Journal articles: '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.

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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.

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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.

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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.

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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.

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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.

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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.

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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.

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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.

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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.

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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.

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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.

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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.

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11

Lago, Daniel, and Farzad Rahnema. "Benchmark description of a stylized three-dimensional European Pressurized Reactor (EPR) problem." Progress in Nuclear Energy 93 (November 2016): 18–46. http://dx.doi.org/10.1016/j.pnucene.2016.07.009.

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12

Ebbesmeyer, Peter, Ju¨rgen Gausemeier, Holger Krumm, Thorsten Molt, and Thomas Gruß. "Virtual Web Plant: An Internet-Based Plant Engineering Information System." Journal of Computing and Information Science in Engineering 1, no. 3 (June 1, 2001): 257–60. http://dx.doi.org/10.1115/1.1405155.

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During the development of the European Pressurized Water Reactor Project (EPR)—an innovative design concept for a new type of pressurized water reactor—large amounts of up-to-date engineering data (i.e., CAD data, planning documentation) have to be made available to all international project partners for presentation and development. This paper describes the web-based tool Virtual Web Plant (VWP), a tool to integrate three-dimensional models from various CAD plant design tools and to display them interactively. The user is hereby able to navigate easily through both the plant structure and the project documentation. The work presented in this paper is part of a Virtual Reality Research Project of the Heinz Nixdorf Institute and the Siemens AG KWU.

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Abrefah, Rex, Prince Atsu, and Robert Sogbadji. "Neutronic safety analysis of proposed reactor technologies for Ghana’s nuclear power plant using the MCNP code." Nuclear Technology and Radiation Protection 34, no. 3 (2019): 238–42. http://dx.doi.org/10.2298/ntrp190108029a.

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In pursuance of sufficient, stable and clean energy to solve the ever-looming power crisis in Ghana, the Nuclear Power Institute of the Ghana Atomic Energy Commission has on the agenda to advise the government on the nuclear power to include in the country's energy mix. After consideration of several proposed nuclear reactor technologies, the Nuclear Power Institute considered a high pressure reactor or vodo-vodyanoi energetichesky reactor as the nuclear power technologies for Ghana's first nuclear power plant. As part of technology assessments, neutronic safety parameters of both reactors are investigated. The MCNP neutronic code was employed as a computational tool to analyze the reactivity temperature coefficients, moderator void coefficient, criticality and neutron behavior at various operating conditions. The high pressure reactor which is still under construction and theoretical safety analysis, showed good inherent safety features which are comparable to the already existing European pressurized reactor technology.

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14

Ding, Yongjian. "Schutzzielorientiertes Design der Sicherheitsleittechnik." atp magazin 56, no. 05 (April 28, 2014): 54–61. http://dx.doi.org/10.17560/atp.v56i05.2248.

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Safety instrumentation and control in nuclear power plants has to be systematically designed and revisable in order to meet the highest safety requirements. A safety objective oriented design process is presented in combination with the defence in depth philosophy. The approach is applicable for the construction of new plants like the European Pressurized Reactor (EPR) and for the comprehensive refurbishment of existing plants. Its success has been demonstrated in several domestic and international I&C projects.

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15

Fischer, Manfred. "ICONE11-36196 Severe accident mitigation and core melt retention in the European Pressurized Reactor (EPR)." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2003 (2003): 151. http://dx.doi.org/10.1299/jsmeicone.2003.151.

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16

Vignon, Dominique. "Franco-German nuclear cooperation: from the ‘common product’ to the first European pressurized water reactor." Nuclear Engineering and Design 187, no. 1 (January 1999): 1–3. http://dx.doi.org/10.1016/s0029-5493(98)00292-1.

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17

Du Toit, Maria Hendrina, and Vishana Vivian Naicker. "Developing a Full-Core MCNP6 and RELAP5 Model of the European Pressurized Reactor Using NWURCS." Nuclear Science and Engineering 191, no. 3 (July 9, 2018): 291–304. http://dx.doi.org/10.1080/00295639.2018.1468153.

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18

Simons, A., and C. Bauer. "Life cycle assessment of the European pressurized reactor and the influence of different fuel cycle strategies." Proceedings of the Institution of Mechanical Engineers, Part A: Journal of Power and Energy 226, no. 3 (March 14, 2012): 427–44. http://dx.doi.org/10.1177/0957650912440549.

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19

Jiang, Jin-yang, Ying-jun Yu, Hong-yan Chu, Wei Sun, and Yun Gao. "Influence of Concrete Properties on Molten Core-Concrete Interaction: A Simulation Study." Advances in Materials Science and Engineering 2016 (2016): 1–10. http://dx.doi.org/10.1155/2016/5380835.

