CFD-Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions: Influence of Turbulence Model and Mesh Refinement on the Vessel Thermal Loading During PTS Transient

2006 ◽  
Author(s):  
A. Martin ◽  
S. Benhamadouche ◽  
G. Bezdikian ◽  
F. Beaud ◽  
F. Lestang
Keyword(s):  
Heat Transfer ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Fluid Temperature ◽  
Hydraulic Analysis ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Thermal Hydraulic Analysis ◽  
Reactor Pressure

Integrity evaluation methods for nuclear Reactor Pressure Vessels (RPVs) under Pressurised Thermal Shock (PTS) loading are applied by French Utility. They are based on the analysis of the behavior of relatively shallow cracks under loading PTS conditions due to the emergency cooling during SBLOCA (Small Break Loss of Coolant Accident) transients. This paper presents the Research and Development program started at E.D.F on the Computational Fluid Dynamic (CFD) determination of the cooling phenomena of a PWR vessel during a Pressurised Thermal Shock. The numerical results are obtained with the thermal-hydraulic tool Code_Saturne, in combination with the thermal-solid code SYRTHES to take into account the coupled effect of heat transfer between the fluid flow and the vessel. Based on the global and local Thermal-hydraulic analysis of a Small Break Loss of Coolant Accident transient, the paper presents mainly a parametric study which helps to understand the main phenomena that can lead to better estimating the margin factors. The geometry studied represents a third of a PWR pressure vessel and the configuration investigated is related to the injection of cold water in the vessel during a SBLOCA transient. Conservative initial and boundary conditions for the CFD calculation are derived from the global Thermal-hydraulic analysis. Both the fluid behavior and its impact on the solid part formed by cladding and base metal are considered. The main purpose of the numerical thermal-hydraulic studies is to accurately estimate the distribution of fluid temperature in the down comer and the heat transfer coefficients on the inner RPV surface for a fracture mechanics computation which will subsequently assess the associated RPV safety margin factors.

CFD-Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions: Influence of Temperature Safety Injection and Fluid-Structure Thermal Coupling

2005 ◽  
Author(s):  
A. Martin ◽  
D. Monfort ◽  
G. Bezdikian ◽  
F. Beaud ◽  
F. Lestang ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Fluid Temperature ◽  
Hydraulic Analysis ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Thermal Hydraulic Analysis ◽  
Reactor Pressure

Integrity evaluation methods for nuclear Reactor Pressure Vessels (RPVs) under Pressurised Thermal Shock (PTS) loading are applied by French Utility. They are based on the analysis of the behavior of relatively shallow cracks under loading PTS conditions due to the emergency cooling during SBLOCA (Small Break Loss of Coolant Accident) transients. This paper presents the Research and Development program started at E.D.F on the Computational Fluid Dynamic (CFD) determination of the cooling phenomena of a PWR vessel during a Pressurised Thermal Shock. The numerical results are obtained with the thermal-hydraulic tool Code_Saturne, in combination with the thermal-solid code SYRTHES to take into account the coupled effect of heat transfer between the fluid flow and the vessel. Based on the global and local Thermal-hydraulic analysis of a Small Break Loss of Coolant Accident transient, the paper presents mainly a parametric study whose purpose is to understand the main phenomena which can lead to better estimating the margin factors. The geometry studied represents a third of a PWR pressure vessel and the configuration investigated is related to the injection of cold water in the vessel during a SBLOCA transient. Conservative initial and boundary conditions for the CFD calculation are derived from the global Thermal-hydraulic analysis. Both the fluid behavior and its impact on the solid part formed by cladding and base metal are considered. On the whole, the main purpose of the numerical thermal-hydraulic studies is to accurately estimate the distribution of fluid temperature in the down comer and the heat transfer coefficients on the inner RPV surface for a fracture mechanics computation which will subsequently assess the associated RPV safety margin factors.

