The Use of Local Approach to Fracture in Reactor Pressure Vessel Structural Integrity Assessment: Synthesis of a Cooperative Research Program Between EDF, CEA, Framatome and AEA Technology

2009 ◽  
pp. 284-284-31 ◽  
Author(s):  
D Moinereau ◽  
JM Frund ◽  
J Brochar ◽  
MP Valeta ◽  
B Marini ◽  
...  
Keyword(s):  
Research Program ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Cooperative Research ◽  
Local Approach ◽  
Integrity Assessment ◽  
Cooperative Research Program ◽  
Reactor Pressure

Constraint Effect On Fracture Behavior of Underclad Crack in Reactor Pressure Vessel

2021 ◽  
Author(s):  
Masaki Shimodaira ◽  
Tohru Tobita ◽  
Hisashi Takamizawa ◽  
Jinya Katsuyama ◽  
Satoshi Hanawa
Keyword(s):  
Fracture Toughness ◽  
Surface Crack ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Compact Tension ◽  
Reactor Pressure Vessel ◽  
Local Approach ◽  
Integrity Assessment ◽  
Plastic Constraint ◽  
Reactor Pressure

Abstract According to JEAC4206-2016, in the structural integrity assessment of a reactor pressure vessel (RPV), the fracture toughness (KJc) should be higher than the stress intensity factor at the crack tip of a postulated underclad crack (UCC) near the inner surface of the RPV during a pressurized thermal shock event. Previous analytical studies show that the plastic constraint for UCC is lower than that for surface crack. Consequently, the apparent KJc for UCC is expected to be higher than that for surface crack. In this study, we performed three-point bending fracture toughness tests and finite element analyses (FEAs) for RPV steel containing a UCC or a surface crack to quantitatively investigate the effect of cladding on the plastic constraint and subsequent KJc evaluation. From the tests, we found that the apparent KJc for the UCC was considerably higher than that for the surface crack. Such a high KJc could be explained by the lower plastic constraint parameters, such as T-stress and Q-parameter, of the UCC compared with those for the surface crack. Additionally, local approach analysis showed that the KJc for the UCC was significantly higher than the master curve estimated from the fracture toughness tests using compact tension specimens.

Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Direct Coupling Between Probabilistic and Finite-Element Codes to Model Sensitivity to Key Thermo-Hydraulic Variability

2004 ◽  
Author(s):  
Etienne de Rocquigny ◽  
Yoan Chevalier ◽  
Silvia Turato ◽  
Eric Meister
Keyword(s):  
Finite Element ◽  
Pressure Vessel ◽  
Mechanical Model ◽  
Analytical Approach ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Direct Coupling ◽  
Integrity Assessment ◽  
Key Variables ◽  
Reactor Pressure

The structural integrity assessment of a nuclear Reactor Pressure Vessel (RPV) during accidental conditions such as loss-of-coolant accident (LOCA) is a major safety concern. Besides conventional deterministic calculations to justify as a nuclear operator the RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Probabilistic analyses become most interesting when some key variables, albeit conventionally taken at conservative values, can be modelled more accurately through statistical variability. In the context of low failure probabilities, this requires however a specific coupling effort between a specific probabilistic analysis method (e.g. Form-Sorm method) and the thermo-mechanical model to be reasonable in computing time. In this paper, the variability of a key variable — the mid-transient cooling temperature, tied to a climate-dependent tank — has been modelled, in some flaw configurations (axial sub-clad) for a French vessel. In a first step, a simplified analytical approach was carried out to assess its sensitivity upon the thermo-mechanical phenomena; hence, a direct coupling had to be implemented to allow a probabilistic calculation on the finite-element mechanical model, taking also into account a failure event properly defined through minimisation of the instantaneous failure margin during the transient. Comparison with the previous (indirectly-coupled) studies and the simplified analytical approach is drawn, demonstrating the interest of this new modelling effort to understand and order the sensitivity of the probability of crack initiation to the key variables. While being noticeable in the cases studied, sensitivity to the safety injection temperature variability proves to be less than the choice of the toughness model. Finally, regularity of the thermo-mechanical model is evidenced by the coupling exercise, suggesting that a modified response-surface based method could replace direct coupling for further investigation.

Deterministic and Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Benchmark Comparisons

2003 ◽  
Author(s):  
Silvia Turato ◽  
Vincent Venturini ◽  
Eric Meister ◽  
B. Richard Bass ◽  
Terry L. Dickson ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Safety Concern ◽  
Nuclear Regulatory Commission ◽  
Reactor Pressure Vessel ◽  
Oak Ridge ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Reactor Pressure

The structural integrity assessment of a nuclear Reactor Pressure Vessel (RPV) during accidental conditions, such as loss-of-coolant accident (LOCA), is a major safety concern. Besides Conventional deterministic calculations to justify the RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Since in the USA the probabilistic fracture mechanics analyses are accepted by the Nuclear Regulatory Commission (NRC), a benchmark has been realized between EDF and Oak Ridge Structural Assessments, Inc. (ORSA) to compare the models and the computational methodologies used in respective deterministic and probabilistic fracture mechanics analyses. Six cases involving two distinct transients imposed on RPVs containing specific flaw configurations (two axial subclad, two circumferential surface-breaking, and two axial surface-braking flaw configurations) were defined for a French vessel. In two separate phases, deterministic and probabilistic, fracture mechanics analyses were performed for these six cases.

Reactor Pressure Vessel Integrity in Light of the Evolution of Materials Science and Engineering

2005 ◽  
Vol 473-474 ◽  
pp. 287-292
Author(s):  
Péter Trampus
Keyword(s):  
Pressure Vessel ◽  
Structural Integrity ◽  
Materials Science ◽  
Reactor Pressure Vessel ◽  
Special Focus ◽  
Science And Engineering ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Materials Science And Engineering ◽  
Reactor Pressure

Structural integrity of the reactor pressure vessel of pressurized water reactors is one of the key safety issues in nuclear power operation. Integrity may be jeopardized during operational transients. The problem is compounded by radiation damage of the vessel structural materials. Structural integrity assessment as an interdisciplinary field is primarily based on materials science and fracture mechanics. The paper gives an overview on the service induced damage processes and associated changes of mechanical properties, the prediction of degradation and the assessment of the entire component against brittle fracture with a special focus on how the evolution of materials science and engineering has contributed to reactor vessel structural integrity assessment.

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.

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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