scholarly journals Application of Inelastic Method and Its Comparison with Elastic Method for the Assessment of In-Box LOCA Event on EU DEMO HCPB Breeding Blanket Cap Region

2021 ◽  
Vol 11 (19) ◽  
pp. 9104
Author(s):  
Anoop Retheesh ◽  
Francisco A. Hernández ◽  
Guangming Zhou
Keyword(s):  
Failure Modes ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Inelastic Analysis ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Institute Of Technology ◽  
Elastic Design ◽  
The Given ◽  
Karlsruhe Institute

The Helium Cooled Pebble Bed (HCPB) breeding blanket, being developed by the Karlsruhe Institute of Technology (KIT) and its partners is one of the two driver blanket candidates to be selected for the European demonstration fusion power plant (EU DEMO). The in-box Loss of Coolant Accident (LOCA) is a postulated initiating event of the breeding blanket (BB) that must be accounted within the design basis. In this paper, the BB cap region is analyzed for its ability to withstand an in-box LOCA event. Initially, an assessment is performed using conventional elastic design codes for nuclear pressure vessels. However, it is thought that the elastic rules are not ‘equipped’ to assess the material damage modes which are essentially inelastic. Therefore, a non-linear inelastic analysis is further performed to better understand the damage in the material. Two predominant inelastic failure modes are thought to be relevant and addressed: exhaustion of ductility and plastic flow localization. While the design of HCPB BB has been predominantly based on the elastic design-by-analysis studies, results from the present study show that the elastic rules may be overly conservative for the given material and loading and could lead to inefficient designs. To our knowledge, this study is the first attempt to investigate the structural integrity of the European DEMO blankets under in-box LOCA conditions using the inelastic methods.

Nonlinear Analysis in Pressure Vessel Design Codes: Recommendations for Codified Rules Improvements

2018 ◽  
Author(s):  
Claude Faidy
Keyword(s):  
Material Properties ◽  
Failure Modes ◽  
Plastic Instability ◽  
Pressure Vessels ◽  
Inelastic Analysis ◽  
Stress Classification ◽  
Piping Systems ◽  
Plastic Shakedown ◽  
Stress Criteria

During the past 30 years the main rules to design pressure vessels were based on elastic analyses. Many conservatisms associated to these different elastic approaches are discussed in this paper, like: stress criteria linearization for 3-D components, stress classification in nozzle areas, plastic shake down analysis, fatigue analysis, Ke evaluation, and pipe stress criteria for elastic follow-up due to thermal expansion or seismic loads... This paper will improve existing codified rules in nuclear and non-nuclear Codes that are proposed as alternatives to elastic evaluation for different failure modes and degradation mechanisms: plastic collapse, plastic instability, tri-axial local failure, rupture of cracked component, fatigue and Ke, plastic shakedown. These methods are based on limit loads, monotonic or cyclic elastic-plastic analyses. Concerned components are mainly vessels and piping systems. No existing Code is sufficiently detailed to be easily applied; the needs are stress analysis methods through finite elements, material properties including material constitutive equations and criteria associated to each methods and each failure modes. A first set of recommendation to perform these inelastic analysis will be presented to improve existing codes on an international harmonized way, associated to all material properties and criteria needed to apply these modern methods. An international draft Code Case is in preparation.

Piping System Structural Integrity Simulation for Post LOCA Water Hammer Loads

2002 ◽  
Author(s):  
B. W. Manning ◽  
T. Stevens ◽  
G. Morandin ◽  
R. G. Sauve´ ◽  
R. Richards ◽  
...  
Keyword(s):  
Structural Damage ◽  
Structural Integrity ◽  
Water Hammer ◽  
Piping System ◽  
Linear Modeling ◽  
Linear Elastic ◽  
Beam Elements ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Non Linear

The Canadian Nuclear Safety Commission (CNSC) required as part of the operating license for Ontario Power Generation’s Darlington Nuclear Generating Station, that the structural integrity of the piping following a loss of coolant accident (LOCA) be demonstrated. This is necessary to ensure that no subsequent pressure boundary failures will impede the ability to maintain fuel cooling. The injection of cold emergency coolant following a LOCA creates the potential for the occurrence of condensation-induced water hammers (CIWH) in the primary heat transport (PHT) system piping. Classical linear elastic piping analysis using the class 1 NB-3656 rules of the ASME Boiler & Pressure Vessel Code failed to demonstrate the adequacy of the piping and/or its supports that were designed using the linear elastic rules of subsection NF for nine of the twelve piping models that comprise the PHT system. A decision was made to undertake a state-of-the-art non-linear explicit analysis in order to qualify the piping. Strain rather than stress limits would be applied similar to those being developed by ASME for nuclear packaging undergoing accidental impact during transportation. In order to address the feasibility of this approach, a non-linear analysis was performed on a portion of one of the piping systems. The piping was modeled as shells and again as beam elements with and without detailed modeling of the supports. After these initial simulations, it was determined that the piping could be modeled with simplified beam elements, however, the supports would require a more detailed modeling in order to determine the extent of support damage and the effect the supports have on the integrity of the piping system itself. This paper addresses the non-linear modeling of the piping models and discusses the modeling details, assumptions and analysis results. This approach is shown to be a useful alternative for predicting the extent of structural damage that can be expected by a Level D event such as a condensation induced water hammer following a loss of coolant accident.

