Fracture mechanics in design and service: ‘living with defects’ - Fracture mechanics as an aid to design and operation of nuclear plant

1981 ◽  
Vol 299 (1446) ◽  
pp. 227-236 ◽  
Keyword(s):  
Fracture Mechanics ◽  
Material Properties ◽  
Pressure Vessels ◽  
Nuclear Plant ◽  
Plant Design ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Plant Safety ◽  
Usual Criterion ◽  
Design Protection

Fracture mechanics analyses are an important part of nuclear plant design, supplementing the conventional design protection against failure to cover the possibility of the presence of crack-like defects. The degree of detail and accuracy required for a particular application depends on the possible consequences of a failure and whether the assessment is concerned with plant safety or with aspects of reliability. In the former case, a conservative approach is necessary and the prevention of initiation is the usual criterion. This approach is typified by the safety assessment applied to pressurized water reactor pressure vessels, which is outlined and discussed in relation to elastic plastic approaches and the importance of plant transient conditions, material properties (especially in weldments) and possible defect distributions. Fracture mechanics can help in defining quality control and quality assurance procedures, including both requirements for mechanical property appraisal and nondestructive testing. The latter aspects extend into operation, in respect of monitoring of plant conditions, surveillance of changes in material properties and the use of periodic inspection and plant condition monitoring techniques. A number of examples are quoted and recommendations made to permit improved fracture mechanics assessments.

Uncertainties in Pressurized Thermal Shock Analyses

2019 ◽  
Author(s):  
M. Niffenegger ◽  
O. Costa Garrido ◽  
D. F. Mora ◽  
G. Qian ◽  
R. Mukin ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Pressurized Water Reactor ◽  
Ansys Fluent ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Safety Margins ◽  
Pressurized Thermal Shock

Abstract Integrity assessment of reactor pressure vessels (RPVs) can be performed either by deterministic fracture mechanics (DFM) or/and by probabilistic fracture mechanics (PFM) analyses. In European countries and Switzerland, only DFM analyses are required. However, in order to establish the probabilistic approach in Switzerland, the advantages and shortcomings of the PFM are investigated in the frame of a national research project. Both, the results from DFM and PFM depend strongly on the previous calculated thermal-hydraulic boundary conditions. Therefore, complete integrity analyses involving several integrated numerical codes and methods were performed for a reference pressurized water reactor (PWR) RPV subjected to pressurized thermal shock (PTS) loads. System analyses were performed with the numerical codes RELAP5 and TRACE, whereas for structural and fracture mechanics calculations, the FAVOR and ABAQUS codes were applied. Additional computational fluid dynamics analyses were carried out with ANSYS/FLUENT, and the plume cooling effect was alternatively considered with GRS-MIX. The results from the different analyses tools are compared, to judge the expected overall uncertainty and reliability of PTS safety assessments. It is shown that the scatter band of the stress intensities for a fixed crack configuration is rather significant, meaning that corresponding safety margins should be foreseen. The conditional probabilities of crack initiation and RPV failure might also differ, depending on the considered random parameters and applied rules.

Inspection Optimization Justification for PWR Main Steam and Feedwater Nozzles Using Probabilistic Fracture Mechanics

2019 ◽  
Author(s):  
C. Lohse ◽  
D. J. Shim ◽  
D. Somasundaram ◽  
R. Grizzi ◽  
G. L. Stevens ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Regulatory Commission ◽  
Pressure Vessels ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Asme Code ◽  
Sensitivity Studies ◽  
Main Steam ◽  
Regulatory Commission ◽  
Technical Basis

Abstract Pressurized water reactor (PWR) steam generator (SG) main steam and feedwater nozzles are classified as ASME Code, Section XI, Class 2, Category C-B, pressure retaining welds in pressure vessels. Current ASME Code requirements specify that the nozzle-to-shell welds (Item No. C2.21 & C2.32) and nozzle inner radius sections (Item C2.22) are to be examined very 10 years. An evaluation was performed to establish a technical basis for optimized inspection frequencies for these items. The work included a review of inspection history and results, a survey of components in the PWR fleet (which included both U.S. and overseas plants), selection of representative main steam and feedwater nozzle configurations and operating transients for stress analysis, evaluation of potential degradation mechanisms, and flaw tolerance evaluations consisting of probabilistic and deterministic fracture mechanics analyses. The results of multiple inspection scenarios and sensitivity studies were compared to the U.S. Nuclear Regulatory Commission (NRC) safety goal of 10−6 failures per year.

