Home / Store / Journals / Electronic Articles / Nuclear Technology / Volume 177 / Number 3 / Pages 382-394
Kyoung-Ho Kang, Hyun-Sik Park, Seok Cho, Nam-Hyun Choi, In-Cheol Chu, Byong-Jo Yun, Kyung-Doo Kim, Yeon-Sik Kim, Won-Pil Baek, Ki-Yong Choi
Nuclear Technology / Volume 177 / Number 3 / Pages 382-394
Format:electronic copy (download)
A postulated steam generator tube rupture (SGTR) event of the APR1400 (Advanced Power Reactor 1400 MWe) was experimentally investigated with the thermal-hydraulic integral effect test facility ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation). The SGTR accident is one of the design-basis accidents having a significant impact on safety from the viewpoint of radiological release. To simulate a SGTR accident of the APR1400, the SGTR-HL-04 and the SGTR-HL-05 tests were performed by simulating double-ended ruptures of a single U-tube and five U-tubes at the hot side of the ATLAS steam generator. Following the reactor trip induced by a high steam generator level signal, the primary-system pressure decreased and the secondary-system pressure increased until the main steam safety valves were opened to reduce the secondary-system pressure. A mild change of the water level in the core was observed, which was attributed to the small break sizes of the present tests compared with conventional loss-of-coolant-accident tests. No excursion in the cladding temperature was observed in either test. The break area affected the timing of the major events observed in the tests. Lessened heat transfer to the secondary side caused by earlier actuation of the safety injection pumps had more influence on the secondary pressure of the affected steam generator than the break flow. The break flow was discharged as single-phase water in both tests.
Your cart is empty.
Home|Invoice Payment|Nuclear Links