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

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 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

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.

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.

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.

PACTEL integral test facility – Description of versatile applications

2006 ◽  
Vol 33 (11-12) ◽  
pp. 994-1009 ◽  
Author(s):  
Heikki Purhonen ◽  
Markku Puustinen ◽  
Vesa Riikonen ◽  
Riitta Kyrki-Rajamäki ◽  
Juhani Vihavainen
Keyword(s):  
Test Facility ◽  
Integral Test

scholarly journals Qualification of TRACE V5.0 Code against Fast Cooldown Transient in the PKL-III Integral Test Facility

2013 ◽  
Vol 2013 ◽  
pp. 1-11 ◽  
Author(s):  
Eugenio Coscarelli ◽  
Alessandro Del Nevo ◽  
Francesco D'Auria
Keyword(s):  
Heat Transfer ◽  
Test Facility ◽  
Steam Generators ◽  
Integral Test ◽  
Pressurized Thermal Shock ◽  
Vessel Integrity ◽  
Main Steam Line ◽  
Main Steam ◽  
Secondary Side ◽  
Reactor Pressure

The present paper deals with the analytical study of the PKL experiment G3.1 performed using the TRACE code (version 5.0 patch1). The test G3.1 simulates a fast cooldown transient, namely, a main steam line break. This leads to a strong asymmetry caused by an increase of the heat transfer from the primary to the secondary side that induces a fast cooldown transient on the primary side-affected loop. The asymmetric overcooling effect requires an assessment of the reactor pressure vessel integrity considering PTS (pressurized thermal shock) and an assessment of potential recriticality following entrainment of colder water into the core area. The aim of this work is the qualification of the heat transfer capabilities of the TRACE code from primary to secondary side in the intact and affected steam generators (SGs) during the rapid depressurization and the boiloff in the affected SG against experimental data.

Experimental Investigation on Gas Mixing and Stratification in Containment Influenced by External Cooling

2020 ◽  
Author(s):  
Ying Li ◽  
Wei Xue ◽  
Xuewu Cao ◽  
Lili Tong
Keyword(s):  
Large Scale ◽  
Initial Pressure ◽  
Stable Stratification ◽  
Test Facility ◽  
Gas Mixing ◽  
Additional Time ◽  
Integral Test ◽  
External Cooling ◽  
Comprehensive Test ◽  
Accident Conditions

Abstract The distribution of hydrogen inside the containment is a key issue in assessing the evolution of the postulated accident. For safety analysis and codes validation purposes, a large scale comprehensive test facility has been built to investigate the containment thermal-hydraulic characteristics under accident conditions. In this paper, a test was performed to experimentally investigate the distribution of the hydrogen inside the containment and the influence of the external cooling on gas mixing and stratification. The paper presents the experimental results of the integral test performed in this facility. During the experiments, helium was used to simulate hydrogen. Helium and steam are released together and allowed to take additional time to form a relatively stable stratification, then followed by external cooling. The initial pressure of the experiments is around 0.1MPa(a) and the initial Froude number is around 333. The results showed that a helium-enriched stratification emerged in the upper containment due to the density difference after the injection. External cooling caused condensation and intense convective flow. As a result, an overall increase in helium concentration was observed with a decrease in concentration gradient.

The RELAP5 Simulation of 2″ LOCA of APEX Facility

2010 ◽  
Author(s):  
Heng Xie
Keyword(s):  
Test Facility ◽  
State University ◽  
Pressurized Water Reactor ◽  
Reduced Pressure ◽  
Pressurized Water ◽  
Sequence Of Events ◽  
Comparison Results ◽  
Integral Systems ◽  
Oregon State University ◽  
Good Agreement

The RELAP5/SCDAP Mod3.2(am5) code is employed to simulate the OSU-AP1000-05 test conducted in the Advanced Plant Experimental (APEX) test facility at Oregon State University (OSU). The APEX-1000 test facility is an one-fourth height, one-half time scale, and reduced pressure integral systems facility to simulate the Westinghouse Advanced Passive 1000 MW (AP1000) pressurized water reactor. OSU-AP1000-05 is a two-inch break at the bottom of cold leg #4 with 3 out of 4 ADS-4 valves of OSU-APEX-1000 facility. RELAP5 predictions are compared to the experimental data generated by the test. The comparison shows good agreement between the predicted and measured sequence of events of some key parameters during the transient. From the comparison results, it could be preliminary concluded that the RELAP5/SCDAP Mod3.2(am5) code are suitable to simulate the small LOCA of APEX.

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