Structural integrity evaluation for interference-fit flywheels in reactor coolant pumps of nuclear power plants

2005 ◽  
Vol 19 (11) ◽  
pp. 1988-1997 ◽  
Author(s):  
June-soo Park ◽  
Ha-cheol Song ◽  
Ki-seok Yoon ◽  
Taek-sang Choi ◽  
Jai-hak Park
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
Interference Fit ◽  
Reactor Coolant

Pre-Operational Seismic Walk-Through of NPPs in India

2002 ◽  
Author(s):  
R. S. Soni ◽  
R. K. Mishra ◽  
M. K. Agrawal ◽  
G. R. Reddy ◽  
H. S. Kushwaha ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Seismic Event ◽  
Structural Integrity ◽  
Functional Performance ◽  
Design Intent ◽  
Detailed Procedure ◽  
Construction Activity ◽  
Major Emphasis

In nuclear power plants, it is essential to design the various safety and safety related systems and components of the plant in such a manner that they maintain their structural integrity as well as serve their functional performance during a seismic event. The pre-operational seismic walk-through helps in ensuring the installation of various seismic supports as per design intent, identifying the areas where supports are inadequate, identifying the interaction concerns between the systems of various safety classes and locating the various undesired loose, untied / unanchored components, tools, etc. used during the construction activity. A detailed procedure for the pre-operational seismic walk-through of the NPPs was therefore, prepared. Since the types and locations of seismic supports for the various systems and components of the plant had been already reviewed, the major emphasis during the walk-through was laid on their proper installation.

Design and Maintenance in an IAEA Technical Guidelines on Fluid-Structure Interaction

2014 ◽  
Author(s):  
Fumio Inada ◽  
Tomomichi Nakamura ◽  
Takashi Nishihara ◽  
Shigehiko Kaneko ◽  
Manwoong Kim ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
General Description ◽  
Fluid Structure Interactions ◽  
Flow Induced Vibration ◽  
Energy Agency ◽  
Fluid Structure ◽  
Component Systems

In nuclear power plants, fluid structure interactions (FSI) occurring in component systems can cause excessive forces or stresses to the structures resulting in mechanical damages that may eventually threaten the structural integrity. FSI in the guidelines includes flow-induced vibration, water hammer, and pipewhip. It can also include movement, deformation, or fracture of equipments by tsunami etc. They can be issues of design and maintenance. Authors cannot find complete guidelines to correspond to the FSI phenomena which can be important in the design and maintenance of nuclear power plants. Based on the background, International Atomic Energy Agency (IAEA) has drafted guidelines on FSI. This paper summarizes general description of FSI as well as design and maintenance against FSI.

Benchmark Analysis on Probabilistic Fracture Mechanics Analysis Codes Considering Multiple Cracks and Crack Initiation in Aged Piping of Nuclear Power Plants

2014 ◽  
Author(s):  
Yinsheng Li ◽  
Kazuya Osakabe ◽  
Genshichiro Katsumata ◽  
Jinya Katsuyama ◽  
Kunio Onizawa ◽  
...  
Keyword(s):  
Risk Assessment ◽  
Fracture Mechanics ◽  
Crack Initiation ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
Multiple Cracks ◽  
Benchmark Analysis ◽  
Probabilistic Fracture Mechanics

In recent years, cracks have been detected in piping systems of nuclear power plants. Many of them are multiple cracks in the same welded joints. Therefore, structural integrity evaluation and risk assessment considering multiple cracks and crack initiation in aged piping have become increasingly important. Probabilistic fracture mechanics (PFM) is a rational methodology in structural integrity evaluation and risk assessment of aged piping in nuclear power plants. Two PFM codes, PASCAL-SP and PRAISE-JNES, have been improved or developed in Japan for the structural integrity evaluation and risk assessment considering the age related degradation mechanisms of pipes. Although the purposes to develop these two codes are different, both have almost the same basic functions to obtain the failure probabilities of pipes. In this paper, a benchmark analysis was conducted considering multiple cracks and crack initiation, in order to confirm their reliability and applicability. Based on the numerical investigation in consideration of important influence factors such as crack number, crack location, crack distribution and crack detection probability of in-service inspection, it was concluded that the analysis results of these two codes are in good agreement.

