scholarly journals Fracture Analysis and Fatigue Strength Calculation of Anchor Bolt Used in Circulating Water Pump in Nuclear Power Plant

Scanning ◽  
2021 ◽  
Vol 2021 ◽  
pp. 1-8
Author(s):  
Qiang Chen ◽  
Shuai Zu ◽  
Yinhui Che ◽  
Dongxiong Feng ◽  
Yang Li
Keyword(s):  
Fatigue Strength ◽  
Tensile Stress ◽  
Safety Factor ◽  
Nuclear Power ◽  
Power Plants ◽  
Fracture Analysis ◽  
Water Pump ◽  
Axial Vibration ◽  
Circulating Water ◽  
Tool Marks

A circulating water pump is a key equipment of cooling systems in nuclear power plants. Several anchor bolts were broken at the inlet rings of the same type of pumps. The bolts were turned by a special material for seawater corrosion protection. There were obvious turning tool marks at the root of the thread, which was considered as the source of the crack. The fatigue crack extended to the depth of the bolt, causing obvious radiation stripes on the fracture surface, which was a typical fatigue fracture. Obvious overtightening characteristics were found at the head of the broken bolt. Fracture and energy spectrum analysis showed that the bolt was not corroded. The axial vibration of the pump was measured. The static tensile stress along the bolt axis caused by the preload, the axial tensile stress caused by the axial vibration, and the torsional stress were calculated, respectively. According to the fatigue strength theory, the composite safety factor of the bolt fatigue strength was 1.37 when overtightening at 1.2 times the design torque, which was less than the allowable safety factor of 1.5-1.8, so the bolt was not safe, which further verified the conclusion of fracture analysis. The reason for the low safety factor was caused by the overtightening force. The improvement method was to control the bolt preload or increasing the bolt diameter.

Study on 3-D Flow Field of Concrete Volute Conduit of Circulating Water Pump in Nuclear Power Plant

2011 ◽  
Vol 99-100 ◽  
pp. 350-353
Author(s):  
Xiao Bing Sun ◽  
Xu Bin Qiao
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Stokes Equations ◽  
Three Dimensional ◽  
Gradual Change ◽  
Water Pump ◽  
Pump System ◽  
Circulating Water ◽  
Operating Modes

As the largest unit capacity of nuclear power plant at present, the flow conduit of circulating water pump in EPR1750 nuclear power plant is a volute conduit, which is a cast-in-situ conceret structure with complexly gradual change cavity. Therefore, the hydraulic efficiency of circulating water pump is not only related with the design of pump leaves, but also closely related to the design of volute and the complicated spatial type of intake and outtake conduits. With the pump leaves and the intake and outtake conduits of conceret volute as the research model, based on computational fluid dynamics (CFD)and the three dimensional Reynolds averaged Navier-Stokes equations, an analytic model suitable for computation is established to simulate the three-dimensional steady flow in the whole pumping system under different operating modes. By use of the commercial fluid-computation softer ANSYS, the distribution of basic physic quantities in the fluid field inside the pump and the conduits is obtained. The analysis and prediction of the performance of pump system are made, and the spatial type design of intake and outtake conduits is evaluated. The calculation results can be referenced to improve the design of pump systems in the similar projects.

Investigation of Effect of Pre-Strain on Very High-Cycle Fatigue Strength of Austenitic Stainless Steels

2008 ◽  
Author(s):  
Takeshi Ogawa ◽  
Motoki Nakane ◽  
Kiyotaka Masaki ◽  
Shota Hashimoto ◽  
Yasuo Ochi ◽  
...  
Keyword(s):  
Fatigue Strength ◽  
Nuclear Power ◽  
Power Plants ◽  
Austenitic Stainless Steels ◽  
Fatigue Tests ◽  
Structural Components ◽  
Very High Cycle Fatigue ◽  
Tension And Compression ◽  
Strained Materials ◽  
Very High

