PWR Nuclear Power Plants Reactor Vessel and Coolant System Cast Duplex Stainless Steel Elbows: New Results in Integrity Analysis

2003 ◽  
Author(s):  
G. Bezdikian ◽  
C. Faidy ◽  
P. Cambefort ◽  
D. Moinereau
Keyword(s):  
Finite Element ◽  
Stainless Steel ◽  
Duplex Stainless Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Reactor Vessel ◽  
Integrity Assessment ◽  
Reactor Coolant

The Reactor Pressure Vessel and Reactor coolant materials (hot and cold CAST elbows) are major components for integrity evaluation of nuclear plant units. The French Utility (Electricite de France) has engaged a few years ago an important program regarding the integrity assessment of RPV and cast duplex stainless steel elbows based on large real database. This paper deals with the verification of the integrity of the Reactor Vessel component by finite element mechanical studies, in all conditions of loading in relation with RTNDT (Reference Nil Ductility Transition Temperature), and considering all parameters. An overall review of actions will be presented describing the French approach regarding the assessment of nuclear RPV. The latest results obtained are based on generic integrity analyses for all categories of situations (normal upset emergency and faulted conditions), particularly in case of PTS, until the end of lifetime, postulating longitudinal shallow subclad flaws. For the Reactor Coolant Elbows, the results of structural integrity analyses, beginning with elastic computations and completed with three-dimensional finite element elastic-plastic computations for envelope cases, are compared with in-service inspection real flaw characterisation and the results are compared to the margin on loading condition with the criteria included in the code.

Large EDF Tests on Aged Cast Duplex Stainless Steel Components: Part I — Reduced Scale Tests

2015 ◽  
Author(s):  
Patrick Le Delliou ◽  
Sébastien Saillet ◽  
Georges Bezdikian
Keyword(s):  
Stainless Steel ◽  
Duplex Stainless Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Evolution ◽  
Thermal Ageing ◽  
Life Extension ◽  
Term Operation ◽  
Integrity Assessment ◽  
Safety Margins

Thermal ageing of cast duplex stainless steel primary loops components (elbows, pump casings and branch connections) is a concern for long-term operation of EDF nuclear power plants. The thermal ageing embrittlement results from the micro-structural evolution of the ferrite phase (spinodal decomposition), and can reduce the fracture toughness properties of the steel. In addition, it is necessary to consider manufacturing quality and the possible occurrence of casting defects such as shrinkage cavities. In a context of life extension, it is important to assess the safety margins to crack initiation and crack propagation instability. This paper presents several tests conducted by EDF on aged cast duplex stainless steel NPP components, respectively on two-third scale elbows and welded mock-ups. The main characteristics of the tests are recalled, the results are presented, and finally, the lessons drawn are summarized. These tests and their detailed analyses contribute to validate and justify the methodology used by EDF in the integrity assessment of in-service cast duplex stainless steel components.

The Application of Concurrent Installation of the RCS Cross Over Leg and the Reactor Vessel Internals

2010 ◽  
Author(s):  
K.-W. Park ◽  
J.-H. Bae ◽  
S.-H. Park
Keyword(s):  
Feasibility Study ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Reactor Vessel ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Study Results ◽  
Under Construction

The reactor vessel internals (RVI) of a pressurized water reactor (PWR) must be installed precisely in the reactor vessel (RV) according to the requirements for levelness, orientation and vertical alignments for its proper functions and structural integrity. For the precise installation, deformation of the RV should be controlled during the RVI installation. Traditionally, the RVI has been installed in the RV after the completion of welding work for large bore pipings in the reactor coolant system (RCS). To reduce installation time, the concurrent installation of the RVI and RCS pipings is investigated. This paper describes the feasibility study on the concurrent installation including the Finite Element Method (FEM) analyses of the RV deformation due to the welding and heat treatment of the pipings. Based on the feasibility study results, the optimum schedule of the RVI installation in parallel with the installation of the cross-over leg pipings (reactor coolant pump inlet pipings) and confirmation measurement locations are developed. Thereby the concurrent installation will be applied to the nuclear power plants under construction in Korea, and it is expected to reduce installation period of 2 months compared to the traditional sequential installation method.

