Analysis of an ultrasonic level device for in-core Pressurized Water Reactor coolant detection

1981 ◽  
Author(s):  
K. R. Johnson
Keyword(s):  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Reactor Coolant

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

scholarly journals Behavior of Zinc Incorporation into Fuel Deposits in Borated and Lithiated Water with Dissolved Zinc

Materials ◽  
2020 ◽  
Vol 13 (19) ◽  
pp. 4317
Author(s):  
Do Haeng Hur ◽  
Kyeong-Su Kim ◽  
Hee-Sang Shim ◽  
Jinsoo Choi ◽  
Kyu Min Song
Keyword(s):  
Nuclear Fuel ◽  
Lattice Constant ◽  
Iron Content ◽  
Pressurized Water Reactor ◽  
Fuel Cladding ◽  
Pressurized Water ◽  
Water Reactor ◽  
Reactor Coolant ◽  
Zinc Addition ◽  
Nuclear Fuel Cladding

The objective of this study was to investigate the behavior of zinc incorporation into newly forming fuel deposits and pre-formed deposits in a simulated pressurized water reactor coolant including 1000 ppm of boron and 2 ppm of lithium at 328 °C. Zinc was incorporated into fuel deposits that were being newly nucleated and grown on nuclear fuel cladding tubes in a zinc-containing coolant. The zinc incorporation resulted in a decrease in the lattice constant of the deposits, which was attributed to the decrease in larger iron content and the corresponding incorporation of smaller zinc in the deposits. However, zinc incorporation was not found, even after the fuel deposits pre-formed before zinc addition were subsequently exposed to the 60 ppb of zinc coolant for 500 h.

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