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In a severe nuclear power plant accident, the molten core can be released into the reactor pit and interact with sacrificial concrete. In this paper, a simulation study is presented that aims to address the influence of sacrificial concrete properties on molten core-concrete interaction (MCCI). In particular, based on the MELCOR Code, the ferrosiliceous concrete used in European Pressurized Water Reactor (EPR) is taken into account with respect to the different ablation enthalpy and Fe2O3 and H2O contents. Results indicate that the concrete ablation rate as well as the hydrogen generation rate depends much on the concrete ablation enthalpy and Fe2O3 and H2O contents. In practice, the ablation enthalpy of sacrificial concrete is the higher the better, while the Fe2O3 and H2O content of sacrificial concrete is the lower the better.

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20

Qi, Shuang, Wenxin Xiang, Lixun Cai, Xiaokun Liu, Yonggang Wang, Fangmao Ning, Lei Qi, Weiwei Yu, and Jinhua Shi. "Study on Micro-Structure and Tensile Mechanical Properties of Dissimilar Metal Weld Joint Connecting Steam Generator Nozzle and Safe-End." Crystals 11, no. 12 (November 26, 2021): 1470. http://dx.doi.org/10.3390/cryst11121470.

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The safe-end of a steam generator (SG) nozzle dissimilar metal weld (DMW) for pressurized water reactors (PWRs) is the weakest point of failure which is crucial for the safe operation of a nuclear power station. Related to materials micro-structures, a uniaxial stress–strain relationship is the basic input parameter for nuclear power plant design, safety evaluation, and life management. In this paper, the micro-structure and tensile mechanical properties of a DMW of a European pressurized water reactor (EPR) were studied. Vickers hardness tester, optical microscope, and electron back scatter diffraction were used to analyze the micro-structure of the DMW joint. In addition, the residual strain of the DMW joint base material, heat-affected zone, weld metal, and fusion boundary region were studied. Based on digital image correlation (DIC) technology, tensile mechanical properties of the DMW joint were obtained. The results show that an accurate tensile stress–strain relationship of dissimilar metal welded joints can be obtained by using the DIC technique, the weld is the relatively weak link, and the residual strain is concentrated in the heat-affected zone. This study provides valuable engineering information regarding nuclear power plant design, in-service performance testing, and structural analysis and evaluation of welds containing defects.

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21

Fischer, U. "The European pressurized water reactor Result of the French–German cooperation of experienced NPP suppliers and operators." Nuclear Engineering and Design 187, no. 1 (January 1999): 15–23. http://dx.doi.org/10.1016/s0029-5493(98)00254-4.

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22

Eglin, Wolfgang, Ulrich Krugmann, Horst A. Weisshäupl, and Werner Scholtyssek. "Assessment of the Severe Accident Mitigation Features of the European Pressurized Water Reactor by Cooperative Research and Development." Nuclear Technology 126, no. 2 (May 1999): 143–52. http://dx.doi.org/10.13182/nt99-a2963.

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23

Lucas, D., D. Bestion, E. Bodèle, P. Coste, M. Scheuerer, F. D'Auria, D. Mazzini, et al. "An Overview of the Pressurized Thermal Shock Issue in the Context of the NURESIM Project." Science and Technology of Nuclear Installations 2009 (2009): 1–13. http://dx.doi.org/10.1155/2009/583259.

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Within the European Integrated Project NURESIM, the simulation of PTS is investigated. Some accident scenarios for Pressurized Water Reactors may cause Emergency Core Coolant injection into the cold leg leading to PTS situations. They imply the formation of temperature gradients in the thick vessel walls with consequent localized stresses and the potential for propagation of possible flaws present in the material. This paper focuses on two-phase conditions that are potentially at the origin of PTS. It summarizes recent advances in the understanding of the two-phase phenomena occurring within the geometric region of the nuclear reactor,that is, the cold leg and the downcomer, where the “PTS fluid-dynamics" is relevant. Available experimental data for validation of two-phase CFD simulation tools are reviewed and the capabilities of such tools to capture each basic phenomenon are discussed. Key conclusions show that several two-phase flow subphenomena are involved and can individually be simulated at least at a qualitative level, but the capability to simulate their interaction and the overall system performance is still limited. In the near term, one may envisage a simplified treatment of two-phase PTS transients by neglecting some effects which are not yet well controlled, leading to slightly conservative predictions.

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Nakata, Alexandre Ezzidi, Masanori Naitoh, and Chris Allison. "NEED OF A NEXT GENERATION SEVERE ACCIDENT CODE." JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA 21, no. 3 (November 12, 2019): 119. http://dx.doi.org/10.17146/tdm.2019.21.3.5630.

