3-Dimensional Finite Element T-Stress Calculation for Reactor Pressure Vessel Nozzle Blend Radius Semi-Elliptical Surface Cracks

2015 ◽  
Author(s):  
Minghao Qin ◽  
Shu Tang ◽  
Francis Ku ◽  
Daniel V. Sommerville ◽  
Hal Gustin
Keyword(s):  
Finite Element ◽  
Stress Intensity ◽  
Plane Strain ◽  
Plane Stress ◽  
Pressure Vessel ◽  
Crack Tip ◽  
Reactor Pressure Vessel ◽  
T Stress ◽  
Crack Tip Constraint ◽  
Reactor Pressure

T-stress is used as an indicator of the condition of crack tip constraint. In current fracture mechanics engineering applications in the U.S. nuclear industry, T-stress generally has been ignored during the calculation of applied stress intensity factors (SIF). Consideration of this crack tip constraint component could affect the evaluation of material fracture behavior, under either plane strain or plane stress or plane strain and plane stress combination. When the T-stress shows that the condition of crack front constraint is not plane strain, incorporation of T-stress may allow reduction of unnecessary conservatisms in such calculations. Under this condition, the allowable stress intensity factor is modified by increasing it above the KIc value, and it potentially increases the predicted allowable flaw sizes. In this paper, T-stress has been calculated using 3-D finite element analyses (FEA) with a typical semi-elliptical crack in a reactor pressure vessel (RPV) nozzle blend radius. Both thermal and internal pressure load cases are considered. To verify this finite element analysis approach, this method is applied to comparable literature models. The FEA results are consistent with closed-form solutions for T-stress calculation.

scholarly journals Integrity Evaluation of a Reactor Pressure Vessel Based on a Sequential Abaqus-FRANC3D Simulation Method

2021 ◽  
Vol 2021 ◽  
pp. 1-12
Author(s):  
M. Annor-Nyarko ◽  
Hong Xia
Keyword(s):  
Finite Element ◽  
Stress Intensity ◽  
Pressure Vessel ◽  
Simulation Method ◽  
Reactor Pressure Vessel ◽  
Injection System ◽  
Intensity Factors ◽  
3D Fem ◽  
Ageing Management ◽  
Reactor Pressure

The safety-risk pressurized thermal shock (PTS) have on a reactor pressure vessel (RPV) is one of the most important studies for the lifetime ageing management of a reactor. Several studies have investigated PTS induced by postulated accidents and other anticipated transients. However, there is no study that analyzes the effect of PTS induced by one of the most frequent anticipated operational occurrences—inadvertent operation of the safety injection system. In this paper, a sequential Abaqus-FRANC3D simulation method is proposed to study the integrity status of an ageing pressurized water reactor subjected to PTS induced by inadvertent actuation of the safety injection system. A sequential thermal-mechanical coupling analysis is first performed using a three-dimensional reactor pressure vessel finite element model (3D-FEM). A linear elastic fracture mechanics submodel with a postulated semielliptical surface crack is then created from the 3D-FEM. Subsequently, the submodel is used to evaluate the stress intensity factors based on the M-integral approach coupled within the proposed simulation method. Finally, the stress intensity factors (SIFs) obtained using the proposed method are compared with the conventional extended finite element method approach, and the result shows a good agreement. The maximal thermomechanical stress concentration was observed at the inlet nozzle-inner wall intersection. In addition, The ASME fracture toughness of the reactor vessel’s steel compared with SIFs show that the PTS event and crack configuration analysed may not pose a risk to the integrity of the RPV. This work serves as a critical reference for the ageing management and fatigue life prediction of reactor pressure vessels.

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.

