results have been verified reasonable and appropriate by comparing with the operating experience data from comparable international PWRs. This methodology has been applied to HPR1000 successfully for Generic Design Assessment (GDA) in the UK and the European Utility Requirements for LWR Nuclear Power Plants (EUR) and can also be widely applied for other PWRs with slight adjustment.
APA, Harvard, Vancouver, ISO, and other styles
You might also be interested in the extended bibliographies on the topic 'European Pressurized Reactor' for particular source types:
Journal articles
We offer discounts on all premium plans for authors whose works are included in thematic literature selections. Contact us to get a unique promo code!

To the bibliography