The austenitic stainless steels have excellent mechanical and chemical characteristics and these materials are widely used for the main structural components in the nuclear power plants. A part of structural components using these materials is considered to have strain-history by machining, welding and etc in the process of manufacturing and these parts would be hardened because these materials have a remarkable work-hardening property. On the other hand, conventional studies for the fatigue strength used to be investigated by the results of fatigue tests applying normal specimens without the effect of hardening by pre-strain. This paper describes the effect of large pre-strain on very high cycle fatigue strength of the materials in consideration for the evaluation of strength of actual structures in the nuclear power plants. In order to achieve this purpose, the fatigue tests were carried out with strain hardened specimens. The material served in this study was type SUS316NG. Up to ±20% pre-strain was introduced to the round bar shaped materials by tension and compression load test, and the materials were mechanically machined to the hourglass shaped smooth specimens. On the other hand, the pre-strain of some specimens were introduced after machining so as to study the influence of roughness of the surface of the specimens for the fatigue property. Fatigue tests were conducted by ultrasonic and rotating-bending fatigue test machines and conditions were decided by preliminary examinations to control temperature elevation of the specimen during the fatigue test. The S-N curves obtained from fatigue tests show that increase in magnitude of the pre-strain cause increase in the fatigue strength of the material and this relationship is independent of type of the pre-strains of tension and compression. Though all specimens were fractured by the surface initiated fatigue crack, only one specimen was fractured by the internal crack and so-called “fish-eye” was observed on the fracture surface. However, the internal fracture of the SUS316NG does not cause sudden drop of the fatigue strength. Also, the Vickers hardness tests were carried out to discuss the relationship between fatigue strength and hardness of the pre-strained materials. It is found that the increase in fatigue limit of the pre-strained materials strongly depend on the hardness derived from the indentation size equals to the scale of stage I fatigue crack.

A STUDY ON THE FRETTING FATIGUE LIFE OF ZIRCALOY ALLOYS

2008 ◽  
Vol 22 (11) ◽  
pp. 851-856
Author(s):  
JAE-DO KWON ◽  
DAE-KYU PARK ◽  
SEUNG-WAN WOO ◽  
YOUNG-SUCK CHAI
Keyword(s):  
Fatigue Life ◽  
Fatigue Strength ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Fatigue Behavior ◽  
Fretting Fatigue ◽  
Mechanical Tests ◽  
Fatigue Tests ◽  
Cyclic Damage

Studies on the strength and fatigue life of machines and structures have been conducted in accordance with the development of modern industries. In particular, fine and repetitive cyclic damage occurring in contact regions has been known to have an impact on fretting fatigue fractures. The main component of zircaloy alloy is Zr , and it possesses good mechanical characteristics at high temperatures. This alloy is used in the fuel rod material of nuclear power plants because of its excellent resistance. In this paper, the effect of the fretting damage on the fatigue behavior of the zircaloy alloy is studied. Further, various types of mechanical tests such as tension and plain fatigue tests are performed. Fretting fatigue tests are performed with a flat-flat contact configuration using a bridge-type contact pad and plate-type specimen. Through these experiments, it is found that the fretting fatigue strength decreases by about 80% as compared to the plain fatigue strength. Oblique cracks are observed in the initial stage of the fretting fatigue, in which damaged areas are found. These results can be used as the basic data for the structural integrity evaluation of corrosion-resisting alloys considering the fretting damages.

Performance of Equipment Qualification Safety Factor of Periodic Safety Review for Qinshan III CANDU NPP

2014 ◽  
Author(s):  
Shen Xiaoyao
Keyword(s):  
Safety Factor ◽  
Nuclear Power ◽  
Power Plants ◽  
Phase Iii ◽  
Safety Review ◽  
Plant Safety ◽  
Number Of Factors ◽  
Commercial Operation ◽  
Periodic Safety ◽  
Equipment Qualification

Qinshan phase III is the only CANDU nuclear power plant (NPP) in China, which has two 728MW units, and starts commercial operation on Dec. 31st, 2002 and July 24th, 2003 respectively. According to the Periodic Safety Review of Nuclear Power Plants (NS-G-2.10) issued by IAEA in 2003 and the corresponding Chinese edition HAD103/11 issued by NNSA in 2006, the first PSR of Qinshan III should be carried out in 2012 and 2013 after 10 years of commercial operation. Comprehensive assessment of plant safety is a complex task and is facilitated by dividing it into a number of factors. The equipment qualification (EQ) factor is an important one of them. In this paper, the equipment qualification safety factor review of PSR is carried out for Qinshan III. Firstly, the project background is described. Then objectives, scopes and main review elements of EQ factor review are summarized. Also, the EQ factor review process is emphasized. Finally, the project team and outcome of EQ factor review are given.

Fracture Analysis of Pneumatic Control Valve Rod Used for Nuclear Power Plants

2014 ◽  
Vol 1004-1005 ◽  
pp. 1359-1364
Author(s):  
Lei Yang ◽  
Yan Li ◽  
Lu Zhang ◽  
Yun Ting Lai ◽  
Zhi Feng Luo ◽  
...  
Keyword(s):  
Electron Microscopy ◽  
Scanning Electron Microscopy ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Impact Test ◽  
Control Valve ◽  
Fracture Analysis ◽  
Pneumatic Control ◽  
Scanning Electron

The fracture of a carbon steel pneumatic control valve rod used for Nuclear Power Plants was analyzed in terms of the microstructure, inclusions and fractogragh by means of tensile test, charpy V-Notch impact test, optical microscopy, scanning electron microscopy and energy dispersive spectroscopy. The results indicated that the fracture of the rod is induced by the unqualified chemical composition: a large number of inclusions which distribute in grain boundaries reduce the material plasticity and toughness, and eventually cause fracture.