Structural Integrity of Penetration Nozzle of Reactor Pressure Vessel Head of PWR Plant

2008 ◽  
Author(s):  
Kiminobu Hojo ◽  
Naoki Ogawa ◽  
Yoichi Iwamoto ◽  
Kazutoshi Ohoto ◽  
Seiji Asada ◽  
...  
Keyword(s):  
Residual Stress ◽  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Intensity Factor ◽  
Stress Field ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Complex Structure ◽  
Reactor Pressure Vessel ◽  
Reactor Pressure

A reactor pressure vessel (RPV) head of PWR has penetration holes for the CRDM nozzles, which are connected with the vessel head by J-shaped welds. It is well-known that there is high residual stress field in vicinity of the J-shaped weld and this has potentiality of PWSCC degradation. For assuring stress integrity of welding part of the penetration nozzle of the RPV, it is necessary to evaluate precise residual stress and stress intensity factor based on the stress field. To calculate stress intensity factor K, the most acceptable procedure is numerical analysis, but the penetration nozzle is very complex structure and such a direct procedure takes a lot of time. This paper describes applicability of simplified K calculation method from handbooks by comparing with K values from finite element analysis, especially mentioning crack modeling. According to the verified K values in this paper, fatigue crack extension analysis and brittle fracture evaluation by operation load were performed for initial crack due to PWSCC and finally structural integrity of the penetration nozzle of RPV head was confirmed.

scholarly journals Effect of cladding on the stress intensity factors in the reactor pressure vessel

2000 ◽  
Vol 199 (1-2) ◽  
pp. 101-111 ◽  
Author(s):  
S.N. Choi ◽  
K.S. Jang ◽  
J.S. Kim ◽  
J.B. Choi ◽  
Y.J. Kim
Keyword(s):  
Stress Intensity ◽  
Stress Intensity Factors ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Intensity Factors ◽  
Reactor Pressure

Assessment of a Stress-Free Temperature Model for Residual Stresses in Surface Cladding of a Reactor Pressure Vessel

2017 ◽  
Author(s):  
B. Richard Bass ◽  
Paul T. Williams ◽  
Terry L. Dickson ◽  
Hilda B. Klasky
Keyword(s):  
Residual Stress ◽  
Finite Element ◽  
Thermal Expansion ◽  
Residual Stresses ◽  
Base Metal ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Normal Operation ◽  
Inner Surface ◽  
Reactor Pressure

This paper describes further results from an ongoing study of a simplified engineering model that is intended to account for effects of clad residual stresses on the propensity for initiation of preexisting inner-surface flaws in a commercial nuclear reactor pressure vessel (RPV). The deposition of stainless steel cladding during fabrication of an RPV generates residual stresses in the cladding and the heat affected zone of the under-lying base metal. In addition to residual stress, thermal strains are generated by the differential thermal expansion (DTE) of the cladding and base material due to temperature changes during normal operation. A simplified model used in the ORNL-developed FAVOR probabilistic fracture mechanics (PFM) code accounts for the clad residual stress by incorporating a stress-free temperature (SFT) approach. At the stress-free temperature (Ts-free), the model assumes there is no thermal strain, i.e., the thermal expansion stresses and clad residual stresses offset each other. For normal cool-down transients applied to the RPV, interactions of the latter stresses generate additional crack driving forces on shallow, internal surface-breaking flaws near the clad/base metal interface; those flaws tend to dominate the RPV failure probability computed by FAVOR. In a previous report from this study (PVP2015-45086), finite element analysis was used to compare the stresses and stress-intensity factors (SIF) during a cool-down transient for two cases: (1) the existing SFT model of FAVOR, and (2) directly applied RPV clad residual stress (CRS) distribution obtained from empirical (hole-drilling) measurements made at room temperature on an RPV that was never put into service. However, those analyses were limited in scope and focused on a single flaw orientation. In this updated study, effects of CRS on the SIF histories computed for both circumferential and axial flaw orientations subjected to a cool-down transient were determined from an extended set of finite element analyses. Specifically, comparisons were made between results from applying CRS experimental data to ABAQUS two-dimensional, inner-surface flaw models and those generated by the FAVOR SFT model. It is demonstrated that the FAVOR-recommended SFT value of 488 °F produces conservatively high values of SIF relative to the use of CRS profiles in the ABAQUS models. For the vessel and flaw geometry and transient under study, the circumferential flaw (360° continuous) required a decrease of SFT down to 390 °F to match the CRS SIF histories. For the infinite axial flaw model, a decrease down to 300 °F matched the CRS SIF histories. Future plans are described to develop more general conclusions regarding the FAVOR model.

