Numerical Investigations on the Interaction of Cracks in Quasi-Laminar Crack Fields

2018 ◽  
Author(s):  
Christian Swacek ◽  
Xaver Schuler ◽  
Michael Seidenfuss
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Base Material ◽  
Pressure Vessels ◽  
Loading Conditions ◽  
Critical Crack ◽  
Ongoing Research ◽  
Specific Loading ◽  
Mechanical Methods ◽  
Crack Tips

Non-destructive testing measurements in the Belgian nuclear power plants Doel 3 and Tihange 2 revealed a high quantity of indications in the upper and lower core shells of the reactor pressure vessels. The most likely explanation is that the indications are hydrogen flakes positioned in segregated zones of the base material of the pressure vessel. These hydrogen flakes have a laminar and quasi-laminar orientation to the pressure retaining surface. Under mechanical loading the crack tips undergo predominantly mixed mode loading conditions, where the induced stress and strain fields of the single crack tips influence each other. Due to these specific loading conditions, the assumptions for classical standardized fracture mechanical methods are not met. Currently, there is no verified concept for the evaluation of such kind of crack fields. Therefore the mechanical behavior of components with laminar crack fields and the interaction of cracks in such crack fields are investigated in an ongoing research project. Relevant parameters to describe crack fields in terms of crack size, crack location and crack orientation are derived from literature and own nondestructive measurements. Damage mechanical approaches are used in finite element calculations to investigate the interaction of cracks. Advanced damage mechanical models will be used to investigate crack initiation, crack growth and coalescence of cracks in crack fields. According to the results, representative parameters for crack fields will be derived and critical crack formations determined. The results will be evaluated and compared with state of the art approaches and standards.

Evaluation of the Crack Interaction and Failure Behavior of Components With Crack Fields Using a Damage Mechanical Approach

2019 ◽  
Author(s):  
Christian Swacek ◽  
Patrick Gauder ◽  
Michael Seidenfuss
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Base Material ◽  
Pressure Vessels ◽  
Failure Behavior ◽  
Multiple Cracks ◽  
Critical Crack ◽  
Ongoing Research ◽  
Integrity Assessment ◽  
Crack Tips

Abstract In 2012 non-destructive testing measurements (NDT) of the reactor pressure vessels (RPV) in the Belgian Nuclear Power Plants Doel 3 and Tihange 2 revealed a high quantity of indications in the upper and lower core shells. The most likely explanation is that the measured indications are hydrogen flakes positioned in segregated zones in the base material of the pressure vessels. These hydrogen flakes have a laminar and quasi-laminar orientation with an inclination up to 15° to the pressure retaining surface. Under internal pressure, the crack tips undergo predominantly mixed mode loading conditions, where the induced stress and strain fields of the single crack tips influence each other. The safety assessment of crack afflicted pressurized components is performed by fracture mechanical approaches. For the evaluation of multiple cracks in crack fields, state of the art codes and standards apply interaction criteria and grouping methods, to determine a representative crack, which has to be evaluated. In this paper, the interaction of cracks in crack fields is numerically and experimentally evaluated. Damage mechanical models based on the Rousselier- and the Beremin model are used to investigate numerically the interaction of cracks in crack fields. Experimental data from ferritic flat tensile specimens afflicted with cracks are used to verify the numerical results. The damage mechanical calculations reveal critical crack arrangements due to coalescence behavior and cleavage fracture probability. These results and ongoing research intends the derivation of interaction criteria for cracks in crack fields. The interaction criteria will be used for the definition of a representative flaw for a conservative integrity assessment of crack afflicted components.

scholarly journals Importance of Advanced Planning of Manufacturing for Nuclear Industry

2016 ◽  
Vol 7 (2) ◽  
pp. 42-49
Author(s):  
Nick Shykinov ◽  
Robert Rulko ◽  
Dariusz Mroz
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Regulatory Compliance ◽  
Pressure Vessels ◽  
Nuclear Industry ◽  
Project Schedule ◽  
Nuclear Facilities ◽  
Project Completion ◽  
Advanced Planning ◽  
And Performance

