Contents lists available at ScienceDirect Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes Radiation shielding calculations for a 3D ITER benchmark model using TRIPOLI-4 ® Monte Carlo code Yi-Kang Lee DEN, Service d’étude des réacteurs et de mathématiques appliquées (SERMA), CEA, Université Paris-Saclay, F-91191 Gif-sur-Yvette, France ARTICLE INFO Keywords: Iter Neutronics TRIPOLI-4 ® Monte Carlo code Variance reduction Display tool ABSTRACT The International Thermonuclear Experimental Reactor (ITER) is currently under construction in France. From the radiation shielding and radiation protection points of view, an intense and large neutron source with energy around 14.1 MeV will be generated from the D-T plasma zone during the ITER operation and diverse gamma-ray sources from neutron activation of the reactor structure materials and coolant should be considered for the reactor operation and maintenance. To decrease the radiation impacts caused by these neutron and gamma sources, iterative designs and nuclear analyses of ITER components are currently performed with three di- mensional Monte Carlo radiation transport calculations. Due to the important dimensions of ITER, the thick tokamak blanket modules, and the diagnostic and functional port openings, variance reduction techniques are essential in these Monte Carlo neutron transport calculations. To verify the reactor components design models and to check the radiation transport calculation results, advanced graphic features of the calculation tool are also necessary. With the growing interest in using the TRIPOLI-4 ® Monte Carlo radiation transport code for ITER applications, the aim of this paper is to study the feasibility to use variance reduction features of TRIPOLI-4 code on a 3D ITER benchmark model which is a 40° toroidal segment including 5796 vol cells. The calculation results reported in this paper include the axial and radial profiles of the inboard TF coil heating and the neutron flux attenuation through the equatorial port plugs and shield. The performance of the TRIPOLI-4 graphic tool under its parallel computing mode was also evaluated. 1. Introduction The International Thermonuclear Experimental Reactor (ITER) is currently under construction in southern France. The ITER tokamak complex is the world's largest fusion device, with a plasma major radius (R) of 6.2 m and a plasma volume of 840 m 3 . The fusion energy output power of this reactor is 500 MW. That means, under the deuterium- tritium (D–T) fusion operation, an intense and large neutron source with energy around 14.1 MeV will be generated from the plasma zone during the reactor operation. For the after-operation maintenance, di- verse gamma-ray sources from neutron activation of the ITER structure materials must be considered in order to follow the radiation protection regulations. That is why the ITER nuclear analyses are fundamental technical challenges associated with the design, fabrication, assembly, and integration of the ITER Tokamak components. These tasks have to be compromised in order to follow the cost and schedule targets for the project [1,2]. The priority issues on current ITER nuclear analyses include the toroidal field coils (TFC) heating [3,4], the radiation streaming from neutral beam ports, the shut-down dose rates in different locations including the cryostats, and the impact of cooling water activation on radiation safety etc. To estimate these physical quantities, determina- tion of neutron and gamma flux in different ITER tokamak components is essential. Using traditional radiation transport codes developed under the nuclear fission backgrounds, diverse efforts are required for both de- terministic and Monte Carlo codes to correctly calculate the neutron and gamma flux in ITER. From the published studies, it was confirmed that, due to the highly anisotropic scattering of fusion neutrons, the unoptimized multi-group nuclear data including their angular quad- rature can be crucial for deterministic calculation results [5]. Conse- quently, the continuous-energy Monte Carlo radiation transport calcu- lations were mostly performed on current ITER nuclear analyses. The deterministic transport calculations were often used to generate var- iance reduction parameters for the Monte Carlo transport calculations [4]. From the plasma zone, ITER fusion neutrons should be transported through different radiation shielding structures including first wall (FW), blanket modules, vacuum vessel (VV) and TF magnets for both the inboard (I/B) and outboard (O/B) directions (see Fig. 1). On the I/B https://doi.org/10.1016/j.fusengdes.2018.03.036 Received 1 August 2017; Received in revised form 14 March 2018; Accepted 15 March 2018 E-mail address: yi-kang.lee@cea.fr. Fusion Engineering and Design xxx (xxxx) xxx–xxx 0920-3796/ © 2018 Elsevier B.V. All rights reserved. Please cite this article as: Lee, Y., Fusion Engineering and Design (2018), https://doi.org/10.1016/j.fusengdes.2018.03.036