Evaluation of Deformation and Fracture Behaviors of Nuclear Components Using a Simulated Specimen Under Excessive Seismic Loads

2018 ◽  
Author(s):  
Jin Weon Kim ◽  
Ik Hyun Song ◽  
Hyeong Do Kweon ◽  
Jong Sung Kim ◽  
Yun Jae Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Safety Margin ◽  
Compact Tension ◽  
System Structure ◽  
Seismic Loads ◽  
Cyclic Loads ◽  
Deformation And Failure ◽  
Load Line Displacement ◽  
Deformation And Fracture

This study designed a specimen that can simulate deformation and crack initiation in system, structure, and components (SSCs) of nuclear power plants (NPPs) under excessive seismic loads, and conducted ultimate strength tests using this specimen at room temperature (RT) and 316°C. The specimen designed was a compact tension (CT) type with a round notch, and both SA312 TP316 stainless steel (SS) and SA508 Gr.3 Cl.1 low-alloy steel (LAS) were used in the experiment. Displacement-controlled cyclic loads with constant and random amplitudes were applied as input loads for the test. One set of input loads consisted of 20 cycles, and the input amplitudes of load-line displacement (LLD) were determined to induce the maximum elastic stress of 6∼ 42Sm on the specimen, where Sm is allowable design stress intensity. The input LLD had a triangular waveform and was fully reversed for both types of amplitude. During the test, multiple sets of input cyclic loads, with increasing amplitude of input LLD, were applied to the specimen until a crack was initiated. The results demonstrated that the specimen used in this study adequately simulates the deformation and failure behaviors of SSCs under excessive seismic loads. In addition, the samples in both materials failed under cyclic load levels that were several times higher than those of design basis earthquake (DBE). The SA316 TP316 SS specimen had a greater safety margin under excessive seismic loading conditions than SA508 Gr.3 Cl.1 LAS specimen, regardless of test temperature.

Evaluation of Deformation and Failure Behaviors of Nuclear Piping Components Under Beyond Design Basis Seismic Loads Using a Simulated Specimen

2020 ◽  
Vol 142 (5) ◽  
Author(s):  
Jin Weon Kim ◽  
Ik Hyun Song ◽  
Heong Do Kweon ◽  
Jong Sung Kim ◽  
Yun Jae Kim
Keyword(s):  
Stress Intensity ◽  
Large Amplitude ◽  
Nuclear Power ◽  
Cyclic Load ◽  
Power Plants ◽  
Evaluation Procedure ◽  
Seismic Loads ◽  
Cyclic Loads ◽  
Deformation And Failure ◽  
Design Basis

Abstract This study designed a specimen that simulates the deformation and failure behaviors of the piping components in nuclear power plants (NPPs) under excessive seismic loads beyond the design basis, and conducted ultimate-strength tests using this specimen at room temperature (RT) and 316 °C. SA312 TP316 stainless steel (SS) and SA508 Gr.3 Cl.1 low-alloy steel (LAS) were used in the experiments. Displacement-controlled cyclic loads with constant and random amplitudes of load-line displacement (LLD) were applied as input loads. A set of input cyclic loads consisted of 20 cycles, and the LLD amplitudes of the cyclic load were determined to induce the maximum membrane plus bending stress intensity of 6–42Sm on the specimen, where Sm is the allowable design stress intensity. Multiple sets of input cyclic loads, with increasing amplitude of LLD, were applied to the specimen until cracking initiated. The results demonstrate that the simulated specimen adequately showed the ratcheting deformation and fatigue-induced cracking of piping components under displacement-controlled excessive seismic loads. In addition, samples of both materials failed under displacement-controlled cyclic load levels that were several times higher than those of the design basis earthquake (DBE). The SA316 TP316 SS had greater resistance to failure under large-amplitude cyclic loads than did SA508 Gr.3 Cl.1 LAS. For both materials, resistance to failure was lower at 316 °C than at RT. This study confirmed that the evaluation procedure of the ASME design code predicted the fatigue failure of specimens very conservatively under large-amplitude cyclic loads simulating displacement-controlled excessive seismic loads.

