Pressurized water nuclear reactors

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  • The employed method is easy to implement in practical applications and moreover, the SMC exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. Results show that the proposed controller for the load following operation is so effective that the xenon oscillations are kept bounded in the given region.

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  • Self-pressurization analysis of the natural circulation integral nuclear reactor through a new dynamic model is studied. Unlike conventional pressurized water reactors, this reactor type controls the system pressure using saturated coolant water in the steam dome at the top of the pressure vessel. Selfpressurization model is developed based on conservation of mass, volume, and energy by predicting the condensation that occurs in the steam dome and the flashing inside the chimney using the partial differential equation.

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  • The fatigue crack growth behavior of Type 347 (S347) and Type 347N (S347N) stainless steel was evaluated under the operating conditions of a pressurized water reactor (PWR). These two materials showed different fatigue crack growth rates (FCGRs) according to the changes in dissolved oxygen content and frequency. Under the simulated PWR conditions for normal operation, the FCGR of S347N was lower than that of S347 and insensitive to the changes in PWR water conditions.

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  • The aim of this work is to simulate the thermohydraulic consequences of a main steam line break and to compare the obtained results with Rossendorf Coolant Mixing Model (ROCOM) 1.1 experimental results. The objective is to utilize data from steady-state mixing experiments and computational fluid dynamics (CFD) calculations to determine the flow distribution and the effect of thermal mixing phenomena in the primary loops for the improvement of normal operation conditions and structural integrity assessment of pressurized water reactors.

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  • Monte Carlo calculation method can be used for resolving particle transport in matter, and particularly the transport of neutrons in the environment of the reactor core. The method has become more efficient because of high accuracy of updated nuclear data and fast development of advanced super-computing system. In this work, we would like to present calculations for kinematic characteristics of neutron transport in a typical configuration of the pressurized water reactor (PWR) fuel assembly based on the Monte-Carlo simulation method.

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  • The management of hydrogen safety and prevention of overpressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity.

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  • This article discusses the evolution of the collective dose for several types of reactors, mainly based on publications from the NEA and the IAEA.

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  • This paper analyses the implications of the SSR and consequential increase of the frequency on the nuclear power plant safety. The Simulink MATLAB® model of the steam turbine and governor system and RELAP5 computer code of the pressurized water reactor are used in the analysis.

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  • Low-cycle fatigue (LCF) tests were performed for Alloy 690 and 316 SS in a simulated pressurized water reactor (PWR) environment. Alloy 690 showed about twice longer LCF life than 316 SS at the test condition of 0.4% amplitude at strain rate of 0.004%/s. Observation of the oxide layers formed on the fatigue crack surface showed that Cr and Ni rich oxide was formed for Alloy 690, while Fe and Cr rich oxide for 316 SS as an inner layer.

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  • The present simulation results obtained using the two different outflow boundary conditions were discussed through a comparison with the predictions using a simple nonflashing model neglecting the effects of phase change. In addition, the applicability of each of the non-flashing water discharge and saturated water flashing models for the confirmatory assessments of new SG designs was examined.

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  • In the PWR pressure water reactor (PWR), stainless steel is used in many important parts in both primary and secondary water circuits. There are not enough necessary condition to experiment in extremly conditons of nuclear reactor, such as high temperature, high pressure in radiation environment in Vietnam.

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  • Design Philosophies: Radiological release to the environment is prevented by maintaining R/B pressure negative(-63Pa). Radioactive substance in PCV are removed and captured in the filter device.R/B Closed Cooling Water System (RCW) / Reactor Service Water System (RSW) How are the many Heat exchangers cooled down?

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  • This paper presents the verification and validation (V&V) of the STREAM/RAST-K 2.0 code system for a pressurized water reactor (PWR) analysis. A lattice physics code STREAM and a nodal diffusion code RAST-K 2.0 have been developed by a computational reactor physics and experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for an accurate two-step PWR analysis. The calculation modules of each code were already verified against various benchmark problems, whereas this paper focuses on the V&V of linked code system.

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  • This study was found by simulating public information on Lightbridge’s fuel design for pressurized water reactors. This study explores the temperature profile and maximum stress within a simple (first generation design) hypothetical nuclear explosive device of four unique scenarios over time. Analyzing the transient development of both the temperature profile and maximum stress not only establishes a technical limit on the 238Pu content, but also establishes a time limit for which each scenario would be useable.

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  • In the SG (steam generator) of PWR (pressurized water reactor) for a nuclear plant, hundreds of Ushaped tubes are used for the heat exchanger system. They interact with primary pressurized cooling water flow, generating flow-induced vibration in the secondary flow region. A simplified U-tube model is proposed in this study to apply for experiment and its counterpart computation. Using the commercial code, ANSYS-CFX, we first verified the Moody chart, comparing the straight pipe theory with the results derived from CFD (computational fluid dynamics) analysis.

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  • This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on UO2-M5® fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities.

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  • In the LSTF test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on the upper core plate in the upper plenum. Results of the uncertainty analysis with RELAP5/MOD3.3 code clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

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  • The VVER-1200/V491 was a selected candidate for the Ninh Thuan I Nuclear Power Plant. However, in the Feasibility Study Safety Analysis Report (FS-SAR) of the VVER-1200/V491, the core loading pattern of this reactor was not provided. To assess the safety features of the VVER1200/V491, finding the core loading patterns and verifying their safety characteristics are necessary.

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  • The objective of this work is to experimentally investigate the effect of the oxygen potential on the fuel and FP chemical behaviour in conditions representative of a severe accident. More specifically, the speciation of Cs, Mo and Ba is investigated.

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  • CHAPTER 56 NUCLEAR POWER William Kerr Department of Nuclear Engineering University of Michigan Ann Arbor, Michigan 56.1 HISTORICAL PERSPECTIVE 56.1.1 The Birth of Nuclear Energy 56.1.2 Military Propulsion Units 56.1.3 Early Enthusiasm for Nuclear Power 56.1.4 U.S. Development of Nuclear Power CURRENT POWER REACTORS, AND FUTURE PROJECTIONS 56.2. 1 Light- Water-Moderated Enriched-Uranium-Fueled Reactor 56.2.2 Gas-Cooled Reactor 56.2.3 Heavy-Water-Moderated Natural-Uranium-Fueled Reactor 56.2.4 Liquid-Metal-Cooled Fast Breeder Reactor 56.2.

    pdf24p hadalabo 29-09-2010 55 6   Download



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