The Fatigue and Aging Effects on the Materials of the Nuclear Power Plant Components and Systems

2010 ◽  
Author(s):  
Sajjad Akbar ◽  
M. Shahid Khalil
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Service Condition ◽  
Reactor Pressure Vessel ◽  
Aging Effects ◽  
Operational Life ◽  
The Future ◽  
Aging Phenomena ◽  
Reactor Pressure

The first generation of nuclear power plants (NPP) parts, components, equipment and systems have completed their predicted trouble free operational life. Now their materials are posing fatigue and aging effects for the future useful applications in the NPP. For the reliable future operations, this study is made for the management of the fatigue and aging effects on the components/systems. The neutron irradiation and normal aging phenomena will cause deterioration in the properties of materials by increasing brittleness-ductile transition temperature. This would affect the reactor pressure vessel as well. For the design purpose as well as estimating the permissible service life of a reactor pressure vessel, it is essential to have reliable data regarding this effect. Considerable efforts have been made and data search carried out on this work. The routine and schedule maintenance programs are the prerequisite means of managing fatigue and aging effects. Therefore, these are monitored through collecting data on plant service condition, inspection, surveillance, testing and by regularly monitoring such programs. This paper presents the future course of actions, which are necessary in the enhancement of the useful and safe NPP operations throughout the world.

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.

scholarly journals Occupational ALARA Planning for Reactor Pressure Vessel Dismantling at Kori Unit 1

2020 ◽  
Vol 17 (15) ◽  
pp. 5346
Author(s):  
Juyoul Kim ◽  
Batbuyan Tseren
Keyword(s):  
Radiation Exposure ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Occupational Safety ◽  
Health Management ◽  
Primary Component ◽  
Reactor Pressure Vessel ◽  
Safety And Health ◽  
Reactor Pressure

Assessing workers’ safety and health during the decommissioning of nuclear power plants (NPPs) is an important procedure in terms of occupational radiation exposure (ORE). Optimizing the radiation exposure through the “As Low As Reasonably Achievable (ALARA)” principle is a very important procedure in the phase of nuclear decommissioning. Using the VISIPLAN 3D ALARA planning tool, this study aimed at assessing the radiological doses to workers during the dismantling of the reactor pressure vessel (RPV) at Kori NPP unit 1. Fragmentation and segmentation cutting processes were applied to cut the primary component. Using a simulation function in VISIPLAN, the external exposure doses were calculated for each work operation. Fragmentation involved 18 operations, whereas segmentation comprised 32 operations for each fragment. Six operations were additionally performed for both hot and cold legs of the RPV. The operations were conducted based on the radioactive waste drum’s dimensions. The results in this study indicated that the collective doses decreased as the components were cut into smaller segments. The fragmentation process showed a relatively higher collective dose compared to the segmentation operation. The active part of the RPV significantly contributed to the exposure dose and thus the shielding of workers and reduced working hours need to be considered. It was found that 60Co contained in the stainless steel of the reactor vessel greatly contributed to the dose as an activation material. The sensitivity analysis, which was conducted for different cutting methods, showed that laser cutting took a much longer time than plasma cutting and contributed higher doses to the workers. This study will be helpful in carrying out the occupational safety and health management of decommissioning workers at Kori NPP unit 1 in the near future.

Upgrading the Operational Safety of Nuclear Power Plants Through the TACIS Nuclear Safety Assistance Programme

2004 ◽  
Author(s):  
M. Bie`th ◽  
R. Ahlstrand ◽  
C. Rieg ◽  
P. Trampus
Keyword(s):  
European Commission ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Safety ◽  
Reactor Pressure Vessel ◽  
Nuclear Safeguards ◽  
Integrity Assessment ◽  
Reactor Pressure

The European Union’ TACIS programme was established for the New Independent States since 1991. One priority for TACIS funding is nuclear safety. The European Commission has made available a total of € 944 million for nuclear safety programmes covering the period 1991–2003. The TACIS nuclear safety programme is devoted to the improvement of the safety of Soviet designed nuclear installations in providing technology and safety culture transfer. The Joint Research Center (JRC) of the European Commission is carrying out works in the following areas: • On-Site Assistance for TACIS Nuclear Power Plants; • Design Safety and Dissemination of TACIS results; • Reactor Pressure Vessel Embrittlement for VVER in Russia and Ukraine; • Regulatory Assistance; • Industrial Waste Management and Nuclear Safeguards. This paper gives an overview of the Scientific and Technical support that JRC is providing for the programming and the implementation of the TACIS nuclear safety programmes. In particular, two new projects are being implemented to get an extensive understanding of the VVER reactor pressure vessel embritttlement and integrity assessment.

