scholarly journals Micromagnetic Microstructure- and Stress-Independent Materials Characterization in Reactor Safety Research

Materials ◽  
2021 ◽  
Vol 14 (18) ◽  
pp. 5258
Author(s):  
Cyril Zimmer ◽  
Yashashwini Nikhitha Rallabandi ◽  
Klaus Szielasko ◽  
Christian Eichheimer ◽  
Michael Luke ◽  
...  
Keyword(s):  
Service Life ◽  
Nuclear Power ◽  
Power Plants ◽  
Material Behavior ◽  
Testing System ◽  
Reactor Safety ◽  
Safety Research ◽  
Aging Mechanisms ◽  
Prediction Of Mechanical Properties ◽  
Stress Dependent

Reactor safety research aims at the safe operation of nuclear power plants during their service life. In this respect, Fraunhofer IZFP’s micromagnetic multiparameter, microstructure, and stress analysis (3MA) has already made a significant contribution to the understanding of different aging mechanisms of component materials and their characterization. The basis of 3MA is the fact that microstructure and mechanical stress determine both the mechanical and magnetic material behavior. The correlation between features of magnetic and mechanical material behavior enables the micromagnetic prediction of mechanical properties and stress, both of which can decisively influence the service life. The Federal Ministry for Economic Affairs and Energy (BMWi) funded this research, handling the mutually superimposed microstructural and stress-dependent influences, a substantial challenge, especially under practical conditions. This superposition leads to ambiguities in the micromagnetic features. The 3MA testing system has been extended by more sophisticated evaluation methods being able to cope with more complex datasets. Investigations dealing with the expansion of the feature extraction and machine learning methods have led to a more precise distinction between microstructural and stress-dependent influences. This approach provides the basis for future applications in reactor safety.

Investigation of the Influence of Operational Loading Regimes on the Service Life of Nuclear Power Plants

2020 ◽  
pp. 319-330
Author(s):  
Nikolay A. Makhutov ◽  
Mikhail M. Gadenin ◽  
Sergey V. Maslov ◽  
Dmitry O. Reznikov ◽  
Sergey N. Pichkov ◽  
...  
Keyword(s):  
Service Life ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants

Research of Technology for Repair of Heat Exchangers of Nuclear Power Plants by Laser Welding

2021 ◽  
Vol 313 ◽  
pp. 94-105
Author(s):  
A. Bernatskyi ◽  
V. Sydorets ◽  
O.M. Berdnikova ◽  
I. Krivtsun ◽  
O. Kushnarova
Keyword(s):  
Service Life ◽  
Laser Welding ◽  
Welded Joints ◽  
Heat Exchangers ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Mechanical Test ◽  
Austenitic Steels ◽  
Depth Of Penetration

Extending the lifetime of energy facilities is extremely important today. This is especially true of nuclear power plants, the closure (or modernization) of which poses enormous financial and environmental problems. High-quality repair of reactors can significantly extend their service life. One of the critical parts is heat exchangers, the tubes of which quite often fail. Sealing, as a type of repair of heat exchanger tubes by the plugs, is promising provided that the joint quality is high. Practical experience in the use of welding to solve this problem has shown the need to search technological solutions associated with increasing the depth of penetration and reducing the area of thermal effect. The aim of the work was to develop a highly efficient technology for repair and extension of service life of heat exchangers of nuclear power plants based on the results of studying the technological features of laser welding of joints of dissimilar austenitic steels AISI 321 and AISI 316Ti in the vertical spatial position. Based on the results of the analysis of mechanical test data, visual and radiographic control, impermeability tests and metallographic studies of welded joints, the appropriate modes of laser welding of plugs have been determined. The principal causes of defects during laser welding of annular welded joints of dissimilar stainless steels are determined and techniques for their elimination and prevention of their formation are proposed. Based on the results of the research, technological recommendations for laser welding of plugs in the heat exchange tube of the collector are formulated, which significantly improves the technology of repair of steam generators of nuclear power plants and extends the service life of reactors.