Finite Element Analysis of Local Inelastic Strain Based on Stress Redistribution Locus

2008 ◽  
Author(s):  
Shunji Kataoka ◽  
Takuya Sato
Keyword(s):  
Finite Element Analysis ◽  
Finite Element ◽  
Thermal Loading ◽  
Failure Modes ◽  
Elevated Temperatures ◽  
Pressure Vessels ◽  
Stress Redistribution ◽  
Strain Diagram ◽  
Element Analysis ◽  
Inelastic Analysis

Creep-fatigue damage is one of the dominant failure modes for pressure vessels and piping used at elevated temperatures. In the design of these components the inelastic behavior should be estimated accurately. An inelastic finite element analysis is sometimes employed to predict the creep behavior. However, this analysis needs complicated procedures and many data that depend on the material. Therefore the design is often based on a simplified inelastic analysis based on the elastic analysis result, as described in current design codes. A new, simplified method, named, Stress Redistribution Locus (SRL) method, was proposed in order to simplify the analysis procedure and obtain reasonable results. This method utilizes a unique estimation curve in a normalized stress-strain diagram which can be drawn regardless of the magnitude of thermal loading and constitutive equations of the materials. However, the mechanism of SRL has not been fully investigated. This paper presents results of the parametric inelastic finite element analyses performed in order to investigate the mechanism of SRL around a structural discontinuity, like a shell-skirt intersection, subjected to combined secondary bending stress and peak stress. This investigation showed that SRL comprises a redistribution of the peak and secondary stress components and that although these two components exhibit independent redistribution behavior, they are related to each other.

Development of a Method to Assess Impact Hazards

2015 ◽  
Author(s):  
Peter Gill ◽  
Adam Toft
Keyword(s):  
Structural Integrity ◽  
Nuclear Reactors ◽  
Primary Circuit ◽  
Plant Safety ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Particle Hydrodynamics ◽  
Smoothed Particle ◽  
Circuit Components ◽  
Safety Systems

The structural integrity of pressure retaining primary circuit components and of the containment boundary is of great significance for the safety justification of all nuclear reactors. In this regard, it is important to understand the vulnerability of safety related components to potential accidents. Whilst the direct consequences of pipe or pressure vessel failure, for example in terms of the extent of loss of coolant, may be tolerable, the indirect consequences of failure may not. A Loss of Coolant Accident (LOCA) may indirectly damage plant safety systems, including the containment boundary, due to the effect of missiles generated by the LOCA. This paper describes a study to develop a modern numerical modelling technique to estimate damage by missiles. Smoothed Particle Hydrodynamics (SPH) has been applied to simulate the acceleration of the missile due to the fluid jet.

Creep Evaluation for a PWR Reactor Pressure Vessel Lower Head Under Severe Accident Conditions Considering Sustained Internal Pressure

2018 ◽  
Author(s):  
YongJian Gao ◽  
Ming Cao ◽  
YinBiao He
Keyword(s):  
Internal Pressure ◽  
Failure Modes ◽  
Structural Integrity ◽  
Creep Damage ◽  
Pressure Vessels ◽  
Mitigation Strategies ◽  
Field Analysis ◽  
Severe Accident ◽  
Lower Head ◽  
Reactor Pressure

In-Vessel Retention (IVR) is one of appropriate severe accident mitigation strategies for AP1000 Nuclear Power Plant (NPP), and assurance of prevention against to thermal failure and structural failure of Reactor Pressure Vessels (RPV) is the prerequisite of IVR. A Finite Element Model fora RPV considering lower head melting was established, the creep calculation was carried out after the temperature field analysis, and the stress-strain responses for different times were obtained. By means of choosing representative evaluation sections and applying the Accumulative Damage Theory based on Larson-Miller Parameter, the Creep Damage calculations and evaluations were conducted. The results showed that the failure modes associated with creep rupture would not happen under IVR condition when a certain amount of internal pressure sustained. The approaches employed in this paper could be utilized in structural integrity evaluation of RPV under IVR for other new type NPPs.

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.

Effects of Residual Stress by Weld Overlay Cladding and PWHT on the Structural Integrity of RPV During PTS

2007 ◽  
Author(s):  
Makoto Udagawa ◽  
Jinya Katsuyama ◽  
Kunio Onizawa
Keyword(s):  
Residual Stress ◽  
Structural Integrity ◽  
Crack Size ◽  
Operating Conditions ◽  
Residual Stress Field ◽  
Weld Overlay ◽  
Loss Of Coolant Accident ◽  
System Pressure ◽  
Loss Of Coolant ◽  
Main Steam

In order to assess the structural integrity of a reactor pressure vessel (RPV), it is assumed that a surface crack resides through the cladding at the inner surface of the vessel. It is, therefore, important to precisely evaluate stress intensity factor (SIF) under the residual stress field due to weld overlay cladding and post-weld heat treatment (PWHT). In this work, numerical simulation based on thermal-elastic-plastic-creep analysis using finite element method was performed to evaluate residual stress distribution near the cladding layer produced by weld overlay cladding and PWHT. The tensile residual stress of about 400 MPa occurs in the cladding at room temperature after the PWHT. The residual stress distributions under the normal operating conditions (system pressure and temperature) of RPV were also evaluated. The effect of residual stress and evaluation methods on SIF behavior for various crack size were studied under typical pressurized thermal shock (PTS) conditions such as small break loss of coolant accident (SBLOCA), main steam line break (MSLB) and large break loss of coolant accident (LBLOCA). It is clarified from comparison of this weld simulation with the other simple methods that SIF is affected by residual stress by weld overlay cladding and PWHT.

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.

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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