scholarly journals Major Achievements and Prospect of the ATLAS Integral Effect Tests

2012 ◽  
Vol 2012 ◽  
pp. 1-18 ◽  
Author(s):  
Ki-Yong Choi ◽  
Yeon-Sik Kim ◽  
Chul-Hwa Song ◽  
Won-Pil Baek
Keyword(s):  
Large Scale ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Hydraulic Test ◽  
Validation Data ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Test Loop ◽  
Loss Of Coolant ◽  
Integral Effect

A large-scale thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been operated by KAERI. The reference plant of ATLAS is the APR1400 (Advanced Power Reactor, 1400 MWe). Since 2007, an extensive series of experimental works were successfully carried out, including large break loss of coolant accident tests, small break loss of coolant accident tests at various break locations, steam generator tube rupture tests, feed line break tests, and steam line break tests. These tests contributed toward an understanding of the unique thermal-hydraulic behavior, resolving the safety-related concerns and providing validation data for evaluation of the safety analysis codes and methodology for the advanced pressurized water reactor, APR1400. Major discoveries and lessons found in the past integral effect tests are summarized in this paper. As the demand for integral effect tests is on the rise due to the active national nuclear R&D program in Korea, the future prospects of the application of the ATLAS facility are also discussed.

scholarly journals Validation of Advanced Computer Codes for VVER Technology: LB-LOCA Transient in PSB-VVER Facility

2012 ◽  
Vol 2012 ◽  
pp. 1-15 ◽  
Author(s):  
A. Del Nevo ◽  
M. Adorni ◽  
F. D'Auria ◽  
O. I. Melikhov ◽  
I. V. Elkin ◽  
...  
Keyword(s):  
Natural Circulation ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Mass Inventory ◽  
Computer Codes ◽  
Set Up ◽  
Advanced Computer

The OECD/NEA PSB-VVER project provided unique and useful experimental data for code validation from PSB-VVER test facility. This facility represents the scaled-down layout of the Russian-designed pressurized water reactor, namely, VVER-1000. Five experiments were executed, dealing with loss of coolant scenarios (small, intermediate, and large break loss of coolant accidents), a primary-to-secondary leak, and a parametric study (natural circulation test) aimed at characterizing the VVER system at reduced mass inventory conditions. The comparative analysis, presented in the paper, regards the large break loss of coolant accident experiment. Four participants from three different institutions were involved in the benchmark and applied their own models and set up for four different thermal-hydraulic system codes. The benchmark demonstrated the performances of such codes in predicting phenomena relevant for safety on the basis of fixed criteria.

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.

Analysis of loss of coolant accident without ECCS and DHRS in an integral pressurized water reactor using RELAP/SCDAPSIM

2021 ◽  
Vol 134 ◽  
pp. 103648
Author(s):  
Katarzyna Skolik ◽  
Chris Allison ◽  
Judith Hohorst ◽  
Mateusz Malicki ◽  
Marina Perez-Ferragut ◽  
...  
Keyword(s):  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Effect of Nitrogen Release From Accumulators on PWR LOCA Analysis

2002 ◽  
Author(s):  
Woon-Shing Yeung ◽  
Ramu K. Sundaram
Keyword(s):  
Coming Out ◽  
Computer Code ◽  
Nitrogen Release ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Concentration Difference ◽  
Dissolved Nitrogen ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Release Model

The accumulator in a Pressurized Water Reactor (PWR) is generally pressurized with inert nitrogen cover gas, and the accumulator water will be saturated with nitrogen. Nitrogen released due to system depressurization during a Loss-of-Coolant Accident (LOCA) transient, consists of the nitrogen that is in the gas phase as well as nitrogen coming out of the liquid from a dissolved state. The effect of nitrogen release from the accumulator on the accident sequence is generally not explicitly addressed in typical LOCA analyses. This paper presents an analytical nitrogen release model and its incorporation into the RELAP5/MOD3 computer code. The model predicts the amount of nitrogen release as a function of concentration difference between the actual and equilibrium conditions, and can track its subsequent transport through the downstream reactor coolant system in a LOCA transient. The model is compared to data from discharge tests with a refrigerant type fluid, pressurized with nitrogen. The results demonstrate that the model is able to calculate the release of the dissolved nitrogen as designed. The modified computer code has been applied to analyze the discharge from a typical PWR accumulator. The results are compared to those obtained without the nitrogen release model. The effect of nitrogen release on major system parameters, including accumulator level, accumulator flow rate, and accumulator pressure, is discussed.

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