scholarly journals Nuclear Power Plant Prestressed Concrete Containment Vessel Structure Monitoring during Integrated Leakage Rate Testing Using Fiber Bragg Grating Sensors

2017 ◽  
Vol 7 (4) ◽  
pp. 419 ◽  
Author(s):  
Jinke Li ◽  
Kaixing Liao ◽  
Xianglong Kong ◽  
Shengyuan Li ◽  
Xinwang Zhang ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Fiber Bragg Grating ◽  
Prestressed Concrete ◽  
Nuclear Power ◽  
Leakage Rate ◽  
Fiber Bragg Grating Sensors ◽  
Structure Monitoring ◽  
Rate Testing ◽  
Vessel Structure

scholarly journals Structural Monitoring and Response Analysis of Prestressed Concrete Primary Containment of a Nuclear Power Plant

2014 ◽  
Vol 86 ◽  
pp. 554-559
Author(s):  
Y.T. Praveenchandra ◽  
Raghupati Roy ◽  
N. Madhusudana Rao ◽  
Arvind Shrivastava
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Prestressed Concrete ◽  
Nuclear Power ◽  
Structural Monitoring ◽  
Response Analysis

LEAPOR: A Computer Code for Leakage-Rate Calculations for Cracks in Cooling Water Piping Systems

2017 ◽  
Author(s):  
Paul T. Williams ◽  
B. Richard Bass ◽  
Terry L. Dickson ◽  
Hilda B. Klasky
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Flow Boiling ◽  
Cooling Water ◽  
Computer Code ◽  
Leakage Rate ◽  
Vapor Generation ◽  
Subcooled Liquid ◽  
Non Equilibrium

This paper describes the development of a new computer code called Leak Analysis of Piping - Oak Ridge (LEAPOR) which calculates estimates for the leakage rate of water escaping from postulated through-wall cracks in a piping segment of a nuclear power plant cooling water system. The ability of nuclear power plant control and safety systems to detect a piping leak prior to breakage is a fundamental requirement of the leak-before-break concept. The design and assessment of leak-detection systems, therefore, requires the determination of through-wall crack leakage rates covering a significant range of operating and flow conditions. For the primary use case of pressurized water reactors, the coolant is subcooled liquid-phase water at high pressures and temperatures, and the leakage flow regimes can range from adiabatic flow boiling (“flashing”) with non-equilibrium vapor generation inside the crack to orifice flow of a subcooled liquid with vapor generation occurring outside of the pipe. The thermohydraulic Henry-Fauske model (with extensions) for non-equilibrium flashing flow through “tight cracks” has been implemented into LEAPOR. A primary driver in the development of LEAPOR has been that its Software Quality Assurance (SQA) requirements included evaluations for correctness, consistency, completeness, accuracy, source code readability, and testability. The new code should be prepared to successfully meet the criteria of formal SQA audits. The attributes of maintainability, portability, and extensibility also informed LEAPOR’s layered software architectural design. The paper presents the results of verification and validation studies carried out with LEAPOR where verification by benchmark comparisons to the results of an independently developed leak rate code and validation against experimental data are described.

scholarly journals A CASE STUDY ON PRE-TENSIONING & POST TENSIONING SYSTEMS OF A PRESTRESSED CONCRETE

2020 ◽  
Vol 5 (2) ◽  
pp. 249-254
Author(s):  
Veerat Srilaxmi ◽  
K. Manju ◽  
M. Vijaya
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Prestressed Concrete ◽  
Nuclear Power ◽  
Shear Capacity ◽  
Depth Ratio ◽  
Concrete Pile ◽  
Post Tensioning

In Present days the pretensioning and post tensioning systems are very popularly using in the major constructions of a structure. The Pre-tensioning and Post-tensioning both methods are used under pre-stressing process. In which has few edges over the orthodox non-stressed structures like greater span to depth ratio, higher moment and shear capacity. These methods are generally adopted in the making of PSC girders, sleepers, Bridges, Slabs in buildings, Concrete Pile, Repair and Rehabilitations, Nuclear Power Plant etc.

scholarly journals Maintenance of Prestressed Concrete Containment Vessels in a Nuclear Power Plant

2016 ◽  
Vol 14 (8) ◽  
pp. 464-474
Author(s):  
Yoshihiro Yamaguchi ◽  
Takashi Kitagawa ◽  
Hideki Tanaka ◽  
Yasumichi Koshiro ◽  
Hideo Takahashi ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Prestressed Concrete ◽  
Nuclear Power
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