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.

Scaling Analysis and Facility Design for Stability Investigation During Accidents in a PWR-Type SMR

2016 ◽  
Author(s):  
Yikuan Yan ◽  
Shanbin Shi ◽  
Mamoru Ishii
Keyword(s):  
Sea Water ◽  
Flow Instability ◽  
Natural Circulation ◽  
Clean Energy ◽  
Water Desalination ◽  
Normal Operation ◽  
Test Facility ◽  
Small Scale ◽  
Scaling Analysis ◽  
Two Phase

Small modular reactor (SMR) concept has been developed as one of the key solutions for the growing demand of safe and clean energy. SMR designs can be applied extensively in areas such as sea water desalination and small-scale power generation etc. Unlike conventional light water reactors, most SMRs greatly simplify the structure of reactor pressure vessel, usually eliminate pumps and use natural circulation to cool down the core and transfer energy. However, flow instability may easily occur and affect the entire two phase natural circulation, which is of great importance for the start-up and normal operation process of BWR-type SMRs. For PWR-type SMRs, two-phase natural circulation could exist during accidents such as small break loss of coolant accident (SBLOCA) and loss of heat sink. Current research aims to experimentally investigate potential flow instabilities related to natural circulation for a PWR-type SMR during the accidents. For current research, the NuScale reactor design is selected as the research prototype. In this paper, the design and scaling analysis of a scaled PWR-type experimental facility are provided. In order to experimentally study the natural circulation behavior of PWR-type SMR during accidental scenarios, detailed scaling analyses are necessary to ensure that the scaled phenomena could be obtained in a laboratory test facility. A three-level scaling method is used to get the scaling ratios derived from various non-dimensional numbers. An ideally scaled facility is first accomplished based on derived scaling ratios. RELAP5 simulations of both steady state and transient cases for the ideally scaled facility are performed and compared to the prototype to ensure the accuracy of the scaling analysis. Then the ideally scaled facility is modified under engineering considerations and an engineering scaled facility is designed. Similar RELAP5 analyses are performed on the engineering scaled facility and the results match well with those in the prototype and ideally scaled facility.

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

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 Analysis of Passive Heat Exchanger Under Station Blackout Accident

2017 ◽  
Author(s):  
Liu Yu-sheng ◽  
Xu Chao ◽  
Zhuang Shao-xin ◽  
Li Cong-xin ◽  
Zhang Pan
Keyword(s):  
Heat Exchanger ◽  
Nuclear Power ◽  
Natural Circulation ◽  
Cooling Capacity ◽  
Similarity Criteria ◽  
Test Facility ◽  
Passive Safety ◽  
Scaling Analysis ◽  
Station Blackout ◽  
Source Number

Station blackout accidents increasingly become the focus of research in the field of nuclear safety after Japan’s Fukushima nuclear plant accident in March 2011. Core decay heat under station blackout condition will be transferred by natural circulation occurring between core and passive heat exchanger for the nuclear plants incorporated passive safety design concept such as AP1000 or CAP1400. As a result, response of safety systems will differ in accident sequence and kind between passive safety plant and traditional plant. What is more, cooling capacity of passive heat exchanger (PHX) which takes on heat sink has significant effect on performance of natural circulation in passive safety system. The safety need that characteristics of passive safety plant should be verified through integral experiment facility makes scaling analysis important in design or modification of experiment facility. Furthermore, scaling analysis of natural circulation phenomena under station blackout accident plays an important role in design verification, safety review verification or thermo-hydraulic program development. It not only determines the similar similarity criteria between the nuclear power plant prototype and test facility, but also provides technical basis for selecting different experiment schemes. As a part of scaling analysis on natural circulation phenomena for station blackout, the cooling capacity of PHX in test facility should be scaled properly and reasonably with conservatism. Therefore, scaling of passive residual heat removal (PRHR) heat exchanger under station blackout accident is investigated analytically in this paper. The analytical model for natural circulation in passive heat exchanger is established based on the performance characteristics of PRHR system in passive plant. By proper hypothesis and simplification, the governing equations for PHX are normalized using steady-state solutions, initial or boundary conditions. The similarity criteria that should be preserved between PHXs in test facility and prototype are finally obtained from non-dimensionalized equations. Furthermore, the distortion analysis for PHE design is also investigated based on the similarity criteria for selected scaling factors and parameters. The safety analysis based on models of nuclear power plant prototype and test facility is conducted on transient performance of designed PHX with PHX of prototype. The results show that: heat source number is the dominant similarity criteria for PHXs design under SBO condition. Requirements of Richardson number and friction number could be satisfied by resistance adjusting on test loop. The performance of PHX designed following heat source number requirement can better represent the transient response characteristics of prototype under SBO condition.

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.

scholarly journals CATHARE Assessment of Natural Circulation in the PKL Test Facility during Asymmetric Cooldown Transients

2012 ◽  
Vol 2012 ◽  
pp. 1-10 ◽  
Author(s):  
Anis Bousbia Salah ◽  
Jacques Vlassenbroeck
Keyword(s):  
Natural Circulation ◽  
Test Facility ◽  
Steam Generators ◽  
Calculation Results ◽  
The Impact

Results of the CATHARE code calculations related to asymmetric cooldown tests in the PKL facility are presented. The test under consideration is the G2.1 experiment performed within the OECD/NEA PKL-2 project. It consists of carrying out a cooldown under natural circulation conditions in presence of two (out of four) emptied Steam Generators (SGs) and isolated on their secondary sides. The main goal of the current study is to assess the impact of a chosen cooldown strategy upon the occurrence of a Natural Circulation Interruption (NCI) in the inactive (i.e., noncooling) loops. For this purpose, three G2.1 test runs were investigated. The calculation results emphasize, mainly, the effect of the cooldown strategy, and the conditions that could lead to the occurrence of the NCI phenomenon.

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