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Two international severe accident benchmark problems have been performed recently by using several existing parametric severe accident codes: The Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF) and the Benchmark of the In-Vessel Melt Retention (IVMR) Analysis of a VVER-1000 Nuclear Power Plant (NPP). The BSAF project was organized by the Nuclear Power Engineering Center (NUPEC) of the Institute of Applied Energy (IAE) in Japan for the three Boiling Water Reactors (BWRs) of the Fukushima NPP. The IVMR Project was organized by the Joint Research Center (JRC) of the European Commission (EC) in Holland (Europe) for a Pressurized Water Reactor (PWR). The obtained results of both projects have shown very large discrepancies between the used severe accident codes for both reactor types BWR and PWR. Consequently, the results for a real plant analysis by these integral codes, may not be correct after the beginning of core melt. Discrepancies of results of ex-vessel phenomena in the containment between the codes are in general larger. Therefore, there is a strong need for a reliable new generation mechanistic severe accident code which can simulate severe accident scenarios from an initiating event till containment failure with better accuracy not only for existing light water reactors but also for new generation IV reactor types. SAMPSON mechanistic ex-vessel modules coupled with SCDAPSIM and a new thermal-hydraulic module ASYST-ISA with particularly newly developed options for the reactor coolant system (RCS) and material properties applicable to new reactor deigns, is proposed as a best etimate new generation severe accident code for several reasons which are described in this paper.Keywords: Severe accident, SAMPSON, SCDAPSIM, ASYST-ISA, Steam explosion, Hydrogen detonation

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Bassi, C., and M. Marques. "Reliability Assessment of 2400 MWth Gas-Cooled Fast Reactor Natural Circulation Decay Heat Removal in Pressurized Situations." Science and Technology of Nuclear Installations 2008 (2008): 1–16. http://dx.doi.org/10.1155/2008/287376.

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As the 2400 MWth gas-cooled fast reactor concept makes use of passive safety features in combination with active safety systems, the question of natural circulation decay heat removal (NCDHR) reliability and performance assessment into the ongoing probabilistic safety assessment in support to the reactor design, named “probabilistic engineering assessment” (PEA), constitutes a challenge. Within the 5th Framework Program for Research and Development (FPRD) of the European Community, a methodology has been developed to evaluate the reliability of passive systems characterized by a moving fluid and whose operation is based on physical principles, such as the natural circulation. This reliability method for passive systems (RMPSs) is based on uncertainties propagation into thermal-hydraulic (T-H) calculations. The aim of this exercise is finally to determine the performance reliability of the DHR system operating in a “passive” mode, taking into account the uncertainties of parameters retained for thermal-hydraulical calculations performed with the CATHARE 2 code. According to the PEA preliminary results, exhibiting the weight of pressurized scenarios (i.e., with intact primary circuit boundary) for the core damage frequency (CDF), the RMPS exercise is first focusing on the NCDHR performance at these T-H conditions.

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26

Fischer, Manfred. "The severe accident mitigation concept and the design measures for core melt retention of the European Pressurized Reactor (EPR)." Nuclear Engineering and Design 230, no. 1-3 (May 2004): 169–80. http://dx.doi.org/10.1016/j.nucengdes.2003.11.034.

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27

Curca-Tivig, Florin. "Emergency Core Cooling Mode of the European Pressurized Water Reactor: The Evolution from Combined Injection to Cold-Leg Injection." Nuclear Technology 124, no. 1 (October 1998): 65–81. http://dx.doi.org/10.13182/nt98-a2909.

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28

du Toit, Maria Hendrina, and Vishana Vivian Naicker. "Neutronic design of homogeneous thorium/uranium fuel for 24 month fuel cycles in the European pressurized reactor using MCNP6." Nuclear Engineering and Design 337 (October 2018): 394–405. http://dx.doi.org/10.1016/j.nucengdes.2018.07.023.

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29

Gross, Maximilian, Faisal Sedeqi, Diana-María Amaya-Dueñas, Marc P. Heddrich, and S. Asif Ansar. "Characterisation of a 10-Layer SOC Stack Under Pressurised CO₂ Electrolysis Operation." ECS Meeting Abstracts MA2022-02, no. 49 (October 9, 2022): 1950. http://dx.doi.org/10.1149/ma2022-02491950mtgabs.