Virtual Web Plant: An Internet-Based Plant Engineering Information System

2000 ◽  
Author(s):  
Peter Ebbesmeyer ◽  
Jürgen Gausemeier ◽  
Holger Krumm ◽  
Thorsten Molt ◽  
Thomas Gruß
Keyword(s):  
Object Oriented ◽  
3D Models ◽  
Plant Design ◽  
Pressurized Water Reactor ◽  
Plant Structure ◽  
Pressurized Water ◽  
Web Based ◽  
Water Reactor ◽  
Engineering Information ◽  
New Type

Abstract The European Pressurized Water Reactor Project (EPR) is based on an innovative design concept for a new type of pressurized water reactor. The development of this concept will be carried out by a consortium of international partners and customers based in Germany and France. During the development of the EPR large amounts of up-to-date engineering data (i.e. CAD data, planning documentation) have to be made available to all project partners for presentation and development. This paper describes the web-based tool Virtual Web Plant (VWP), a tool to integrate 3D models from various CAD plant design tools and to display them interactively. The tool receives the data through the Internet. We describe the special advantages of an object-oriented database for the storage of the graphical data are shown. Through the application of object-oriented databases, it is possible to define various views of the logical plant structure, for example. The user is hereby able to navigate easily through both the plant structure and the project documentation. The work presented in this paper is part of a Virtual Reality Research Project of the Heinz Nixdorf Institute and the Siemens AG KWU.

Consideration of Special Effects for the Application of an Optimized Fracture Mechanics Approach for the RPV Assessment: Project CAMERA

2020 ◽  
Author(s):  
Florian Obermeier ◽  
Hieronymus Hein ◽  
Johannes May ◽  
Julia Kobiela ◽  
Marco Kaiser
Keyword(s):  
Fracture Toughness ◽  
Fracture Mechanics ◽  
Transition Region ◽  
Crack Extension ◽  
Master Curve ◽  
Pressure Vessels ◽  
Ductile Material ◽  
Pressurized Water ◽  
Special Effects ◽  
Material Tests

Abstract A large database for fracture toughness data according to the standard ASTM E1921 in the brittle and brittle to ductile transition region was generated within the former research programs CARINA and CARISMA for materials used in western pressurized water reactor (PWR) reactor pressure vessels (RPV). These programs confirmed successfully the application of the Master Curve approach for German RPVs. With respect to the RPV proof of safety during plant operation, in particular for integrity assessments considering the event of a Pressurized Thermal Shock (PTS) further relevant issues appeared in terms of the use of suitable fracture toughness curves for components and of the quantification of safety margins for the irradiated material state: • Validity of fracture toughness curve and Master Curve respectively in the irradiated state at higher test temperatures (ductile material region) • Verification of the WPS effect (warm pre-stress) in the irradiated state for representative loading paths and materials • Impact of material inhomogeneities on the fracture toughness • Transferability of specimen-specific effects on the safety-related RPV integrity assessment To address these open issues a follow-up project was initiated called CAMERA (Consideration of special effects for the application of an optimized fracture mechanics approach for the RPV safety assessment). One main aspect within this program is to assess the possibility of describing the whole fracture toughness curve including the upper shelf based on empirical T0 correlations and the effect of a load case related warm pre-loading. Material tests are carried out and evaluated according to ASTM E1820 to get the crack extension resistance as a function of stable crack extension (J-R curves) in the transition to the ductile region and according to ASTM E1921 (WPS) in the lower shelf region and transition region. The least-square fit curves (J-R) as well as the effect of warm pre-stress are determined for several types of irradiated and unirradiated RPV base and weld metals. The results serve to expand the application window of the Master Curve concept to the ductile region and to complete the fracture toughness curve over the entire temperature range from lower shelf until operating temperature also considering the load scenario based on the effect of warm pre-stress. This paper will summarize and discuss the results of the material tests performed.

Deterministic and Probabilistic Fracture Mechanics Analysis for Structural Integrity Assessment of Pressurized Water Reactor Pressure Vessel

2016 ◽  
Vol 138 (3) ◽  
Author(s):  
Kuan-Rong Huang ◽  
Chin-Cheng Huang ◽  
Hsiung-Wei Chou
Keyword(s):  
Structural Reliability ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Benchmark Problems ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Oak Ridge ◽  
Integrity Assessment ◽  
Reactor Pressure