Nuclear Reactor Pressure Vessel Integrity Assessment: Enhancement Provided by 3D Modeling Flaw Stability Calculations

2013 ◽  
Author(s):  
Adolfo Arrieta-Ruiz ◽  
Eric Meister ◽  
Henriette Churier
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Integrity Assessment ◽  
Potential Damage ◽  
Vessel Integrity ◽  
Reactor Pressure

Structural integrity of the Reactor Pressure Vessel (RPV) is one of the main considerations regarding safety and lifetime of Nuclear Power Plants (NPP) since this component is considered as not reasonably replaceable. Brittle fracture risk associated with the embrittlement of RPV steel in irradiated areas is the main potential damage. In France, deterministic integrity assessment for RPV is based on the crack initiation stage. The stability of an under-clad postulated flaw in the core area is currently evaluated under a Pressurized Thermal Shock (PTS) through a fracture mechanics simplified method. One of the axes of EDF’s implemented strategy for NPP lifetime extension is the improvement of the deterministic approach with regards to the input data and methods so as to reduce conservatisms. In this context, 3D finite element elastic-plastic calculations with flaw modelling have been carried out recently in order to quantify the enhancement provided by a more realistic approach in the most severe events. The aim of this paper is to present both simplified and 3D modelling flaw stability evaluation methods and the results obtained by running a small break LOCA event.

French PWR Nuclear Power Plants Steam Generators and Reactor Pressure Vessel Heads Aging Management Criteria and Replacement Strategy

2004 ◽  
Author(s):  
Georges Bezdikian
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Life Time ◽  
Reactor Pressure Vessel ◽  
Steam Generators ◽  
Life Management ◽  
Power Stations ◽  
Reactor Pressure

The approach used by the French utility, concerning the Aging Management system of the Steam Generators (SG) and Reactor Pressure Vessel Heads, applied on 58 PWR NPPs, involves the verification of the integrity of the component and the Life Management of each plant to guarantee in the first step the design life management and in the second step to prepare long term life time in operation, taking into account the degradation of Alloy 600 material and the replacement of these materials by components made with Alloy 690. The financial stakes associated with maintaining the lifetime of nuclear power stations are very high; thus, if their lifetime is shortened by about ten years, dismantling and renewal would be brought forward which would increase their costs by several tens of billions of Euros. The main objectives are: • to maintain current operating performances (safety, availability, costs, security, environment) in the long term, and possibly improve on some aspects; • wherever possible, to operate the units throughout their design lifetime, 40 years, and even more if possible. This paper shows the program to follow the aging evaluation with application of specific criteria for SG and for Vessel Heads, and the replacement of the Steam Generators and Vessel Heads at the best period. The strategy of Steam Generators Replacement are developed and Vessel Head program of monitoring and replacement are detailed.

On the Analysis of Clustering in an Irradiated Low Alloy Reactor Pressure Vessel Steel Weld

2017 ◽  
Vol 23 (2) ◽  
pp. 376-384 ◽  
Author(s):  
Kristina Lindgren ◽  
Krystyna Stiller ◽  
Pål Efsing ◽  
Mattias Thuvander
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
Number Density ◽  
Cluster Number ◽  
Vessel Steel ◽  
Time Operation ◽  
Reactor Pressure

AbstractRadiation induced clustering affects the mechanical properties, that is the ductile to brittle transition temperature (DBTT), of reactor pressure vessel (RPV) steel of nuclear power plants. The combination of low Cu and high Ni used in some RPV welds is known to further enhance the DBTT shift during long time operation. In this study, RPV weld samples containing 0.04 at% Cu and 1.6 at% Ni were irradiated to 2.0 and 6.4×1023 n/m2 in the Halden test reactor. Atom probe tomography (APT) was applied to study clustering of Ni, Mn, Si, and Cu. As the clusters are in the nanometer-range, APT is a very suitable technique for this type of study. From APT analyses information about size distribution, number density, and composition of the clusters can be obtained. However, the quantification of these attributes is not trivial. The maximum separation method (MSM) has been used to characterize the clusters and a detailed study about the influence of the choice of MSM cluster parameters, primarily on the cluster number density, has been undertaken.

Evaluation of Mechanical Properties of the Reactor Pressure Vessel Materials Changes by Small Punch Test Application

2013 ◽  
Author(s):  
Jana Petzová ◽  
Martin Březina ◽  
Ľudovít Kupča
Keyword(s):  
Mechanical Properties ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Mechanical Tests ◽  
Reactor Pressure Vessel ◽  
Small Punch Test ◽  
Small Punch ◽  
Punch Test ◽  
Reactor Pressure

The reactor pressure vessel (RPV) is the most important component of nuclear power plants. RPV steel near the reactor core is subject of irradiation degradation due to the fast neutron flux. Irradiation processes are rather complex but after all the damage of the steel crystal lattice lead to the changes of RPV mechanical properties as well as the shift of the transition temperature to higher values. Hence, monitoring of the RPV material irradiation changes must be proved during the all nuclear power plant (NPP) operation. The new surveillance specimen programs (SSP) at all Slovak NPPs reactors included, among the standard mechanical tests, also new types of evaluation mechanical properties due to method Small Punch Test (SPT).

