Georges Bezdikian

Analysis of a bending test on a full-scale PWR hot leg elbow containing a surface crack

EDF, in co-operation with Framatome, has conducted a large research programme on the mechanical behaviour of thermally aged cast duplex stainless steel elbows, which are part of the main primary circuit of French PWRs. One important task of this programme consisted of testing a full-scale PWR hot leg elbow. The elbow contained a semi-elliptical circumferential notch machined on the outer surface of the intrados, as well as casting defects located on the flanks. To simulate the end-of-life condition of the component regarding material toughness, it had previously undergone a 2400-h ageing heat treatment at 400°C. The test preparation and execution, as well as the material characterization programme, were handled by MPA. The test was conducted under constant internal pressure and in-plane bending (opening mode) at 200°C. For safety reasons, it took place at an open air-site: the Meppen military test ground. At the maximum applied moment (6000 kN m), the notch did not initiate. This paper presents the results of the experiment and the results of the fracture mechanics analysis, based on finite element calculations.

SMILE: Validation of the Warm-Pre-Stress Effect With a Cylindrical Thick-Walled Specimen

The Reactor Pressure Vessel (RPV) is an essential component, which is liable to limit the lifetime duration of PWR plants. The assessment of defects in RPV subjected to pressurized thermal shock (PTS) transients made at an European level generally does not necessarily consider the beneficial effect of the load history (Warm Pre-stress, WPS). The SMILE project — Structural Margin Improvements in aged embrittled RPV with Load history Effects — aims to give sufficient elements to demonstrate, to model and to validate the beneficial WPS effect. It also aims to harmonize the different approaches in the national codes and standards regarding the inclusion of the WPS effect in a RPV structural integrity assessment. The project includes significant experimental work on WPS type experiments with C(T) specimens and a PTS type transient experiment on a large component. This paper deals with the results of the PTS type transient experiment on a component-like specimen subjected to WPS- loading, the so called Validation Test, carried out within the framework of work package WP4. The test specimen consists of a cylindrical thick walled specimen with a thickness of 40mm and an outer diameter of 160mm, provided with an internal fully circumferential crack with a depth of about 15mm. The specified load path type is Load-Cool-Unload-Fracture (LCUF). No crack initiation occurred during cooling (thermal shock loading) although the loading path crossed the fracture toughness curve in the transition region. The benefit of the WPS-effect by final reloading up to fracture in the lower shelf region, was shown clearly. The corresponding fracture load during reloading in the lower shelf region was significantly higher than the crack initiation values of the original material in the lower shelf region. The post test fractographic evaluation showed that the fracture mode was predominantly cleavage fracture also with some secondary cracks emanating from major crack.Copyright

Life Management of Reactor Coolant Piping — Probabilistic Calibration of Partial Safety Factors in Flaw Acceptance Criteria: Application to Cast Duplex Stainless Steel Components

The RSE-M Code provides rules and requirements for in-service inspection of the components of the French PWR power plant. The Code gives non mandatory guidance for analytical evaluation of flaws, comprising fracture mechanics analyses based on engineering methods, flaw acceptance criteria and codification of material characteristics. Based on a probabilistic calibration methodology, partial safety factors on the main random variables involved in flaw assessments (loading, crack size, yield strength and material toughness) are given in Appendix 5.5 of the Code for each category of operating conditions (A, C or D) and for the possible failure modes (ductile tearing or brittle fracture). These partial safety factors should be used with the material characteristics specified in Appendix 5.6 of the Code, to insure the consistency of the methodology. The criteria of the RSE-M Code have been implemented for the acceptance of generic flaws in cast duplex stainless steel elbows of the Reactor Coolant System. Statistical analyses of the mechanical properties of the base-metal have been carried out to get their characteristic values consistent with the Code criteria: • tensile properties comprising yield strength, ultimate tensile strength and non dimensional reference true stress - true strain curves taking into account thermal ageing, • value of the J-integral in the ductile regime at the onset of crack extension (J0.2 after 0.2 mm of crack extension), J-Δa curves in the ductile regime taking into account thermal ageing in the hot leg conditions, • fatigue crack growth rates. The results show that the aged cast duplex stainless steel elbows satisfy the Code criteria for each category of operating conditions.Copyright

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

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.Copyright

Reactor Pressure Vessel and Reactor Coolant Circuit Cast Duplex Stainless Steel Components: Contribution of the Expertise for Life Management Studies

The life management of French Nuclear Power Plants is a major stake from an economic and a technical point of view considering the aging management assessment of the key components of the plant. The actual life evaluation is the results of prediction of life assessment from important program of expertises for the 3-loop PWR and 4-loop PWR plants in operation. To optimize the strategic in order to achieve the best possible performance and to prepare the technical and economical choice and decision, the paper presents the association of life management strategy and the program of expertises considering: • the identification of degradation for different components and prediction criteria proposed, • the large database from cast reactor coolant and component removed from nuclear power plants and expertised to confirm the prediction, • the life evaluation of RPV with radiation surveillance program based on the expertises of irradiation capsules, it is particularly shown how the expertise is in the center of the strategic choice. The French utility has organized the life management of Nuclear Plant in function of several program of expertises of knowledge on the long term experience feedback and the maintenance program for life. This paper shows updated on RPV and reactor coolant equipments activities engaged by utility on: • periodic maintenance and volume of expertise, • Alternative maintenance actions, • Large volume of expertises and how are managed these results to predict the aging management.Copyright

