Abstract
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.
Highlights
ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation) is a large-scale integral effect test facility with a reference plant of APR1400 (Advanced Power Reactor, 1400 MWe), which is under construction in Korea [1]
In 2007, ATLAS was used for a wide range of integral effect tests on the reflood phase of a large break loss of coolant accident (LOCA) to resolve the safety issues of the APR1400 raised by a regulatory body [3]
The three-level scaling methodology which consists of integral scaling, boundary flow scaling, and local phenomena scaling proposed by Ishii et al [7] was applied to the design of ATLAS
Summary
ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation) is a large-scale integral effect test facility with a reference plant of APR1400 (Advanced Power Reactor, 1400 MWe), which is under construction in Korea [1]. Afterwards, at the beginning of 2008, ATLAS was modified to have a configuration for simulating the direct vessel injection (DVI) line break accidents of the APR1400. Sensitivity tests for different DVI line break sizes were performed and an integral effect database was established for various break sizes of 5%, 25%, 50%, and 100% [4]. In order to provide an integral effect database for SBLOCA of the APR1400, sensitivity tests for different break sizes of the cold leg have been conducted. When the major integral effect tests on LOCA series accidents were completed, typical non-LOCA events were conducted from the second half of the year 2010, including the feed line break (FLB) and steam line break (SLB). Tests were performed either in conservative or in best-estimated initial and boundary conditions to support the nuclear industry and validate the safety analysis codes, respectively
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