Cumulative radiation embrittlement is one of the main causes for the degradation of pressurized water reactor (PWR) reactor pressure vessels (RPVs) over their long-term operations. To assess structural reliability of degraded reactor vessels, the FAVOR code from the Oak Ridge National Laboratories of the U.S. is employed to perform probabilistic fracture analysis for existing Taiwan domestic PWR reactor vessels with consideration of irradiation embrittlement effects. The plant specific parameters of the analyzed reactor vessel associated with assumed design transients are both considered as the load conditions in this work. Furthermore, two overcooling transients of steam generator tube rupture (SGTR) and pressurized thermal shock (PTS) addressed in the USNRC/EPRI benchmark problems are also taken into account. The computed low failure probabilities can demonstrate the structural reliability of the analyzed reactor vessel for its both license base and extended operations. This work is helpful for the risk assessment and aging management of operating PWR RPVs and can also be referred as its regulatory basis in Taiwan.

Virtual Reality for Power Plant Design

1999 ◽  
Author(s):  
Peter Ebbesmeyer ◽  
Peter Gehrmann ◽  
Michael Grafe ◽  
Holger Krumm
Keyword(s):  
Virtual Reality ◽  
Plant Design ◽  
Pressurized Water Reactor ◽  
Virtual Reality System ◽  
Virtual Reality Technology ◽  
Support Design ◽  
Pressurized Water ◽  
Design Review ◽  
Development Processes ◽  
Future Work

Abstract We describe DesiRe, a Design Review system based on virtual reality technology. DesiRe is a production virtual reality system that is used to support design review processes during the development of The European Pressurized Water Reactor (EPR). This paper starts with a brief description of the EPR project. We then focus on the design review processes that are an important part of the EPR development processes. We discuss why the use of virtual reality technology significantly improves the design review processes. The subsequent part of the paper introduces in detail the implementation issues and application issues of the virtual reality system DesiRe currently being used by the EPR engineers in their design review meetings. The end of the paper gives an outlook on future work and a brief conclusion. The work presented in this paper is part of an ongoing virtual reality research project at the Heinz Nixdorf Institut and Siemens AG KWU.

Study of Equipment Survivability Under Severe Accident Conditions

2016 ◽  
Author(s):  
Jinquan Yan ◽  
Shanhu Xue ◽  
Lin Tian ◽  
Wei Lu
Keyword(s):  
Nuclear Power ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Plant Safety ◽  
New Development ◽  
Nuclear Power Plant Safety ◽  
Accident Conditions ◽  
Accident Sequences

To improve nuclear power plant safety, severe accident prevention and mitigation for both new development and existing plants are generally required by various nuclear safety authorities worldwide. Although great efforts have been made, how to ensure equipment survivability under severe accident conditions is still a concern. This paper depicts an approach to demonstrate the equipment survivability under severe accident conditions by taking passive pressurized water reactor CAP1400 as an instance, including screening of severe accident sequences, determination of bounding environment conditions within containment, equipments identification used for severe accident mitigation and proposed test plan.

Flaw Growth Prediction and Fitness-for-Service Assessment of a PWR Nuclear Plant Steam Generator U-Tube

2002 ◽  
Author(s):  
Robert S. Vecchio ◽  
Sri K. Sam Sinha
Keyword(s):  
Finite Element ◽  
Fracture Mechanics ◽  
Crack Growth ◽  
Steam Generator ◽  
Degradation Mechanism ◽  
Element Model ◽  
Nuclear Regulatory Commission ◽  
Nuclear Plant ◽  
Pressurized Water ◽  
Nondestructive Examination

Primary water stress corrosion cracking (PWSCC) continues to be a dominant degradation mechanism affecting the service life of steam generators in several operating pressurized water reactor (PWR) nuclear plants. Recently, in one operating nuclear plant, a steam generator U-tube ruptured catastrophically while the unit was on-line. Although the plant operators were able to shutdown the reactor without significant release of radiation, the Nuclear Regulatory Commission (NRC) and the owner utility launched a full-scale investigation of the incident. The owner utility requested that a crack growth analysis and engineering evaluation of the tube rupture be performed, as well as assess the fitness-for-service of the generator for continued operation. This paper presents a summary of elastic-plastic finite element and fracture mechanics analyses performed for a steam generator U-tube, subjected to crack initiation at the inner diameter of the tube in the apex region. Residual stresses were computed from a finite element model of the tube simulating the mechanical bending process with the use of an anvil. Fracture mechanics and crack growth evaluations were performed to predict the time required for a pre-existing flaw at the inside diameter of the tube to propagate through-wall. Additionally, a fitness-for service assessment was performed in order to permit a degraded tube to remain in service, given an initial flaw size as determined by nondestructive examination.

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