Large EDF Tests on Aged Cast Duplex Stainless Steel Components: Part II — Full Scale Tests

2015 ◽  
Author(s):  
Patrick Le Delliou ◽  
Sébastien Saillet ◽  
Georges Bezdikian
Keyword(s):  
Stainless Steel ◽  
Duplex Stainless Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Thermal Ageing ◽  
Full Scale ◽  
Life Extension ◽  
Term Operation ◽  
Integrity Assessment ◽  
Safety Margins

Thermal ageing of cast duplex stainless steel primary loops components (elbows, pump casings and branch connections) is a concern for long-term operation of EDF nuclear power plants. The thermal ageing embrittlement results from the microstructural evolution of the ferrite phase (spinodal decomposition), and can reduce the fracture toughness properties of the steel. In addition, it is necessary to consider manufacturing quality and the possible occurrence of casting defects such as shrinkage cavities. In a context of life extension, it is important to assess the safety margins to crack initiation and crack propagation instability. This paper presents two tests conducted by EDF on aged cast duplex stainless steel NPP components, respectively on a full-scale elbow and a branch connection. The main characteristics of the tests are recalled, the results are presented, and finally, the lessons drawn are summarized. These tests and their detailed analyses contribute to validate and justify the methodology used by EDF in the integrity assessment of in-service cast duplex stainless steel components.

Elastic-Plastic Fracture Mechanics Analysis of Cast Stainless Steel Pipes (Influence of Axial Load on Z-Factor of Pipes With Circumferential Crack Under Bending Load)

2013 ◽  
Author(s):  
Masayuki Kamaya ◽  
Kiminobu Hojo
Keyword(s):  
Stainless Steel ◽  
Axial Load ◽  
Power Plants ◽  
Structural Integrity ◽  
Term Operation ◽  
Steel Pipes ◽  
Integrity Assessment ◽  
Plastic Failure ◽  
Elastic Plastic ◽  
Bending Load

Since the ductility of cast austenitic stainless steel pipes decreases due to thermal aging embrittlement after long term operation, not only plastic collapse failure but also unstable ductile crack propagation (elastic-plastic failure) should be taken into account for the structural integrity assessment of cracked pipes. In the ASME Section XI, the load multiplier (Z-factor) is used to derive the elastic-plastic failure of the cracked components. The Z-factor of cracked pipes under bending load has been obtained without considering the axial load. In this study, the influence of the axial load on Z-factor was quantified through elastic-plastic failure analyses under various conditions. It was concluded that the axial load increased the Z-factor; however, the magnitude of the increase was not significant, particularly for the main coolant pipes of PWR nuclear power plants.

Structural integrity evaluation for interference-fit flywheels in reactor coolant pumps of nuclear power plants

2005 ◽  
Vol 19 (11) ◽  
pp. 1988-1997 ◽  
Author(s):  
June-soo Park ◽  
Ha-cheol Song ◽  
Ki-seok Yoon ◽  
Taek-sang Choi ◽  
Jai-hak Park
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
Interference Fit ◽  
Reactor Coolant

Overview of Recent NUGENIA Activities Particularly in Relation to Structural Integrity

2015 ◽  
Author(s):  
John Sharples ◽  
Elisabeth Keim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Nuclear Research ◽  
Joint Research ◽  
Non Profit ◽  
Integrity Assessment ◽  
Industry Research ◽  
Implementation And Dissemination ◽  
Technical Area

NUGENIA, an international non-profit association founded under Belgian legislation and launched in March 2012, is dedicated to nuclear research and development (R&D) with a focus on Generation II and III power plants. NUGENIA is the integrated framework between industry, research and safety organisations for safe, reliable and competitive nuclear power production, and is aimed at running an open innovation marketplace, to promote the emergence of joint research and to facilitate the implementation and dissemination of R&D results. The technical scope of NUGENIA consists of eight technical areas. One of these areas, Technical Area 4, is associated with the structural integrity assessment of systems, structures and components. A brief overview of recent NUGENIA activities in general is provided in this paper and a specific focus is given on developments in relation to Technical Area 4.