Abstract In the context of energy demands by growing economies, climate changes, fossil fuel pricing volatility, and improved safety and performance of nuclear power plants, many countries express interest in expanding or acquiring nuclear power capacity. In the light of the increased interest in expanding nuclear power the supply chain for nuclear power projects has received more attention in recent years. The importance of the advanced planning of procurement and manufacturing of components of nuclear facilities is critical for these projects. Many of these components are often referred to as long-lead items. They may be equipment, products and systems that are identified to have a delivery time long enough to affect directly the overall timing of a project. In order to avoid negatively affecting the project schedule, these items may need to be sourced out or manufactured years before the beginning of the project. For nuclear facilities, long-lead items include physical components such as large pressure vessels, instrumentation and controls. They may also mean programs and management systems important to the safety of the facility. Authorized nuclear operator training, site evaluation programs, and procurement are some of the examples. The nuclear power industry must often meet very demanding construction and commissioning timelines, and proper advanced planning of the long-lead items helps manage risks to project completion time. For nuclear components there are regulatory and licensing considerations that need to be considered. A national nuclear regulator must be involved early to ensure the components will meet the national legal regulatory requirements. This paper will discuss timing considerations to address the regulatory compliance of nuclear long-lead items.

Alternative RTNDT for Linde 80 Weld Materials

2002 ◽  
Author(s):  
K. K. Yoon ◽  
J. B. Hall
Keyword(s):  
Fracture Toughness ◽  
Transition Temperature ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Screening Criteria ◽  
Master Curve Method ◽  
Asme Code ◽  
Curve Method

The ASME Boiler and Pressure Vessel Code provides fracture toughness curves of ferritic pressure vessel steels that are indexed by a reference temperature for nil ductility transition (RTNDT). The ASME Code also prescribes how to determine RTNDT. The B&W Owners Group has reactor pressure vessels that were fabricated by Babcock & Wilcox using Linde 80 flux. These vessels have welds called Linde 80 welds. The RTNDT values of the Linde 80 welds are of great interest to the B&W Owners Group. These RTNDT values are used in compliance of the NRC regulations regarding the PTS screening criteria and plant pressure-temperature limits for operation of nuclear power plants. A generic RTNDT value for the Linde 80 welds as a group was established by the NRC, using an average of more than 70 RTNDT values. Emergence of the Master Curve method enabled the industry to revisit the validity issue surrounding RTNDT determination methods. T0 indicates that the dropweight test based TNDT is a better index than Charpy transition temperature based index, at least for the RTNDT of unirradiated Linde 80 welds. An alternative generic RTNDT is presented in this paper using the T0 data obtained by fracture toughness tests in the brittle-to-ductile transition temperature range, in accordance with the ASTM E1921 standard.

Lifetime Management of EDF PWR Pressure Vessels and Pipings

2004 ◽  
Author(s):  
F. Hedin ◽  
J. C. Legendre
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cost Benefit Analysis ◽  
Cost Benefit ◽  
Pressure Vessels ◽  
Destructive Testing ◽  
Key Points ◽  
Cast Stainless Steel ◽  
Non Destructive

Lifetime management of EDF PWR vessels and pipings are one of the main technical key points of safety and competitivness. This paper describes the EDF global approach in this field, which is applied to the nuclear fleet i.e 58 nuclear power plants, and particularly to the first 34 three loops, as far as lifetime is concerned: • operating procedures and routine maintenance, special maintenance and ten years safety reassessment, • engineering analysis, based on feed back experience, scientific knowledge, degradations mechanisms, causes and consequences management, • operating loadings decrease, • complementary deterministic and cost-benefit analysis, • fit for service justifications, • anticipation strategy to prepare future, based on Non Destructive Testing investigations, ability to repair and/or to replace components, in situ expertises, ... Some examples are given: lifetime management of reactor vessels heads and bottom penetrations of pressure vessels, fit for service of cast stainless steel primary pipings, primary nozzles and auxiliary pipings special maintenance.