Fracture Mechanics Fatigue Evaluation of a Flowline Clamp Connector Using Finite Element Modeling of a Crack

2019 ◽  
Author(s):  
Curtis Sifford ◽  
Ali Shirani
Keyword(s):  
Finite Element Analysis ◽  
Stress Intensity Factor ◽  
Finite Element ◽  
Fatigue Life ◽  
Stress Intensity ◽  
Fracture Mechanics ◽  
Intensity Factor ◽  
Pressure Vessel ◽  
Crack Tip ◽  
Element Analysis

Abstract This paper presents the application of the rules from ASME Section VIII, Division 3 of the ASME Boiler and Pressure Vessel Code for a fracture mechanics evaluation to determine the damage tolerance and fatigue life of a flowline clamp connector. The guidelines from API 579-1 / ASME FFS-1 Fitness-For-Service for the stress analysis of a crack-like flaw have been considered for this assessment. The crack tip is modeled using a refined mesh around the crack tip that is referred to as a focused mesh approach in API 579-1 / ASME FFS-1. The driving force method is used as an alternative to the failure assessment diagram method to account for the influence of crack tip plasticity. The J integral is determined using elastic-plastic finite element analysis and converted to an equivalent stress intensity factor to be compared to the fracture toughness of the material. The fatigue life is calculated using the Paris Law equation and the stress intensity factor calculated from the finite element analysis. The allowable number of design cycles is determined using the safety factors required from Division 3 of the ASME Pressure Vessel Code.

Fracture Mechanics Fatigue Evaluation of a Flowline Clamp Connector Using Finite Element Modeling of a Crack

2018 ◽  
Author(s):  
Curtis Sifford ◽  
Ali Shirani
Keyword(s):  
Finite Element Analysis ◽  
Stress Intensity Factor ◽  
Finite Element ◽  
Fatigue Life ◽  
Stress Intensity ◽  
Fracture Mechanics ◽  
Intensity Factor ◽  
Pressure Vessel ◽  
Crack Tip ◽  
Element Analysis

This paper presents the application of the rules from ASME Section VIII, Division 3 of the ASME Boiler and Pressure Vessel Code for a fracture mechanics evaluation to determine the damage tolerance and fatigue life of a flowline clamp connector. The guidelines from API 579-1 / ASME FFS-1 Fitness-For-Service for the stress analysis of a crack-like flaw have been considered for this assessment. The crack tip is modeled using a refined mesh around the crack tip that is referred to as a focused mesh approach in API 579-1 / ASME FFS-1. The driving force method is used as an alternative to the failure assessment diagram method to account for the influence of crack tip plasticity. The J integral is determined using elastic-plastic finite element analysis and converted to an equivalent stress intensity factor to be compared to the fracture toughness of the material. The fatigue life is calculated using the Paris Law equation and the stress intensity factor calculated from the finite element analysis. The allowable number of design cycles is determined using the safety factors required from Division 3 of the ASME Pressure Vessel Code.

Development of a Three-Dimensional Green’s Function and Its Application to the Fatigue Evaluation of Reactor Pressure Vessel

2006 ◽  
Author(s):  
M. Y. Ahn ◽  
J. C. Kim ◽  
Y. S. Chang ◽  
J. B. Choi ◽  
Y. J. Kim ◽  
...  
Keyword(s):  
Finite Element Analysis ◽  
Finite Element ◽  
Fatigue Life ◽  
Green’S Function ◽  
Green's Function ◽  
Pressure Vessel ◽  
Three Dimensional ◽  
Reactor Pressure Vessel ◽  
Element Analysis ◽  
Reactor Pressure

The design of major nuclear components for the prevention of fatigue failure has been achieved on the basis of ASME codes, which are usually very conservative. However, it is necessary to make it more accurate for the continued operation beyond the design life. In this paper, 3-dimensional stress and fatigue analyses reflecting entire geometry have been carried out. The number of operating transient data obtained from a monitoring system were filtered and analyzed. Then, Green’s function which transfers temperature gradient into the corresponding thermal stress is proposed and applied to critical locations of a reactor pressure vessel. The validity of proposed Green’s function is approved by comparing the result with corresponding 3-D finite element analysis results. Also, the amount of conservatism included in design transients in comparison with real transients is analyzed. The results for 3-D finite element analysis are also compared with corresponding 2-D finite element analysis results, and a considerable amount of difference was observed in terms of fatigue life. Therefore, it is expected that the proposed evaluation scheme adopting real operating data and Green’s function can provide more accurate fatigue life evaluation for a reactor pressure vessel.

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