Effect of Thermal Aging on the Deformation and Failure Behaviors of Cast Austenitic Stainless Steels Under Excessive Cyclic Loads

2019 ◽  
Author(s):  
Jin Weon Kim ◽  
Sang Eon Kim ◽  
Yun Jae Kim
Keyword(s):  
Stainless Steels ◽  
Thermal Aging ◽  
Nuclear Power ◽  
Cyclic Load ◽  
Power Plants ◽  
Austenitic Stainless Steels ◽  
System Structure ◽  
Cyclic Loads ◽  
Deformation And Failure ◽  
Negative Effect

Abstract This study conducts failure tests using a simulated specimen to investigate the effect of thermal aging on the deformation and failure behaviors of system, structure, and components (SSCs) of nuclear power plants (NPPs) made of cast austenitic stainless steels (CASSs) under excessive seismic loads. Both unaged and thermally aged CF8A CASSs were used for the experiment, and the large cyclic loads in the form of displacement-control and load-control were applied at a quasi-static displacement rate. Displacement-controlled tests were performed at room temperature (RT) and 316°C and load-controlled tests were performed at RT. The results show that the deformation behaviors of aged CF8A CASS under both types of cyclic load are almost the same as those of unaged CF8A CASS. The thermal aging slightly promotes the failure of CF8A CASS under displacement-controlled cyclic loads, but the failure of specimen still occurs under the cyclic load levels several times higher than the load of the design basis earthquake. Under load-controlled cyclic loads, thermal aging retards the failure of CF8A CASS. Consequently, the thermal aging has no apparent negative effect on the deformation and failure behaviors of CASSs under large cyclic loads, even if it considerably changes the strength, ductility, and fracture toughness of CASSs.

Seismic Fragility of Piping Nozzles in Nuclear Power Plants: A Case for Updating the Current State-of-Practice

2021 ◽  
Author(s):  
Abhinav Gupta ◽  
Ankit Dubey ◽  
Sunggook Cho
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Seismic Analysis ◽  
Nuclear Industry ◽  
Life Extension ◽  
Seismic Loads ◽  
Thermal Loads ◽  
Loss Of Coolant Accident ◽  
Piping Systems ◽  
Recent Developments

Abstract Nuclear industry spends enormous time and resources on designing and managing piping nozzles in a plant. Nozzle locations are considered as a potential location for possible failure that can lead to loss of coolant accident. Industry spends enormous time in condition monitoring and margin management at nozzle locations. Margins against seismic loads play a significant role in the overall margin management. Available margins against thermal loads are highly dependent upon seismic margins. In recent years, significant international collaboration has been undertaken to study the seismic margin in piping systems and nozzles through experimental and analytical studies. It has been observed that piping nozzles are highly overdesigned and the margins against seismic loads are quite high. While this brings a perspective of sufficient safety, such excessively high margins compete with available margins against thermal loads particularly during the life extension and subsequent license renewal studies being conducted by many plants around the world. This paper focuses on identifying and illustrating two key reasons that lead to excessively conservative estimates of nozzle fragilities. First, it compares fragilities based on conventional seismic analysis that ignores piping-equipment-structure interaction on nozzle fragility with the corresponding assessment by considering such interactions. Then, it presents a case that the uncertainties considered in various parameters for calculating nozzle fragility are excessively high. The paper identifies a need to study the various uncertainties in order to achieve a more realistic quantification based on recent developments in our understanding of the seismic behavior of piping systems.

scholarly journals Probability Modeling Taking into Account Nonlinear Processes of a Deformation and Fracture for the Equipment of Nuclear Power Plants

2020 ◽  
Author(s):  
Nikolay Andreevich Makhutov ◽  
Mikhail Matveevich Gadenin ◽  
Igor Alexandrovich Razumovskiy ◽  
Sergey Valerievich Maslov ◽  
Dmitriy Olegovich Reznikov
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nonlinear Processes ◽  
Probability Modeling ◽  
Deformation And Fracture

Over Limit Oxygen Effect on Crack Propagation of WWER 440 RPV Steel in Primary Circuit Water

2009 ◽  
Author(s):  
Miroslava Ernestova ◽  
Anna Hojna
Keyword(s):  
Crack Propagation ◽  
Nuclear Power ◽  
Power Plants ◽  
Water Environment ◽  
Nuclear Research ◽  
Compact Tension ◽  
Oxygen Effect ◽  
Primary Circuit ◽  
Nuclear Research Institute ◽  
Primary Water

