Scaling analysis of core pressure drop in reduced height integral test facility

Kerntechnik ◽  
2018 ◽  
Vol 83 (3) ◽  
pp. 178-180
Author(s):  
P. Ju ◽  
B. Long ◽  
L. Li ◽  
Q. Su ◽  
X. Wu ◽  
...  
Keyword(s):  
Pressure Drop ◽  
Test Facility ◽  
Scaling Analysis ◽  
Integral Test ◽  
Reduced Height

Simulating Analysis for Small Break LOCA Test Facilities by Applying the Volume Scaling With Reduced Pressure and Reduced Height

2001 ◽  
Author(s):  
S. K. Moussavian ◽  
M. A. Salehi
Keyword(s):  
Experimental Data ◽  
Pressure Drop ◽  
Test Facility ◽  
Integral Test ◽  
Reduced Pressure ◽  
Simple Loop ◽  
Loss Of Coolant Accident ◽  
Reduced Height ◽  
Simulating Analysis ◽  
Loss Of Coolant

Abstract In this paper first we briefly define the different scaling schemes and scaling logic in which we use these schemes to simulate the Small-Break Loss Of Coolant Accident (SB-LOCA) in test facilities. The simple loop of the test facility is considered and the mass, momentum and energy conservation equations are used for the derivation of the scaling model. The variations of mass flow rate, pressure drop and the void fraction in the loop as functions of the time scale or the inventories are obtained. Finally, the calculated results from the simulating schemes are compared with the experimental data previously obtained in an integral test facility.

Scaling Analysis of the Pressurizer Surge Line in Nuclear Plant Integral Test Facilities

2012 ◽  
Author(s):  
Cheng-Cheng Deng ◽  
Hua-Jian Chang ◽  
Ben-Ke Qin ◽  
Han Wang ◽  
Lian Chen
Keyword(s):  
Reactor Core ◽  
Nuclear Plant ◽  
Test Facility ◽  
Scaling Analysis ◽  
Two Phase ◽  
Integral Test ◽  
Loss Of Coolant Accident ◽  
Initial Injection ◽  
Surge Line ◽  
Core Cooling

During small break loss of coolant accident (SBLOCA) of AP1000 nuclear plant, the behavior of pressurizer surge line has an important effect on the operation of ADS valves and the initial injection of IRWST, which may happen at a time when the reactor core reaches its minimum inventory. Therefore, scaling analysis of the PRZ surge line in nuclear plant integral test facilities is important. Four scaling criteria of surge line are developed, which respectively focus on two-phase flow pattern transitions, counter-current flow limitation (CCFL) behavior, periodic draining and filling and maintaining system inventory. The relationship between the four scaling criteria is discussed and comparative analysis of various scaling results is performed for different length scale ratios of test facilities. The results show that CCFL phenomenon and periodic draining and filling behavior are relatively more important processes and the surge line diameter ratios obtained by the two processes’ scaling criteria are close to each other. Thus, an optimal scaling analysis considering both CCFL phenomenon and periodic draining and filling process of PRZ surge line is given, which is used to provide a practical reference to choose appropriate scale of the surge line for the ACME (Advanced Core-cooling Mechanism Experiment) test facility now being built in China.

Scaling analysis and evaluation for the design of integral test facility of HPR1000 containment (PANGU)

2021 ◽  
Vol 373 ◽  
pp. 111035
Author(s):  
Ji Xing ◽  
Zhuo Liu ◽  
Weimin Ma ◽  
Yidan Yuan ◽  
Zhongning Sun ◽  
...  
Keyword(s):  
Test Facility ◽  
Scaling Analysis ◽  
Integral Test ◽  
Analysis And Evaluation

Scaling methodology for a reduced-height reduced-pressure integral test facility to investigate direct vessel injection line break SBLOCA

2008 ◽  
Vol 238 (9) ◽  
pp. 2197-2205 ◽  
Author(s):  
Byoung Uhn Bae ◽  
Keo Hyoung Lee ◽  
Yong Soo Kim ◽  
Byong Jo Yun ◽  
Goon Cherl Park
Keyword(s):  
Test Facility ◽  
Integral Test ◽  
Reduced Pressure ◽  
Reduced Height ◽  
Injection Line