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One promising way of facing recent challenges to slow down the climate crisis or to reduce dependencies on fossil energy sources, e.g. natural gas, is using renewable methane and other e-fuels for storage and distribution via existing infrastructure. Solid oxide cell (SOC) reactors play an important role in the conversion of sustainable electric power into chemicals as they can be obtained from combined steam and CO2 co-electrolysis for syngas production. The pressurised electrolysis operation is a key factor for increasing the system efficiency of PtX-processes, including balance-of-plant (BoP) components, electrochemical reactors and high pressure downstream processes. In general, the yield of CO2 electrochemical reduction at atmospheric and pressurised conditions in high temperature co-electrolysis is still controversially discussed. Previously, several SOC short stacks were thoroughly analysed in pressurised steam- and co-electrolysis operation in a test-rig environment. These experimental results indicate marginal influence of pressure on the performance of electrolyte supported cells (ESC). In contrast, electrochemical impedance spectroscopy (EIS) suggests that pressurisation of pure CO2 electrolysis significantly reduces the fuel electrode impedance contribution, especially at lower temperatures around 700 °C [1,2]. This work aims to experimentally determine the kinetic behaviour of pure CO2 electrolysis by varying operating conditions like pressure, temperature, reactant conversion and feed gas composition. The investigation of kinetic parameters during these experiments could complement the formerly described research. Furthermore, the kinetic expressions can be used when studying co-electrolysis operation to identify the shares of: (i) the reverse water-gas-shift (rWGS) and (ii) the CO2 electrochemical reduction. Polarisation curves were dynamically recorded and different current densities were evaluated in steady-state operation. Additionally, EIS measurements were performed at open circuit voltage (OCV), as well as under different current densities. The kinetic parameters were estimated by curve-fitting analysis of the experimental results. The resulting expressions will be implemented in the in-house modelling framework, TEMPEST, based on [3,4] with the aim to increase the accuracy of modelling high-temperature CO2 electrolysis and co-electrolysis systems. [1] Riedel, M., Heddrich, M. P., & Friedrich, K. A. (2020). Experimental Analysis of the Co-Electrolysis Operation under Pressurized Conditions with a 10 Layer SOC Stack. Journal of The Electrochemical Society, 167(2), 024504, DOI: 10.1149/1945-7111/ab6820. [2] Riedel, M. (2020, October 20–23). Experimental analysis of SOE stacks under pressurized co- and CO2 electrolysis operation [Paper presentation]. 14th European SOFC & SOE Forum, Lucerne, Switzerland. [3] Tomberg, M., Santhanam, S., Heddrich, M. P., Ansar, A., & Friedrich, K. A. (2019). Transient Modelling of Solid Oxide Cell Modules and 50 kW Experimental Validation. ECS Transactions, 91(1), 2089, DOI: 10.1149/09101.2089ecst. [4] Srikanth, S., Heddrich, M. P., Gupta, S., & Friedrich, K. A. (2018). Transient reversiblesolid oxide cell reactor operation–Experimentally validated modeling and analysis. Applied Energy, 232, 473-488, DOI: 10.1016/j.apenergy.2018.09.186.

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Terzuoli, F., M. C. Galassi, D. Mazzini, and F. D'Auria. "CFD Code Validation against Stratified Air-Water Flow Experimental Data." Science and Technology of Nuclear Installations 2008 (2008): 1–7. http://dx.doi.org/10.1155/2008/434212.

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Pressurized thermal shock (PTS) modelling has been identified as one of the most important industrial needs related to nuclear reactor safety. A severe PTS scenario limiting the reactor pressure vessel (RPV) lifetime is the cold water emergency core cooling (ECC) injection into the cold leg during a loss of coolant accident (LOCA). Since it represents a big challenge for numerical simulations, this scenario was selected within the European Platform for Nuclear Reactor Simulations (NURESIM) Integrated Project as a reference two-phase problem for computational fluid dynamics (CFDs) code validation. This paper presents a CFD analysis of a stratified air-water flow experimental investigation performed at the Institut de Mécanique des Fluides de Toulouse in 1985, which shares some common physical features with the ECC injection in PWR cold leg. Numerical simulations have been carried out with two commercial codes (Fluent and Ansys CFX), and a research code (NEPTUNE CFD). The aim of this work, carried out at the University of Pisa within the NURESIM IP, is to validate the free surface flow model implemented in the codes against experimental data, and to perform code-to-code benchmarking. Obtained results suggest the relevance of three-dimensional effects and stress the importance of a suitable interface drag modelling.

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Ciurluini, Cristiano, Fabio Giannetti, Alessandro Del Nevo, and Gianfranco Caruso. "Study of the EU-DEMO WCLL Breeding Blanket Primary Cooling Circuits Thermal-Hydraulic Performances during Transients Belonging to LOFA Category." Energies 14, no. 6 (March 11, 2021): 1541. http://dx.doi.org/10.3390/en14061541.

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The Breeding Blanket (BB) is one of the key components of the European Demonstration (EU-DEMO) fusion reactor. Its main subsystems, the Breeder Zone (BZ) and the First Wall (FW), are cooled by two independent cooling circuits, called Primary Heat Transfer Systems (PHTS). Evaluating the BB PHTS performances in anticipated transient and accident conditions is a relevant issue for the design of these cooling systems. Within the framework of the EUROfusion Work Package Breeding Blanket, it was performed a thermal-hydraulic analysis of the PHTS during transient conditions belonging to the category of “Decrease in Coolant System Flow Rate”, by using Reactor Excursion Leak Analysis Program (RELAP5) Mod3.3. The BB, the PHTS circuits, the BZ Once Through Steam Generators and the FW Heat Exchangers were included in the study. Selected transients consist in partial and complete Loss of Flow Accident (LOFA) involving either the BZ or the FW PHTS Main Coolant Pumps (MCPs). The influence of the loss of off-site power, combined with the accident occurrence, was also investigated. The transient analysis was performed with the aim of design improvement. The current practice of a standard Pressurized Water Reactor (PWR) was adopted to propose and study actuation logics related to each accidental scenario. The appropriateness of the current PHTS design was demonstrated by simulation outcomes.

