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http://dx.doi.org/10.1016/j.net.2017.03.004

ROSA/LSTF Test and RELAP5 Analyses on PWR Cold Leg Small-Break LOCA with Accident Management Measure and PKL Counterpart Test  

Takeda, Takeshi (Nuclear Safety Research Center, Japan Atomic Energy Agency)
Ohtsu, Iwao (Nuclear Safety Research Center, Japan Atomic Energy Agency)
Publication Information
Nuclear Engineering and Technology / v.49, no.5, 2017 , pp. 928-940 More about this Journal
Abstract
An experiment using the $Prim{\ddot{a}}rkreisl{\ddot{a}}ufe$ Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg smallbreak loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.
Keywords
Counterpart Test; Loss-of-Coolant Accident; Pressurized Water Reactor; $Prim{\ddot{a}}rkreisl{\ddot{a}}ufe$ Versuchsanlage; RELAP5 Code; Rig of Safety Assessment/Large-Scale Test; Facility;
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