Master Curve Approach for Some Japanese Reactor Pressure Vessel Steels

2008 ◽  
Author(s):  
Minoru Tomimatsu ◽  
Takashi Hirano ◽  
Seiji Asada ◽  
Ryoichi Saeki ◽  
Naoki Miura ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Master Curve ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steels ◽  
The World ◽  
Reactor Pressure ◽  
Plant Components

The Master Curve Approach for assessing fracture toughness of reactor pressure vessel (RPV) steels has been accepted throughout the world. The Master Curve Approach using fracture toughness data obtained from RPV steels in Japan has been investigated in order to incorporate this approach into the Japanese Electric Association (JEA) Code 4206, “Method of Verification Tests of the Fracture Toughness for Nuclear Power Plant Components”. This paper presents the applicability of the Master Curve Approach for Japanese RPV steels.

Warm Pre-Stress Tests of the Irradiated RPV Base Material of Beznau Unit 1

2016 ◽  
Author(s):  
Florian Obermeier ◽  
Stefan Heußner ◽  
Heinz Hägeli ◽  
Herbert Schendzielorz ◽  
Marco Kaiser ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Base Material ◽  
Reactor Pressure Vessel ◽  
Stress Effect ◽  
Loading Conditions ◽  
Stress Tests ◽  
Reactor Pressure

According to the pertinent regulations, the integrity of a reactor pressure vessel (RPV) of a nuclear power plant is to be assessed by fracture mechanics for postulated flaws under most severe loading conditions. In such an analysis usually loss of coolant accidents are assumed to cause highest possible loading of the structural material of the RPV. This is due to the fact that such a pressurized thermal shock (PTS) event during which cold emergency coolant is injected into the primary system generates additional thermal stresses in the RPV wall. Based on the applicable regulation, the initiation of postulated flaws is to be excluded by the comparison of the calculated crack tip loading and the fracture toughness of the particular material. This kind of assessment was motivation of various research projects in the last decades addressing both evaluation approaches and experimental testing. A crucial result in this context is the existence of the so-called warm pre-stress effect (WPS) on the resulting fracture toughness. Generally, it is known as the increase of the apparent fracture toughness of a flaw in a specimen or structure after loading at high temperatures, generally in the upper shelf region, followed by a reloading at a lower temperature. This represents the typical loading scenario postulated for the assessment of a RPV during a PTS event. Experiments were performed to quantify this effect in the case of the irradiated reactor pressure vessel base material of the nuclear power plant Beznau unit 1. This paper presents the results of the Master Curve tests to determine the reference temperature (according to ASTM E 1921) and the design and testing of the warm pre-stress experiments using irradiated 10×10 mm reconstituted single edge notch bend (SE(B)) specimens. The design of these warm pre-stress tests was based on the loading transients for postulated surface and sub-surface flaws investigated within the scope of the assessment of the Beznau unit 1 reactor pressure vessel against brittle failure. Finite element simulations were performed to transfer the loading conditions at the crack tip of the RPV determined during the brittle fracture safety assessment onto the SE(B) specimen. The simulation results were used to control the loading conditions as a function of time and temperature during the experimental tests. The fracture toughness values of the warm pre-stress specimens were finally compared with the original fracture toughness values determined in the absence of a warm pre-stress effect to demonstrate the increase of the safety margin when the warm pre-stress effect is taken into account.

Development of an Integrity Evaluation System on the Basis of Cooperative Virtual Reality Environment for Nuclear Power Plant

2004 ◽  
Author(s):  
J. C. Kim ◽  
J. B. Choi ◽  
Y. H. Choi
Keyword(s):  
Virtual Reality ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Evaluation System ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Virtual Reality Environment ◽  
Reactor Pressure

Since early 1950’s fracture mechanics has brought significant impact on structural integrity assessment in a wide range of industries such as power, transportation, civil and petrochemical industries, especially in nuclear power plant industries. For the last two decades, significant efforts have been devoted in developing defect assessment procedures, from which various fitness-for-purpose or fitness-for-service codes have been developed. From another aspect, recent advances in IT (Information Technologies) bring rapid changes in various engineering fields. IT enables people to share information through network and thus provides concurrent working environment without limitations of working places. For this reason, a network system based on internet or intranet has been appeared in various fields of business. Evaluating the integrity of structures is one of the most critical issues in nuclear industry. In order to evaluate the integrity of structures, a complicated and collaborative procedure is required including regular in-service inspection, fracture mechanics analysis, etc. And thus, experts in different fields have to cooperate to resolve the integrity problem. In this paper, an integrity evaluation system on the basis of cooperative virtual reality environment for reactor pressure vessel which adapts IT into a structural integrity evaluation procedure for reactor pressure vessel is introduced. The proposed system uses Virtual Reality (VR) technique, Virtual Network Computing (VNC) and knowledge based programs. This system is able to support 3-dimensional virtual reality environment and to provide experts to cooperate by accessing related data through internet. The proposed system is expected to provide a more efficient integrity evaluation for reactor pressure vessel.

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.

Reactor Pressure Vessel Modification Approaches

2005 ◽  
Author(s):  
Ronald J. Payne ◽  
Stephen Levesque
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
State Of The Art ◽  
Electrical Power ◽  
Evaluation Criteria ◽  
Reactor Pressure Vessel ◽  
Alloy 600 ◽  
Electrical Power Research Institute ◽  
Reactor Pressure ◽  
Plant Components

Stress corrosion cracking of Alloy 600 has lead to the modification and replacement of many nuclear power plant components. Among these components are the Bottom Mounted Nozzles (BMN) of the Reactor Pressure Vessel (RPV). Modifications of these components have been performed on an emergent basis. Since that time, Framatome ANP has developed state-of-the-art modification methods for the repair of BMNs using the Electrical Power Research Institute (EPRI) managed Materials Reliability Program (MRP) attributes for an ideal repair as a basis for evaluation of modification concepts. These attributes were used to evaluate the optimal modification concepts and develop processes and tooling to support future modification activity. This paper details the BMN configurations, modification evaluation criteria, several modification concepts, and the development of the tooling to support the optimal modification scenarios.

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