Design Process for an Advanced Reactor Coolant Pump for a 1400 MW Nuclear Power Plant

2005 ◽  
Author(s):  
Claus Knierim ◽  
Sven Baumgarten ◽  
Jochen Fritz ◽  
Michael T. Coon
Keyword(s):  
Fluid Dynamics ◽  
Hydraulic System ◽  
Nuclear Power ◽  
Pressurized Water Reactor ◽  
Power Station ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Computational Fluid Dynamics Cfd ◽  
Specific Speed

As part of the planning activities for a new 1400 MW nuclear power station with a pressurized water reactor, a new hydraulic system had to be designed for the reactor coolant pumps (RCPs). Starting from the design principles and the main dimensions of an existing pump a new diffuser and impeller had to be designed for the specified requirements which provided a specific speed of almost ns = 100 rpm (≈ 5100 in US units). The authors describe how impeller and diffuser of the hydraulic system were gradually optimized with the aid of computational fluid dynamics (CFD). The system had to meet demanding requirements, thus it was decided to build a model pump (on a scale ≈ 1:2) to demonstrate that the new pump would satisfy the specified duty parameters. Based on the spectra of tests performed on the model pump the authors discuss the resulting pump characteristics (Q, H, η, NPSH).

Experimental Research on the Fluid Induced Forces of Clearance Flow in Canned Motor Reactor Coolant Pump

2018 ◽  
Author(s):  
Rui Xu ◽  
Yaoyu Hu ◽  
Yun Long ◽  
Junlian Yin ◽  
Dezhong Wang
Keyword(s):  
Large Mass ◽  
Stiffness Coefficient ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Various Boundary Conditions ◽  
Clearance Flow ◽  
Mass Coefficient ◽  
Canned Motor

Reactor coolant pump is one of the key equipment of the coolant loop in a pressurized water reactor system. Its safety relies on the characteristics of the rotordynamic system. For a canned motor reactor coolant pump, the liquid coolant fills up the clearance between the metal shields of the rotor and stator inside the canned motor, forming a clearance flow. The fluid induced forces of the clearance flow in canned motor reactor coolant pump and their effects on the rotordynamic characteristics of the pump are experimentally analyzed in this work. A vertical experiment rig has been established for the purpose of measuring the fluid induced forces of the clearance. Fluid induced forces of clearance flow with various whirl frequencies and various boundary conditions are obtained through the experiment. Results show that clearance flow brings large mass coefficient into the rotordynamic system and the direct stiffness coefficient is negative under the normal operating condition. The rotordynamic stability of canned motor reactor coolant pump does not deteriorate despite the existence of significant cross-coupled stiffness coefficient from the fluid induced forces of the clearance flow.

scholarly journals Visualization and Finding Ways to Intensify the Processes of Non-isothermal Mixing Flows

2018 ◽  
Vol 3 (3) ◽  
pp. 240
Author(s):  
Sataev A.A. ◽  
Duntsev A.V.
Keyword(s):  
Fluid Dynamics ◽  
Cold Water ◽  
Density Structure ◽  
Pressurized Water Reactor ◽  
Thermal Reliability ◽  
Reactor Plant ◽  
Pressurized Water ◽  
Real Geometry ◽  
Computational Fluid Dynamics Cfd ◽  
Isothermal Mixing

Simulation of mixing flows of different temperature, density structure has important implications for the assessment of thermal reliability of reactor plants, thermo-cyclic pulsations, and safety analysis. To study the mixing model was used for the mixing, which was visualized by using imaging methods. The injection of cold water into the hot volume was examined, which simulates the flow of the coolant in the pressurized-water reactor. The obtained results have given the basis for further analysis of non-isothermal mixing flows. However, the model is still far from the real geometry of the reactor plant. The construction of a reactor reduced model with a simulation of one loop of a coolant flow with low settings has been developed for a more detailed study of the processes of non-isothermal mixing flows has planned. In the future, these data will be used in the programs of computational fluid dynamics (CFD). 