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Lamirand, Vincent, Pavel Frajtag, Daniel Godat, Oskari Pakari, Axel Laureau, Adolfo Rais, Mathieu Hursin, Grégory Hursin, Carlo Fiorina, and Andreas Pautz. "The COLIBRI experimental program in the CROCUS reactor: characterization of the fuel rods oscillator." EPJ Web of Conferences 225 (2020): 04020. http://dx.doi.org/10.1051/epjconf/202022504020.

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The present article presents the mechanical characterization of the fuel rods oscillator developed for the purposes of the COLIBRI experimental program in CROCUS. COLIBRI aims at investigating the radiation noise related to fuel vibrations. The main motivation is the increased amplitudes in the neutron noise distributions recorded in ex- and in-core detectors that have been observed in recent years in Siemens pre-Konvoi type of pressurized water reactors. Several potential explanations have been put forward, but no definitive conclusions could yet be drawn. Among others, changes in fuel assembly or pin vibration patterns, due to recent modifications of assembly structural designs, were pointed out as a possible cause. Computational dynamic tools are currently developed within the Horizon 2020 European project CORTEX, to help with understanding the additional noise amplitude. The COLIBRI program is used for their validation. An in-core device was designed, tested, and licensed between 2015 and 2019 for fuel rods oscillation in CROCUS, in successive steps from out-of-pile tests with dummy fuel rods to critical in-core tests. The characterization of its mechanical behavior is presented, in air and in water, and as a function of the load, for safety and experimental purposes. The device allows simultaneously oscillating up to 18 fuel rods. The maximum oscillation amplitude is 5 mm, while the maximum allowed frequency is 2 Hz, i.e. in the frequency range in which the induced neutron flux fluctuations are most pronounced in nuclear power plants.

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Chu, Hong Yan, Jin Yang Jiang, Wei Sun, and Ming Zhong Zhang. "Mechanical Properties and Damage Evolution of Siliceous Concrete Subjected to Elevated Temperatures." Key Engineering Materials 711 (September 2016): 488–95. http://dx.doi.org/10.4028/www.scientific.net/kem.711.488.

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Siliceous concrete (SC) is applied in European Pressurized Water Reactor that is a key component of the third generation nuclear power plant. This paper investigates the mechanical properties and damage evolution of SC (with and without polypropylene fibers) exposed to high temperatures. The mass loss, compressive strength, splitting tensile strength and spalling sensitivity of SC before and after being heated to 200, 400, 600, 800, and 1000 °C are investigated. The ultrasonic testing technique was used to assess the thermal damage, by evaluating the variations of the ultrasonic wave velocity (UWV) for different temperature levels. According to the available literature, a new relationship between damage and UWV was proposed to establish a damage evolution model of SC. The results indicated that: (a) specimens without polypropylene (PP) fibers suffered severe spalling in the range 380-400°C and 470-510°C, while no spalling took place in the specimens with PP fibers in the whole range 25-1000°C; (b) the damage evolution with and without polypropylene fibers was similar, and could adequately be described by means of a Weibull distribution model.

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Veres, Mihály, Ede Hertelendi, György Uchrin, Eszter Csaba, István Barnabás, Péter Ormai, Gábor Volent, and István Futó. "Concentration of Radiocarbon and Its Chemical Forms in Gaseous Effluents, Environmental Air, Nuclear Waste and Primary Water of a Pressurized Water Reactor Power Plant in Hungary." Radiocarbon 37, no. 2 (1995): 497–504. http://dx.doi.org/10.1017/s0033822200030976.

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We measured airborne releases of 14C from the Paks Pressurized Water Reactor (PWR) Nuclear Power Plant (NPP). Two continuous stack samplers collect 14C in 14CO2 and 14CnHm chemical forms. 14C activities were measured using two techniques; environmental air samples of lower activities were analyzed by proportional counting, stack samples were measured by liquid scintillation counting. 14C concentration of air in the stack varies between 80 and 200 Bqm−3. The average normalized yearly discharge rates for 1988–1993 were 0.74 TBqGW−1ey−1 for hydrocarbons and 0.06 TBqGW−1ey−1 for CO2. The discharge rate from Paks Nuclear Power Plant is about four times higher than the mean discharge value of a typical Western European PWR NPP. The higher 14C production may be apportioned to the higher level of nitrogen impurities in the primary coolant. Monitoring the long-term average excess from the NPP gave D14C = 3.5‰ for CO2 and D14C = 20‰ for hydrocarbons. We determined 14C activity concentration in the primary coolant to be ca. 4 kBq liter−1. The 14C activity concentrations of spent mixed bed ion exchange resins vary between 1.2 and 5.3 MBqkg−1 dry weight.

