Calculation and Measurement Method of Evaluating the Leakage of Radioactive Nitrogen 16N7 in Steam Generators of Nuclear Reactors of KLT-40 Type

2021 ◽  
Vol 41 (4) ◽  
pp. 16-30
Author(s):  
A.P. Elokhin ◽  
S.N. Fedorchenko
Keyword(s):  
Measurement Method ◽  
Nuclear Reactors ◽  
Steam Generators

Extrinsic Calibration of Portable Manipulators

1992 ◽  
Author(s):  
Steven B. Shooter ◽  
Charles F. Reinholtz
Keyword(s):  
Closed Form ◽  
Nuclear Reactors ◽  
Video Feedback ◽  
Sufficient Information ◽  
Steam Generators ◽  
Extrinsic Calibration ◽  
Tool Positioning ◽  
Orthogonal Plane ◽  
Position And Orientation ◽  
Operator Control

Abstract Portable manipulators are installed for operation and then removed upon completion of their task. Typical applications of portable manipulators include the inspection of nuclear reactors, inspection and repair of nuclear steam generators and asbestos removal in buildings. In such operations, it is difficult to precisely position the manipulator at a fixed location within its workplace, yet this is critical for accurate tool positioning. It can be possible, however, to position the tool tip at several points in the environment using video feedback and manual operator control of the manipulator. This provides sufficient information to determine the position and orientation of the manipulator base frame with respect to the environment, hereafter referred to as extrinsic calibration. Following extrinsic calibration, subsequent moves of the manipulator can be automated. This paper describes a closed-form method for performing extrinsic calibration by contacting the tool to a total of six places on three orthogonal plane surfaces of reference.

Comparative Thermal Analyses Between Theoretical Mode and RELAP5 Code Simulation for OTSG of a Small PWR

2020 ◽  
Author(s):  
Baihui Jiang ◽  
Zhiwei Zhou ◽  
Zhaoyang Xia ◽  
Qian Sun
Keyword(s):  
Heat Transfer ◽  
Heat Exchange ◽  
Nuclear Reactors ◽  
Heat Transfer Analysis ◽  
Tube Length ◽  
Outlet Temperature ◽  
Steam Generators ◽  
Inclined Tube ◽  
Heat Transfer Calculation ◽  
Relap5 Code

Abstract Due to the low nuclear safety risk, low initial investment cost and short construction period, integrated small nuclear reactors have received wide attention from all over the world. As advanced new type of nuclear reactors, the technologies of integrated small nuclear reactors are in the process of exploration and development. Steam generators are used as the heat transfer system for energy exchange between the primary and secondary circuit in reactors, and their heat transfer analysis is very important for reactor design and development. Due to the simple structure, strong heat exchange capacity and timely load following, Once-Through Steam Generators (OTSGs) are the mainly used steam generators in the design of integrated small nuclear reactors. RELAP5/MOD4.0 is a commercial software developed by Innovative System Software, LLC for transient analysis of light water reactors (LWR). After years of development and improvement, RELAP5 has been a basic tool for analysis and calculation of various simulators of nuclear power plants. However, when RELAP5 models steam generators, only structural models related to straight pipes can be established, which is very inconvenient for the heat transfer research of Once-Through Steam Generators. In this paper, Once-Through Steam Generators with specified structural parameters are taken as the research object. The heat transfer calculation is performed on the simplified inclined tube models by RELAP5 code and the theoretical calculation of the spiral tube heat transfer models is also carried out. Comparing the steam outlet temperature on the primary and secondary sides, the heat exchange power, the average heat transfer coefficient and the tube length of different heat exchange zones under given primary and secondary side inlet fluid conditions, it is confirmed that the RELAP5 heat transfer calculation is verified for simplifying Once-Through Steam Generators with inclined tube models.

Heat-Transfer at Supercritical Pressures

2010 ◽  
Author(s):  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Power Plants ◽  
Cooling System ◽  
Nuclear Reactors ◽  
Turbine Blades ◽  
Convection Heat Transfer ◽  
Steam Generators ◽  
Capital Costs ◽  
The 1950S

The first works devoted to the problem of heat transfer at supercritical pressures started as early as the 1930s. E. Schmidt and his associates investigated free-convection heat transfer to fluids at the near-critical point with the objective of developing a new effective cooling system for turbine blades in jet engines. In the 1950s, the idea of using supercritical “steam”-water appeared to be rather attractive for steam generators / turbines to increase thermal efficiency of fossil-fired power plants. Intensive work on this subject was mainly performed in the former USSR and in the USA in the 1950s–1980s. Therefore, the most investigated flow geometry at supercritical pressures is circular tubes with water as the coolant. Currently, using supercritical “steam” in fossil-fired power plants is the largest industrial application of fluids at supercritical pressures. At the end of the 1950s and the beginning of the 1960s, some studies were conducted to investigate the possibility of using supercritical water as a coolant in nuclear reactors. Several concepts of nuclear reactors were developed. However, this idea was abandoned for almost 30 years, and then regained momentum in the 1990s as a means to improve the performance of water-cooled nuclear reactors. Main objectives of using supercritical water in nuclear reactors are increasing the efficiency of modern nuclear power plants, which is currently 30–35%, to circa 43–50%, and decreasing operational and capital costs by eliminating steam generators, steam separators, steam dryers, etc. Therefore, objectives of the current paper are to assess the work that was performed and to understand specifics of heat transfer at supercritical pressures.

Influence of Imperfections on the Collapse of Thick Tubes Subjected to External Pressure

2009 ◽  
Author(s):  
Valentino Di Marcello
Keyword(s):  
Numerical Investigation ◽  
Nuclear Reactors ◽  
Oil Industry ◽  
External Pressure ◽  
Initial Stresses ◽  
Primary System ◽  
Steam Generators ◽  
Water Pipelines ◽  
Conservative Attitude ◽  
Computer Codes

Cylindrical shells pressurized from outside are required for several engineering applications, and a growing need of tubes with significant thickness has been recently experienced in the oil industry (very deep water pipelines) and in the frame of integrated primary system nuclear reactors (steam generators). Their collapse behaviour has been explored little if at all, both experimentally and numerically, as witnessed by the extremely conservative attitude that codes assume for very thick tubes. A numerical investigation has been performed in this context at the Politecnico di Milano, which was originally intended as a support for requesting a relaxation of ASME regulations. In fact, in 2007 Code Case N-759 [1] was approved, permitting significant thickness saving in the tube design. Nevertheless, the numerical investigation was pursued to assess the influence of different parameters, such as eccentricity, initial stresses and material hardening, on the collapse of tubes with diameter to thickness ratios D/t<20. Results are thought to be useful under at least two respects: first, they provide some understanding on the collapse behaviour in a thickness range so far unexplored; secondly, they give an indication on the assumptions on which computer codes ought to be based when numerical analyses are required.

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