Ways for Extending the Reactor Pressure Vessel Lifetime

2008 ◽  
Author(s):  
Milan Brumovsky
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Integrity Assessment ◽  
Low Leakage ◽  
Vessel Integrity ◽  
Direct Use ◽  
Reactor Pressure ◽  
Lifetime Extension

Reactor pressure vessels are components that usually determine the lifetime of the whole nuclear power plant and thus also its efficiency and economy. There are several ways how to ensure conditions for reactor pressure vessel lifetime extension, mainly: - pre-operational, like: • optimal design of the vessel; • proper choice of vessel materials and manufacturing technology; - operational, like: • application of low-leakage core; • increase of water temperature in ECCS; • insertion of dummy elements; • vessel annealing; • decrease of conservatism during reactor pressure vessel integrity assessment e.g. using direct use of fracture mechanics parameters, like “Master Curve” approach. All these ways are discussed in the paper and some qualitative as well as quantitative evaluation is given.

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.

Comparison of PTS Guides for Reactor Pressure Vessel Integrity Assessment

2013 ◽  
Author(s):  
Milan Brumovsky
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Detailed Comparison ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Integrity Assessment ◽  
Asme Code ◽  
Vessel Integrity ◽  
Degradation Of Materials ◽  
Reactor Pressure

Integrity of reactor pressure vessels (RPV) are of the most importance for safety of the whole NPP. From all potential regimes, Pressurized Thermal Shock (PTS) regimes during emergency cooling conditions are the most severe and most important. Several nuclear codes are based in similar approaches but their procedures differ and are based on national knowledge and approach to fracture mechanics as well as non-destructive methods of reactor pressure vessel testing. The paper will compare requirements and procedures for PTS evaluation in accordance with RCC-M code in France [2], KTA in Germany [3], Russian original code PNAEG from 1989 [5] and new procedure from 2004 for WWER vessels [4], as well as VERLIFE procedure and IAEA-NULIFE VERLIFE [6] procedure for WWER RPVs and finally ASME Code requirements [1] including US NRC RG approach. Detailed comparison of individual parameters in calculations are compared — material properties, degradation of materials, calculated defects size and form, fracture mechanics approach, warm pre-stressing possibility etc.

Comparison of PTS Guides for Reactor Pressure Vessel Integrity Assessment

2013 ◽  
Author(s):  
Milan Brumovsky
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Detailed Comparison ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Integrity Assessment ◽  
Asme Code ◽  
Vessel Integrity ◽  
Degradation Of Materials ◽  
Reactor Pressure

Integrity of reactor pressure vessels (RPV) are of the most importance for safety of the whole NPP. From all potential regimes, Pressurized Thermal Shock (PTS) regimes during emergency cooling conditions are the most severe and most important. Several nuclear codes are based in similar approaches but their procedures differ and are based on national knowledge and approach to fracture mechanics as well as non-destructive methods of reactor pressure vessel testing. The paper will compare requirements and procedures for PTS evaluation in accordance with RCC-M code in France [2], KTA in Germany [3], Russian original code PNAEG from 1989 [5] and new procedure from 2004 for WWER vessels [4], as well as VERLIFE procedure and IAEA-NULIFE VERLIFE [6] procedure for WWER RPVs and finally ASME Code requirements [1] including US NRC RG approach. Detailed comparison of individual parameters in calculations are compared — material properties, degradation of materials, calculated defects size and form, fracture mechanics approach, warm pre-stressing possibility etc.

Assessment of the Reactor Pressure Vessel Integrity in PTS Conditions: Presentation and Demonstration of SECURE — A Numerical Tool Developed by EDF to Realise Automatically Integrity Assessments

2008 ◽  
Author(s):  
A. Dahl ◽  
C. Messelier-Gouze
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Pressure Vessel ◽  
Full Data ◽  
Integrity Assessment ◽  
Nuclear Plants ◽  
Numerical Tool ◽  
Vessel Integrity ◽  
Reactor Pressure

The reactor pressure vessel (RPV) is an essential component liable to limit the lifetime duration of PWR nuclear power plants. For the RPV assessment and Lifetime evaluation of the nuclear plants, French Utility applies a series of calculations including different approaches, more or less refined. These approaches combine different types of data : neutron physics, vessel geometry, material data, thermal and mechanical transients. SECURE, a numerical tool based on EDF finite element code, code_Aster, and full data bases was developed by EDF in order to make automatically RPV integrity assessment with simplified approaches (1D computations) and more refined approaches (2D and 3D computations). Then, with this tool, RPV integrity studies are made more quickly and with more reliability. This paper describes SECURE tool and presents an application with more and less refined methods on a 3D thermal computation of a transient. This application shows that this tool allowed to do easily refined numerical methods which improve the final margins. This tool, developed and validate by EDF Research and development division, must now be used by engineering division.

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.