Background on Low Temperature Overpressure Protection System Setpoints for Pressure-Temperature Curves

2020 ◽  
Author(s):  
Jeffrey C. Poehler ◽  
Gary L. Stevens ◽  
Anees A. Udyawar ◽  
Amy Freed
Keyword(s):  
Low Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pressure Vessels ◽  
Pressurized Water Reactors ◽  
Documentation System ◽  
Pressurized Water ◽  
System Pressure ◽  
Water Reactors

Abstract ASME Code, Section XI, Nonmandatory Appendix G (ASME-G) provides a methodology for determining pressure and temperature (P-T) limits to prevent non-ductile failure of nuclear reactor pressure vessels (RPVs). Low-Temperature Overpressure Protection (LTOP) refers to systems in nuclear power plants that are designed to prevent inadvertent challenges to the established P-T limits due to operational events such as unexpected mass or temperature additions to the reactor coolant system (RCS). These systems were generally added to commercial nuclear power plants in the 1970s and 1980s to address regulatory concerns related to LTOP events. LTOP systems typically limit the allowable system pressure to below a certain value during plant operation below the LTOP system enabling temperature. Major overpressurization of the RCS, if combined with a critical size crack, could result in a brittle failure of the RPV. Failure of the RPV could make it impossible to provide adequate coolant to the reactor core and result in a major core damage or core melt accident. This issue affected the design and operation of all pressurized water reactors (PWRs). This paper provides a description of an investigation and technical evaluation regarding LTOP setpoints that was performed to review the basis of ASME-G, Paragraph G-2215, “Allowable Pressure,” which includes provisions to address pressure and temperature limitations in the development of P-T curves that incorporate LTOP limits. First, high-level summaries of the LTOP issue and its resolution are provided. LTOP was a significant issue for pressurized water reactors (PWRs) starting in the 1970s, and there are many reports available within the U.S. Nuclear Regulatory Commission’s (NRC’s) documentation system for this topic, including Information Notices, Generic Letters, and NUREGs. Second, a particular aspect of LTOP as related to ASME-G requirements for LTOP is discussed. Lastly, a basis is provided to update Appendix G-2215 to state that LTOP setpoints are based on isothermal (steady-state) conditions. This paper was developed as part of a larger effort to document the technical bases behind ASME-G.

Determination of Copper in A533b Steel for the Assessment of Radiation Embrittlement Using Laser-Induced Breakdown Spectroscopy

1996 ◽  
Vol 50 (3) ◽  
pp. 306-309 ◽  
Author(s):  
Wolfgang E. Ernst ◽  
Dave F. Farson ◽  
D. Jason Sames
Keyword(s):  
Laser Beam ◽  
Fiber Optics ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Laser Induced Breakdown Spectroscopy ◽  
Radiation Embrittlement ◽  
Breakdown Spectroscopy ◽  
Laser Induced Breakdown

Determination of radiation embrittlement in nuclear reactor pressure vessels is crucial to assessing safe operative lifetimes for many aging nuclear power plants. Conservative nuclear fluence estimates and trace impurity diagnosis of the weldment material are the basis of radiation embrittlement analysis. Copper is thought to be a key impurity contributing to radiation embrittlement. In this paper, the application of laser-induced breakdown spectroscopy (LIBS) as a means to assess radiation embrittlement by the detection and quantification of copper in A553b steel was investigated. A LIBS configuration completely coupled by fiber optics was attempted, but because of low laser power and fiber losses, fiber-optic delivery of the laser beam was unsuccessful. Consequently, hard optics (lenses and mirrors) were employed for laser beam delivery. The plasma emission was delivered successfully via fiber optics to the detection apparatus. Copper measurements were made from custom-fabricated steel samples. Comparison of the LIBS results to an independent atomic absorption spectrophotometry (AAS) analysis showed LIBS to be of comparable accuracy, especially in low-level copper samples.

