Hot Fuel Element Thermal-Hydraulic Modeling in the Jules Horowitz Reactor Nominal and LOFA Conditions
2015 (English)In: Proceedings of the 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16), 2015Conference paper (Refereed)
The newest European high performance material testing reactor, the Jules Horowitz Reactor, is under construction at CEA Cadarache research center in France. The reactor will support the existing and future nuclear reactor technologies and will start operation at the end of the decade.
The current CEA methodology for simulating the thermal-hydraulic behavior of the reactor gives reliable results. Today the CATHARE2 code simulates the full reactor with a simplified approach for the core and the boundary conditions are transferred into the three-dimensional FLICA4 core simulation. However this procedure needs to be further improved and simplified to shorten the computational time and to give more accurate core level data. Specific CFD calculations will better identify the thermal-hydraulics phenomena and optimize the meshing/model of the improved procedure.
This article presents the current one-coupled thermal-hydraulic modeling of the reactor utilizing the system code CATHARE2 and the core analysis code FLICA4 and describes the more realistic new hot fuel element modeling by using CFD code STAR-CCM+ including conjugate heat transfer. Finally, the results from the both modeling are compared in the hot channel in the nominal condition and in the case of LOFA.
This study has improved the thermal-hydraulic knowledge of the complex hot fuel element and the most prominent finds are presented. In addition, the possible improvements for the more realistic CATHARE2’s core model are proposed. In all simulations the safety criteria were satisfied, the reactor stayed in the single-phase regime and overall integrity of the fuel plate was ensured.
Place, publisher, year, edition, pages
Jules Horowitz Reactor, CATHARE2, FLICA4, STAR-CCM+, loss of flow accident
IdentifiersURN: urn:nbn:se:kth:diva-173599ScopusID: 2-s2.0-84962636900OAI: oai:DiVA.org:kth-173599DiVA: diva2:853805
16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16)
QC 20150915. QC 201602262015-09-152015-09-152016-02-26Bibliographically approved