scholarly journals A modern data measurement system to study and test thermionic heat to electricity converters

2018 ◽  
Vol 4 (3) ◽  
pp. 179-183
Author(s):  
Andrey Kirillov ◽  
Valeriy Yarygin
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Test System ◽  
Temperature Fields ◽  
Test Facility ◽  
Measurement Systems ◽  
Data Measurement ◽  
Functional Capabilities ◽  
Thermionic Nuclear Power ◽  
A Current

Studies and tests are conducted to determine the performance of thermionic nuclear power plants (TNPP) a stage in which is pre-irradiation testing of laboratory thermionic converters (TIC) with flat and cylindrically shaped electrodes using test facilities fitted with automated data measurement systems (DMS). The TIC volt-ampere characteristics (VAC) are measured in the DMS jointly with the measured test section and experimental test facility temperature fields. The structure and the characteristics of a DMS based on products from ICP DAS Co., Ltd are presented. A developed VAC measurement program providing the operator with a convenient graphic interface and enabling adjustment of the measurement parameters has been considered. The VAC recording errors in the process of measurements have been determined using TIC simulators. The error in the VAC diffusion portion on a simulator (with a current of less than 3 A) is not more than 1%. Thanks to the use of modern components, the developed DMS offers extended functional capabilities for measuring the thermocouple signals in an experimental electrophysical test facility. The DMS structure provides for the convenience of scaling (through a larger number of measuring channels) and makes it possible to add modules from other manufacturers. The experience of operating this DMS will be used to develop the DMS for an in-pile test system designed for similar functions.

The reactor research and test facility

2020 ◽  
pp. 69-79
Author(s):  
Andrey S. KIRILLOV ◽  
Aleksandr P. PYSHKO ◽  
Andrey A. ROMANENKO ◽  
Valery I. YARYGIN
Keyword(s):  
Measurement System ◽  
Nuclear Power ◽  
Power Plants ◽  
Test Facility ◽  
Physical Parameters ◽  
Pumping System ◽  
Current State ◽  
History Of ◽  
Special Measurement ◽  
Space Nuclear Power

The paper describes an overview of the history of development and the current state of JSC “SSC RF-IPPE” reactor research and test facility designed for assembly, research and full-scale life energy tests of space nuclear power plants with a thermionic reactor. The leading specialists involved in development and operation of this facility are represented. The most significant technological interfaces and upgrade operations carried out in the recent years are discussed. The authors consider the use of an oil-free pumping system as part of this facility during degassing and life testing. Proposed are up-to-date engineering solutions for development of the automated special measurement system designed to record NPP performance, including volt-ampere characteristics together with thermophysical and nuclear physical parameters of a ground prototype of the space nuclear power plant. Key words: reactor research and test facility, thermionic reactor, life energy tests, oil-free pumping system, automated special measurement system, volt-ampere characteristics.

Flow and Temperature Fields Measurement Inside Rod Bundle by the Combined Use of PIV and LIF Technique

2018 ◽  
Author(s):  
Xing Li ◽  
Sichao Tan ◽  
Zhengpeng Mi ◽  
Peiyao Qi ◽  
Yunlong Huang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Great Influence ◽  
Reactor Core ◽  
Transfer Characteristic ◽  
Index Of Refraction ◽  
Temperature Fields ◽  
Spacer Grid ◽  
Rod Bundle ◽  
Combined Use

Thermal hydraulic research of reactor core is important in nuclear energy applications, the flow and heat transfer characteristics of coolant in reactor fuel assembly has a great influence on the performance and safety of nuclear power plants. Particle image velocimetry (PIV) and Laser induced fluorescence (LIF) are the instantaneous, non-intrusive, whole-field fluid mechanics measuring method. In this study, the simultaneous measurement of flow field and temperature field for a rod bundle was conducted using PIV and LIF technique. A facility system, utilizing the matching index of refraction approach, has been designed and constructed for the measurement of velocity and temperature in the rod bundle. In order for further study on complex heat and mass transfer characteristic of rod bundle, the single-phase experiments on the heating conditions are performed. One of unique characteristics of the velocity and temperature distribution downstream the spacer grid was obtained. The experimental results show that the combined use of PIV and LIF technique is applied to the measurement of multi-physical field in a rod bundle is feasible, the measuring characteristics of non-intrusive ensured accuracy of whole field data. The whole field experimental data obtained in rod bundle benefits the design of spacer grid geometry.

scholarly journals Simulation of MASPn Experiments in MISTRA Test Facility with COCOSYS Code

2008 ◽  
Vol 2008 ◽  
pp. 1-7 ◽  
Author(s):  
Mantas Povilaitis ◽  
Egidijus Urbonavičius
Keyword(s):  
Boundary Conditions ◽  
Nuclear Power ◽  
Power Plants ◽  
Volume Fraction ◽  
Characteristic Length ◽  
Initial Conditions ◽  
Test Facility ◽  
Work Package ◽  
Acceptable Agreement