Constraint-Dependent J-R Curves of a Dissimilar Metal Welded Joint for Connecting Pipe-Nozzle of Nuclear Pressure Vessel

2015 ◽  
Vol 137 (2) ◽  
Author(s):  
J. Wang ◽  
G. Z. Wang ◽  
F. Z. Xuan ◽  
S. T. Tu
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Welded Joint ◽  
Pressure Vessels ◽  
Stress Constraint ◽  
Dissimilar Metal ◽  
Local Strength ◽  
T Stress

In this paper, the J-R curves of two cracks (A508 HAZ crack 2 and A508/Alloy52Mb interface crack 3) located at the weakest region in an Alloy52M dissimilar metal welded joint (DMWJ) for connecting pipe-nozzle of nuclear pressure vessel have been measured by using single edge-notched bend (SENB) specimens with different crack depths a/W (different constraint). Based on the modified T-stress constraint parameter τ*, the equations of constraint-dependent J-R curves for the crack 2 and crack 3 were obtained. The predicted J-R curves using different constraint equations derived from the three pairs of crack growth amount all agree with the experimental J-R curves. The results show that the modified T-stress approach for obtaining constraint-dependent J-R curves of homogeneous materials can also be used for the DMWJs with highly heterogeneous mechanical properties (local strength mismatches) in nuclear power plants. The use of the constraint-dependent J-R curves may increase the accuracy of structural integrity design and assessment for the DMWJs of nuclear pressure vessels.

Magnetic Testing of Power Plant Steel Deterioration

2014 ◽  
Vol 792 ◽  
pp. 183-188
Author(s):  
István Mészáros ◽  
János Ginsztler
Keyword(s):  
Power Plant ◽  
Thermal Shock ◽  
Power Plants ◽  
Materials Science ◽  
Magnetic Measurements ◽  
Magnetic Measurement ◽  
Base Material ◽  
Time Estimation ◽  
Pressure Vessels ◽  
Engineering Structures

Nowadays, there is increasing importance of the remaining life time estimation of engineering structures. In this work the thermal shock fatigue process induced deterioration of the three different power plant steels was investigated. The tested steels are widely used as steam pipeline base material of power plants. The applied thermal shock fatigue test can model the material degradation due to long term service in high temperature environment. A special AC magnetometer was designed and used for the magnetic measurements at the Department of Materials Science and Engineering of BUTE. In this paper a new high sensitivity magnetic measurement is presented for controlling the thermal shock fatigue deterioration. This measurement technique was developed for non-destructive testing of pipelines and pressure vessels of steam power plants.

2004 Edition of Japanese Fitness-for-Service Code for Nuclear Power Plants

2006 ◽  
Author(s):  
Koichi Kashima ◽  
Tomonori Nomura ◽  
Koji Koyama
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pressure Vessels ◽  
Industrial Safety ◽  
National Rule ◽  
Class 1 ◽  
Asme Code ◽  
Service Inspection ◽  
Core Shroud

JSME (Japan Society of Mechanical Engineers) published the first edition of a FFS (Fitness-for-Service) Code for nuclear power plants in May 2000, which provided rules on flaw evaluation for Class 1 pressure vessels and piping, referring to the ASME Code Section XI. The second edition of the FFS Code was published in October 2002, to include rules on in-service inspection. Individual inspection rules were prescribed for specific structures, such as the Core Shroud and Shroud Support for BWR plants, in consideration of aging degradation by Stress Corrosion Cracking (SCC). Furthermore, JSME established the third edition of the FFS Code in December 2004, which was published in April 2005, and it included requirements on repair and replacement methods and extended the scope of specific inspection rules for structures other than the BWR Core Shroud and Shroud Support. Along with the efforts of the JSME on the development of the FFS Code, Nuclear and Industrial Safety Agency, the Japanese regulatory agency approved and endorsed the 2000 and 2002 editions of the FFS Code as the national rule, which has been in effect since October 2003. The endorsement for the 2004 edition of the FFS Code is now in the review process.

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