Safety Assessment of Reactor Pressure Vessel Integrity for Loss of Coolant Accident Conditions

2011 ◽  
Vol 134 (1) ◽  
Author(s):  
Dieter Beukelmann ◽  
Wenfeng Guo ◽  
Wieland Holzer ◽  
Robert Kauer ◽  
Wolfgang Münch ◽  
...  
Keyword(s):  
Finite Element ◽  
Pressure Vessel ◽  
Safety Margin ◽  
Reactor Pressure Vessel ◽  
Test Facility ◽  
Temperature Distributions ◽  
Loss Of Coolant Accident ◽  
System Pressure ◽  
Loss Of Coolant ◽  
Reactor Pressure

One of the critical issues for reactor pressure vessel (RPV) structural integrity is related to the pressurized thermal shock (PTS) event. Therefore, within the framework of safety assessments special emphasis is given to the effect of PTS-loadings caused by the nonuniform azimuthal temperature distribution due to cold water plumes or stripes during emergency coolant injection. This paper describes the method used to predict the thermal mechanic boundary conditions (system pressure and wall temperature). Using a system code the pressure and global temperature distributions were calculated, systematically varying the leak size and the location of the coolant water injection. Spatial and temporal temperature distributions in the main circulation pipes and at the RPV wall were predicted by mixing analyses with a computational fluid dynamics (CFD) code. The model used for these calculations was validated by post-test calculations of a UPTF (upper plenum test facility) experiment simulating cold leg injection during a small break loss of coolant accident (LOCA). Comparison with measured temperatures showed that the modeling used is suitable to obtain enveloping results. Fracture mechanics analyses were carried out for circumferential flaw sizes in the weld joint near the core region and between the RPV shell and the flange, as well as for axial flaws in the nozzle corner. Stress intensity factors KI were calculated numerically using the finite element program ansys and analytically on the basis of weight and polynomial influence functions using stresses obtained from elastic finite element analyses. Benchmark tests revealed good agreement between the results from numerical and analytical calculations. For all regions of the RPV investigated and the most severe transients it was demonstrated that a large safety margin against brittle crack initiation exists and brittle fracture of the RPV can be excluded.

Safety Assessment of Reactor Pressure Vessel Integrity for Loss of Coolant Accident Conditions

2010 ◽  
Author(s):  
D. Beukelmann ◽  
W. Guo ◽  
W. Holzer ◽  
R. Kauer ◽  
W. Mu¨nch ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Finite Element ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Test Facility ◽  
Worst Case ◽  
Temperature Distributions ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Reactor Pressure

One of the critical issues for Reactor Pressure Vessel (RPV) structural integrity is related to the Pressurized Thermal Shock (PTS) event. Therefore, within the framework of safety assessments special emphasis is given to the effect of PTS-loadings caused by the non-uniform azimuthal temperature distribution due to cold water plumes or stripes during emergency coolant injection. The paper describes the method used to predict the thermal mechanic boundary conditions (system pressure, wall temperature). Using a system code the pressure and global temperature distributions were calculated, systematically varying the leak size and the location of the coolant water injection. Local and temporal temperature distributions in the main circulation pipes and at the RPV wall were predicted by mixing analyses with a Computational Fluid Dynamics (CFD) code. The model used for these calculations was validated by post-test calculations of a UPTF (Upper Plenum Test Facility) experiment simulating cold leg injection during a small break Loss of Coolant Accident (LOCA). Comparison with measured temperatures showed that the modelling used is suitable to obtain bounding results. Fracture mechanics analyses were carried out for circumferential flaw sizes in the weld joint near the core region and between the RPV shell and the flange, as well as for axial flaws in the nozzle corner. Stress intensity factors KI were calculated numerically using the finite element program ANSYS and analytically on the basis of weight and polynomial influence functions using stresses obtained from elastic finite element analyses. Benchmark tests revealed good agreement between the results from numerical and analytical calculations. In order to determine the worst case loading conditions a wide spectrum of thermal-hydraulic transients was considered. Since the resulting load paths decrease with lower temperatures after a maximum, the warm prestress (WPS) effect was employed. The fracture toughness curve determined by deeply notched specimens with high constraint is not representative of the nozzle corner due to the considerable loss of constraint at LOCA conditions. Hence the influence of constraint on fracture toughness was accounted applying the constraint modified master curve concept and the relationship between the T-stress and the reference temperature T0. According to ASME Code Cases N-629 and N-631 the reference temperatures RTNDT and RTT0 can be used alternatively for the adjustment of the KIC-curve. Therefore both the RTNDT- and the RTT0-concept were considered. For all regions of the RPV investigated and the most severe transients it was demonstrated that a large safety margin against crack initiation exists and brittle fracture of the RPV can be excluded.