CFD Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions: Influence of Turbulence Model and Mesh Refinement on the Vessel Thermal Loading During PTS Transient

2011 ◽  
Vol 133 (3) ◽  
Author(s):  
A. Martin ◽  
S. Benhamadouche ◽  
G. Bezdikian ◽  
F. Beaud ◽  
F. Lestang
Keyword(s):  
Heat Transfer ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Hydraulic Analysis ◽  
Loss Of Coolant Accident ◽  
Pressurized Thermal Shock ◽  
Loss Of Coolant ◽  
Thermal Hydraulic Analysis ◽  
Reactor Pressure

Integrity evaluation methods for nuclear reactor pressure vessels (RPVs) under pressurized thermal shock (PTS) loading are applied by French Utility. They are based on the analysis of the behavior of relatively shallow cracks under loading PTS conditions due to the emergency cooling during small break loss of coolant accident (SBLOCA) transients. This paper presents the research and development program started at EDF on the computational fluid dynamics (CFD) determination of the cooling phenomena of a PWR vessel during a pressurized thermal shock. The numerical results are obtained with the thermal-hydraulic tool Code̱Saturne, in combination with the thermal-solid code SYRTHES to take into account the coupled effect of heat transfer between the fluid flow and the vessel. Based on the global and local thermal-hydraulic analysis of a small break loss of coolant accident transient, this paper presents mainly a parametric study that helps to understand the main phenomena that can lead to better estimating the margin factors. The geometry studied represents a third of a PWR pressure vessel, and the configuration investigated is related to the injection of cold water in the vessel during a SBLOCA transient. Conservative initial and boundary conditions for the CFD calculation are derived from the global thermal-hydraulic analysis. Both the fluid behavior and its impact on the solid part formed by cladding and base metal are considered. The main purpose of the numerical thermal-hydraulic studies is to accurately estimate the distribution of fluid temperature in the downcomer and the heat transfer coefficients on the inner RPV surface for a fracture mechanics computation, which will subsequently assess the associated RPV safety margin factors.

Use of a CFD-Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressurised Thermal Shock Conditions: Thermal-Hydraulic Studies of a Safety Injection in a PWR Plant — Methodology and Present Limitation

2007 ◽  
Author(s):  
A. Martin ◽  
S. Bosse ◽  
F. Lestang
Keyword(s):  
Heat Transfer ◽  
Thermal Shock ◽  
Thermal Loading ◽  
Development Program ◽  
Pressure Vessels ◽  
Surveillance Program ◽  
Fluid Temperature ◽  
Two Phase ◽  
Safety Margins ◽  
Reactor Pressure

Integrity evaluation methods for nuclear Reactor Pressure Vessels (RPVs) under Pressurised Thermal Shock (PTS) loading are applied by French Utility. They are based on the analysis of the behaviour of cracks under PTS loading conditions due to the emergency cooling during PTS transient like SBLOCA. This paper explains the Research and Development program started at Electricite´ De France about the cooling phenomena of a PWR vessel after a Pressurised Thermal Shock. The numerical results are obtained with the E.D.F ThermalHydraulic code (Code_Saturne) coupled with the thermal-solid code SYRTHES to take into account the conjugate heat transfer on the cooling of the vessel. We first explain the global methodology with a progress report on the state of the art of the tools available to simulate the different scenari displayed within the frame of the plant life project in order to reassess the integrity of the RPV, taking into account the evolution of some input data, such as the new value of end of life (EOL) fluence, the feedback results of surveillance program and the evolution of the functional requirements. The main results are presented and are related to the evaluation of the RPV integrity during a Small Break Loss Of Coolant Accident transient for 900 and 1300 MWe nuclear plant. On the whole, the main purpose of the numerical CFD studies is to accurately estimate the distribution of fluid temperature in the down comer and the heat transfer coefficients on the inner RPV surface for a fracture mechanics computation which will subsequently assess the associated RPV safety margins. In a second time, a new analysis is performed to assess an accurate temperature distribution in the RPV. Indeed, from a physical phenomena point of view, the EDF thermalhydraulic tool Code_Saturne is now qualified in order to assess single phase transient but in the case where the cold legs are partially filled with steam, it becomes a two-phase problem and new important effects occur, such as condensation due to the emergency core cooling injections of sub-cooled water. Thus, an advanced prediction of RPV thermal loading during these transients requires sophisticated two-phase, local scale, 3D codes. In that purpose, a program has been set up to extend the capabilities of the Neptune_CFD two-phase solver which is the tool able to solve two phase flow configuration. In a same time, A simplified approach has showed that for a type of transient weakly uncovered, a free surface calculation was sufficient to respect the necessary criteria of safety. A Qualification study was carried out on the Hybiscus experimental E.D.F facility, representing a cold leg with ECC injection and a third down comer. Temperature profiles have been compared and are presented and analysed here, showing encouraging results.