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Šadek, Siniša, Davor Grgić, and Zdenko Šimić. "Application of ASTEC, MELCOR, and MAAP Computer Codes for Thermal Hydraulic Analysis of a PWR Containment Equipped with the PCFV and PAR Systems." Science and Technology of Nuclear Installations 2017 (2017): 1–16. http://dx.doi.org/10.1155/2017/8431934.

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The integrity of the containment will be challenged during a severe accident due to pressurization caused by the accumulation of steam and other gases and possible ignition of hydrogen and carbon monoxide. Installation of a passive filtered venting system and passive autocatalytic recombiners allows control of the pressure, radioactive releases, and concentration of flammable gases. Thermal hydraulic analysis of the containment equipped with dedicated passive safety systems after a hypothetical station blackout event is performed for a two-loop pressurized water reactor NPP with three integral severe accident codes: ASTEC, MELCOR, and MAAP. MELCOR and MAAP are two major US codes for severe accident analyses, and the ASTEC code is the European code, joint property of Institut de Radioprotection et de Sûreté Nucléaire (IRSN, France) and Gesellschaft für Anlagen und Reaktorsicherheit (GRS, Germany). Codes’ overall characteristics, physics models, and the analysis results are compared herein. Despite considerable differences between the codes’ modelling features, the general trends of the NPP behaviour are found to be similar, although discrepancies related to simulation of the processes in the containment cavity are also observed and discussed in the paper.

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Cheymol, G., L. Maurin, L. Remy, V. Arounassalame, H. Maskrot, S. Rougeault, V. Dauvois, et al. "Tests under irradiation of optical fibers and cables devoted to corium monitoring in case of severe accident in a Nuclear Power Plant." EPJ Web of Conferences 225 (2020): 08006. http://dx.doi.org/10.1051/epjconf/202022508006.

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The DISCOMS project, which stands for “DIstributed Sensing for COrium Monitoring and Safety”, considers the potential of distributed sensing technologies, based on remote instrumentations and Optical Fiber Sensing cables embedded into the concrete floor under the reactor vessel, to monitor the status of this third barrier of confinement. This paper focuses on the selection and testing of singlemode (SM) optical fibers with limited RIA (Radiation Induced Attenuation) to be compliant with remote distributed instruments optical budgets, the ionizing radiation doses to sustain, and their reduction provided by the concrete basemat shielding. The tests aimed at exposing these fibers and the corresponding sensitive optical cables, to the irradiation doses expected during the normal operation of the reactor (up to 60 years for the European Pressurized Reactor), followed by a severe accident. Several gamma and mixed (neutron-gamma) irradiations were performed at CEA Saclay facilities: POSÉÏDON irradiator and ISIS reactor, up to a gamma cumulated dose of about 2 MGy and fast neutron fluence (E > 1 MeV) of 6 x 1015 n/cm2. The first gamma test permitted to assess the RIA at various optical wavelengths, and to select three radiation tolerant singlemode fibers (RIA < 5 dB/100 m, at 1550 nm operating wavelength). The second one was performed on voluminous strands of sensitive cables encapsulating the selected optical fibers, up to approximately the same accumulated dose, at two temperatures: 30°C and 80°C. A significant increase of the RIA, without any saturation tendency, appeared for fibers inserted into cables, correlated with the increase of the hydroxyl attenuation peak at 1380 nm. Molecular hydrogen generated by the radiolysis of compounds of the cable is at the origin of this phenomenon. A third gamma irradiation run permitted to measure the radiolytic hydrogen production yield of some compounds of a dedicated temperature cable sample. The efficiency of a carbon coating layer over the silica cladding, acting as a barrier against hydrogen diffusion, was also successfully confirmed. Finally, the efficiency of this carbon coating layer has also been tested under neutron irradiation, then qualified as a protection barrier against hydrogen diffusion in the optical fiber cores.

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Dolan, P., and A. Dolan. "A review of the European pressurised water reactor (EPR)." Nuclear Energy 43, no. 4 (2004): 205–8. http://dx.doi.org/10.1680/nuen.43.4.205.41193.

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Teichel, Holger, and Xavier Pouget-Abadie. "How the European Pressurised Water Reactor fulfils the utility requirements." Nuclear Engineering and Design 187, no. 1 (January 1999): 9–13. http://dx.doi.org/10.1016/s0029-5493(98)00253-2.

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Maillart, Hervé. "Design of the I&C for the European pressurised water reactor." Nuclear Engineering and Design 187, no. 1 (January 1999): 135–41. http://dx.doi.org/10.1016/s0029-5493(98)00260-x.