Three Dimensional Finite Element Analyses of Welding Residual Stresses of a Repaired Weld

2018 ◽  
Author(s):  
Mingya Chen ◽  
Weiwei Yu ◽  
Fei Xue ◽  
Francis Ku ◽  
Zhilin Chen ◽  
...  
Keyword(s):  
Finite Element ◽  
Residual Stresses ◽  
Nuclear Power ◽  
Power Plants ◽  
Three Dimensional ◽  
Fe Model ◽  
Dimensional Finite Element ◽  
Surge Line ◽  
Line Slope ◽  
Three Dimensional Finite Element

The objective of this study is to correct installation non-conformance of a surge line using the excavation and re-weld method which is widely used in nuclear power plants. The surge line with a backslope was not at the required design level after initial installation. In order to solve the problem, a repairing technology is shown as follows: the weld was successively excavated and welded again while the surge line slope was corrected with the help of jacks. Because many of the degradation mechanisms relevant to power plant components can be accelerated by the presence of welding residual stresses (WRS), the WRS caused by the repairing process need to be studied. In this paper, the WRS simulation technique employed in this project is sophisticated. It utilizes a 3-D finite element (FE) model, and simulates the weld sequencing and excavation. Moreover, the WRS simulation performed in this project not only uses the un-axisymmetric model, but also considers the deformation caused by the external jacking loads. The results show that the repairing process is effective, and strain damage induced by the welding repair is also acceptable.

Plastic Limit Loads for Through-Wall Cracked Pipes Using Finite Element Limit Analysis

2006 ◽  
Vol 321-323 ◽  
pp. 724-728
Author(s):  
Nam Su Huh ◽  
Yoon Suk Chang ◽  
Young Jin Kim
Keyword(s):  
Finite Element ◽  
Limit Analysis ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Limit Load ◽  
Plastic Behavior ◽  
Plastic Limit ◽  
Integrity Assessment ◽  
Perfectly Plastic ◽  
Elastic Perfectly Plastic

The present paper provides plastic limit load solutions for axial and circumferential through-wall cracked pipes based on detailed three-dimensional (3-D) finite element (FE) limit analysis using elastic-perfectly plastic behavior. As a loading condition, both single and combined loadings are considered. Being based on detailed 3-D FE limit analysis, the present solutions are believed to be valuable information for structural integrity assessment of cracked pipes.

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.

Effect of The Sigma Phase on the Mechanical Properties of a Cast Duplex Stainless Steel during the Ageing Treatment at 850°C

2013 ◽  
Vol 684 ◽  
pp. 325-329 ◽  
Author(s):  
Tian Liang ◽  
Xiao Qiang Hu ◽  
Xiu Hong Kang ◽  
Dian Zhong Li
Keyword(s):  
Mechanical Properties ◽  
Stainless Steel ◽  
Duplex Stainless Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Sigma Phase ◽  
Volume Fraction ◽  
Optical Microscope ◽  
Eutectoid Reaction ◽  
Secondary Austenite

With about equal amount of austenite and ferrite in volume fraction, duplex stainless steel (DSS) is in advantage of mechanical properties and corrosive behaviors. Hence it is widely applied to the heavy castings for nuclear power plants inshore, such as impellers, pumps and valves. However, lots of cracks usually occur in these castings during manufacturing processes, because it is susceptible to precipitate the brittle intermetallic compound of sigma phase when the castings are exposed from 600 to 1000oC. In this work, the precipitation of sigma phase was observed by optical microscope (OM) and scanning electron microscope (SEM) in a cast DSS named as MAS/6001, which aged at 850oC from 5 to 300 minutes. The effect of sigma phase on the mechanical properties was analyzed by the tensile at room temperature and impact tests at -10°C. The results show that sigma phase in the MAS/6001 steel precipitated simultaneously with the secondary austenite, which obeyed the eutectoid reaction. The interfaces between austenite or secondary austenite and sigma phase were the locations where cracks generated from the void aggregation. Cracks are susceptible to propagate along or cross these interfaces, and to promote the sigma phase breaking-off, which severely deteriorated the mechanical properties.

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