Development of Web-Based Fatigue Life Evaluation System for Reactor Pressure Vessel

2007 ◽  
Vol 120 ◽  
pp. 25-30 ◽  
Author(s):  
J.C. Kim ◽  
Jae Boong Choi ◽  
Yoon Suk Chang ◽  
Young Jin Kim ◽  
Youn Won Park ◽  
...  
Keyword(s):  
Fatigue Life ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Pressure Vessel ◽  
Web Based ◽  
Life Evaluation ◽  
Fatigue Life Evaluation ◽  
Reactor Pressure

While the demand on electric power is consistently increasing, public concerns and regulations for the construction of new nuclear power plants are getting restrict, and also operating nuclear power plants are gradually ageing. For this reason, the interest on lifetime extension for operating nuclear power plants by applying lifetime management system is increasing. The 40-year design life concept was originally introduced on the basis of economic and safety considerations. In other words, it was not determined by technological evaluations. Also, the transient design data which were applied for fatigue damage evaluation were overly conservative in comparison with actual transient data. Therefore, the accumulation of fatigue damage may result in a big difference between the actual data and the design data. The lifetime of nuclear power plants is mostly dependent on the fatigue life of a reactor pressure vessel, and thus, the exact evaluation of fatigue life on a reactor pressure vessel is a crucial factor in determining the extension of operating life. The purpose of this paper is to introduce a real-time fatigue monitoring system for an operating reactor pressure vessel which can be used for the lifetime extension. In order to satisfy the objectives, a web-based transient acquisition system was developed, thereby, real-time thermal-hydraulic data were reserved for 18 operating reactor pressure vessels. A series of finite element analyses was carried out to obtain the stress data due to actual transient. The fatigue life evaluation has been performed based on the stress analysis results and, finally, a web-based fatigue life evaluation system was introduced by combining analysis results and on-line monitoring system. Comparison of the stress analysis results between operating transients and design transients showed a considerable amount of benefits in terms of fatigue life. Therefore, it is anticipated that the developed web-based system can be utilized as an efficient tool for fatigue life estimation of reactor pressure vessel.

The Research on Ultrasonic Testing Technology for Bottom Part of Reactor Pressure Vessel

2017 ◽  
Author(s):  
Ge Liang ◽  
Xu Xuguang ◽  
Cai Jiafan ◽  
Cheng Zhaoyu
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Pressure Vessel ◽  
Inspection System ◽  
Lower Head ◽  
Automatic Docking ◽  
Multi Mode ◽  
Reactor Pressure

Reactor Pressure Vessel (hereinafter RPV) is the core component of PWR nuclear power plants. The nuclear power plants operation experience shows that the dangerous defects could produce in the material of Bottom Mounted Instrument and J type weld (hereinafter BMI). In early nuclear power plants, the BMI is made of corrosion-sensitive materials (such as Alloy 600). The stress corrosion crack is easy to grow up and destroy the integrity of RPV in operation. Therefore, inspection for BMI during the service stage is necessary to ensure the integrity of the primary circuit pressure boundary. The BMI is general made of Alloy 600 or Alloy 690 and penetrate the reactor pressure vessel lower head vertically to form J type weld on the inner wall of the RPV lower head. The inner diameter is generally 15 to 16 mm and the wall thickness is about 15 mm. In order to improve the welding performance, the dissimilar metal J-type weld groove is usually overlaid with nickel-based pre-heap edge. The component structural and material characteristics have brought great difficulties to the test technology research as well as the location. In this paper, the main failure modes of BMI are introduced and the technical requirements of in-service inspections are presented. The technical scheme and inspection system are designed according to the above requirements and weld structural characteristics. Both UT and ET methods are used for the inspection. UT is based on the time of flight diffraction (hereinafter TOFD) technique. According to the structural and material characteristics of the BMI, many researches have been done on block with simulated defect, and characteristic rule of defect indication signals in different areas are summarized. Then, the optimal design parameters of the TOFD probe are gained and the disadvantages of ultrasonic inspection are overcome. The eddy current inspection system adopts the inspection technology with the Bobbin probe and the rotary point probe combined together. The use of its surface and near the surface of the detection ability has become a powerful complement to ultrasonic testing. From the test result, TOFD probe has high defect height measurement accuracy. Measurement error isn’t more than 1mm and mostly is positive deviation. The surface defects with a height of not less than 2 mm can be measured. Considering the complex geometry of BMI, the multi-mode multi-axis scanning program and remote automatic docking device are used in the system. The multi-mode multi-axis scanning program can be achieved axial and circumferential rotation of the probe with the axial positioning accuracy within 2mm and circumferential positioning accuracy within 3°. The remote automatic docking device can be used to connect the bottom of the mechanical device and the top of the BMI in fast and seamless security to avoid the mechanical device bumping of the component. The probe connecting shaft is made of flexible material, which avoids the violent bumping of the tested parts and greatly enhances the safety of the equipment operation. The system simulates the actual scanning environment by establishing three-dimensional model and testing it on the simulation body.

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