B2' Stress Indices for Elbows Using Non-Linear Finite Element Analysis

2004 ◽  
Author(s):  
K. Venkataramana ◽  
V. Bhasin ◽  
K. K. Vaze ◽  
H. S. Kushwaha
Keyword(s):  
Plastic Deformation ◽  
Finite Element ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Kinematic Hardening ◽  
Material Behavior ◽  
Stress Index ◽  
Stress Indices

Nuclear power plants are designed to withstand earthquake loads without severe damage under service level D conditions. Under earthquake induced reversing dynamic load, nuclear power plant components may undergo plastic deformation. Plastic deformation in class I nuclear power plant piping systems is limited by Equation (9) of ASME Boiler & Pressure Vessel Code [14], Section III, NB-3652. In the year 2000, the ASME B&PV Code was revised to accommodate reversing dynamic loading in which the failure mode is fatigue ratcheting, instead of plastic collapse. This modified equation [16] contains B2′ index, which is given as a fraction of B2 index where, B2 is defined for monotonic loading [17]. In this study a new definition is proposed for calculating B2′ stress index which is given by B2′ = MCLcyclicRange,straightpipe/MCLcyclicRange,component, where MClcyclicRange is the range of collapse moment. Incremental elastic-plastic nonlinear finite element analyses are performed considering both material and geometric nonlinearities. Kinematic hardening, isotropic hardening and elastic-perfectly plastic material models have been used to model the material behavior during plastic deformation. Load deflection curves are obtained and from these curves collapse loads for monotonic and cyclic loading are determined. B2 and B2′ stress indices are computed for elbows using the proposed equation. The computed stress indices are compared with ASME Code values.

Development of Russian Regulatory Basis for License Renewal/Life Extension of Nuclear Power Plants

2002 ◽  
Author(s):  
Jaroslav Pachner ◽  
Inna V. Kaliberda
Keyword(s):  
Power Plant ◽  
Service Life ◽  
Nuclear Power ◽  
Power Plants ◽  
Life Extension ◽  
Safety Standard ◽  
Regulatory Strategy ◽  
Regulatory Guidelines ◽  
License Renewal ◽  
Regulatory Organization

By 2010, fifteen Russian nuclear power plant (NPP) units will reach the end of their scheduled service life of 30 years. An extension of NPP operation beyond the scheduled service life, which is provided for by the Russian safety standard OPB-88/97, requires a renewal of the plant operating license by the regulatory organization Gosatomnadzor (GAN). This paper presents an overview of an IAEA project aimed at assisting GAN in the development of a set of regulatory guidelines for NPP license renewal/life extension. The overview includes a description of the Russian regulatory strategy for life extension of NPPs and IAEA activities to facilitate drafting and application of the regulatory guidelines using international experience.

scholarly journals International Standard Problems and Small Break Loss-of-Coolant Accident (SBLOCA)

2008 ◽  
Vol 2008 ◽  
pp. 1-22 ◽  
Author(s):  
N. Aksan
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Lessons Learned ◽  
Reactor Safety ◽  
Hydrogen Distribution ◽  
Integral Test ◽  
International Standard ◽  
Hydraulic Behaviour ◽  
Nuclear Reactor Safety

Best-estimate thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants and also used in the design of advance reactors. Evaluation of the capabilities and the performance of these codes can be accomplished by comparing the code predictions with measured experimental data obtained on different test facilities. OECD/NEA Committee on the Safety of Nuclear Installations (CSNI) has promoted, over the last twenty-nine years, some forty-eight international standard problems (ISPs). These ISPs were performed in different fields as in-vessel thermal-hydraulic behaviour, fuel behaviour under accident conditions, fission product release and transport, core/concrete interactions, hydrogen distribution and mixing, containment thermal-hydraulic behaviour. 80% of these ISPs were related to the working domain of principal working group no.2 on coolant system behaviour (PWG2) and were one of the major PWG2 activities for many years. A global review and synthesis on the contribution that ISPs have made to address nuclear reactor safety issues was initiated by CSNI-PWG2 and an overview on the subject of small break LOCA ISPs is given in this paper based on a report prepared by a writing group. In addition, the relevance of small break LOCA in a PWR with relation to nuclear reactor safety and the reorientation of the reactor safety program after TMI-2 accident are shortly summarized. The experiments in four integral test facilities, LOBI, SPES, BETHSY, ROSA IV/LSTF and the recorded data during a steam generator tube rupture transient in the DOEL-2 PWR (Belgium) were the basis of the five small break LOCA related ISP exercises, which deal with the phenomenon typical of small break LOCAs in Western design PWRs. Some lessons learned from these small break LOCA ISPs are identified in relation to code deficiencies and capabilities, progress in the code capabilities, possibility of scaling, and various additional aspects. ISPs are providing unique material and benefits for some safety-related issues.

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