Results and Discussion on Scaling Analysis for the NHR200-II Integral Test Facility Under Single Phase Natural Circulation

2013 ◽  
Author(s):  
Yang Liu ◽  
Qianqian Jia ◽  
Haijun Jia ◽  
Lei Wu
Keyword(s):  
Sea Water ◽  
Natural Circulation ◽  
Density Ratio ◽  
Test Facility ◽  
Scaling Analysis ◽  
Flow Area ◽  
Integral Test ◽  
Primary Loop ◽  
New Energy ◽  
Scale Ratio

The SMR has become great interest to researchers in recent years. The integrated nuclear heating reactor NHR200 was developed by Institute of Nuclear and New Energy Technology (INET) in Tsinghua University. On the basis of the NHR200, the NHR200-II is developed. The thermal-hydraulic parameters in the primary loop such as temperature and pressure are increased, which make it more suitable for industrial steam supply and sea water desalinization. Because the primary loop of NHR200-II operates with fully natural circulation, it is necessary to investigate the natural circulation characteristics by experiment under new thermal parameters. In order to determine the characteristics scale of the test facility, the scaling analysis is needed. The similarity groups, including Richardson number, etc. are determined from the dimensionless fluid and solid governing equations. The test facility uses the fluid with same property as the prototype to reduce the distortion. The axial length scale ratio of the test facility is set to be 1:1. The analysis results of the core surface average heat flux density ratio and the flow area ratio are 1:1 and 1:210 respectively. At the end of this paper, the results of the scaling analysis are present.

Integral Effect Test for a Inadvertent Opening of POSRV Using the ATLAS Facility

2010 ◽  
Author(s):  
Seok Cho ◽  
Ki-Yong Choi ◽  
Hyun-Sik Park ◽  
Kyoung-Ho Kang ◽  
Yeon-Sik Kim ◽  
...  
Keyword(s):  
Nuclear Industry ◽  
Test Facility ◽  
System Response ◽  
Integral Test ◽  
Hydraulic Test ◽  
Pressurized Water ◽  
Reduced Height ◽  
Test Loop ◽  
Integral Effect ◽  
Effect Test

A thermal-hydraulic integral effect test facility for advanced pressurized reactors (PWRs), ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been operated by KAERI (Korea Atomic Energy Research Institute). The reference plant of the ATLAS is a 1400 MWe-class evolutionary pressurized water reactor (PWR), the APR1400 (Advanced Power Reactor 1,400 MWe), which was developed by the Korean industry. The ATLAS has a 1/2 reduced height and a 1/288 volume scaled integral test facility with respect to the APR1400. It has a maximum power capacity of 10% of the scaled nominal core power, and it can simulate full pressure and temperature conditions of the APR1400. The ATLAS could be used to provide experimental data on design-basis accidents including the reflood phase of a large break loss of coolant accident (LBLOCA), small break LOCA (SBLOCA) scenarios including the DVI line and cold leg breaks, a steam generator tube rupture, a main steam line break, a feed line break, etc. An inadvertent opening of POSRV test (SB-POSRV-02) was carried out as one of the SBLOCA spectra. The main objectives of this experimental test were not only to provide physical insight into the system response of the APR1400 reactor during a transient situation but also to present integral effect data for the validation of the SPACE (Safety and Performance Analysis Computer Code), which is now under development by the Korean nuclear industry.

