Mechanical Characteristics of Nuclear Main Steam Line Penetration Fabricated by Electric Melting Additive Manufacturing

2017 ◽  
Author(s):  
Xiaoming He ◽  
Xiaoning Zhang ◽  
Yongdong Wang ◽  
Yiqing Wang ◽  
Dong Ning
Keyword(s):  
Additive Manufacturing ◽  
Computer Aided Design ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Steam Line ◽  
Electric Melting ◽  
Industry Applications ◽  
Main Steam Line ◽  
Main Steam ◽  
Net Shaping

Additive Manufacturing (AM) can fabricate 3D near-net-shaped functional parts using unit materials, such as powder or wire. Additive manufacturing’s computer-aided design offers superior geometrical flexibility. The near-net shaping capability also reduces materials waste and increase manufacturing efficiency significantly. These benefits make AM desirable for critical industry applications, such as art, aerospace, ground transportation, and medical. Confident utilization of the technology requires thorough understanding of the AM materials, ensuring both structural integrity and performance requirements are met or exceeded. Safety and economics are essential to nuclear power plant. In this study, mechanical properties of a ferritic steel fabricated by electric melting additive manufacturing (EMAM) technique are studied and compared with ASME SA-336 Gr.F12, which applied to nuclear main steam line penetration, the results are systematically presented and discussed. Key technical issues of application of AM to manufacturing nuclear components are also discussed.

Thermal Hydraulic Performance Analysis for Passive Containment Cooling System of AP1000 in the Event of LOCA

2017 ◽  
Author(s):  
Talha Bin Mujahid ◽  
Yu Yu ◽  
Bin Wang ◽  
Muhammad Ali Shahzad ◽  
Fenglei Niu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Safety Analysis ◽  
Hydraulic Performance ◽  
Steam Line ◽  
Loss Of Coolant ◽  
Main Steam Line ◽  
Main Steam

The design of a nuclear reactor containment building is of key importance in order to enhance the safety of a nuclear power plant. Owing to nuclear accidents such as TMI, Chernobyl and Fukushima, more and more attention is paid to the passive concept in nuclear power development. In order to improve the safety of new generation nuclear power plant, passive systems are widely used, passive containment cooling system in AP1000 is one of the typical example of such kinds of systems. It’s function is to transfer the heat produced in the containment to the atmosphere and keep the pressure in the vessel below the threshold under such accidents as Loss of coolant (LOCA), main steam line break (MSLB), etc. The system operates based on natural circulations inside the steel vessel and in the air baffle outside the containment, and the cooling water is sprayed to the steel surface to enhance the heat transfer process. A proper model simulating the system behavior is needed for system design and safety analysis, and a multivolume lumped parameter approach is employed in order to analyze the containment integrity and to study the long term response of postulated Loss of coolant (LOCA) accidents and Main steam line break (MSLB) accidents. However, the temperature and pressure distributions cannot be described detailed by such model, which is important to study the T-H characteristics in the containment. In this paper LOCA has been simulated on MATLAB using a given pipe break size and the response of containment is analyzed. Furthermore, the results are compared with the results in the Westinghouse Design Control Document 2002. Then the thermal hydraulic performance is studied, the factors such as the air temperature, containment pressure and mass flow rate of the coolant and their effects on the containment are analyzed. This research is done to get further insight on the safety analysis of reactor containment regarding maximum temperature and stress calculation inside the containment.

On-Line Workload Assessment by the PET Method

2009 ◽  
Author(s):  
Gueorgui Petkov ◽  
Vladislav Georgiev ◽  
Emil Stefanov
Keyword(s):  
Nuclear Power ◽  
Reliability Assessment ◽  
Steam Line ◽  
Simulator Training ◽  
On Line ◽  
Nasa Tlx ◽  
Assessment Performance ◽  
Main Steam Line ◽  
Main Steam ◽  
Operator Workload

The paper presents benchmarking workload results obtained by a method for operator workload assessment - NASA TLX and a method for human reliability assessment - Performance Evaluation of Teamwork - PET. Based on the archives of nuclear power plant full-scope simulator training in the accident “Main Steam Line Tube Rupture at the WWER-1000 Containment” the capacities of the two methods for indirect and direct or even on-line workload assessment are presented and discussed.

