CFD-Tool for Thermal-Hydraulics Pressurised Thermal Shock Analysis: Qualification of the Code_Saturne

2002 ◽  
Author(s):  
A. Martin ◽  
S. Bellet
Keyword(s):  
Thermal Shock ◽  
Cold Water ◽  
Life Time ◽  
Test Case ◽  
Fluid Temperature ◽  
Thermal Hydraulics ◽  
Numerical Tools ◽  
Physical Phenomena ◽  
Numerical Program ◽  
Reactor Pressure

This paper explains the numerical program concerning the new thermalhydraulic Code_Saturne qualification for Safety Injection studies. Within the frame of the plant life time project, an analysis has shown that the most severe loading conditions are generated by a pressurised injection of cold water in the downcomer of a Reactor Pressure Vessel. For this kind of transients, a thermal hydraulics study has to be carried out in order to better adjust the accurate distribution of the fluid temperature in the downcomer. For that, the numerical tools have to be able to simulate the physical phenomena present during the Pressurised Thermal Shock. (PTS). For this qualification task, we have investigated one configuration related to an injection of cold water particularly in cold leg but also in a downcomer. One experiment test case has been studied and this paper gives a comparison between experiment and numerical results in terms of temperature field.

CFD-Tools Qualification for Thermal-Hydraulics Pressurized Thermal Shock Analysis

2003 ◽  
Vol 125 (4) ◽  
pp. 418-424 ◽  
Author(s):  
A. Martin ◽  
S. Bellet
Keyword(s):  
Thermal Shock ◽  
Cold Water ◽  
Life Time ◽  
Test Cases ◽  
Fluid Temperature ◽  
Thermal Hydraulics ◽  
Pressurized Thermal Shock ◽  
Numerical Tools ◽  
Physical Phenomena ◽  
Numerical Program

This paper explains the numerical program concerning the E.D.F. CFD tools qualification for Safety Injection studies. Within the frame of the plant life time project, an analysis has shown that the most severe loading conditions are generated by a pressurized injection of cold water in the downcomer of a Reactor Pressure Vessel. For this kind of transients, a thermal hydraulics study has to be carried out in order to better adjust the accurate distribution of the fluid temperature in the downcomer. For that, the numerical tools have to be able to simulate the physical phenomena present during the Pressurized Thermal Shock (PTS). For this qualification task, we have investigated several configurations related to an injection of cold water particularly in cold leg but also in a downcomer. Two experiment test cases have been studied and this paper gives a comparison between experiment and numerical results in terms of temperature field.

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.

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.

XFEM and Standard Fracture-Mechanics Analyses of the Reactor Pressure Vessel Gösgen Based on Thermal-Hydraulics Input Data From KWU-MIX and CFD Analyses

2019 ◽  
Author(s):  
Alexander Mutz ◽  
Tomas Nicak ◽  
Richard Trewin ◽  
Ingo Cremer
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Brittle Failure ◽  
Input Data ◽  
Cold Water ◽  
Reactor Pressure Vessel ◽  
Thermal Hydraulics ◽  
Mechanics Analysis ◽  
Reactor Pressure

Abstract The integrity of a reactor pressure vessel (RPV) has to be ensured throughout its entire life in accordance with the applicable regulations. Typically an assessment of the RPV against brittle failure needs to be conducted by taking into account all possible loading cases. One of the most severe loading cases, which can potentially occur during the operating time, is the loss-of-coolant accident, where cold water is injected into the RPV at operating conditions. High pressure in combination with a thermal shock of the ferritic pressure vessel wall caused by the injection of cold water leads to a considerable load at the belt-line area known as Pressurized Thermal Shock (PTS). Usually the assessment against brittle failure is based on a deterministic fracture-mechanics analysis, in which common parameters like J-integral or stress intensity factor are employed to calculate the load path for an assumed (postulated) flaw during the PTS event. As an alternative to this standard approach a fracture mechanics assessments based on eXtended Finite Element Method (XFEM) approach can be performed. The most important input data for the fracture-mechanics analysis is the transient thermal-hydraulics (TH) load of the RPV during the emergency cooling. Such data can be calculated by analytical fluid-mixing codes verified on experiments, such as KWU-MIX, or by numerical Computational Fluid Dynamics (CFD) tools after suitable validation. In KWU-MIX, which is the standard used for TH calculations within PTS analyses, rather conservative analytical models for the quantification of mixing and, depending on the water level, condensation processes in the downcomer (including simplified stripe and plume formations) are utilized. On the contrary, the numerical CFD tools can provide best-estimate results due to the possibility to consider more realistically the stripe and plume formations as well as the geometry of the RPV in detail. In a previous paper [1] results of standard and XFEM analyses of the RPV Gösgen 1 based on thermal-hydraulics input data from KWU-MIX were presented. This paper presents new results based on thermal-hydraulics input data from CFD. The new results are compared with those from [1] in order to show additional safety margins obtained by using thermal-hydraulics input data from CFD.