An issue of the stratified atmospheres in the containments of nuclear power plants is still unresolved; different experiments are performed in the test facilities like TOSQAN and MISTRA. MASPn experiments belong to the spray benchmark, initiated in the containment atmosphere mixing work package of the SARNET network. The benchmark consisted of MASP0, MASP1 and MASP2 experiments. Only the measured depressurisation rates during MASPn were available for the comparison with calculations. When the analysis was performed, the boundary conditions were not clearly defined therefore most of the attention was concentrated on MASP0 simulation in order to develop the nodalisation scheme and define the initial and boundary conditions. After achieving acceptable agreement with measured depressurisation rate, simulations of MASP1 and MASP2 experiments were performed to check the influence of sprays. The paper presents developed nodalisation scheme of MISTRA for the COCOSYS code and the results of analyses. In the performed analyses, several parameters were considered: initial conditions, loss coefficient of the junctions, initial gradients of temperature and steam volume fraction, and characteristic length of structures. Parametric analysis shows that in the simulation the heat losses through the external walls behind the lower condenser installed in the MISTRA facility determine the long-term depressurisation rate.

A study on accumulator analysis for the valve performance evaluation system of nuclear power plants

Kerntechnik ◽  
2021 ◽  
Vol 86 (1) ◽  
pp. 39-44
Author(s):  
K. Ryu ◽  
T. Lee ◽  
D. Baek ◽  
J. Park ◽  
N. Kim
Keyword(s):  
Latent Heat ◽  
Temperature Difference ◽  
Evaluation System ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
High Capacity ◽  
Test System ◽  
Performance Evaluation System ◽  
High Temperature Steam

Abstract To evaluate the valves used in the nuclear power plants are working properly under the required conditions, the performance and capacity test should be performed. In the test system, the accumulator was employed to control the large amount of high pressure and high temperature steam generated in the boiler precisely. In the accumulating process, the steam is often condensed. In order to prevent condensation, it is needed to install heaters and preheat the accumulator. However, if the size of the accumulator becomes large, the installation of the heater may not be easy. Therefore, when the test is conducted, the system was preheated by the latent heat generated from the phase change. Insufficient thermal insulation may cause temperature differences and it can cause mechanical problems in the accumulator structure. If insulation is sufficient, the temperature difference is indicated by the height. As the cooled condensate moves downwards, the condensate is discharged by the drain valve control and the temperature difference of the structure can be disappeared. The results of this paper can be applied to the conceptualization of equipment that uses latent heat and for the design of high-precision steam experimental devices or the design of high-capacity steam utilization systems.

Setup of the Supercritical CO2 Test Facility “SCARLETT” for Basic Experimental Investigations of a Compact Heat Exchanger for an Innovative Decay Heat Removal System

2018 ◽  
Vol 4 (3) ◽  
Author(s):  
Wolfgang Flaig ◽  
Rainer Mertz ◽  
Joerg Starflinger
Keyword(s):  
Heat Exchanger ◽  
Supercritical Co2 ◽  
Nuclear Power ◽  
Power Plants ◽  
Heat Removal ◽  
Test Facility ◽  
Experimental Investigations ◽  
Decay Heat ◽  
Heat Removal System ◽  
Removal System

Supercritical fluids show great potential as future coolants for nuclear reactors, thermal power, and solar power plants. Compared to the subcritical condition, supercritical fluids show advantages in heat transfer due to thermodynamic properties near the critical point. A specific field of interest is an innovative decay heat removal system for nuclear power plants, which is based on a turbine-compressor system with supercritical CO2 as the working fluid. In case of a severe accident, this system converts the decay heat into excess electricity and low-temperature waste heat, which can be emitted to the ambient air. To guarantee the retrofitting of this decay heat removal system into existing nuclear power plants, the heat exchanger (HE) needs to be as compact and efficient as possible. Therefore, a diffusion-bonded plate heat exchanger (DBHE) with mini channels was developed and manufactured. This DBHE was tested to gain data of the transferable heat power and the pressure loss. A multipurpose facility has been built at Institut für Kernenergetik und Energiesysteme (IKE) for various experimental investigations on supercritical CO2, which is in operation now. It consists of a closed loop where the CO2 is compressed to supercritical state and delivered to a test section in which the experiments are run. The test facility is designed to carry out experimental investigations with CO2 mass flows up to 0.111 kg/s, pressures up to 12 MPa, and temperatures up to 150 °C. This paper describes the development and setup of the facility as well as the first experimental investigation.