CFD-Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions for Lifetime Evaluation: Methodology and Qualification Task of the EDF Numerical Tools

2003 ◽  
Author(s):  
A. Martin ◽  
F. Beaud ◽  
F. Ternon Morin ◽  
T. Veneau
Keyword(s):  
Heat Transfer ◽  
Thermal Shock ◽  
Development Program ◽  
Pressure Vessels ◽  
Numerical Results ◽  
Evaluation Methodology ◽  
Fluid Temperature ◽  
Safety Margins ◽  
Numerical Tools ◽  
Reactor Pressure

Integrity evaluation methods for nuclear Reactor Pressure Vessels (RPVs) under Pressurised Thermal Shock (PTS) loading are applied by French Utility. They are based on the analysis of the behaviour of relatively shallow cracks under loading PTS conditions due to the emergency cooling during SBLOCA transients. This paper presents the Research and Development program started at E.D.F on the CFD determination of the cooling phenomena of a PWR vessel during a Pressurised Thermal Shock. The numerical results are obtained with the thermalhydraulic tools N3S and Code_Saturne, in combination with the thermal-solid code SYRTHES to take into account the coupled effect of heat transfer between the fluid flow and the vessel. We first explain the recent improvement of the thermalhydraulic analysis with the global definition of the SBLOCA transient and the local analysis in the downcomer. Then, the qualification task of the EDF numerical tools is described. In order to reach this purpose, we have investigated several configurations related to an injection of cold water and focused our analysis particularly in the cold leg but also in a the downcomer. Two experiment test cases have been studied. A comparison between experiment and numerical results in terms of temperature field is presented. On the whole, the main purpose of the numerical thermalhydraulic studies is to accurately estimate the distribution of fluid temperature in the downcomer and the heat transfer coefficients on the inner RPV surface for a fracture mechanics computation which will subsequently assess the associated RPV safety margins.

Re-Evaluation of the Structural Integrity of the Chinese Qinshan 300-MWe Reactor Pressure Vessel Under Pressurized Thermal Shock Using the Master Curve Method

2015 ◽  
Author(s):  
Yupeng Cao ◽  
Yinbiao He ◽  
Hu Hui ◽  
Hui Li ◽  
Fuzhen Xuan
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Master Curve ◽  
Structural Integrity ◽  
Master Curve Method ◽  
Loss Of Coolant Accident ◽  
Pressurized Thermal Shock ◽  
Loss Of Coolant ◽  
Curve Method ◽  
Reactor Pressure

Pressurized thermal shock (PTS) is a potential major threat to the structural integrity of the reactor pressure vessel (RPV) in a nuclear power plant. A comprehensive structural integrity analysis of the Chinese Qinshan 300-MWe RPV subjected to PTS events including the small break loss-of-coolant accident (SB-LOCA) and large break loss-of-coolant accident (LB-LOCA) transients was performed by Shanghai nuclear engineering and design institute (SNERDI). The J-integral values at the deepest and the near cladding-base interface points of the crack were calculated with the linear elastic material model. And the RTPTS values were determined by the tangent approach. In the case that the RTNDT at or beyond the RPV design life may exceed the RTPTS according to the previous analysis procedure, the objective of this paper is to apply the Master Curve method to the re-evaluation of the integrity of this RPV, taking account of constraint and crack length effects. The over-conservatism in the previous evaluation is identified by comparing the new calculation with the previous one. The new RTPTS values are increased to varied extents for the different loading transients.

