Stable and Unstable Crack Growth of A508 Class 3 Plates Subjected to Combined Force of Thermal Shock and Tension

1989 ◽  
Vol 111 (3) ◽  
pp. 234-240 ◽  
Author(s):  
G. Yagawa ◽  
Y. Ando ◽  
K. Ishihara ◽  
T. Iwadate ◽  
Y. Tanaka
Keyword(s):  
Thermal Shock ◽  
Nuclear Power ◽  
Power Plants ◽  
Fracture Behavior ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Test Results ◽  
Unstable Crack ◽  
Pressurized Thermal Shock ◽  
Series Of Experiments

An urgent problem for nuclear power plants is to assess the structural integrity of the reactor pressure vessel under pressurized thermal shock. In order to estimate crack behavior under combined force of thermal shock and tension simulating pressurized thermal shock, two series of experiments are demonstrated: one to study the effect of material deterioration due to neutron irradiation on the fracture behavior, and the other to study the effect of system compliance on fracture behavior. The test results are discussed with the three-dimensional elastic-plastic fracture parameters, J and Jˆ integrals.

Structural Integrity for the RPV of French NPP during the LRA

2016 ◽  
Vol 853 ◽  
pp. 453-457
Author(s):  
Ming Ya Chen ◽  
Wei Wei Yu ◽  
Jin Hua Shi ◽  
Rong Shan Wang ◽  
Lv Feng ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
The Other ◽  
Light Water Reactor ◽  
Original Design ◽  
Light Water ◽  
Pressurized Thermal Shock ◽  
The Usa ◽  
Reactor Pressure

Most of the French Nuclear Power Plants (NPPs) are currently embarking upon efforts to renew their operating license, while the pressurized thermal shock (PTS) events and environmentally assisted fatigue (EAF) pose potentially significant challenges to the structural integrity of the reactor pressure vessel (RPV) which has the potential to be NPP life-limiting conditions. In the assessment of the PTS events, the deterministic fracture mechanics (DFM) is still used as the basic mechanics in most countries except for the USA. While the maximum nil-ductility-transition temperature (RTNDT) is about 80°C for 54 French RPVs after 40 years operation, the maximum allowable RTNDT is only about 70 oC and 80 oC for the typical PTS events in the IAEA and NEA reports, respectively. On the other hand, the effects of light water reactor (LWR) environmental (other than moderate environment in the code) were not considered in the original design, while the effects of LWR environmental are needed to be considered in the LRA according to the USA regulations. In this paper, the challenges of the PTS and EAF are discussed, and some suggestions are also given for the LRA

Pressure Tests for Thickness Management of Wall Thinning Tees

2012 ◽  
Author(s):  
Kaina Teshima ◽  
Yoichi Iwamoto ◽  
Kiminobu Hojo ◽  
Tomoyuki Oka ◽  
Kunihiro Kobayashi ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Outer Diameter ◽  
Test Results ◽  
Minimum Thickness ◽  
T Joints ◽  
Wall Thinning ◽  
Pressure Tests ◽  
Design Pressure

Although the minimum thickness of pipe wall required (tsr) of T-joints (tees) of class 2, 3 and lower classes of nuclear power plants in Japan is calculated from the design pressure and temperature, there is no rule or standard of wall thinning T-joints for thickness management. This paper describes the pressure tests procedure and six test results with parameters of T-joint geometry such as outer diameter D, thickness T and T/D to establish structural integrity of wall thinning T-joints. Based on the fracture surface observation, a ductile crack initiation of each test mock-ups was confirmed.

Optimizing Maintenance Strategy of a Reactor Pressure Vessel Using 3D-CFD and FEM Based Probabilistic Pressurized Thermal Shock Analysis

2017 ◽  
Author(s):  
Xiaoyong Ruan ◽  
Toshiki Nakasuji ◽  
Kazunori Morishita
Keyword(s):  
Spatial Distribution ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Cooling System ◽  
Reactor Pressure Vessel ◽  
Pressurized Thermal Shock ◽  
Core Cooling ◽  
Reactor Pressure

The structural integrity of a reactor pressure vessel (RPV) is important for the safety of a nuclear power plant. When the emergency core cooling system (ECCS) is operated and the coolant water is injected into the RPV due to a loss-of-coolant accident (LOCA), the pressurized thermal shock (PTS) loading takes place. With the neutron irradiation, PTS loading may lead a RPV to fracture. Therefore, it is necessary to evaluate the performance of RPV during PTS loading to keep the reactor safety. In the present study, optimization of RPV maintenance is considered, where two different attempts are made to investigate the RPV integrity during PTS loading by employing the deterministic and probabilistic methodologies. For the deterministic integrity evaluation, 3D-CFD and finite element method (FEM) simulations are performed, and stress intensity factors (SIFs) are obtained as a function of crack position inside the RPV. As to the probabilistic integrity evaluation, on the other hand, a more accurate spatial distribution of SIF on the RPV is calculated. By comparing the distribution thus obtained with the fracture toughness included as a part of the master curve, the dependence of fracture probabilities on the position inside the RPV is obtained. Using the spatial distribution of fracture probabilities in RPV, the priority of the inspection and maintenance is finally discussed.