Experience with operating nuclear power plants worldwide reveals that many failures may be attributed to fatigue associated with mechanical loading due to vibration and with corrosion effect due to exposure to high-temperature environment. In order to clarify the simultaneous influence on reactor pressure vessel (RPV) material testing of ferritic steel 15Ch2MFA used for RPV of WWER 440 was performed at Nuclear Research Institute (NRI) autoclaves. Cyclic and constant loadings were applied to Compact Tension (CT) specimens in WWER primary water environment at 290°C and simultaneous effect of different oxygen levels (< 20 ppb, 200 ppb, 2000 ppb) on crack propagation has been evaluated. Obtained crack growth rates are compared with ASME XI Code and VERLIFE curves and crack behaviour is discussed.

Failure Probability Analysis of Aged Piping Using Probabilistic Fracture Mechanics Methodology Considering Seismic Loads

2016 ◽  
Author(s):  
Yoshihito Yamaguchi ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Failure Probability ◽  
Power Plants ◽  
Fragility Curves ◽  
Safety Margin ◽  
Probability Analysis ◽  
Seismic Loads ◽  
Probabilistic Fracture Mechanics ◽  
Age Related ◽  
Piping Systems

Several nuclear power plants in Japan have been operating for more than 30 years and cracks due to age-related degradations have been detected in some piping systems during in-service inspections. Furthermore, several of them have experienced severe earthquakes in recent years. Therefore, failure probability analysis and fragility evaluation for piping systems, taking both age-related degradations and seismic loads into consideration, has become increasingly important for the structural integrity evaluation and the seismic probabilistic risk assessment. Probabilistic fracture mechanics (PFM) is recognized as a rational methodology for failure probability analysis and fragility evaluation of aged piping, because it can take the scatters and uncertainties of influence parameters into account. In our Japan Atomic Energy Agency (JAEA), a PFM analysis code PASCAL-SP was developed for aged piping considering age-related degradations. In this study, we improved PASCAL-SP for the fragility evaluation taking both age-related degradations and seismic loads into account. The details of the improvement of PASCAL-SP are explained and some example analysis results of failure probabilities, fragility curves and a preliminary investigation on seismic safety margin are presented in this paper.

Seismic Safety Margin Assessment for Thin-Wall Cylindrical Tanks Utilizing Ultimate Strength Test

2010 ◽  
Author(s):  
Toru Iijima ◽  
Masaki Nakagawa ◽  
Akira Shibuya ◽  
Katsumi Ebisawa ◽  
Hiroyuki Kameda
Keyword(s):  
Ultimate Strength ◽  
Seismic Design ◽  
Nuclear Power ◽  
Power Plants ◽  
Safety Margin ◽  
Strength Test ◽  
Thin Wall ◽  
Seismic Capacity ◽  
Seismic Safety ◽  
Cylindrical Tanks

In Japan, the Seismic Design Guideline for Nuclear Power Plants revised in 2006 requires that residual risk for earthquakes beyond design base be considered. Moreover, in the Niigata-ken Chuetsu-oki earthquake (NCOE:2007), the earthquake motion exceeded the seismic design condition of the Kashiwazaki-Kariwa nuclear power plants (KK NPPs). In response to these issues, there is a growing demand for quantitative clarification of seismic safety margin. The Japan Nuclear Energy Safety Organization (JNES) started a study on the seismic safety margin. JNES defined the term “seismic safety margin” in this study. The seismic safety margin is based on the probability distribution of seismic response and seismic capacity of equipment. Regarding the seismic capacity, JNES has carried out seismic capacity tests on various types of equipment whose malfunction would significantly affect core damage frequency in terms of seismic probabilistic safety assessment (seismic PSA). Our seismic safety margin is effective to understand the quantitative margin related to the failure of equipment. JNES applied the concept of seismic safety margin to thin wall cylindrical tanks utilizing ultimate strength test results, and compared it with the design margin. This paper reports examples of the seismic margin evaluation of cylindrical tank.