Measurements of Stress During Thermal Shock in Clad Reactor Pressure Vessel Material Using Time-Resolved In-Situ Synchrotron X-Ray Diffraction

2018 ◽  
Author(s):  
Sam Oliver ◽  
Chris Simpson ◽  
Andrew James ◽  
Christina Reinhard ◽  
David Collins ◽  
...  
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
X Ray Diffraction ◽  
Time Resolved ◽  
Vessel Steel ◽  
X Ray ◽  
Reactor Pressure

Nuclear reactor pressure vessels must be able to withstand thermal shock due to emergency cooling during a loss of coolant accident. Demonstrating structural integrity during thermal shock is difficult due to the complex interaction between thermal stress, residual stress, and stress caused by internal pressure. Finite element and analytic approaches exist to calculate the combined stress, but validation is limited. This study describes an experiment which aims to measure stress in a slice of clad reactor pressure vessel during thermal shock using time-resolved synchrotron X-ray diffraction. A test rig was designed to subject specimens to thermal shock, whilst simultaneously enabling synchrotron X-ray diffraction measurements of strain. The specimens were extracted from a block of SA508 Grade 4N reactor pressure vessel steel clad with Alloy 82 nickel-base alloy. Surface cracks were machined in the cladding. Electric heaters heat the specimens to 350°C and then the surface of the cladding is quenched in a bath of cold water, representing thermal shock. Six specimens were subjected to thermal shock on beamline I12 at Diamond Light Source, the UK’s national synchrotron X-ray facility. Time-resolved strain was measured during thermal shock at a single point close to the crack tip at a sample rate of 30 Hz. Hence, stress intensity factor vs time was calculated assuming K-controlled near-tip stress fields. This work describes the experimental method and presents some key results from a preliminary analysis of the data.

CFD-Tools for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions: Thermal-Hydraulic Methods and Main Industrial Results on the French 900 MWe PWR RPV

2009 ◽  
Author(s):  
A. Martin ◽  
F. Lestang ◽  
S. Bellet ◽  
D. Guichard ◽  
C. Vit ◽  
...  
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Cold Water ◽  
Reactor Pressure Vessel ◽  
Fluid Temperature ◽  
Time Duration ◽  
Two Phase ◽  
Safety Margins ◽  
Reactor Pressure ◽  
Fast Fracture

For the Reactor Pressure Vessel (RPV) assessment and lifetime evaluation of the nuclear plants, French Utilities apply a series of calculations including thermal-hydraulic, thermo mechanical and fracture mechanics studies in order to study the Pressurized Thermal Shock (PTS) in the downcomer caused by the safety injection. Within the frame of the plant lifetime project, integrity assessments of the French 900 MWe (3-loops) series RPV have been performed. A gain for safety margins to fast fracture of the RPV can be found with a 3D modeling of thermal-hydraulics loads. From a physical phenomena point of view, the results of the system code analysis (CATHARE computation) of the PTS transient induce two kinds of scenarios: single phase and two-phase flows in the cold leg. In the case where the cold legs are partially filled with steam, it becomes a two-phase problem and new important effects occur. Thus, an advanced prediction of RPV thermal loading during these transients requires sophisticated two-phase, local scale, 3D codes. For that purpose, a program has been set up to extend the capabilities of the NEPTUNE_CFD two-phase solver which is the tool able to solve two-phase flow configuration. At the same time, a simplified approach has shown that for this kind of scenario where the cold leg is weakly uncovered, a free surface calculation (without phase change) was sufficient to respect the necessary criteria of safety. Considering the time duration of 3D computation and the large number of cases, EDF and AREVA-NP decided to share the effort. The two teams use the NEPTUNE_CFD code (coupled with the thermal solid SYRTHES code) for thermal-hydraulic computations. The thermo mechanical code used is CALORI. According to this approach and to reduce the CPU time, two computations have been performed for 2″ and 3″ Small Break Loss Of Coolant Accident (SBLOCA) on a one-third RPV model. Computations on a complete RPV model have been performed to demonstrate the relevance of the one-third RPV model. The studies have been performed by two independent teams from EDF and AREVA-NP. The investigated configuration corresponds to the injection of cold water in the RPV during a penalizing representation of a primary break transient and its impact on the solid part formed by cladding and base metal. Numerical results are given in terms of fluid temperature in the cold legs and in the downcomer. The obtained numerical description of the transient is used as boundary conditions for a full mechanical computation of the stresses. The results show that such a complete thermal-hydraulic and mechanical 3-dimensional analysis improves the evaluation of the consequences of the loading on the stress fields and eventually the margins to fast fracture of the RPV. The good agreement observed between a one-third RPV model and a complete RPV model results confirms the validity of the approach.

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