Main Steam Line Technical Design Improvement Using a More Detailed Piping Analysis

2002 ◽  
Author(s):  
Zsolt Revesz
Keyword(s):  
Nuclear Power ◽  
Steam Line ◽  
Piping System ◽  
Seismic Loads ◽  
Structural Elements ◽  
Configuration Change ◽  
Simple Support ◽  
Main Steam Line ◽  
Main Steam ◽  
Computational Resources

The paper describes a case in which with more detailed analysis of the seismic loads and with more elaborated, more accurate model the Main Steam Line of a 1000 MWe class nuclear power plant with BWR has been redesigned. Re-design and re-qualification has resulted in a new piping configuration with significantly reduced number of installed snubbers, and in a more simple support system. The analysis of the piping system has been repeated over twenty years after the original work, whereas the main differences consisted of changes in the piping code and the use of the abundant computational resources available today — besides revision of the earthquake analysis of the buildings. Final recommendation consisted in two suggested configuration change packages. One with a significant reduction in the number of earthquake restraints snubbers) with no need to further analyze elements of the support structure before implementation. A second one with an additional possibility to reduce the number of snubbers, but requiring analysis of piping support structures to see if the higher loads are still allowable or else if a reasonable hardware changes would reinforce structural elements to carry the new load.

Evaluation of LBB characteristics of candidate materials for main steam line piping in Korea nuclear power plants

2020 ◽  
Vol 188 ◽  
pp. 104226
Author(s):  
Seokmin Hong ◽  
Jongmin Kim ◽  
Maan-Won Kim ◽  
Hong-Deok Kim ◽  
Bong-Sang Lee ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Steam Line ◽  
Main Steam Line ◽  
Main Steam

Fretting-Wear Damage at Hydraulic Snubber Pins

2005 ◽  
Author(s):  
Robert Engel
Keyword(s):  
Nuclear Power ◽  
Degradation Mechanism ◽  
Power Level ◽  
Measurement Data ◽  
Steam Line ◽  
Fretting Wear ◽  
Vibration Measurement ◽  
Flow Induced Vibration ◽  
Main Steam Line ◽  
Main Steam

During the refueling outage at the Leibstadt Nuclear Power Plant in August 2002 and 2003 some severely worn hydraulic snubber pins were found at a main steam line support inside the turbine building. Possible causes of the failure were considered to be increased loads due to the implementation of a snubber reduction program in the 1996 to 1998 time period, or flow-induced vibration from the extended power uprate implemented in four steps from 1998 to 2002. During the 2003 outage, a vibration measurement program as well as a root cause analysis were started to check the integrity of the whole system and to reduce or eliminate the degradation mechanism. The dynamic measurement data obtained during the 2003 power ascension showed that the vibration acceptance criteria were met. However, the main steam line support with the worn snubber pins experienced a horizontal relative movement between the pins and the clamps. This started at a power level corresponding to an intermediate power uprate step. Four different pragmatic local design solutions were considered to prevent the severe wear.

Oconee Nuclear Power Station Main Steam Line Break Analysis for Steam Generator Tube Stress Evaluation

2004 ◽  
Vol 148 (1) ◽  
pp. 48-55 ◽  
Author(s):  
Jan S. Muransky ◽  
John G. Shatford ◽  
Craig E. Peterson ◽  
Gregg B. Swindlehurst
Keyword(s):  
Nuclear Power Station ◽  
Steam Generator ◽  
Nuclear Power ◽  
Steam Line ◽  
Power Station ◽  
Steam Generator Tube ◽  
Stress Evaluation ◽  
Main Steam Line ◽  
Main Steam
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