The Research and Comparison of Thermal-Hydraulic Test Facilities for Pressurized Water Reactors (PWRs)

2017 ◽  
Author(s):  
Yuan Lu ◽  
Changzhi Xiao
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Nuclear Safety ◽  
Test Facility ◽  
Scaling Analysis ◽  
Volume Data ◽  
Integral Test ◽  
Hydraulic Test ◽  
Pressurized Water ◽  
Wide Range

Recently, nuclear power safety draws more attention after Fukushima nuclear accident, for which it is essential to construct a large number of test facilities simulating possible issues occurred in the reactor. The thermal-hydraulic test facility is extensively used to simulate thermal-hydraulic response during a loss of coolant accident (LOCA) or an operational transient which can minimize the nuclear safety accidents. This paper focus on the research of thermal-hydraulic test facilities of PWRs in different countries. All of facilities were designed by scaling analysis method. Meanwhile, a wide range of data comprising of power data, pressure data, volume data, configuration and a series of nuclear safety test data is compared in over ten test facilities. Based on above comparable data and relevant research, the main conclusions are as follows: Scaling analysis lays a solid foundation for the design and construction of scale-down nuclear reactor thermal-hydraulic test facilities. This would provide the reference for choosing scaling concepts in the reactor integral test facility.

Influence of Electric Fields on Flow of Liquid Crystal Mixture in a Circular Tube With Electrode Surface

2003 ◽  
Author(s):  
Tetsuhiro Tsukiji ◽  
Tsuyoshi Mitani
Keyword(s):  
Liquid Crystal ◽  
Pressure Drop ◽  
Electric Fields ◽  
Pulse Width Modulation ◽  
Circular Tube ◽  
Pulse Frequency ◽  
Open Loop ◽  
Test Facility ◽  
Pulse Frequency Modulation ◽  
Crystal Mixture

Liquid crystal is one of functional fluids to control an apparent viscosity using an electric field intensity. It is also called ER (Electro-rheological) fluids. In the present experiment a liquid crystal mixture made of some kinds of the nematic liquid crystal is used. The responses of the pressure drop are examined when the liquid crystal mixture flows in a circular tube with the electrode walls on some parts of the inner surface of the tube for the constant flow rates. The four pair of the electrode is used and the voltages are added in the peripheral direction. When the voltages are applied on the liquid crystal mixture and removed, the pressure responses of the inlet of the circular tube are measured with the pressure transducer. On the other hand, the pulse-wave voltages are added to the electrodes to control the pressure drop using the pulse width modulation or the pulse frequency modulation. The diameter of the circular tube is 1.0mm. The isotropic-nematic transition is 90.0°C and smectic-nematic transition is −44.0°C for the liquid crystal mixture. The open-loop test facility with the liquid crystal mixture is set in a pyrostat to keep the temperature constant.

scholarly journals Consistent Posttest Calculations for LOCA Scenarios in LOBI Integral Facility

2012 ◽  
Vol 2012 ◽  
pp. 1-16 ◽  
Author(s):  
F. Reventós ◽  
P. Pla ◽  
C. Matteoli ◽  
G. Nacci ◽  
M. Cherubini ◽  
...  
Keyword(s):  
Experimental Data ◽  
Mass Flow ◽  
Scaling Factor ◽  
Test Facility ◽  
Data Bases ◽  
Integral Test ◽  
The Core ◽  
Active Elements ◽  
Accuracy Data

Integral test facilities (ITFs) are one of the main tools for the validation of best estimate thermalhydraulic system codes. The experimental data are also of great value when compared to the experiment-scaled conditions in a full NPP. The LOBI was a single plus a triple-loop (simulated by one loop) test facility electrically heated to simulate a 1300 MWe PWR. The scaling factor was 712 for the core power, volume, and mass flow. Primary and secondary sides contained all main active elements. Tests were performed for the characterization of phenomenologies relevant to large and small break LOCAs and special transients in PWRs. The paper presents the results of three posttest calculations of LOBI experiments. The selected experiments are BL-30, BL-44, and A1-84. They are LOCA scenarios of different break sizes and with different availability of safety injection components. The goal of the analysis is to improve the knowledge of the phenomena occurred in the facility in order to use it in further studies related to qualifying nodalizations of actual plants or to establish accuracy data bases for uncertainty methodologies. An example of procedure of implementing changes in a common nodalization valid for simulating tests occurred in a specific ITF is presented along with its confirmation based on posttests results.

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