Full-Scale Test on an Aged Cast Duplex Stainless Steel Lateral Connection: Results and Analysis

Some components (elbows, pump casings and lateral connections) of the primary loop of French PWRs are made of static cast duplex stainless steels. This kind of steel may age even at relatively low temperatures (in the temperature range of PWR service conditions), depending on the material composition. An important consequence of this ageing process is the decrease in the ductility and fracture toughness of the material. It is feared that an embrittlement, associated with the occurrence of casting defects, may increase the risk of failure. Therefore, an extensive programme has been launched by EDF in co-operation with Framatome ANP, in order to determine acceptability criteria for operating cast stainless steel components. This programme relies on a large R&D effort, involving metallurgical studies, large-scale experiments, development of specific finite element tools and J-estimation schemes, and research of methods to assess the ageing state of in-service components. This paper presents the main characteristics and results of an experiment conducted on an aged cast 45 degree lateral connection. This connection contained a machined notch at the acute corner and was tested under internal pressure. The chemical composition was chosen to obtain a fast thermal ageing and low fracture toughness properties. During the test, the defect initiated and grew subsequently by ductile tearing. The test showed that it was possible to obtain a significant amount of stable crack growth (about 2.5 mm) despite the low toughness properties of the aged material. The pressure reached at the end of the test was about twice the in-service pressure. A detailed fracture mechanics analysis, based on finite element calculations, was performed. These calculations fairly simulated the overall behaviour of the tested structure, gave a conservative prediction of the crack initiation pressure and well predicted the crack size associated with the maximum pressure. These tests and their detailed analyses contribute to validate and justify the methodology used in the integrity assessment of in-service cast duplex stainless steel components.Copyright

French Nuclear Power Plants Life Management Ageing Management Process: Matrix of Degradation Evaluation for Strategic Decision Combined With Components

The life management of French Nuclear Power Plants is an important issue and a major stake considering the aging management assessment of the key components of the plant, both from an economic and a technical point of view. The actual life evaluation is: • The first 3-loop PWR plants have 25 years in operation, • The first 4-loop PWR plants have 19 years in operation. To optimize the best and strategic choice in order to achieve the best possible performance and to prepare the technical and economical choice and decision considering: • the mode of degradation for different components, • the industrial capacity and capability to replace or to repair components, • the life evaluation of components with time limit in operation, the French utility has organized the life management of Nuclear Plant in function of several program of knowledge on degradation mode from the large and long term experience feedback and the maintenance program for life management based on normal and periodic maintenance actions and exceptional maintenance carried out on strategic components. This paper shows the table of degradation mode and different actions engaged by utility on: • In-Service Inspection, periodic maintenance, • Alternative maintenance actions, • Exceptional program of maintenance based on several mitigation and strategic replacement of components. This paper provides the reader with an overview of how advanced information processing techniques to the improvement of in-service inspections, condition-based maintenance, and asset management.Copyright

French PWR Nuclear Power Plants Steam Generators and Reactor Pressure Vessel Heads Aging Management Criteria and Replacement Strategy

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.© 2004 ASME

RPV Integrity Assessment: Comparison Between 3 Loops and 4 Loops Studies

For the French utility (Electricite de France–EDF), Nuclear Energy represents 75% of generation of the total electric energy in France. Total nuclear electricity were generated mainly from Nuclear Power plants stations, 34 PWR NPPs 3-loop Reactors- 900 MWe, 20 PWR NPPs 4-loop Reactors- 1300 MWe and 4 PWR NPPs 4-loop Reactors- 1450 MWe. The 3-loop Reactor Pressure Vessel (RPV) integrity assessment, applied on 34 PWR NPPs Reactors, involved the verification of the integrity of the component under the most severe conditions of situation, and the result obtained was the justification of the 900 MWe RPV life management for at least 40 years and to prepare the projection beyond 40 years. Since 2000, in the continuity of these results, the studies were carried out on the 20 PWR NPPs 4-loop 1300 MWe Reactor Pressure Vessels, and the recent results obtained show the demonstration of the integrity of the RPV, in the most severe conditions of loading in relation with RTNDT (Reference Nil Ductility Transition Temperature), and other major parameters. This approach is based on specific mechanical safety studies on the RPV to demonstrate the absence of the risk of failure by brittle fracture. For these mechanical studies the major input data are necessary: 1 - the fluence distribution and the values of 3-loop and 4-loop RPV, 2 - RTNDT during the lifetime in operation, 3 - the temperature distribution values in the downcomer and the PTS evaluation. The main results must show significant margins against initiation of brittle fracture for all of 3-loop and 4-loop RPV. The flaws considered in this approach are shallow flaws beneath the cladding (subclad flaws) or in the first layer of cladding. The major tasks and expertises engaged by EDF are: • more precise assessment of the fluence calculations, • better knowledge of the vessel material properties, including the effect of radiation, • the NDE inspection program on the core zone. The comparison of the results are developed in this paper: • for the fluence evaluation and the optimisation of the fuel management, • the data gathered from radiation specimen capsules, removed from the vessels (radiation surveillance program), • and the thermal-hydraulic and mechanical calculations based on finite element thermal-hydraulic and 3D elastic-plastic mechanical computations.Copyright

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

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.© 2003 ASME