Ab initio calculations and iodine kinetic modeling in the reactor coolant system of a pressurized water reactor in case of severe nuclear accident

2012 ◽  
Vol 990 ◽  
pp. 194-208 ◽  
Author(s):  
Bertrand Xerri ◽  
Sébastien Canneaux ◽  
Florent Louis ◽  
Julien Trincal ◽  
Frédéric Cousin ◽  
...  
Keyword(s):  
Ab Initio Calculations ◽  
Ab Initio ◽  
Kinetic Modeling ◽  
Nuclear Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Reactor Coolant ◽  
Coolant System ◽  
Severe Nuclear Accident

Study of RuO4 decomposition in dry and moist air

2007 ◽  
Vol 95 (11) ◽  
Author(s):  
Christian Mun ◽  
L. Cantrel ◽  
Charles Madic
Keyword(s):  
Nuclear Accident ◽  
Pressurized Water Reactor ◽  
Moist Air ◽  
Pressurized Water ◽  
Ruthenium Tetroxide ◽  
Water Reactor ◽  
The Stability ◽  
Radionuclide Release ◽  
Severe Nuclear Accident

During a hypothetical severe nuclear accident on a pressurized water reactor (PWR), it is of primary importance to assess potential radionuclide release into the environment, and thus to better understand the volatile ruthenium tetroxide stability, in the containment building, due to its high radiotoxicity. The stability of RuO

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.

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