Crack Arrest in Water-Cooled Reactor Pressure Vessels During Loss-of-Coolant Accident Conditions

2009 ◽  
pp. 422-422-10
Author(s):  
TU Marston ◽  
E Smith ◽  
KE Stahlkopf
Keyword(s):  
Crack Arrest ◽  
Pressure Vessels ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Accident Conditions ◽  
Reactor Pressure

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.

French RPV PTS Assessment: An Overview of EDF Research and Development in Progress on Thermalhydraulic, Materials and Mechanical Aspects

2003 ◽  
Author(s):  
Dominique Moinereau ◽  
Jean-Michel Frund ◽  
Henriette Churier-Bossennec ◽  
Georges Bezdikian ◽  
Alain Martin
Keyword(s):  
Structural Integrity ◽  
Three Dimensional ◽  
Crack Arrest ◽  
Pressure Vessels ◽  
Short Description ◽  
Constraint Effect ◽  
Materials Properties ◽  
Integrity Assessment ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

A significant extensive Research & Development work is conducted by Electricite´ de France (EDF) related to the structural integrity re-assessment of the French 900 and 1300 MWe reactor pressure vessels in order to increase their lifetime. Within the framework of this programme, numerous developments have been implemented or are in progress related to the methodology to assess flaws during a pressurized thermal shock (PTS) event. The paper contains three aspects: a short description of the specific French approach for RPV PTS assessment, a presentation of recent improvements on thermalhydraulic, materials and mechanical aspects, and finally an overview of the present R&D programme on thermalhydraulic, materials and mechanical aspects. Regarding the last aspect on present R&D programme, several projects in progress will be shortly described. This overview includes the redefinition of some significant thermalhydraulic transients based on some new three-dimensional CFD computations (focused at the present time on small break LOCA transient), the assessment of vessel materials properties, and the improvement of the RPV PTS structural integrity assessment including several themes such as warm pre-stress (WPS), crack arrest, constraint effect ....

Data Analysis for RPV Life Management Studies of Temperature Measurements From Monitoring of Safety Injection Fluid in Refueling Water Storage Tank

2007 ◽  
Author(s):  
Albannie Cagnac ◽  
Silvia Turato ◽  
Fre´de´ric Lestang ◽  
Franc¸ois Planckeel
Keyword(s):  
Storage Tank ◽  
Input Parameter ◽  
Water Storage ◽  
Water Storage Tank ◽  
Density Distributions ◽  
Life Management ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Important Input ◽  
Reactor Pressure

In Electricite´ de France (EDF), it’s used to taking as a conservative constant value for the temperature of Refueling Water Storage Tank (RWST) in the deterministic and probabilistic analyses for Reactor Pressure Vessel (RPV) life management. The water contained in this storage tank supplies Security Injection during accidental conditions, such as loss-of-coolant accident (LOCA), so the temperature of this water is a very important input parameter for fracture mechanics analyses. In the continuation of [1], the aim of our study is to evaluate the variability of this temperature. Since 1999, EDF has been collecting the RWST temperature for several sites and several nuclear plant units. Using statistical analyses, this study aims at first to identify the most important RPV exploitation life events that influence this temperature and, at secondly, to obtain statistical density distributions in order to describe its variability in the most realistic way possible.

Recapturing Net Positive Suction Head Margins in Boiling Water Reactor Emergency Core Cooling Systems

2017 ◽  
Author(s):  
Alan J. Bilanin ◽  
Andrew E. Kaufman ◽  
Warren J. Bilanin
Keyword(s):  
Cooling System ◽  
Cooling Water ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Cooling Systems ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling ◽  
Reactor Pressure

Boiling Water Reactor pressure suppression pools have stringent housekeeping requirements, as well as restrictions on amounts and types of insulation and debris that can be present in the containment, to guarantee that suction strainers that allow cooling water to be supplied to the reactor during a Loss of Coolant Accident remain operational. By introducing “good debris” into the cooling water, many of these requirements/restrictions can be relaxed without sacrificing operational readiness of the cooling system.

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