Modelling and simulation of the source term for a sodium cooled fast reactor under hypothetical severe accident conditions; Final report of a coordinated research project

Arokiaswamy, J. A.*; Batra, C.*; Chang, J. E.*; Garcia, M.*; Herranz, L. E.*; Klimonov, I. A.*; Kriventsev, V.*; Li, S.*; Liegeard, C.*; Mahanes, J.*; et al.

IAEA-TECDOC-2006, 380 Pages, 2022/00

https://www.iaea.org/publications/15086/modelling-and-simulation-of-the-source-term-for-a-sodium-cooled-fast-reactor-under-hypothetical-severe-accident-conditions

The IAEA coordinated research project on "Radioactive Release from the Prototype Sodium Cooled Fast Reactor under Severe Accident Conditions" was devoted to realistic numerical simulation of fission products and fuel particles inventory inside the reference sodium cooled fast reactor volumes under severe accident conditions at different time scales. The scope of analysis was divided into three parts, defined as three work packages (WPs): (1) in-vessel source term estimation; (2) primary system/containment system interface source term estimation; and, (3) in-containment phenomenology analysis. Comparison of the results obtained in WP-1 indicates that the release fractions of noble gases and cesium radionuclides, and fractions of radionuclides released to the cover gas are in a good agreement. In the analysis using a common pressure history in WP-2, the results were in good agreement indicating that the accuracy of the analysis method of each institution is almost the same. The standalone case, which uses a set of pre-defined release fractions, was defined for WP-3 which enables to decouple this part of analysis from previous WPs. There is broad consensus among the predicted results by all the participants in WP-3.

Current status of the engineering design of the test modules for the IFMIF

Yamamoto, Michiyoshi; Arbeiter, F.*; Yokomine, Takehiko*; Wakai, Eiichi; Theile, J.*; Garcia, A. S.*; Rapisarda, D. S.*; Casal, N. I.*; Mas, A. S.*; Gouat, P.*; et al.

Fusion Engineering and Design, 88(6-8), p.746 - 750, 2013/10

https://doi.org/10.1016/j.fusengdes.2013.03.062

Under Broader Approach (BA) Agreement between EURATOM and Japan, IFMIF/EVEDA (International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities) has been performed since the middle of 2007. The present EVEDA phase aims at producing a detailed, compete and fully integrated engineering design of IFMIF. The deivery of the Intermediate IFMIF engineering design report is foreseen mid-2013. The main function of the IFMIF is to provide the materials database for the design and licensing of DEMO reactor and further fusion rectors from the material sets irradiated in High Flux Test Modules (HFTM, Startup Monitoring Module), Medium Flux Test Modules (Creep Fatigues Test Module, Tritium Release Test Module, Liquid Breeder Validation Module) and Low Flux Test Modules of IFMIF. This paper is summarizing the current progress of the engineering design of the test modules within the IFMIF/EVEDA project.

Integrated modeling of steady-state scenarios and heating and current drive mixes for ITER

Murakami, Masanori*; Park, J. M.*; Giruzzi, G.*; Garcia, J.*; Bonoli, P.*; Budny, R. V.*; Doyle, E. J.*; Fukuyama, Atsushi*; Hayashi, Nobuhiko; Honda, Mitsuru; et al.

Proceedings of 23rd IAEA Fusion Energy Conference (FEC 2010) (CD-ROM), 8 Pages, 2011/03

http://www-naweb.iaea.org/napc/physics/FEC/FEC2010/html/node76.htm#18822

Current ramps in tokamaks; From present experiments to ITER scenarios

Imbeaux, F.*; Basiuk, V.*; Budny, R.*; Casper, T.*; Citrin, J.*; Fereira, J.*; Fukuyama, Atsushi*; Garcia, J.*; Gribov, Y. V.*; Hayashi, Nobuhiko; et al.

Proceedings of 23rd IAEA Fusion Energy Conference (FEC 2010) (CD-ROM), 8 Pages, 2010/10

http://www-naweb.iaea.org/napc/physics/FEC/FEC2010/html/node62.htm#15641

In order to prepare adequate current ramp-up and ramp-down scenarios for ITER, present experiments from several tokamaks have been analyzed by means of integrated modeling in view of determining relevant heat transport models for these operation phases. The results of these studies are presented and projections to ITER current ramp-up and ramp-down scenarios are done, focusing on the baseline inductive scenario (main heating plateau current of 15 MA). Various transport models have been tested by means of integrated modeling against experimental data from ASDEX Upgrade, C-Mod, DIII-D, JET and Tore Supra, including both Ohmic plasmas and discharges with additional heating/current drive. With using the most successful models, projections to the ITER current ramp-up and ramp-down phases are carried out. Though significant differences between models appear on the electron temperature prediction, the final q-profiles reached in the simulation are rather close.

Simulation of the hybrid and steady state advanced operating modes in ITER

Kessel, C. E.*; Giruzzi, G.*; Sips, A. C. C.*; Budny, R. V.*; Artaud, J. F.*; Basiuk, V.*; Imbeaux, F.*; Joffrin, E.*; Schneider, M.*; Luce, T.*; et al.

Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 8 Pages, 2007/03

http://www-naweb.iaea.org/napc/physics/FEC/FEC2006/html/node78.htm#20378

no abstracts in English