An Experimental Investigation on Thermal-Hydraulic Interaction Between Core Makeup Tank and Safety Injection Tank of an Integral Small Reactor During a Small Break Loss-of-Coolant Accident

2015 ◽  
Author(s):  
Hwang Bae ◽  
Sung Uk Ryu ◽  
Hyo Bong Ryu ◽  
Woo Shik Kim ◽  
Sung-Jae Yi ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Injection System ◽  
Test Facility ◽  
Passive Safety ◽  
Injection Test ◽  
Loss Of Coolant Accident ◽  
Test Loop ◽  
Loss Of Coolant ◽  
Reactor Pressure

A passive injection test was conducted using a core makeup tank (CMT), a safety injection tank (SIT) and an automatic depressurization system (ADS), which consists of a passive safety system (PSS) of the SMART reactor. This paper investigates the thermal-hydraulic interaction between CMT and SIT during sequential injections of coolant from these two tanks to a high-temperature and high-pressure reactor pressure vessel using an integral effect test facility of SMART-ITL (System-Integrated Modular Advanced ReacTor-Integral Test Loop). Both CMT and SIT were connected to the reactor pressure vessel by a pressure balance line (PBL) and injection line (IL). A steady-state condition was maintained for 1,000 seconds before the start of the injection. The major parameters agreed well with the target value. After one of safety injection system line was simulated to be broken, a transient injection test was conducted according to the small-break loss-of-coolant accident (SBLOCA) scenario. Coolant injections from a CMT and SIT were started sequentially by opening quick-opening valves installed on the IL and PBL piping, respectively. Several thermal-hydraulic phenomena such as direct contact condensation, thermal stratification, and coupling effects between the CMT and SIT were locally observed during the SBLOCA scenario. The results show that the adopted passive safety injection system functions well as an emergency core cooling system.

Application of the BEREMIN Model for Evidencing the Warm Pre-Stress Effect in an Irradiated Reactor Pressure Vessel Containing a Semi-Elliptical Subclad Defect Under Small and Large LOCA Conditions

2008 ◽  
Author(s):  
Ph. Gilles ◽  
J.-P. Izard ◽  
J. Devaux
Keyword(s):  
Brittle Fracture ◽  
Pressure Vessel ◽  
Fracture Resistance ◽  
Thermal Fluctuations ◽  
Reactor Pressure Vessel ◽  
Stress Effect ◽  
Load History ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Reactor Pressure

The nuclear power plants lifetime is strongly dependent of the guarantee of the reactor pressure vessel (RPV) integrity. Therefore, the RPV integrity has to be demonstrated under the most severe configuration, namely the Pressurized Thermal Shock induced by the Loss of Coolant Accident induced by a large break in the primary loop. For such a transient, the apparent risk of failure is maximum when the load is decreasing; the fracture resistance decreasing rate being stronger. However, such type of loading generates an increase of the fracture resistance as shown by numerous studies (Chell, 1980 – BEREMIN, 1981 – Smith et al., 2004). This is known as the warm pre-stress (WPS) effect. This beneficial effect on the resistance to brittle fracture is not accounted for in the French RCCM and RSEM codes (RCCM, 2000 – RSEM, 2005). EDF has launched several R&D actions with CEA and AREVA as well as with European partners (SMILE, 2001) to validate and model the WPS effect under RPV representative conditions. Proving the existence of this beneficial load history effect (designated as Warm Pre Stress WPS), in the case of a defective RPV in emergency and faulted conditions is the aim of the present paper. The demonstration is conducted in the case of cleavage fracture using an improved version of the BEREMIN model. As opposite to the classical Fracture Mechanics methodology, this approach allows to account for load history effects on cleavage. The study analyzes the behavior of a semi-elliptical under clad crack in the EoL core shell of a 900 MWe RPV for two loading cases: the large break Loss Of Coolant Accident transient and a small break LOCA inducing thermal fluctuations on the vessel inner wall. The WPS effect is evidenced by comparing the plasticity corrected SIF levels of two loadings for the same value of failure probability: the considered WPS loading and a virtual monotonously increasing load applied at the temperature at which the brittle fracture risk is estimated.