The Application of Concurrent Installation of the RCS Cross Over Leg and the Reactor Vessel Internals

2010 ◽  
Author(s):  
K.-W. Park ◽  
J.-H. Bae ◽  
S.-H. Park
Keyword(s):  
Feasibility Study ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Reactor Vessel ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Study Results ◽  
Under Construction

The reactor vessel internals (RVI) of a pressurized water reactor (PWR) must be installed precisely in the reactor vessel (RV) according to the requirements for levelness, orientation and vertical alignments for its proper functions and structural integrity. For the precise installation, deformation of the RV should be controlled during the RVI installation. Traditionally, the RVI has been installed in the RV after the completion of welding work for large bore pipings in the reactor coolant system (RCS). To reduce installation time, the concurrent installation of the RVI and RCS pipings is investigated. This paper describes the feasibility study on the concurrent installation including the Finite Element Method (FEM) analyses of the RV deformation due to the welding and heat treatment of the pipings. Based on the feasibility study results, the optimum schedule of the RVI installation in parallel with the installation of the cross-over leg pipings (reactor coolant pump inlet pipings) and confirmation measurement locations are developed. Thereby the concurrent installation will be applied to the nuclear power plants under construction in Korea, and it is expected to reduce installation period of 2 months compared to the traditional sequential installation method.

A Study of Some Key Parameters Used in the Environmental Fatigue Evaluation of a Hot Leg Surge Nozzle

2009 ◽  
Author(s):  
Tim F. Wiley ◽  
Tim J. Pournaras ◽  
Chris T. Kupper ◽  
Mark A. Gray ◽  
Seth A. Swamy
Keyword(s):  
Nuclear Power ◽  
Sensor Data ◽  
Maximum Temperature ◽  
Pressurized Water Reactor ◽  
Maximum Temperature Difference ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Surge Line ◽  
Fatigue Evaluation ◽  
Made In

When considering environmentally assisted fatigue (EAF) in the fatigue evaluation of nuclear power plant components, some assumptions made pertaining to plant operation in the design basis fatigue analyses have to be re-evaluated to accommodate potential increase in fatigue usage factors resulting from environmental effects. The surge line was identified in NUREG/CR-6260 [1] to be a representative component for the evaluation of EAF for Pressurized Water Reactor (PWR) plants. For some PWR plants, the hot leg surge nozzle is one of the components evaluated for environmentally assisted fatigue. The hot leg surge nozzle was chosen for this study because the results of the fatigue evaluation are highly dependent on several key parameters, such as maximum temperature difference between the pressurizer and hot leg piping during heatups and cooldowns, the amount of temperature sensor data available along the surge line, availability of thermal event cycle counting, and the frequency and timing of reactor coolant pump starts and stops during heatups and cooldowns. This paper assesses the impacts of the assumptions made in these key parameters on the environmental fatigue evaluation results for a typical hot leg surge nozzle.

Modeling of Steam Explosion in Partially Flooded PWR Reactor Cavity

2005 ◽  
Author(s):  
Bosˇtjan Koncˇar ◽  
Matjazˇ Leskovar ◽  
Leon Cizelj
Keyword(s):  
Steam Explosion ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Fission Products ◽  
General Purpose ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Fit For Purpose ◽  
Computational Fluid Dynamics Cfd ◽  
Reactor Accidents

When the hot molten core comes into contact with the water in the reactor cavity a steam explosion can occur. The steam explosion might be triggered during some scenarios of severe nuclear reactor accidents, when extremely hot molten nuclear fuel interacts with the coolant water. A highly energetic steam explosion in a nuclear power plant could cause the containment failure and the release of radioactive fission products to the environment. The purpose of the performed analysis is to provide a first estimation of the expected pressure loadings on the typical PWR cavity structures during a steam explosion. To achieve this, the fit-for-purpose steam explosion model is proposed, followed by a Computational Fluid Dynamics (CFD) analysis. In the present work two steam explosion scenarios in the partially flooded Pressurized Water Reactor (PWR) cavity were simulated with the general purpose code CFX-5 [1] to estimate pressure loadings on cavity walls.

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