Experimental studies of temperature pulsations during the process of mixing non¬isothermal coolant flows in nuclear reactor equipment components

The investigation on the heat transfer characteristics for supercritical pressure water (SCW) is of value for the development of the supercritical water-cooled nuclear reactor (SCWR). As an important heat transfer enhancement element, heat transfer for SCW in internally-ribbed tubes was still not solved, though lots of experimental studies have been published and a great many heat transfer correlations were proposed. This paper presented an analysis of heat transfer in the internally-ribbed tubes, through comparing heat transfer correlations for SCW gained from different internally-ribbed tubes under the same operating condition. It was found that all existing heat transfer correlations reported could not been well applied for various internally-ribbed tubes with large deviation between prediction results and experimental values, because rib geometry had a great influence on heat transfer of internally-ribbed tubes. On the basis of experimental data collected from open literature for internally-ribbed tubes, a new general calculation correlation of heat transfer coefficient for SCW was developed for various internally-ribbed tubes by combining an optimized empirical correlation for vertically-upward smooth tubes and four dimensionless numbers of rib geometry. The results show that the calculated values of the new present correlation is in reasonable agreement with available experimental data collected. Moreover, the new correlation was verified well by experiment data of two new-type internally-ribbed tubes performed beyond the above experimental database.

Download Full-text

Experimental Studies on the Effect of Drilling Parameters on Monel Alloy

Materials Science Forum ◽

10.4028/www.scientific.net/msf.979.137 ◽

2020 ◽

Vol 979 ◽

pp. 137-141

Author(s):

K. Jayakumar ◽

K. Akash Koundinya ◽

T. Jayakumar ◽

M. Harshal ◽

G. Gopinath

Keyword(s):

Process Parameters ◽

Gas Turbines ◽

Nuclear Reactor ◽

Combustion Engine ◽

Experimental Studies ◽

Tool Material ◽

High Strength ◽

Chemical Components ◽

Optimum Process ◽

Machining Processes

Monel K-500 is a Nickel based super alloy which have superlative properties such as high strength, resistance to corrosion, retention of mechanical properties at service temperature and structural stability. It finds wide application in aerospace application, nuclear reactor, gas turbines, submarines, combustion engine exhaust valve, petro chemical components, heat exchanger, etc. However, during machining of super alloys, difficulties are there due to its low thermal conductivity and work hardening effect. Challenges in conventional machining can be minimized by using suitable machining as well as process parameters. Among the different machining processes, not much work has been initiated on drilling of above super alloys. Normally, drilled holes are used in screws, bolts, shafts, steam pipes, fitting of furniture and other equipments. By considering the applications and difficulties in machining of super alloys, drilling experiment is selected on Monel K-500. Experiments were conducted as per Taguchi’s L9 orthogonal array using process parameters such as different drill tool material, drilling speed, feed and cutting fluids and Surface roughness and MRR values were measured as output responses. Effect of selected process parameters on the above machinability responses were analyzed. Optimum process parameters were identified to improve the machinability of Monel K-500.

Download Full-text

Experimental studies on models of the inside of a nuclear reactor

Meccanica ◽

10.1007/bf02128877 ◽

1972 ◽

Vol 7 (1) ◽

pp. 61-62

Author(s):

A. Mondina ◽

M. Falco

Keyword(s):

Nuclear Reactor ◽

Experimental Studies

Download Full-text

Visualization of the Operational Parameters Archive of a Nuclear Reactor

10.20948/graphicon-2021-3027-413-421 ◽

2021 ◽

Author(s):

Stanislav Ten ◽

Andrey Zagrebayev ◽

Victor Pilyugin

Keyword(s):

Nuclear Reactor ◽

Scientific Visualization ◽

Experimental Studies ◽

Monitoring Systems ◽

Dynamic Visualization ◽

Operational Parameters ◽

Operating Personnel ◽

Protection Systems ◽

Express Analysis

For the most part, the solution of the problem of visualizing any data depends on the structure, size and type of data provided. In this work, the data are archives of RBMK or VVER reactors provided from different control and protection systems of the reactors. Despite the fact that several visualization complexes have already been developed, the visualization task is still relevant due to the necessity to improve the quality of monitoring systems and operational personnel. This paper describes the mathematical apparatus for express analysis of the archive of operational parameters of a VVER nuclear reactor. The developed software makes it possible to carry out express analysis of the VVER reactor archive in terms of plotting altitude or time graphs, as well as using dynamic visualization using the Chernoff faces method, which in turn can provide scientific and practical benefits due to improvement the work quality of operating personnel and conduction analysis of situations that requires additional attention and more detailed analysis. Also, the fundamental concept of this work is the method of scientific visualization, which is widely used in various theoretical and experimental studies. It can be said that the main aim of scientific visualization is to make invisible visible.

Download Full-text

Material Studies for Pulsed High-Intensity Proton Beam Targets

12th International Conference on Nuclear Engineering, Volume 2 ◽

10.1115/icone12-49441 ◽

2004 ◽

Author(s):

Nicholas Simos ◽

Harold Kirk ◽

Hans Ludewig ◽

Peter Thieberger ◽

W.-T. Weng ◽

...

