DOI : https://doi.org/10.18517/ijaseit.10.5.10817

The Analysis of Loss of Forced Flow Event on the HTGR Type Experimental Power Reactor

Andi S. Ekariansyah (1), Surip Widodo (2), - Sudarmono (3), - Susyadi (4), M. Darwis Isnaeni (5), R. Andika P. Dwijayanto (6)
(1) Center for Nuclear Reactor Technology and Safety, National Nuclear Energy Agency (BATAN), Puspiptek Area, South Tangerang, 15318, Indonesia
(2) Center for Nuclear Reactor Technology and Safety, National Nuclear Energy Agency (BATAN), Puspiptek Area, South Tangerang, 15318, Indonesia
(3) Center for Nuclear Reactor Technology and Safety, National Nuclear Energy Agency (BATAN), Puspiptek Area, South Tangerang, 15318, Indonesia
(4) Center for Nuclear Reactor Technology and Safety, National Nuclear Energy Agency (BATAN), Puspiptek Area, South Tangerang, 15318, Indonesia
(5) Center for Nuclear Reactor Technology and Safety, National Nuclear Energy Agency (BATAN), Puspiptek Area, South Tangerang, 15318, Indonesia
(6) Center for Nuclear Reactor Technology and Safety, National Nuclear Energy Agency (BATAN), Puspiptek Area, South Tangerang, 15318, Indonesia
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How to cite (IJASEIT) :
Ekariansyah, Andi S., et al. “The Analysis of Loss of Forced Flow Event on the HTGR Type Experimental Power Reactor”. International Journal on Advanced Science, Engineering and Information Technology, vol. 10, no. 5, Oct. 2020, pp. 1986-91, doi:10.18517/ijaseit.10.5.10817.
Citation Format :
Since 2014, Indonesia's National Atomic Energy Agency (BATAN) has been launching a plan to construct a 10 MWt Experimental Power Reactor (Reaktor Daya Eksperimental / RDE). The RDE design is based on the small-sized pebble-bed high-temperature gas-cooled reactor (HTGR) technology with TRISO fuels. By concept, HTR-10 design, which was developed by the INET of China, is used as the reference design. During the development process, a safety analysis report (SAR) of RDE design has to be prepared to be evaluated by the Indonesia Nuclear Regulatory Agency (BAPETEN). The report contains, among others the description of the RDE accident sequences, which can be only provided by simulations using a certain code. This paper emphasizes the transient analysis, which is simulated using RELAP5/SCDAP/Mod3.4 , which is a thermal-hydraulic code specified for light water coolant systems. The simulated event is the loss of primary coolant mass flow, which is caused by the failure of the primary gas blower motor. The methodology of simulation is first by modelling the RDE nuclear steam supply system to verify steady-state operational parameter of the RDE design. The second step is to simulate the event of loss of flow, which is followed by the failure to shut down the reactor. The simulation results in the decrease of the fuel pebble temperature during the event due to the negative fuel temperature reactivity coefficient and the core heat removal by the cavity cooling. Overall, the RELAP5 code has a limitation in the RDE simulation to define two different non-condensable gases, which reduces the accuracy of the simulation results.

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