The Application of Concurrent Installation of the RCS Cross Over Leg and the Reactor Vessel Internals

2010 ◽  
Author(s):  
K.-W. Park ◽  
J.-H. Bae ◽  
S.-H. Park
Keyword(s):  
Feasibility Study ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Reactor Vessel ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Study Results ◽  
Under Construction

The reactor vessel internals (RVI) of a pressurized water reactor (PWR) must be installed precisely in the reactor vessel (RV) according to the requirements for levelness, orientation and vertical alignments for its proper functions and structural integrity. For the precise installation, deformation of the RV should be controlled during the RVI installation. Traditionally, the RVI has been installed in the RV after the completion of welding work for large bore pipings in the reactor coolant system (RCS). To reduce installation time, the concurrent installation of the RVI and RCS pipings is investigated. This paper describes the feasibility study on the concurrent installation including the Finite Element Method (FEM) analyses of the RV deformation due to the welding and heat treatment of the pipings. Based on the feasibility study results, the optimum schedule of the RVI installation in parallel with the installation of the cross-over leg pipings (reactor coolant pump inlet pipings) and confirmation measurement locations are developed. Thereby the concurrent installation will be applied to the nuclear power plants under construction in Korea, and it is expected to reduce installation period of 2 months compared to the traditional sequential installation method.

Structural Integrity Evaluation of High Strength Anchor Bolt for Supporting Steam Generator and Reactor Coolant Pump in Nuclear Power Plants

2013 ◽  
Author(s):  
Eun-Mo Lim ◽  
Nam-Su Huh ◽  
Hee-Jin Shim ◽  
Chang-Kyun Oh ◽  
Hyun-Su Kim
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
High Strength ◽  
Nominal Stress ◽  
Coolant Pump ◽  
Anchor Bolt ◽  
Reactor Coolant ◽  
Anchor Bolts

In Korea, a fitness-for service evaluation for assuring structural integrity of high strength anchor bolts which support nuclear components such as steam generator and reactor coolant pump, has been one of the important issues in nuclear industry. The main failure mechanism of high strength anchor bolts supporting nuclear components might be degradation due to stress corrosion cracking and brittle fracture. In the present work, the structural integrity of high strength anchor bolts which are used to support steam generator and reactor coolant pump of one of the Korean older vintage nuclear power plants is evaluated by adopting a procedure proposed by Electric Power Research Institute (EPRI) based on an elastic fracture mechanics concept. In this EPRI’s procedure, an accurate estimation of nominal stress acting on the cross section of the bolt is a crucial element since a structural integrity of an anchor bolt is evaluated in the EPRI’s procedure using this nominal stress incorporating reference flaw factors reflecting effects of stress concentration due to bolt thread and reference sized surface crack. In this context, detailed elastic finite element stress analyses are firstly performed on the anchor bolt assemblies to come up with nominal stress in the cross-section of anchor bolt. As for loading condition, bolt pretention as well as normal and faulted loads of the anchor bolts were considered. In addition, the structural integrity of the anchor bolts is demonstrated by comparing nominal stresses of anchor bolts with the maximum allowable stresses obtained by using the EPRI’s reference flaw factors and critical fracture toughness. Furthermore, the accuracy of EPRI’s reference flaw factors which are derived on the assumption that reference sized surface crack is existed on the thread roots is investigated using 3-dimensional elastic finite element fracture mechanics analyses.

The Formulation of Material Characteristics of Austenitic Stainless Steels at Extremely High Temperature

2017 ◽  
Author(s):  
Kenta Shimomura ◽  
Takashi Onizawa ◽  
Shoichi Kato ◽  
Masanori Ando ◽  
Takashi Wakai
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
Austenitic Stainless Steels ◽  
Severe Accident ◽  
Integrity Assessment ◽  
Severe Accidents ◽  
Material Characteristics

This paper describes the formulation of material characteristics of austenitic stainless steels at extremely high temperature which meets in some kinds of severe accidents of nuclear power plants. After the severe accident in Fukushima dai-ichi nuclear power plants, it has been supposed to be very important not only to prevent the occurrence of abnormal conditions, i.e. from the first to the third layer safety, but also to prevent the expansion of the accident conditions, i.e. the fourth layer safety[1] [2]. In order to evaluate the structural integrity under the severe accident condition, material characteristics which can be used in the numerical analyses, such as finite element analysis, were required [3] [4]. However, there were no material characteristics applicable to the structural integrity assessment at extremely high temperature. Therefore, a series of tensile and creep tests was performed for austenitic stainless at extremely high temperature which meets in some kinds of severe accidents of nuclear power plants, namely up to 1000 °C. Based on the acquired data from the tests, monotonic stress-strain equation and creep rupture equation applicable to the structural analysis at extremely high temperature, up to 1000 °C were formulated. As a result, these formulae make it possible to conduct the structural integrity assessment using numerical analysis techniques, such as finite element method.

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