Fracture Toughness of Submerged-Arc Weld Samples from a Decommissioned Magnox Reactor Pressure Vessel: Initial Results

2008 ◽  
pp. 295-295-11 ◽  
Author(s):  
PJE Bischler ◽  
MR Wootton ◽  
CJ Bolton
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Submerged Arc ◽  
Initial Results ◽  
Reactor Pressure

The method for calculating the probability of brittle destruction of NPP equipment in different operating modes with postulated defectiveness

2021 ◽  
Vol 14 (1) ◽  
pp. 34-39
Author(s):  
D. A. Kuzmin ◽  
A. Yu. Kuz’michevskiy
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Pressure Vessel ◽  
Practical Significance ◽  
Reactor Unit ◽  
Operating Modes ◽  
Brittle Destruction ◽  
Reactor Pressure

The destruction of equipment metal by a brittle fracture mechanism is a probabilistic event at nuclear power plants (NPP). The calculation for resistance to brittle destruction is performed for NPP equipment exposed to neutron irradiation; for example, for a reactor plant such as a water-water energetic reactor (WWER), this is a reactor pressure vessel. The destruction of the reactor pressure vessel leads to a beyond design-basis accident, therefore, the determination of the probability of brittle destruction is an important task. The research method is probabilistic analysis of brittle destruction, which takes into account statistical data on residual defectiveness of equipment, experimental results of equipment fracture toughness and load for the main operating modes of NPP equipment. Residual defectiveness (a set of remaining defects in the equipment material that were not detected by non-destructive testing methods after manufacturing (operation), control and repair of the detected defects) is the most important characteristic of the equipment material that affects its strength and service life. A missed defect of a considerable size admitted into operation can reduce the bearing capacity and reduce the time of safe operation from the nominal design value down to zero; therefore, any forecast of the structure reliability without taking into account residual defectiveness will be incorrect. The application of the developed method is demonstrated on the example of an NPP reactor pressure vessel with a WWER-1000 reactor unit when using the maximum allowable operating loads, in the absence of load dispersion in different operating modes, and taking into account the actual values of the distributions of fracture toughness and residual defectiveness. The practical significance of the developed method lies in the possibility of obtaining values of the actual probability of destruction of NPP equipment in order to determine the reliability of equipment operation, as well as possible reliability margins for their subsequent optimization.

Fracture Toughness Characterization of Midland Beltline Weld After High Dose of Irradiation

2006 ◽  
Author(s):  
Mikhail A. Sokolov ◽  
Randy K. Nanstad
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Neutron Fluence ◽  
National Laboratory ◽  
Reactor Pressure Vessel ◽  
High Dose ◽  
Pressure Vessel Steel ◽  
Vessel Steel ◽  
Oak Ridge ◽  
Reactor Pressure

The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory includes a task to investigate the shape of the fracture toughness master curve for reactor pressure vessel steel highly embrittled as a consequence of irradiation exposure, and to examine the ability of the Charpy 41-J shift to predict the fracture toughness shift. As part of this task, a low upper-shelf WF-70 weld obtained from the beltline region of the Midland Unit 1 reactor pressure vessel was characterized in terms of static initiation and Charpy impact toughness in the unirradiated and irradiated conditions. Irradiation of this weld was performed at the University of Michigan Ford Reactor at 288°C to neutron fluence of 3.4×1019 neutron/cm2 in the HSSI irradiation-anneal-reirradiation facility. This reusable facility allowed the irradiation of either virgin or previously irradiated material in a well-controlled temperature regime, including the ability to perform in-situ annealing. This was the last capsule irradiated in this facility before reactor shut down. Thus, the Midland beltline weld was irradiated within the HSSI Program to three fluences — 0.5×1019; 1.0×1019; and 3.4×1019 neutron/cm2. It was anticipated that it would provide an opportunity to address fracture toughness curve shape and Charpy 41-J shift compatibility issues at different levels of embrittlement, including the highest dose considered to be in the range of the current end of life fluence. It was found that the Charpy 41-J shift practically saturated after neutron fluence of 1.0×1019 neutron/cm2. The transition fracture toughness shift after 3.4×1019 neutron/cm2 was only slightly higher than that after 1.0×1019 neutron/cm2. In all cases, transition fracture toughness shifts were lower than predicted by the Regulatory Guide 1.99, Rev. 2 equation.

Investigation of Transition Fracture Toughness Variation within the Thickness of Reactor Pressure Vessel Forgings

2010 ◽  
Vol 7 (1) ◽  
pp. 102431 ◽  
Author(s):  
Enrico Lucon ◽  
Ann Leenaers ◽  
Willy Vandermeulen ◽  
Marc Scibetta ◽  
S. Kalluri ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Toughness Variation ◽  
Reactor Pressure
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