3D FEM Modeling of Crack Initiation in a Reactor Pressure Vessel During a PTS Event (LBLOCA)

2018 ◽  
Author(s):  
Diego Fernando Mora Méndez ◽  
Markus Niffenegger ◽  
Guian Qian ◽  
Michal Jaros ◽  
Bojan Niceno
Keyword(s):  
Stress Intensity Factor ◽  
Finite Element ◽  
Stress Intensity ◽  
Intensity Factor ◽  
Pressure Vessel ◽  
3D Fem ◽  
Integrity Assessment ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Reactor Pressure

To perform the integrity assessment of a reactor pressure vessel (RPV) related to Pressurized Thermal Shock (PTS), we model the RPV using the 3D finite element method (FEM). Accurate prediction of temperature and stress fields is determined by using 2-Phase computational fluid dynamics (CFD) simulation in combination with an appropriate finite element discretization of the RPV wall. The cladding and the ferritic low alloy steel are considered as two separated layers, which can be intersected by superficial cracks. The calculation of the stress intensity factor (SIF) in mode I is based on the linear fracture mechanics theory (LEFM) and hypothetical cracks are located in different locations to consider the most critical cases. In the present study, the sub-modeling technique is implemented to refine the mesh required by the fracture analysis in the region of interest. Three types of cracks are considered: axial, circumferential and inclined. The performed integrity assessment uses the master curve approach. The stress intensity factor in the deepest point of a surface crack was compared with the material’s fracture toughness. In previous studies the integrity of the RPV subjected to medium and small break Loss-of-Coolant Accident (MBLOCA and SBLOCA, respectively) has been evaluated, therefore the concern in this contribution is the large break of Loss-Of-Coolant Accident (LBLOCA). The combination of 3D FEM with CFD simulations allows us to study the influence of the dynamic cooling plume on the stress intensity in more detail than with the conventional one dimensional method.

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.

Bounding Acoustic Load Development for Boiling Water Reactor Shroud and Shroud Supports

2020 ◽  
Author(s):  
Matthew Walter ◽  
Minghao Qin ◽  
Daniel Sommerville
Keyword(s):  
Pressure Vessel ◽  
Reactor Core ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Structural Evaluation ◽  
Loss Of Coolant Accident ◽  
Specific Analysis ◽  
Reactor Pressure

Abstract As part of the license basis of a nuclear boiling water reactor pressure vessel, a sudden loss of coolant accident (LOCA) event needs to be analyzed. One of the loads that results from this event is a sudden depressurization of the recirculation line. This leads to an acoustic wave that propagates through the reactor coolant and impacts several structures inside the reactor pressure vessel (RPV). The authors have previously published a PVP paper (PVP2015-45769) which provides a survey of LOCA acoustic loads on boiling water reactor core shrouds. Acoustic loads are required for structural evaluation of core shrouds; therefore, a defensible load is required. The previous research compiled plant-specific data that was available at the time. Since then, additional data has become available which will add to the robustness of the bounding load methodology that was developed. Investigations are also made regarding the shroud support to RPV weld, which was neglected from the previous study. This will allow a practitioner a convenient method to calculate bounding acoustic loads on all shroud and shroud support welds in the absence of a plant-specific analysis.

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.

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