MAI Benchmark Campaign of International Software for Reactor Pressure Vessel Integrity Assessment

2014 ◽  
Author(s):  
Emilie Dautreme ◽  
Emmanuel Remy ◽  
Roman Sueur ◽  
Jean-Philippe Fontes ◽  
Karine Aubert ◽  
...  
Keyword(s):  
United States ◽  
Pressure Vessel ◽  
Software Tool ◽  
The United States ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Benchmark Study ◽  
Reactor Pressure

Nuclear Reactor Pressure Vessel (RPV) integrity is a major issue concerning plant safety and this component is one of the few within a Pressurized Water Reactor (PWR) whose replacement is not considered as feasible. To ensure that adequate margins against failure are maintained throughout the vessel service life, research engineers have developed and applied computational tools to study and assess the probability of pressure vessel failure during operating and postulated loads. The Materials Ageing Institute (MAI) sponsored a benchmark study to compare the results from software developed in France, Japan and the United States to compute the probability of flaw initiation in reactor pressure vessels. This benchmark study was performed to assess the similarities and differences in the software and to identify the sources of any differences that were found. Participants in this work included researchers from EDF in France, CRIEPI in Japan and EPRI in the United States, with each organization using the probabilistic software tool that had been developed in their country. An incremental approach, beginning with deterministic comparisons and ending by assessing Conditional Probability of crack Initiation (CPI), provided confirmation of the good agreement between the results obtained from the software used in this benchmark study. This conclusion strengthens the confidence in these probabilistic fracture mechanics tools and improves understanding of the fundamental computational procedures and algorithms.

Weld Material Investigations of a WWER-440 Reactor Pressure Vessel: Results From the First Trepan Taken From the Former Greifswald NPP

2009 ◽  
Vol 131 (2) ◽  
Author(s):  
Udo Rindelhardt ◽  
Hans-Werner Viehrig ◽  
Joerg Konheiser ◽  
Jan Schuhknecht
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Reference Temperature ◽  
Monte Carlo Code ◽  
Data Set ◽  
Integrity Assessment ◽  
Inner Surface ◽  
Reactor Pressure

Between 1973 and 1990 four units of the Russian nuclear power plants type WWER-440/230 were operated in Greifswald (former East Germany). Material probes from the pressure vessels were gained in the frame of the ongoing decommissioning procedure. The investigations of this material started with material from the circumferential core weld of unit 1. First, this paper presents results of the reactor pressure vessel (RPV) fluence calculations depending on different loading schemes and on the axial weld position based on the Monte Carlo code TRAMO. The results show that the use of the dummy assemblies reduces the flux by a factor of 2–5 depending on the azimuthal position. The circumferential core weld (SN0.1.4) received a fluence of 2.4×1019 neutrons/cm2 at the inner surface; it decreases to 0.8×1019 neutrons/cm2 at the outer surface. The material investigations were done using a trepan from the circumferential core weld. The reference temperature T0 was calculated with the measured fracture toughness values, KJc, at brittle failure of the specimen. The KJc values show a remarkable scatter. The highest T0 was about 50°C at a distance of 22 mm from the inner surface of the weld. The Charpy transition temperature TT41J estimated with results of subsized specimens after the recovery annealing was confirmed by the testing of standard Charpy V-notch specimens. The VERLIFE procedure prepared for the integrity assessment of WWER RPV was applied on the measured results. The VERLIFE lower bound curve indexed with the Structural Integrity Assessment Procedures for European Industry (SINTAP) reference temperature, RTT0SINTAP, envelops the KJc values. Therefore for a conservative integrity assessment the fracture toughness curve indexed with a RT representing the brittle fraction of a data set of measured KJc values has to be applied.

The Delivery of Technical Training Within the European Style Project

2013 ◽  
Author(s):  
Adam Toft ◽  
John Sharples
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Nuclear Industry ◽  
Technical Training ◽  
Illustrative Case ◽  
Integrity Assessment ◽  
Leak Before Break ◽  
Reactor Pressure

The STYLE project considers structural integrity for lifetime management of non-reactor pressure vessel components of nuclear power plant. The project is funded under the seventh European Commission framework programme. A broad objective of the project is to assess, optimise and develop application of advanced tools for structural integrity assessment of reactor coolant pressure boundary components other than the reactor pressure vessel. One aspect of the STYLE project is intended to address the issue of succession planning within the European nuclear industry. With many key technical experts now approaching retirement it is essential to progress the technical expertise of those at an earlier stage of their career in the industry. The paper describes how technical training has been delivered as an integral part of the STYLE project to support retention of the current level of technical capability in future. Diverse aspects of training are described. These include participation in experimental work, numerical modelling and simulation, application of engineering assessment procedures, leak-before-break, probabilistic fracture mechanics and materials behaviour. An illustrative case study is described, in which trainees received practical instruction in the essential steps for technical justification of a leak-before-break argument.

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