Machining Test Specimens From Harvested Zion RPV Segments

2015 ◽  
Author(s):  
Thomas M. Rosseel ◽  
Mikhail A. Sokolov ◽  
Randy K. Nanstad
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
National Laboratory ◽  
Compact Tension ◽  
Department Of Energy ◽  
Oak Ridge ◽  
Damage Models ◽  
The Us ◽  
Machining Test

The decommissioning of the Zion Nuclear Generating Station (NGS) in Zion, Illinois, presents a special and timely opportunity for developing a better understanding of materials degradation and other issues associated with extending the lifetime of existing nuclear power plants (NPPs) beyond 60 years of service. In support of extended service and current operations of the US nuclear reactor fleet, the Oak Ridge National Laboratory (ORNL), through the Department of Energy (DOE), Light Water Reactor Sustainability (LWRS) Program, is coordinating and contracting with Zion Solutions, LLC, a subsidiary of Energy Solutions, an international nuclear services company, the selective procurement of materials, structures, components, and other items of interest from the decommissioned reactors. In this paper, we will discuss the acquisition of segments of the Zion Unit 2 Reactor Pressure Vessel (RPV), cutting these segments into blocks from the beltline and upper vertical welds and plate material and machining those blocks into mechanical (Charpy, compact tension, and tensile) test specimens and coupons for microstructural (TEM, SEM, APT, SANS and nano indention) characterization. Access to service-irradiated RPV welds and plate sections will allow through wall attenuation studies to be performed, which will be used to assess current radiation damage models [1].

Carbon Fiber Reinforcement of a Water Storage Tank for Beyond Design Basis Loads

2016 ◽  
Author(s):  
Alton Reich ◽  
John Charest
Keyword(s):  
Carbon Fiber ◽  
Storage Tank ◽  
Nuclear Power ◽  
Power Plants ◽  
Water Storage ◽  
Safety Margin ◽  
Repair System ◽  
Buckling Mode ◽  
Water Storage Tank ◽  
Design Basis

The severe damage to the Fukushima nuclear plant occurred as a result of a beyond design basis event. This has prompted a systematic review of safety critical systems at US nuclear power plants to evaluate the existing safety margin based on beyond design basis loads. At one US nuclear power plant it was found that the Refueling Water Storage Tank (RWST) did not have sufficient margin to withstand the defined beyond design basis seismic event. An analysis indicated that the RWST would fail in an elephant foot buckling mode. This paper describes the design and analysis of a Carbon Fiber Reinforced Polymer (CFRP) repair system used to strengthen the RWST to increase the critical buckling stress for the elephant foot buckling mode.

Review of Pipe Wall Thinning Mechanism Study and its National Project in Japan

2011 ◽  
Author(s):  
Hiroshi Miyano ◽  
Naoto Sekimura ◽  
Masayuki Takizawa ◽  
Masaaki Matsumoto
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Pipe Wall ◽  
High Reliability ◽  
Safety Margin ◽  
Fiscal Year ◽  
Mechanism Study ◽  
Wall Thinning ◽  
Piping Systems ◽  
The Government

For nuclear power plants, the four major requirements are 1) high safety, 2) high reliability, 3) economical competitiveness, and 4) minimum environmental impact. However, it is still difficult to completely avoid problems concerning structural materials caused by stress corrosion cracking (SCC) and for piping systems caused by flow accelerated corrosion (FAC) and liquid drop impingement (LDI). Since especially FAC and LDI are uncertain phenomena as pipe wall thinning, there are the piping rupture accident risks on all piping systems under the specific conditions. In Japan, after August 2004, the accident of the secondary pipe rupture in Mihama Power Plant Unit 3, The Kansai Electric Power Co., Inc. (KEPCO), R&D projects on pipe wall thinning phenomena and mechanism have been employed by many organizations. On the other hand, evaluation of the safety and reliability of piping systems of long term operating plants and with the special attention to seismic condition have been requested. It was requested to enable evaluation of pipe wall thinning and its reliability with more accuracy. This project was programmed under the government budget from 2006 to 2010 fiscal year according to the Strategy Load-Map for Ageing Management generated by the society of industry, government and academia [1]. As the milestone for the first half decade of the load-map, the project had these achievements: 1) Establish computer program for FAC simulation, 2) Clarify droplet behavior for LDI prediction, 3) Simplified calculation model of pipe wall thinning for seismic evaluation, 4) Evaluate safety margin of thinned piping by FAC or LDI.

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