Engineering Approach for the Reactor Pressure Vessel Integrity in PTS Conditions: Thermal Hydraulic and Fracture Mechanics Studies of a Safety Injection in a Three Loop Power Plant — French Benchmark With EDF and AREVA-NP Tools

2008 ◽  
Author(s):  
A. Martin ◽  
A. Dahl ◽  
F. Lestang ◽  
S. Bellet ◽  
F. Beaud ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Cold Water ◽  
Life Time ◽  
Reactor Pressure Vessel ◽  
Fluid Temperature ◽  
Circumferential Welds ◽  
3 Dimensional ◽  
Vessel Integrity ◽  
Reactor Pressure

For the Reactor Pressure Vessel (RPV) assessment and Lifetime evaluation of the nuclear plants, French Utility applies a series of calculations including thermal-hydraulic, thermo mechanical and fracture mechanics studies in order to study the Pressurized Thermal Shock (PTS) in the down comer caused by the safety injection. Within the frame of the plant life time project, integrity assessments of the French 900 MWe (3-loops) series reactor pressure vessel (RPV) have been performed. We found that the modeling of thermal-hydraulics loads is a source of gain. Considering the length of local 3D calculation and the large number of cases, E.D.F and AREVA-NP decided to share the effort. However the two chains of software differ: EDF uses Code_Saturne (coupled with the thermal solid code Syrthes) and Cuve 1D and AREVA-NP uses Star_CD and CALORI codes respectively for thermal hydraulic and thermo mechanical computations. According to this approach, comparison between the two chains of tools have been performed. Moreover this action contributes to the verification and the validation of each code in accordance with the OECD Best Practise Guidelines (BPG). The study has been achieved by two independent teams from EDF and AREVA-NP. It should be emphasized that this benchmark helped to strengthen the accuracy of CFD and the adapted methodology (working progress). The investigated configuration corresponds to the injection of cold water in the vessel 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 and velocity fields in the cold legs and in the downcomer. The obtained numerical description of the transient (internal pressure and temperature field within the vessel) is used as boundary conditions for a full mechanical computation of the stresses. This thermal mechanical transient is obtained on a 3D mesh of the vessel, covering the two core shells and their circumferential welds, as well as the internal cladding. The results show that such a complete thermal–hydraulic and mechanic 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 EDF and AREVA-NP results and their accordance with the validation computations, confirm the robustness of the approach.

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.

Fracture-Mechanics Analysis of the Reactor Pressure Vessel Beznau 1 Based on Thermal-Hydraulics Input Data From KWU-MIX and CFD Analyses

2016 ◽  
Author(s):  
Tomas Nicak ◽  
Richard Trewin ◽  
Elisabeth Keim ◽  
Ingo Cremer ◽  
Sebastien Blasset ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Brittle Failure ◽  
Input Data ◽  
Cold Water ◽  
Reactor Pressure Vessel ◽  
Mechanics Analysis ◽  
Cfd Analyses ◽  
Reactor Pressure

The integrity of a reactor pressure vessel (RPV) has to be ensured throughout its entire life in accordance with the applicable regulations. Typically an assessment of the RPV against brittle failure needs to be conducted by taking into account all possible loading cases. One of the most severe loading cases, which can potentially occur during the operating time, is the loss-of-coolant accident, where cold water is injected into the RPV nearly at operating conditions. High pressure in combination with a thermal shock of the ferritic pressure vessel wall caused by the injection of cold water leads to a considerable load at the belt-line area known as Pressurized Thermal Shock (PTS). Usually the assessment against brittle failure is based on a deterministic fracture-mechanics analysis, in which common parameters like J-integral or stress intensity factor are employed to calculate the load path for an assumed (postulated) flaw during the PTS event. The most important input data for the fracture-mechanics analysis is the transient thermal-hydraulics (TH) load of the RPV during the emergency cooling. Such data can be calculated by analytical fluid-mixing codes verified on experiments, such as KWU-MIX, or by numerical Computational Fluid Dynamics (CFD) tools after suitable validation. In KWU-MIX, which is the standard used for TH calculations within PTS analyses, rather conservative analytical models for the quantification of mixing and, depending on the water level, condensation processes in the downcomer (including simplified stripe and plume formations) are utilized. On the contrary, the numerical CFD tools can provide best-estimate results due to the possibility to consider more realistically the stripe and plume formations as well as the geometry of the RPV in detail. In order to quantify the safety margin inherent to the standard approach, two fracture-mechanics analyses of the RPV Beznau 1 based on thermal-hydraulic input data from KWU-MIX and CFD analyses were performed. Subsequently the resulting loading paths were compared between each other and with material properties obtained from the irradiation surveillance program of the RPV to demonstrate the exclusion of brittle-fracture initiation.

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 Tools for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions: Thermalhydraulic Benchmark on the Validation Test Case UPTF

2010 ◽  
Author(s):  
A. Martin ◽  
F. Huvelin ◽  
G. Balard ◽  
S. Bellet ◽  
B. Durand ◽  
...  
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Single Phase ◽  
Physical Phenomenon ◽  
Reactor Pressure Vessel ◽  
Test Facility ◽  
Test Case ◽  
Phase Flow ◽  
Two Phase ◽  
Reactor Pressure

The Reactor Pressure Vessel (RPV) assessment and lifetime evaluation of the nuclear plants leads EDF and AREVA to use CFD tools for the demonstration of the Reactor Pressure Vessel (RPV) integrity. French PWR plants have been studied for a long time and so a large number of computations have been performed in three steps: System — CFD — Mechanical Analysis. This paper focuses on the CFD aspects and especially on the two different CFD tools imposed by the physical phenomena of the transient scenarii. Indeed, the results of the system code analysis (CATHARE computations) of the PTS transient induce two kinds of scenarii: single phase and two-phase flows in the cold leg. According to the approach considered, it was interesting to carry out a global benchmark with the different thermalhydraulic tools. In this frame, Code_Saturne and Star CD CFD tools used for single phase flow scenario configuration and NEPTUNE_CFD used for two phase flow configuration were confronted with UPTF (Upper Plenum Test Facility) experiment results in a single phase scenario. The choice to perform this benchmark on UPTF test case was retained because this integral test was very representative of the main physical phenomenon that is well suited for the validation of the CFD tools used to study the Pressurised Thermal Shock for the RPV integrity.

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.

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