Experimental Investigation of Steam Bubble Condensation in Flowing Subcooled Water With Two Different Injection Nozzle Geometries

2013 ◽  
Author(s):  
Suleiman Al Issa ◽  
Patricia B. Weisensee
Keyword(s):  
Test Section ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Bubble Diameter ◽  
Large Diameter ◽  
Low Pressure ◽  
Test Facility ◽  
Experimental Investigations ◽  
Subcooled Water

A multiphase flow test facility was built at the Department of Nuclear Engineering at the Technical University Munich. The main goal of this facility is to investigate the condensation of steam bubbles injected into a vertical large diameter pipe (104 mm) with flowing subcooled water (6–15 K) at low pressure conditions (1.1–1.45 bar). Current experimental investigations will contribute to a better understanding of subcooled boiling at low pressures, accidental conditions in nuclear power plants and low-pressure research reactors and correlations for the validation of CFD codes. The test section is a 1 m long transparent pipe that is surrounded by an 18×18 cm rectangular “aquarium” filled with distilled water for refraction correction. High-speed camera (HSC) recording was used to gather data about condensing bubbles including bubble diameter, shape and rising velocity. Steam was injected via two different vertical injection nozzles with an inner diameter of 4 and 6 mm, respectively, directly into the center of the test section. The present experiments were carried out at three different steam superficial velocities, water superficial velocities and water temperatures leading to bubble diameters up to 50 mm and bubble relative velocities around 1 m/s. The measurements enabled the calculation of bubble Reynolds and Nusselt numbers and comparison with correlations given in literature. Even though significant differences could be observed between the two injection nozzles with respect to the bubble’s diameter and velocity, the Nusselt and Reynolds numbers are in the same range of values. The bigger bubbles of the 6 mm with respect to the 4 mm nozzle are almost neutralized by the lower rising velocities.

Implementation of an Approach for Characterisation of Radiological Inventory for Decommissioning at Chinon A2 NPP

2014 ◽  
Author(s):  
A. Traichel ◽  
F. Tardy ◽  
M. Mummert
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Statistical Processing ◽  
Analysis Data ◽  
Measured Data ◽  
Primary Circuit ◽  
Measurement Systems ◽  
Definition Of ◽  
Local Sampling

A general overview of the existing radioactive inventory in the plant is necessary for the decommissioning of nuclear power plants. Based on the knowledge about radiological inventory, appropriate decommissioning techniques and procedures can be specifically used. In order to derive the existing radiological activity in the facility a study was carried out to obtain a representative overview of the total radiological situation at the NPP. Within a study a generic methodology for the radiological characterization was developed. This methodology has been applied on the CO2-circuit of the gas-cooled, graphite-moderated reactor Chinon A2 (MAGNOX type). This paper covers the implementation of an approach for characterisation of radiological inventory for decommissioning. The approach aims at the definition of the number and distribution of local sampling, required measurements as well as suitable measurement systems leading to a confident result with minimized effort in sampling. The paper covers two main objectives: 1. Methodology at and 2. Determination of radiological inventory based on measured data. The proposed methodology is a stepwise procedure which offers the possibility for minimizing the number of required measurements/sample analyses. At the first step the underlying system is an “as-simple-as-possible”-example with homogeneous contamination. In a second step the methodology is expanded to a more realistic and complex system, for which additional investigations have to be performed. The determination of the radiological inventory using the methodology has to consider a given confidence level and maximum allowed error. Therefore statistical assessment is widely used in estimations. The result of this first part of study generates the basis for further investigation. This comprises application of methodology to the mentioned technical system. Therefore corresponding measurement and analysis data have to be delivered and proven regarding adequacy for the proposed methodology. From the dataset various measurement systematic and retained parameters could be derived. The accuracy of given measured data was checked by further examination. The result of the performed analysis leads to a statement about the activity in the primary circuit. The result of this study is an comprehensive estimation of the activity by defined statistical processing of analysed data. The result consists moreover of the analysis of the measurement plan and of distribution and deviation within the measured data. Suggestions for further measurement campaigns are provided based on the deviations and inconsistencies of the data. With the help of these suggestions it should be possible to decrease the number of samples and measuring data as well as improve the comparability of separate measurement processes. Particular potential for improvement of the result for inventory can be seen in a deeper analysis of uncertainties, this was realised and will be explained in the paper.

Seismic Hazard and Damages—Avoiding Disaster Through Simulation, Experiment, and Experience

1999 ◽  
Vol 121 (1) ◽  
pp. 30-36 ◽  
Author(s):  
H. Shibata
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Mechanical Engineering ◽  
Design Procedure ◽  
Shaking Table ◽  
Test Facility ◽  
Design Stage ◽  
Engineering Structures ◽  
Aseismic Design ◽  
Piping Systems

This paper deals with the role of proving tests and a large shaking test facility for equipment and piping systems in conjunction with the development of aseismic design in the field of mechanical engineering, especially for nuclear power plants in Japan. To avoid seismic disaster and damage of equipment and piping systems as well as liquid storages, we had to differentiate the seismic design procedure in mechanical engineering from that for building and civil engineering structures. For this process, the dynamic analysis in this field is more significant than for other fields. The author has been trying to develop aseismic design since the design stage of the first nuclear power plant in 1958 based on his experience as a specialist of mechanical vibration. In the early 1970s, shaking tables were developed for this purpose in Japan. The largest one in Japan is a 1000-ton 2-D table. After the 1995 Kobe earthquake, we have been developing a new 1200-ton 3-D shaking table. In the paper, the author discusses the necessity of such a facility and presents a new concept of a numerical shaking table.

Sign in / Sign up

Export Citation Format

Share Document