Dynamic Response of a Nuclear Power Plant Subjected to External Accident Event

2010 ◽  
Author(s):  
Rosa Lo Frano ◽  
Giuseppe Forasassi
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Failure Modes ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Containment System ◽  
Dynamic Loadings ◽  
Water Reactors ◽  
Set Up

In recent times there is a renewed worldwide interest in the development and application of advanced nuclear power plants (NPPs). Decisions on the construction of several NPPs with evolutionary light water reactors have been made (e.g. EPR in Finland and France, AP1000 in China, etc.) and more are under consideration for licensing in several countries. Innovative NPPs are designed to be built with very broad siting conditions; therefore the safety aspects related to the external events might follow new scenarios and failure modes, different from those well known for the currently operated reactors. In this paper, the intent is evaluating the structural integrity of a nuclear containment system subjected to dynamic loadings due to a Design Base Earthquake and an aircraft impact (large size civilian jets or military aircrafts impact), which represent the two most relevant external accidents that should be considered and investigated as part of the basic design of a NPP in particular a III+ and IV Gens. In fact a suitable safety design of the NPP containment system (according to the international safety and design code guidelines, as NRC or IAEA ones), even if designed to meet other design goal, may represent a “built-in protection” to avoid or mitigate the effects of mentioned dynamic loadings. To the purpose a rather sophisticated numerical methodology, adopting finite element (FEM) approach, is employed for studying the overall dynamic behaviour of nuclear reactor and to determine the structural effects of the propagation of dynamic seismic as well as impulsive loads (containment structure response) up to the relevant nuclear components. Therefore representative three-dimensional FEM models of mentioned NPP containment and aircraft structures were set up, and used, in the performed analyses taking also into account the suitable materials behaviour and their related constitutive laws as well as the seismic excitation (determined according to the NRC rules). Moreover the performed analyses and the carried out response analyses of internal components, to both the ground motion and impact loads, were studied to check the considered NPP containment strength reserve in the case of the considered events. The obtained results seem to confirm the possibility to achieve an optimization of the NPP internal components.

PWR Nuclear Power Plants Reactor Vessel and Coolant System Cast Duplex Stainless Steel Elbows: New Results in Integrity Analysis

2003 ◽  
Author(s):  
G. Bezdikian ◽  
C. Faidy ◽  
P. Cambefort ◽  
D. Moinereau
Keyword(s):  
Finite Element ◽  
Stainless Steel ◽  
Duplex Stainless Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Reactor Vessel ◽  
Integrity Assessment ◽  
Reactor Coolant

The Reactor Pressure Vessel and Reactor coolant materials (hot and cold CAST elbows) are major components for integrity evaluation of nuclear plant units. The French Utility (Electricite de France) has engaged a few years ago an important program regarding the integrity assessment of RPV and cast duplex stainless steel elbows based on large real database. This paper deals with the verification of the integrity of the Reactor Vessel component by finite element mechanical studies, in all conditions of loading in relation with RTNDT (Reference Nil Ductility Transition Temperature), and considering all parameters. An overall review of actions will be presented describing the French approach regarding the assessment of nuclear RPV. The latest results obtained are based on generic integrity analyses for all categories of situations (normal upset emergency and faulted conditions), particularly in case of PTS, until the end of lifetime, postulating longitudinal shallow subclad flaws. For the Reactor Coolant Elbows, the results of structural integrity analyses, beginning with elastic computations and completed with three-dimensional finite element elastic-plastic computations for envelope cases, are compared with in-service inspection real flaw characterisation and the results are compared to the margin on loading condition with the criteria included in the code.

Cleavage and Ductile Thermal Shock Fractures of Corner-Cracked Nozzles

1989 ◽  
Vol 111 (3) ◽  
pp. 241-246 ◽  
Author(s):  
G. Yagawa ◽  
K. Ishihara
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Pressure Vessel Steel ◽  
Fracture Parameter ◽  
The Finite Element Method ◽  
Vessel Steel ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

In order to study the structural integrity of the reactor pressure vessel under pressurized thermal shock, both the cleavage and the ductile thermal shock fracture experiments using initially corner-cracked nozzle specimens made of Type A508 class 3 pressure vessel steel were performed. In both experiments, unstable fractures were realized, although the test conditions were very conservative compared to those of real plants. Finally, the three-dimensional and time-dependent fracture parameter obtained with the finite element method was successfully employed to discuss the fracture phenomenon.

Methodology and Warm Prestressing Effect for Pressurized Thermal Shock Analysis in Nuclear Power Plant

2015 ◽  
Vol 750 ◽  
pp. 104-113
Author(s):  
Gui An Qian ◽  
Markus Niffenegger
Keyword(s):  
Thermal Shock ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Pressurized Water Reactor ◽  
Local Approach ◽  
Pressurized Water ◽  
Warm Prestressing ◽  
Load Paths ◽  
Pressurized Thermal Shock ◽  
Integrity Analysis

One potential challenge to the integrity of the reactor pressure vessel (RPV) in a pressurized water reactor is posed by pressurized thermal shock (PTS). Therefore, the safety of the RPV with regard to neutron embrittlement has to be analyzed. In this paper, the procedure and method for the structural integrity analysis of RPV subjected to PTS is presented. The FAVOR code is applied to calculate the probabilities for crack initiation and failure by considering crack distributions based on cracks observed in the Shoreham and PVRUF RPVs in the U.S. A local approach to fracture, i.e. the σ*-A* model is used to predict the warm prestressing (WPS) effect on the RPV integrity. The results show that the remaining stress contributes to the WPS effect, whereas the increase of fracture toughness is not completely attributed to the remaining stress. The modeled load paths predict a material toughness increase of 30-100%.

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 ....

The deterministic structural integrity assessment of reactor pressure vessels under pressurized thermal shock loading

2015 ◽  
Vol 288 ◽  
pp. 130-140 ◽  
Author(s):  
Mingya Chen ◽  
Feng Lu ◽  
Rongshan Wang ◽  
Ping Huang ◽  
Xiangbin Liu ◽  
...  
Keyword(s):  
Thermal Shock ◽  
Shock Loading ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Integrity Assessment ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure
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