The Evaluation of Fretting Fatigue Behavior in Room Temperature for INCOLOY Alloy 800

2007 ◽  
Vol 353-358 ◽  
pp. 89-93 ◽  
Author(s):  
Dae Kyu Park ◽  
Seung Wan Woo ◽  
Yong Tak Bae ◽  
Il Sup Chung ◽  
Young Suck Chai ◽  
...  
Keyword(s):  
Fatigue Strength ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fatigue Behavior ◽  
Fretting Fatigue ◽  
Fatigue Tests ◽  
Chromium Alloy ◽  
Alloy 800 ◽  
Incoloy Alloy

Mechanical breakdown often comes from the fatigue in many structural parts and nuclear power plants. Among the fatigue phenomenon, especially fretting fatigue occurs in mechanical joints showing small relative movements between contact surfaces. Although the research was developed for one hundred years, occurrence mechanism is not clearly identified yet. INCOLOY alloy 800 is a iron-nickel-chromium alloy having excellent resistance to many corrosive aqueous media and high-temperature atmospheres. This alloy is used extensively in the nuclear power plants industry, the chemical industry, the heat-treating industry and the electronic industry. In this paper, the effect of fretting damage on fatigue behavior for INCOLOY alloy 800 was studied. Also, various kinds of mechanical tests such as tension and plain fatigue tests are performed. Fretting fatigue tests were carried out with flat-flat contact configuration using a bridge type contact pad and plate type specimen. Through these experiments, it is found that the fretting fatigue strength decreased about 50% compared to the plain fatigue strength. In fretting fatigue, the oblique micro-cracks at an earlier stage are initiated. These results can be used as basic data in a structural integrity evaluation of heat and corrosion resisting alloy considering fretting damages.

Application of Internal Fire PRA in Elimination of Fire Common Modes

2017 ◽  
Author(s):  
Xiang Liu ◽  
Lihua Huang ◽  
Zeyu Liu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Design Principle ◽  
Vulnerability Analysis ◽  
Water Pump ◽  
Safety Level ◽  
Physical Separation ◽  
Design Basis ◽  
Safety Function ◽  
Service Water

During the design and construction phases of nuclear power plants (NPPs), fire protection design principle requires that the safety classified equipment within the nuclear island, the service water pump building and galleries should meet the safety goal whenever a design basis fire occurs. The design of passive facilities, such as structure of buildings and physical separation of equipment, should prevent one single fire from failing all redundant equipment of the same safety function. When one single fire fails all redundant equipment of the same function, it is called a fire common mode. By performing fire vulnerability analysis of fire areas or fire zones, analysts can identify the possible fire common modes, and confirm if they are acceptable or not through further analysis. For NPPs in operation, the vulnerability analysis can be performed by examining the details of plant specific cables and equipment. The confirmed unacceptable common modes should be modified or protected through changing cable routing or adding fire coating, etc. Moreover, by employing internal fire probabilistic risk analysis (PRA), analysts can categorize the identified fire common modes according to severity, assess the changes in the risk of internal fire after elimination, and evaluate the benefit of performing elimination, which provides reference for engineering modification by showing the order of priority of each modification. This article introduces the process of identifying fire common modes, performing internal fire PRA and providing suggestions of modification, and illustrates the application of internal fire PRA in eliminating fire common modes and enhancing the safety level of NPPs.

scholarly journals Heat Supply for the Power-Biological Complex from TPP and NPP

2021 ◽  
Vol 2096 (1) ◽  
pp. 012150
Author(s):  
E Burdenkova
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Supply ◽  
Nuclear Power ◽  
Power Plants ◽  
Environmental Problem ◽  
Waste Heat ◽  
Thermal Power ◽  
Thermal Power Plants ◽  
Circulating Water

Abstract This work is devoted to the problem of utilization of waste heat from condensers of thermal power plants and nuclear power plants. The waste heat of the condensers of TPPs and NPPs, together with the circulating water, enters the environment, causing its thermal pollution. The use of this heat in an energy-biological complex, for example, in fisheries, increases their efficiency and solves an environmental problem. Compared to ordinary ponds, this energy complex has an almost year-round increase in biomass and accelerated maturation of producers. The article presents a developed methodology that makes it possible to assess the effectiveness of such a fishery. Calculations using this method were carried out for a fish farm raising sturgeons on the basis of the waste heat of a nuclear power plant with a VVER-1200 reactor and a K-1200-6.8/50 turbine

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