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40

Pomponi, Francesco, and Jim Hart. "The greenhouse gas emissions of nuclear energy – Life cycle assessment of a European pressurised reactor." Applied Energy 290 (May 2021): 116743. http://dx.doi.org/10.1016/j.apenergy.2021.116743.

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41

Boettcher, Agnieszka, and Zuzanna Marcinkowska. "Influence of accident tolerant fuel cladding material on the European pressurised reactor core neutronic characteristics." Annals of Nuclear Energy 119 (September 2018): 203–13. http://dx.doi.org/10.1016/j.anucene.2018.05.001.

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42

Addabbo, Carmelo, and Alessandro Annunziato. "The LOBI Integral System Test Facility Experimental Programme." Science and Technology of Nuclear Installations 2012 (2012): 1–16. http://dx.doi.org/10.1155/2012/238019.

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The LOBI project has been carried out in the framework of the European Commission Reactor Safety Research Programme in close collaboration with institutional and/or industrial research organizations of EC member countries. The primary objective of the research programme was the generation of an experimental data base for the assessment of the predictive capabilities of thermal-hydraulic system codes used in pressurised water reactor safety analysis. Within this context, experiments have been conducted in the LOBI integral system test facility designed, constructed, and operated (1979–1991) at the Ispra Site of the Joint Research Centre. This paper provides a historical perspective and summarizes major achievements of the research programme which has represented an effective approach to international collaboration in the field of reactor safety research and development. Emphasis is also placed on knowledge management aspects of the acquired experimental data base and on related online open access/retrieval user functionalities.

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Li, Jinfeng. "Computing Benchmark of Gadolinium-bearing Fuel Pins’ Depletion Skin Effect based on Deterministic and Monte Carlo Methods." Annals of Emerging Technologies in Computing 5, no. 1 (January 1, 2021): 1–12. http://dx.doi.org/10.33166/aetic.2021.01.001.

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Nuclear reactor core depletion and thermal-hydraulics coupling have long been calculation-intensive tasks challenging both nuclear industry development and academic research projects regarding computing budgets of memory and time. Albeit future evolution in smart computation hardware with artificial intelligence and quantum computing facilities embedded could continuously push the predictive modelling limit, the fundamental reactor physics model will still tip the balance in underpinning the prediction accuracy, as evidenced by a benchmark of two computational models in this work for characterising the depletion of highly self-shielded Gadolinium burnable poison-bearing fuel pins in assessing the British first European Pressurised Reactor’s start-up core performance. Specifically, a sub-group multi-annular-ring method is verified to efficiently represent the self-shielded skin effect, which addresses the deficiencies of classic equivalence models. The subgroup method is subsequently applied into a deterministic neutron transport code and a Monte Carlo stochastic code, respectively, for another benchmark. Resulting discrepancies in power peaking factors for the same assembly are less than 2% for the first fuel cycle, the agreement of which well demonstrates the validity of the proposed subgroup model. At the forefront of efforts to quantitatively understand the burnable poisons’ behaviour precisely for fuel optimisation (e.g., mitigating power peaking), this work could also be advantageously used for training purposes in boosting safety philosophy and public engagement in the roadmap for decarbonisation.

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Kronenberg, J. "Uncertainty and sensitivity analysis with respect to the hydrogen production determined by MELCOR 1.8.4 code during a severe accident." Kerntechnik 66, no. 4 (August 1, 2001): 171–76. http://dx.doi.org/10.1515/kern-2001-0075.

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Abstract For the selection and design of measures to scope with the hydrogen which will be produced in the course of a severe accident in LWRs a simple postulate – e. g. 75 % zircaloy oxidation of all fuel rod claddings – is not appropriate. In our opinion, realistic (best estimate) analyses with indication of the uncertainties are substantially more suitable. However, the problem thereby lies in the determination of these uncertainties. The current work presents a method, which determines the uncertainty band for the in-vessel hydrogen production calculated with the MELCOR program using a mathematical-statistic method (SUSA). For the exercise a “Total loss of the alternating current supply” for the planned European Pressurised Water Reactor (EPR) has been chosen as a representative scenario. Besides the actual uncertainties of the computational program concerned with such an analysis also the sensitive parameters can be identified.

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"Defects found in European Pressurized Water Reactor." Physics Today, 2008. http://dx.doi.org/10.1063/pt.5.022126.

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Vukolova, Angelina-Nataliya V., and Andrei A. Rusinkevich. "Radionuclide Composition of Airborne Discharges of European NPPS With WWER, Pressurized Water Reactor, and Boiling Water Reactor Facilities." Journal of Nuclear Engineering and Radiation Science 7, no. 1 (October 1, 2020). http://dx.doi.org/10.1115/1.4047044.

