scholarly journals Long-term operation of a boiling pressure vessel and its internals

2019 ◽  
Vol 52 (4) ◽  
pp. 200-221
Author(s):  
Otso Cronvall
Keyword(s):  
Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Pilot Project ◽  
Computational Time ◽  
Degradation Mechanisms ◽  
Term Operation ◽  
Reactor Pressure

This study concerns the long-term operation (LTO) of a boiling water reactor (BWR) reactor pressure vessel (RPV) and its internals. The main parts of this study are: survey on the susceptibility to degradation mechanisms, and computational time limited ageing analyses (TLAAs). The ageing of nuclear power plants (NPPs) emphasizes the need to anticipate the possible degradation mechanisms. The BWR survey on the susceptibility to these mechanisms uses the RPVs and significant internals of the Olkiluoto power plant units OL1 and OL2 as a pilot project. For the components that screened in, the potential to brittle, ductile or other degradation is determined. This was carried out by applying structural mechanics and fracture mechanics procedures. Only some most significant cases and results are presented here.

Long-Term Operation of BWR RPV and its Internals

2018 ◽  
Author(s):  
Otso Cronvall
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Pilot Project ◽  
Computational Time ◽  
Degradation Mechanisms ◽  
Irradiation Embrittlement ◽  
Screening Process ◽  
Term Operation ◽  
Operational Lifetime

This study concerns the long-term operation (LTO) of a boiling water reactor (BWR) reactor pressure vessel (RPV) and its internals. The main parts of this study are: survey on susceptibility to degradation mechanisms, and computational time limited ageing analyses (TLAAs). The ageing of nuclear power plants (NPPs) emphasises the need to anticipate the possible degradation mechanisms. The BWR survey on susceptibility to these uses the OL1/OL2 RPVs and significant internals as a pilot project. It is not necessary to carry out the TLAAs for all components. Some components were excluded from the TLAAs with a screening process. To do this, it was necessary to determine the component specific load induced stresses, strains and temperature distributions as well as cumulative usage factor (CUF) values. For the screened-in components, the TLAAs covered all significant time dependent degradation mechanisms. These include (but are not limited to): • irradiation embrittlement, • fatigue, • stress corrosion cracking (SCC), and • irradiation accelerated SCC (IASCC). For the components that were screened-in, the potential to brittle, ductile or other degradation was determined. Only some of the most significant cases and results are presented. According to the analysis results, the operational lifetime of the OL1/OL2 RPVs and internals can safely be extended from 40 to 60 years.

Application of Metallographic Methods for Thermal Ageing Evaluation on Samples from Primary Piping of NPP

2017 ◽  
Vol 891 ◽  
pp. 60-66
Author(s):  
Jana Petzová ◽  
Martin Březina ◽  
Miloš Baľák ◽  
Mária Dománková ◽  
Ľudovít Kupča
Keyword(s):  
Material Properties ◽  
Nuclear Power ◽  
Power Plants ◽  
Operating Temperature ◽  
Thermal Ageing ◽  
Primary Circuit ◽  
Degradation Mechanisms ◽  
Reliable Operation ◽  
Term Operation

During a long-term operation of nuclear power plants (NPP), the changes of structural material properties occur. To ensure the safe and reliable operation, it is necessary to monitor and evaluate these changes mainly on components from primary circuit of NPPs. One of the dominant ageing mechanisms of NPP components besides the radiation embrittlement and the fatigue loads is the thermal ageing. The thermal ageing is the temperature, material and time dependent degradation mechanisms due to long-term exposure at the operating temperature of 570 K.This paper describes the project for thermal ageing monitoring at primary piping in NPP Bohunice Unit 3. There are summarized the results obtained from evaluation of original primary piping material.

The role of pressure vessel embrittlement in the long term operation of nuclear power plants

2012 ◽  
Vol 243 ◽  
pp. 63-68 ◽  
Author(s):  
A. Ballesteros ◽  
R. Ahlstrand ◽  
C. Bruynooghe ◽  
U. von Estorff ◽  
L. Debarberis
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Term Operation

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.

“Making It Work”: The Vital Role of Commissioning in Nuclear Power Programs

2016 ◽  
Vol 2 (4) ◽  
Author(s):  
John H. Moore
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Vital Role ◽  
Special Relationship ◽  
Plant Component ◽  
Term Operation

This paper presents an overview of the duties of a nuclear power plant (NPP) commissioning organization and its special relationship with both the early stages of an NPP project and the long-term operation of such power plants. Decisions made early in the design and procurement process need to be understood by commissioning staff, and their implications need to be incorporated into commissioning programs. Commissioning programs also need to take the steps needed to baseline plant-component ageing programs to facilitate the long-term NPP operation.

Surveillance Specimen Program for Lifetime Extension of RPV

2010 ◽  
Author(s):  
Milan Brumovsky ◽  
Milos Kytka
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Thermal Ageing ◽  
Reactor Pressure Vessel ◽  
Life Extension ◽  
Irradiation Embrittlement ◽  
Term Operation ◽  
Reactor Pressure ◽  
Lifetime Extension

Plant life extension (as well as Long Term Operation) to 60 or 80 years of operation also requires a reliable information about the potential irradiation embrittlement (and also thermal ageing) of reactor pressure vessel materials. Such information is usually obtained from testing specimens within the surveillance specimen program that is designed for the design reactor pressure vessel (RPV) life, regularly for 40 years only. Life extension requires modification of such program (if there is still time to perform it) or a design of a new – extended one. Such program should have to contain RPV archive materials that are not in every case available. Thus, combination of archive materials and possible surrogate materials must be taken into account for this program. Some complication can be expected with thermal ageing data as some laboratory tests at higher temperatures must be realized. The paper describes such program for Nuclear Power Plant (NPP) Dukovany, Czech Republic with WWER-440 type reactors.

OECD/NEA Activities to Support Long Term Operation

2010 ◽  
Vol 5 (6) ◽  
pp. 707-711
Author(s):  
Andrei Blahoianu ◽  
◽  
Alejandro Huerta ◽  
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Energy ◽  
State Of The Art ◽  
Integrated Approach ◽  
Energy Agency ◽  
Term Operation ◽  
Plant Life

The Integrity and Aging of Components and Structures Working Group (IAGE) of the Organisation for Economic Cooperation and Development (OECD)/Nuclear Energy Agency (NEA) was established under the Committee on the Safety of Nuclear Installations (CSNI) for three reasons: (i) to advance the current understanding of those aspects relevant to ensuring the integrity of structures, systems, and components ; (ii) to provide for guidance in choosing the optimal ways to handle challenges to the integrity of operating as well as new nuclear power plants, and (iii) to take an integrated approach to design, safety, and nuclear power plant life management. The group operates through annual plenary meetings and technical workshops and by issuing state-of-the-art reports and topical opinion papers. This paper details some recent IAGE activities and products, focusing on those dealing with the degradationmechanisms of metal and concrete components.

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.

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