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.

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.

Recent Improvements in Toughness Prediction of Cast Duplex Stainless Steel Components

2019 ◽  
Author(s):  
Patrick Le Delliou ◽  
Sébastien Saillet
Keyword(s):  
Stainless Steel ◽  
Duplex Stainless Steel ◽  
Power Plants ◽  
Impact Test ◽  
Charpy Impact ◽  
Charpy Impact Test ◽  
Thermal Ageing ◽  
Life Extension ◽  
Term Operation ◽  
Safety Margins

Abstract Thermal ageing of cast duplex stainless steel components 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. One major input of the assessment methodology is the toughness value of the thermally aged component. Recent work conducted at EDF R&D to improve the accuracy and the conservativeness of the toughness prediction has led to the development of new prediction formulae. The toughness prediction relies on three steps: • estimation of the Charpy impact test values at 20 and 320°C using the chemical composition of the steel and the aging conditions (temperature and duration), • estimation of the J-R curve at 20 and 320°C - defined by a power law J = CΔan - thanks to correlations between n and C and the Charpy impact test values, • estimation of the J-R curve at any temperature between 20 and 320°C thanks to interpolation formulae. The paper presents the experimental data used to develop the formulae, the formulae themselves and some elements of validation.

Integrity and Life Assessment of Cast Duplex Stainless Steel Elbows of EDF PWR Main Coolant Piping

2015 ◽  
Author(s):  
Patrick Le Delliou ◽  
Sébastien Saillet
Keyword(s):  
Fracture Toughness ◽  
Stainless Steel ◽  
Duplex Stainless Steel ◽  
Power Plants ◽  
Estimation Method ◽  
Thermal Ageing ◽  
Life Assessment ◽  
Life Extension ◽  
Manufacturing Quality ◽  
Safety Margins

Thermal ageing of cast duplex stainless steel elbows 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 reports the present integrity and life assessment methodologies as carried out by EDF. The approach is based on the in-service inspection and surveillance RSE-M Code and on French regulation requirements for NPPs in operation. This work is supported by an extensive R&D programme on one hand and on field experience analysis on the other hand. The paper details the three main topics of the life assessment methodology: - estimation of the fracture toughness of the steel with predictive formulae using the chemical composition and ageing conditions, - definition of a reference crack size based on an inventory of the manufacturing quality of the elbows, - fracture mechanics evaluation based on the J parameter, computed either by an engineering estimation method or by a finite element analysis. The calculated J parameter is then compared with the estimated fracture toughness of the material. Partial safety coefficients are included in the calculation process as required by the RSE-M Code.

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.

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.

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.

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.

Allowable Flaw Size of Japanese Cast Stainless Steel Pipe Using Probabilistic Fracture Mechanics Method

2017 ◽  
Author(s):  
Shotaro Hayashi ◽  
Mayumi Ochi ◽  
Kiminobu Hojo ◽  
Takahisa Yamane ◽  
Wataru Nishi
Keyword(s):  
Stainless Steel ◽  
Fracture Mechanics ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Thermal Ageing ◽  
Probabilistic Fracture Mechanics ◽  
Stainless Steel Pipe ◽  
Technical Basis ◽  
Cast Stainless Steel

The cast austenitic stainless steel (CASS) that is used for the primary loop pipes of nuclear power plants is susceptible to thermal ageing during plant operation. The Japanese JSME rules on fitness-for-service (JSME rules on FFS)[1] for nuclear power plants specify the allowable flaw depths. However, some of these allowable flaw sizes are small compared with the smallest flaw sizes, which can be detected by nondestructive testing. ASME Section XI Code Case N-838[2] recently specified the maximum tolerable flaw depths for CASS pipes determined by probabilistic fracture mechanics (PFM). In a similar way, the allowable flaw depths of CASS pipes were calculated by PFM analysis code “PREFACE”[3] which considers uncertainty of the mechanical properties of Japanese PWR CASS materials. In order to confirm the validity of PREFACE, the allowable flaw depths calculated by PREFACE were compared with the maximum tolerable flaw depths in the technical basis of Code Case N-838. As a result, although the J calculation method and the embrittlement prediction model of CASS are different, these were qualitatively consistent. In addition, the sensitivity of ferrite content to the allowable flaw depths was investigated.

Investigation of Thermal Ageing in Long Term Operated RPV Materials

2015 ◽  
Author(s):  
Charles C. Eiselt ◽  
Günter König ◽  
Hieronymus Hein ◽  
Maxim Selektor ◽  
Martin Widera
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Elevated Temperatures ◽  
Energy Levels ◽  
Thermal Ageing ◽  
Experimental Investigations ◽  
Low Alloy Steels ◽  
Term Operation ◽  
The Impact

The phenomenon of thermal ageing of low alloy steels comes more into focus in terms of long term operation of nuclear power plants (NPP). Safety-relevant components such as the RPV or the pressurizer have to bear the respective loads at elevated temperatures for longer times. However the mechanical properties of the applied materials might experience certain degradations such as a decrease of the impact energy levels and a shift in the ductile to brittle transition temperature (e.g. T41) leading to higher ductile-brittle reference temperatures and a reduction of material toughness. In terms of a safe long term operation it is important to understand in how far thermal ageing alone, meaning for the RPV without the cumulative damaging effects through neutron irradiation, has detrimental influences on the respective materials of interest. First of all an overview is provided of the current state of the art with respect to thermal ageing by describing influencing mechanisms, its implementation into different nuclear codes, standards and selected experimental investigations in this field. Following this, the test results of the thermal surveillance sets from three German PWRs are presented and discussed. The tested Charpy-V specimens, taken from representative RPV base and weld metals (22NiMoCr3-7 / NiCrMo1UP) as well as their heat affected zones, were exposed to ∼290°C for ∼30 years on the cold leg of the according plants’ main coolant loops. The obtained results are compared with the existing thermal aging data base (baseline and ∼7 years data) of the materials concerned. Finally, the role of thermal ageing particularly with respect to RPV irradiation surveillance will be assessed.

Validation of Constraint Based Methodology in Structural Integrity – VOCALIST: Experimental Programme

2002 ◽  
Author(s):  
Richard Bass ◽  
Ulrich Eisele ◽  
Elisabeth Keim ◽  
Heikki Keinanen ◽  
Ste´phane Marie ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Experimental Investigations ◽  
Integrity Assessment ◽  
Experimental Programme ◽  
Safety Margins ◽  
Materials Used ◽  
Shallow Crack

The aim of VOCALIST (Validation of Constraint-Based Assessment Methodology in Structural Integrity) is to develop and validate innovative procedures for assessing the level of, and possible changes to, constraint-related safety margins in ageing pressure boundary components [1]. An iterative process of experiment and analysis will address this overall objective. The experimental investigations within VOCALIST are performed on three different materials representing the as new state of materials used for components of nuclear power plants as well as a state representing an in service degraded state of RPV materials. Within the experimental programme fracture mechanics specimens with different constraint situations are tested in order to quantify the influence of the constraint on the specimens failure behaviour as a basis for the advanced components integrity assessment. The investigations are performed on small laboratory specimens as well as on biaxially loaded cruciform specimens and large piping components. Within this contribution the experimental programme of VOCALIST is introduced. The investigated materials are characterized in terms of their mechanical properties. Special consideration is given to results of fracture mechanics specimens highlighting the constraint effect via the shallow crack effect and its contribution to a shift of the master curve.

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