Fracture toughness of weld metal samples removed from a decommissioned Magnox reactor pressure vessel

2002 ◽  
Vol 79 (8-10) ◽  
pp. 685-692 ◽  
Author(s):  
C.J Bolton ◽  
P.J.E Bischler ◽  
M.R Wootton ◽  
R Moskovic ◽  
J.R Morri ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Weld Metal ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Reactor Pressure

Applicability of Miniature C(T) Specimen to Fracture Toughness Evaluation for the Irradiated Japanese Reactor Pressure Vessel Steel

2017 ◽  
Author(s):  
Takuji Sugihara ◽  
Takatoshi Hirota ◽  
Hiroyuki Sakamoto ◽  
Kentaro Yoshimoto ◽  
Kazuya Tsutsumi ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Weld Metal ◽  
Pressure Vessel ◽  
Charpy Impact ◽  
Reactor Pressure Vessel ◽  
Base Metals ◽  
Toughness Evaluation ◽  
Fracture Toughness Evaluation ◽  
Reactor Pressure ◽  
Mc Method

Fracture toughness evaluations for irradiated reactor pressure vessel (RPV) steels are essential in the structural integrity assessment of RPVs. In Japanese pressurized water reactor (PWR) plants, fracture toughness tests are conducted for irradiated RPV steels through the surveillance tests and fracture toughness data are obtained. Lately, the Master Curve (MC) approach has become the main stream in fracture toughness evaluation. However, there can be the case that the number of fracture toughness data is not enough for the MC method in some Japanese PWR plants because of limited numbers of fracture toughness specimens contained in the surveillance capsules. On the other hand, for the Japanese PWR plants, a surveillance capsule generally contains a lot of Charpy impact specimens which miniature C(T) (Mini-C(T)) specimens with a size of 4×10×9.6mm can be taken from. Therefore, it is planned that additional fracture toughness tests are performed using Mini-C(T) specimens after the Charpy impact tests to obtain sufficient fracture toughness data for the MC method. Applicability of the Mini-C(T) specimen to the MC evaluation has been studied in a series of international round robin test programs coordinated by Central Research Institute of Electric Power Industry (CRIEPI). In these programs and the related studies, it was demonstrated that the reference temperature (To) can be determined by the Mini-C(T) specimens without any specific difficulties for the unirradiated RPV base metals. In addition, CRIEPI has recently reported on the basis of their studies that the fracture toughness tests could be successfully performed on the typical unirradiated RPV weld metal and the valid To can be determined with the data obtained from the weld metal as well as base metals. However, few papers reported applicability of the Mini-C(T) specimen to the MC evaluation for irradiated RPV steels. In this study, fracture toughness tests using Mini-C(T) specimens were conducted on the irradiated Japanese Industrial Standards (JIS) SFVQ1A steel (equivalent to ASME A508 Cl.3 steel). The Mini-C(T) specimens were machined out from some broken halves of Charpy impact specimens used in a surveillance test of an actual Japanese PWR plant by a wire cut electric spark machine followed by fatigue precracking. After the fracture toughness tests, the evaluation was performed on the obtained fracture toughness data according to the MC method. The effect of specimen size on To was studied and applicability of the Mini-C(T) specimen was discussed by comparing the existing results of fracture toughness tests using the 1/2T-C(T) specimens conducted in the surveillance test. In addition, the issues to obtain valid To for irradiated materials were discussed.

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.

Investigation of the Beltline Welding Seam and Base Metal of the Greifswald WWER-440 Unit 1 Reactor Pressure Vessel

2009 ◽  
Author(s):  
Jan Schuhknecht ◽  
Hans-Werner Viehrig ◽  
Udo Rindelhardt
Keyword(s):  
Weld Metal ◽  
Base Metal ◽  
Pressure Vessel ◽  
Master Curve ◽  
Reactor Pressure Vessel ◽  
Reference Temperature ◽  
Metal Ring ◽  
Welding Seam ◽  
Root Region ◽  
Reactor Pressure

The investigation of reactor pressure vessel (RPV) materials from decommissioned NPPs offers the unique opportunity to scrutinize the irradiation behaviour under real conditions. Material samples taken from the RPV wall enable a comprehensive material characterisation. The paper describes the investigation of trepans taken from the decommissioned WWER-440 first generation RPVs of the Greifswald NPP. Those RPVs represent different material conditions such as irradiated (I), irradiated and recovery annealed (IA) and irradiated, recovery annealed and re-irradiated (IAI). The working program is focussed on the characterisation of the RPV steels (base and weld metal) through the RPV wall. The key part of the testing is aimed at the determination of the reference temperature T0 following the ASTM Test Standard E1921-05 to determine the fracture toughness of the RPV steel in different thickness locations. In a first step the trepans taken from the RPV Greifswald Unit 1 containing the X-butt multilayer submerged welding seam and from base metal ring 0.3.1 both located in the beltline region were investigated. Unit 1 represents the IAI condition. It is shown that the Master Curve approach as adopted in ASTM E1921 is applicable to the investigated original WWER-440 weld metal. The evaluated T0 varies through the thickness of the welding seam. The lowest T0 value was measured in the root region of the welding seam representing a uniform fine grain ferritic structure. Beyond the welding root T0 shows a wavelike behaviour. The highest T0 of the weld seam was not measured at the inner wall surface. This is important for the assessment of ductile-to-brittle temperatures measured on sub size Charpy specimens made of weld metal compact samples removed from the inner RPV wall. Our findings imply that these samples do not represent the most conservative condition. Nevertheless, the Charpy transition temperature TT41J estimated with results of sub size specimens after the recovery annealing was confirmed by the testing of standard Charpy V-notch specimens. The evaluated Charpy-V TT41J shows a better accordance with the irradiation fluence along the wall thickness than the Master Curve reference temperature T0. The evaluated T0 from the trepan of base metal ring 0.3.1 varies through the thickness of the RPV wall. T0 increases from −120°C at the inner surface to −104°C at a distance of 33 mm from it and again to −115°C at the outer RPV wall. The KJc values generally follow the course of the MC, although the scatter is large. The re-embrittlement during 2 campaigns operation can be assumed to be low for the weld and base metal.

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.

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