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Abstract The article presents the analysis of the data on radionuclide composition of airborne discharges of 52 European nuclear power plants (NPPs) with water–water energetic reactor facilities (WWER), pressurized water reactor facilities (PWR), and boiling water reactor facilities (BWR) under normal operation conditions. It contains lists of radionuclides, registered in discharges of researched NPPs, and gives estimation of contributions of radionuclides, forming the discharge, into total activity of discharge and into total effective dose, created by the discharge activity. It was determined that the maximal contribution into discharge activity of all researched NPPs make noble gases, tritium, and carbon-14, while the latter is the main dose-making radionuclide.

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"00/03305 Optimized conventional island design for the European pressurized water reactor (EPR)." Fuel and Energy Abstracts 41, no. 6 (November 2000): 372. http://dx.doi.org/10.1016/s0140-6701(00)94379-0.

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48

Abarca, A., R. Miró, G. Verdú, and J. A. Bermejo. "Analysis of Thermal-Hydraulic Fluctuations in Trillo NPP With CTF/PARCS v. 2.7 Coupled Code." Journal of Nuclear Engineering and Radiation Science 2, no. 2 (February 29, 2016). http://dx.doi.org/10.1115/1.4031660.

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The low-frequency noises are fluctuations in the neutron flux density, in the low-frequency range up to 4 Hz, which generate noise in the neutron instrumentation and could affect the limitation and protection system of the reactor core. Some European pressurized water reactors (PWRs) experienced the effect of low-frequency noise, opening a new research line for the verification of the neutron-kinetics/thermal-hydraulic coupled codes. A CTF/PARCS v. 2.7 simulation study to verify whether periodical fluctuations in the core inlet temperature could activate the core protection system has been done, obtaining the frequency spectrum of the power oscillation amplitudes.

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Katsuyama, Jinya, Kazuya Osakabe, Shumpei Uno, Yinsheng Li, and Shinobu Yoshimura. "Guideline on Probabilistic Fracture Mechanics Analysis for Japanese Reactor Pressure Vessels." Journal of Pressure Vessel Technology 142, no. 2 (February 24, 2020). http://dx.doi.org/10.1115/1.4045874.

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Abstract In Japan, to prevent nil-ductile fracture of reactor pressure vessels (RPVs) due to neutron irradiation embrittlement, deterministic fracture mechanics evaluation in accordance with the codes provided by the Japan Electric Association is performed for assessing the structural integrity of RPVs under pressurized thermal shock (PTS) events considering neutron irradiation embrittlement. In recent years, a structural integrity assessment methodology based on probabilistic fracture mechanics (PFM) has been introduced into the regulations in the United States and a few European countries. PFM is a rational methodology for evaluating the failure frequency of important pressure boundary components by considering the probabilistic distributions of various influence factors related to aged degradation due to the long-term operation. In Japan Atomic Energy Agency (JAEA), a PFM analysis code called PASCAL has been developed to evaluate the failure frequency of RPVs considering neutron irradiation embrittlement and PTS events. In addition, we have developed a guideline for structural integrity assessment of RPVs based on PFM to improve the applicability of PFM in Japan and enable persons who have knowledge on fracture mechanics to perform PFM analyses and evaluate through-wall cracking frequency (TWCF) of RPVs easily. The guideline consists of a main body, explanation, and several supplements. The technical basis for PFM analysis is provided, and the latest knowledge is included in the guideline. In this paper, an overview of the guideline and some typical analysis results obtained based on the guideline and the Japanese database related to PTS evaluation are presented.

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Straetz, Marcel, Joerg Starflinger, Rainer Mertz, and Dieter Brillert. "Cycle Calculations of a Small-Scale Heat Removal System With Supercritical CO2 as Working Fluid." Journal of Nuclear Engineering and Radiation Science 5, no. 1 (January 1, 2019). http://dx.doi.org/10.1115/1.4039884.

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In the case of an accident in a nuclear power plant with combined initiating events (loss of ultimate heat sink and station blackout), an additional heat removal system could transfer the decay heat from the core to an ultimate heat sink (UHS). One specific additional heat removal system, based upon a Brayton cycle with supercritical carbon dioxide (CO2) as working fluid, is currently investigated within the European Union-funded project “sCO2-HeRo” (supercritical carbon dioxide heat removal system). It serves as a self-launching, self-propelling, and self-sustaining decay heat removal system used in severe accident scenarios. Since this Brayton cycle produces more electric power than it consumes, the excess electric power can be used inside the power plant, e.g., for recharging batteries. A small-scale demonstrator is attached to the pressurized water reactor (PWR) glass model at Gesellschaft für Simulatorschulung (GfS), Essen, Germany. In order to design and build this small-scale model, cycle calculations are performed to determine the design parameters from which a layout can be derived.

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Journal articles on the topic 'European Pressurized Reactor'

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