Study on Load Capacity of Main Steam Superpipe Nozzle Used in Nuclear Power Plant

2017 ◽  
Author(s):  
Zhou Gengyu ◽  
Liang Shuhua ◽  
Sun Lin ◽  
Lv Feng
Keyword(s):  
Finite Element ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Load Capacity ◽  
Operating Conditions ◽  
Integrity Assessment ◽  
Inner Radius ◽  
Main Steam

The main steam super pipe used in nuclear power plant is an important safety class2 component. There are several nozzles located on it and linked with main steam safety valves. In the past two decades, the hot extrusion forming technology has been widely used to manufacture the super pipe nozzles. Comparing with traditional insert weldolet, the wall thickness of the extruded nozzle is relative small, and the nozzle inner radius is hard to control precisely in the fabrication process. Due to high temperature working condition and complicated loading conditions, the load capacity of the super pipe extruded nozzle has become an issue of concern for manufacturers and users. This paper presents a structural integrity assessment of a super pipe extruded nozzle. The nozzle stresses due to internal pressure and external loads for different operating conditions are obtained by the three-dimensional finite element analysis. The extruded nozzle is evaluated against the RCCM code Subsection C3200 Service Levels O, B and D stress limits for design, upset and faulted conditions. A parametric sensitivity analysis of the extruded nozzle inner radius size is also carried out. In addition, in order to reduce the calculation effort, an efficient calculation method is developed by using the commercial finite element program ANSYS.

Advances in Residual Stress Modeling and Measurement for the Structural Integrity Assessment of Welded Thermal Power Plant

2008 ◽  
Vol 41-42 ◽  
pp. 391-400 ◽  
Author(s):  
Lyndon Edwards ◽  
Mike C. Smith ◽  
Mark Turski ◽  
Michael E. Fitzpatrick ◽  
P. John Bouchard
Keyword(s):  
Residual Stress ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Residual Stresses ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Economic Life ◽  
Life Assessment ◽  
Integrity Assessment

The safe operation of both thermal and nuclear power plant is increasingly dependent upon structural integrity assessment of pressure vessels and piping. Furthermore, structural failures most commonly occur at welds so the accurate design and remnant life assessment of welded plant is critical. The residual stress distribution assumed in defect assessments often has a deciding influence on the analysis outcome, and in the absence of accurate and reliable knowledge of the weld residual stresses, the design codes and procedures use assumptions that yield very conservative assessments that can severely limit the economic life of some plant. However, recent advances in both the modeling and measurement of residual stresses in welded structures and components open up the possibility of characterising weld residual stresses in operating plant using state-of–the–art fully validated Finite Element simulations. This paper describes research undertaken to predict residual stresses in stainless steel welds in order to provide validated reliable, accurate Structural Integrity assessment of nuclear power plant components

Development of an Integrity Evaluation System on the Basis of Cooperative Virtual Reality Environment for Nuclear Power Plant

2004 ◽  
Author(s):  
J. C. Kim ◽  
J. B. Choi ◽  
Y. H. Choi
Keyword(s):  
Virtual Reality ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Evaluation System ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Virtual Reality Environment ◽  
Reactor Pressure

Since early 1950’s fracture mechanics has brought significant impact on structural integrity assessment in a wide range of industries such as power, transportation, civil and petrochemical industries, especially in nuclear power plant industries. For the last two decades, significant efforts have been devoted in developing defect assessment procedures, from which various fitness-for-purpose or fitness-for-service codes have been developed. From another aspect, recent advances in IT (Information Technologies) bring rapid changes in various engineering fields. IT enables people to share information through network and thus provides concurrent working environment without limitations of working places. For this reason, a network system based on internet or intranet has been appeared in various fields of business. Evaluating the integrity of structures is one of the most critical issues in nuclear industry. In order to evaluate the integrity of structures, a complicated and collaborative procedure is required including regular in-service inspection, fracture mechanics analysis, etc. And thus, experts in different fields have to cooperate to resolve the integrity problem. In this paper, an integrity evaluation system on the basis of cooperative virtual reality environment for reactor pressure vessel which adapts IT into a structural integrity evaluation procedure for reactor pressure vessel is introduced. The proposed system uses Virtual Reality (VR) technique, Virtual Network Computing (VNC) and knowledge based programs. This system is able to support 3-dimensional virtual reality environment and to provide experts to cooperate by accessing related data through internet. The proposed system is expected to provide a more efficient integrity evaluation for reactor pressure vessel.

Critical Section Selection Methodology for the U.S. EPR™ Standard Nuclear Power Plant

2010 ◽  
Author(s):  
Se-Kwon Jung ◽  
Adam Goodman ◽  
Joe Harrold ◽  
Nawar Alchaar
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Structural Design ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Safety Evaluation ◽  
Critical Section ◽  
Selection Methodology ◽  
Critical Sections ◽  
The U.S

This paper presents a three-tier, critical section selection methodology that is used to identify critical sections for the U.S. EPR™ Standard Nuclear Power Plant (NPP). The critical section selection methodology includes three complementary approaches: qualitative, quantitative, and supplementary. These three approaches are applied to Seismic Category I structures in a complementary fashion to identify the most critical portions of the building whose structural integrity needs to be maintained for postulated design basis events and conditions. Once the design of critical sections for a particular Seismic Category I structure is complete, the design for that structure is essentially complete for safety evaluation purposes. Critical sections, taken as a whole, are analytically representative of an “essentially complete” U.S. EPR™ design; their structural design adequacy provides reasonable assurance of overall U.S. EPR™ structural design adequacy.

Overview of the nuclear power plant structural integrity research in Finland

1993 ◽  
Vol 55 (1) ◽  
pp. 3-59 ◽  
Author(s):  
K. Törrönen ◽  
P. Aaltonen ◽  
H. Hänninen ◽  
K. Mäkelä ◽  
P. Karjalainen-Roikonen ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Structural Integrity

scholarly journals Neutronic safety analysis of proposed reactor technologies for Ghana’s nuclear power plant using the MCNP code

2019 ◽  
Vol 34 (3) ◽  
pp. 238-242
Author(s):  
Rex Abrefah ◽  
Prince Atsu ◽  
Robert Sogbadji
Keyword(s):  
High Pressure ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Atomic Energy Commission ◽  
Clean Energy ◽  
Safety Analysis ◽  
Operating Conditions ◽  
The Government

In pursuance of sufficient, stable and clean energy to solve the ever-looming power crisis in Ghana, the Nuclear Power Institute of the Ghana Atomic Energy Commission has on the agenda to advise the government on the nuclear power to include in the country's energy mix. After consideration of several proposed nuclear reactor technologies, the Nuclear Power Institute considered a high pressure reactor or vodo-vodyanoi energetichesky reactor as the nuclear power technologies for Ghana's first nuclear power plant. As part of technology assessments, neutronic safety parameters of both reactors are investigated. The MCNP neutronic code was employed as a computational tool to analyze the reactivity temperature coefficients, moderator void coefficient, criticality and neutron behavior at various operating conditions. The high pressure reactor which is still under construction and theoretical safety analysis, showed good inherent safety features which are comparable to the already existing European pressurized reactor technology.

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