Fracture Mechanics Evaluation of the Integrity of an Inlet Nozzle of a Pressurized Water Reactor Vessel Following a Postulated Loss of Coolant

2009 ◽  
pp. 240-240-13 ◽  
Author(s):  
CB Buchalet
Keyword(s):  
Fracture Mechanics ◽  
Reactor Vessel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Loss Of Coolant

Application of the Failure Assessment Diagram to the Evaluation of Pressure-Temperature Limits for a Pressurized Water Reactor

1985 ◽  
Vol 107 (2) ◽  
pp. 192-196 ◽  
Author(s):  
K. K. Yoon ◽  
J. M. Bloom ◽  
W. A. Pavinich ◽  
H. W. Slager
Keyword(s):  
Fracture Mechanics ◽  
Reactor Vessel ◽  
Normal Operation ◽  
Coolant Temperature ◽  
Operating Pressure ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Failure Assessment Diagram ◽  
Failure Assessment

The failure assessment diagram approach, an elastic-plastic fracture mechanics procedure based on the J-integral concept, was used in the evaluation of pressure-temperature (P-T) limits for the beltline region of the vessel of a pressurized water reactor. The main objective of this paper is to illustrate the application of an alternate fracture mechanics method for the evaluation of pressure-temperature limits, as allowed by the Code of the Federal Regulation 10 CFR 50, Appendix G. The evaluation of P-T limits for the beltline region of a pressurized water reactor was based on the following assumptions: • ASME Pressure Vessel and Piping Code, Section III, Appendix G reference flaw • End-of-life fluence level in the beltline region • Longitudinal flaw in the beltline weld • J-resistance material toughness curves obtained from the U.S. Nuclear Regulatory Commission’s heavy section steel technology (HSST) program • Other material properties obtained from the Babcock & Wilcox Integrated Reactor Vessel Material Surveillance Program The maximum allowable pressure levels were calculated at 33 time points along the given reactor bulk coolant temperature history representing the normal operation of a pressurized water reactor. The results of the calculations showed that adequate margins of safety on operating pressure for the critical weld in the beltline of the pressurized water reactor vessel are assured.

Analysis of loss of coolant accident without ECCS and DHRS in an integral pressurized water reactor using RELAP/SCDAPSIM

2021 ◽  
Vol 134 ◽  
pp. 103648
Author(s):  
Katarzyna Skolik ◽  
Chris Allison ◽  
Judith Hohorst ◽  
Mateusz Malicki ◽  
Marina Perez-Ferragut ◽  
...  
Keyword(s):  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant
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