Keyword(s):

High Performance ◽

Nuclear Reactor ◽

Current Knowledge ◽

Experimental Studies ◽

Integrated System ◽

Critical System ◽

Candidate Target ◽

High Performance Systems ◽

Useful Life ◽

Muon Colliders

Intense beams for muon colliders and neutrino facilities require high-performance target stations of 1–4 MW proton beams. The physics requirements for such a system push the envelope of our current knowledge as to how materials behave under high-power beams for both short and long exposure. The success of an adopted scheme that generates, captures and guides secondary particles depends on the useful life expectancy of this critical system. To address the key technical challenges around the target of these initiatives, a set of experimental studies have either been initiated or being planned that include (a) the response and survivability of target materials intercepting intense, energetic protons, (b) the integrity of beam windows for target enclosures, (c) the effects of irradiation on the long-term integrity of candidate target and focusing element materials, and (d) the performance of the integrated system and the assessment of its useful life. This paper presents an overview of what has been achieved during the various phases of the experimental effort including a tentative plan to continue the effort by expanding the material matrix. The paper also attempts to interpret what the experimental results are revealing and seeks for ways to extrapolate to the required intensities and anticipated levels of irradiation and it discusses the feasibility of the proposed approaches to achieving such high-performance systems. Further it explores the connection of accelerator target systems with reactor systems in order to utilize experience data that the nuclear reactor sector has acquired over the years.

Download Full-text

Experimental Studies on the Use of Liquid Lead in a Molten Salt Nuclear Reactor

Nuclear Technology ◽

10.13182/nt83-a33280 ◽

1983 ◽

Vol 63 (2) ◽

pp. 197-208 ◽

Cited By ~ 4

Author(s):

M. Broc ◽

J. Sannier ◽

G. Santarini

Keyword(s):

Molten Salt ◽

Nuclear Reactor ◽

Experimental Studies ◽

Liquid Lead

Download Full-text

Comparative Evaluation of Three Wireless Sensor Network Transceivers in a High Radiation Environment

EPJ Web of Conferences ◽

10.1051/epjconf/202022508007 ◽

2020 ◽

Vol 225 ◽

pp. 08007

Author(s):

Q. Huang ◽

J. Jiang ◽

Y. Q. Deng

Keyword(s):

Nuclear Power ◽

Nuclear Reactor ◽

Experimental Studies ◽

High Dose Rate ◽

Sensor Nodes ◽

Wireless Sensor ◽

Radiation Environment ◽

High Dose ◽

Rate Condition ◽

High Level

This paper presents on the results of radiation studies for three commonly used wireless sensor nodes based on the following protocols: ZigBee, WirelessHART, ISA 100.11a, and network devices built with commercial off-the-shelf (COTS) components. The level of radiation considered is at par with that experienced at Fukushima Daiichi Nuclear Power Plant after the accident. An experimental setup is developed to monitor behaviors of each wireless device and network real-time under the 60Co gamma radiator at The Ohio State University Nuclear Reactor Lab (OSU-NRL). The experimental results have indicated that the performance of the communication channels and wireless signal parameters do not degrade significant under such radiation. However, all the tested devices and networks can only survive for several hours under the high dose rate condition (20 K Rad/h). The results of these experimental studies have provided useful references to those who design and manufacture COTS-based wireless monitoring systems for use in high level radiation environments.

Download Full-text

Experimental studies of temperature pulsations during the process of mixing non-isothermal coolant flows in nuclear reactor equipment components

Nuclear Energy and Technology ◽

10.3897/nucet.5.39319 ◽

2019 ◽

Vol 5 (3) ◽

pp. 225-229

Author(s):

Sergey M. Dmitriev ◽

Alexandr V. Mamaev ◽

Renat R. Ryazapov ◽

Aleksey Ye. Sobornov ◽

Andrey V. Kotin ◽

...

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Single Phase ◽

Experimental Studies ◽

Coolant Flow ◽

Mixing Zone ◽

Stress Strain ◽

Austenite Steel ◽

Experimental Section ◽

Temperature Pulsations

One of the most important scientific and technical tasks of the nuclear power industry is to assure the reactor equipment life and reliability under random temperature pulsations. High-intensity temperature pulsations appear during the process of mixing non-isothermal coolant flows. Coolant thermal pulsations cause corresponding, sometimes very significant, fluctuations in the temperature stresses of the heat-exchange surface metal, which, added to static loads, can lead to fatigue failure of equipment components. The purpose of this work was to conduct an experimental study of the temperature and stress-strain states of a pipe sample under the influence of local stochastic thermal pulsations caused by the mixed single-phase heat coolant flows. To solve the set problems, an experimental section was created, which made it possible to simulate the process of mixing non-isothermal coolant flows accompanied by significant temperature pulsations. The design of the experimental section allowed us to study the thermohydraulic and life characteristics of pipe samples made of austenite steel (60×5 mm). Some tools were developed for measuring the pipe sample stress-strain state and the coolant flow temperature field in the zone of mixed single-phase media with different temperatures. The measuring tools were equipped with microthermocouples and strain sensors. As a result, we obtained experimental data on temperature pulsations, time-averaged temperature profiles of the coolant flow in the mixing zone as well as statistical and spectral-correlation characteristics of thermal pulsations. Based on the results of measuring the relative strains, the values of fatigue stresses in the mixing zone were calculated. In addition, some devices and methods were elaborated to measure the temperature and stress-strain states of the pipe sample under the influence of local stochastic thermal pulsations. The developed experimental section provided thermal-stress loading of the metal surface at a high level of alternating stress amplitudes causing rapid damage accumulation rates. The results were included in the database to verify the method for assessing the fatigue life of structural materials for nuclear power plants as applied to austenite steel 12Cr18Ni10Ti under the influence of random thermal cyclic loads.

Download Full-text