Pressurized Water Reactor Coolant Pump Locked Rotor Analysis Methodology Using RETRAN-02 and FREY

1999 ◽  
Vol 127 (3) ◽  
pp. 382-388 ◽  
Author(s):  
Lainsu Kao ◽  
Ping-Hue Huang
Keyword(s):  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Water Reactor ◽  
Reactor Coolant ◽  
Analysis Methodology

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.

Thermal Fatigue of Pressurized Water Reactor Coolant System Loop Drain Line Down Horizontal Elbow

2020 ◽  
Author(s):  
Yue Zou ◽  
Brian Derreberry
Keyword(s):  
Thermal Fatigue ◽  
Operating Experience ◽  
Fatigue Cracking ◽  
Screening Methods ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Thermal Transients ◽  
Reactor Coolant ◽  
Coolant System

Abstract Thermal cycling induced fatigue is widely recognized as one of the major contributors to the damage of nuclear plant piping systems, especially at locations where turbulent mixing of flows with different temperature occurs. Thermal fatigue caused by swirl penetration interaction with normally stagnant water layers has been identified as a mechanism that can lead to cracking in dead-ended branch lines attached to pressurized water reactor (PWR) primary coolant system. EPRI has developed screening methods, derived from extensive testing and analysis, to determine which lines are potentially affected as well as evaluation methods to perform evaluations of this thermal fatigue mechanism for the U.S. PWR plants. However, recent industry operating experience (OE) indicate that the model used to predict thermal fatigue due to swirl penetration is not fully understood. There are limitations with the EPRI generic evaluation. In addition, cumulative effects from various thermal transients such as the reactor coolant system (RCS) sampling and excess letdown may also contribute to the failure of RCS branch lines. In this paper, we report direct OE from one of our PWR units where thermal fatigue cracking is observed at the RCS loop drain line close to the welded region of the elbow. A conservative analytical approach that takes into account the influence of thermal stratification, in accordance with ASME Section III Class 1 piping stress formula, is also proposed to evaluate the severity of fatigue damage to the RCS drain line, as a result of various transients. Finally, recommendations are made for future operation and inspection based on results of the evaluation.

Monitoring Pressurized Water Reactor Transients Using Reactor Coolant Pumps

1991 ◽  
Vol 109 (4) ◽  
pp. 325-340 ◽  
Author(s):  
James P. Adams ◽  
Glenn E. McCreery ◽  
Jong H. Kim
Keyword(s):  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Reactor Coolant

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).

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