Results 71 to 80 of about 490,328 (214)

Effect of divertor on tritium breeding in HELIAS

Fusion Engineering and Design
This study presents tritium breeding calculations for the five-period helical-axis advanced stellarator (HELIAS) using the Serpent2 code. To address the absence of a detailed divertor design, a simplified divertor model with a layered homogenized structure was integrated into the parametric geometry of the HELIAS.
Lyytinen, Tommi, Snicker, Antti, Bogaarts, Timo, Warmer, Felix   +3 more
openaire   +2 more sources

Decommissioning Cost Estimate for the Ghana Research Reactor‐1 Facility Using the CERREX‐D2 Software

Science and Technology of Nuclear Installations, Volume 2026, Issue 1, 2026.
The Ghana Research Reactor‐1 facility would certainly be decommissioned, either at the end of its life or by a decision to discontinue its operation. In planning for GHARR‐1’s eventual decommissioning, estimating the overall decommissioning cost is a major activity that needs to be undertaken to support budgeting and decision‐making for the project. To
Emmanuel K. Boafo, Bismark Kwabena Nyamekye, Alfred Gyau, Lilian Hilary, Han Zhang   +4 more
wiley   +1 more source

Tritium technologies for pbli based breeding blankets

, 2013
A lead-lithium based breeding blanket for DEMO and its corresponding Test Blanket Module (TBM) to be tested in ITER is being developed in EU. The same approach is pursued in other countries involved in the development of fusion nuclear technologies.
Ciampichetti, A.
core   +1 more source

Decommissioning of a Nuclear Facility With Sodium‐Cooled Fast Reactor as a Final Stage of Its Life Cycle: Overview of Current Projects

Science and Technology of Nuclear Installations, Volume 2026, Issue 1, 2026.
The nuclear power industry preferably aims to a closed nuclear fuel cycle and a transition to a two‐component structure based on the joint operation of thermal and fast neutron reactors. This approach makes it possible to significantly reduce the volume of radioactive waste and increase the efficiency of using uranium fuel. Sodium fast neutron reactors
Kuanyshbek Toleubekov, Nurzhan Mukhamedov, Artur Surayev, Maxat Bekmuldin, Galina Vityuk, Peter Ivanov   +5 more
wiley   +1 more source

MCINO: a multi-physics coupling and intelligent neutronic optimization code for tritium breeding blanket of fusion reactors

Nuclear Fusion
Tritium self-sufficiency is a critical prerequisite for future fusion reactors. The tritium breeding blanket, as the component responsible for in-vessel tritium generation, requires coordinated neutronic and engineering optimization in order to maximize ...
Shen Qu, Qixiang Cao, Miao Yin, Xinghua Wu, Fengchao Zhao, Long Zhang, Xiaoyu Wang   +6 more
doaj   +1 more source

Numerical Analysis of Bi-Directional Hydrogen Isotopes Transport in First Wall of ITER Helium Coolant Ceramic Breeding Testing Blanket Module

He huaxue yu fangshe huaxue
Helium-Cooled Ceramic Breeding Blanket Module(HCCB-TBM) will be installed on ITER machine to test its tritium breeding performance and thermal removal ability.
WANG Jun, ZHANG Long, WANG Xiao-yu, WU Xing-hua, LI Ru-yan   +4 more
doaj   +1 more source

Tritium production in breeding blankets

, 2013
This chapter describes the equipment referred to as breeding blankets that will be used in the next fusion reactors. The mechanisms occurring in such components are addressed, including simplified diagrams to illustrate their functioning.
Ciampichetti, A.
core   +1 more source

Hypothetical Operation Plan to Minimize Tritium Inventory of Fusion Fuek Cycle

, 2019
1
LEE, IN BEUM
core  

Low technology high tritium breeding blanket concept [PDF]

, 1987
The main function of this low technology blanket is to produce the necessary tritium for INTOR operation with minimum first wall coverage. The INTOR first wall, blanket, and shield are constrained by the dimensions of the reference design and the ...
Hassanein, A. M., Finn, P. A., Smith, D. L., Cha, Y. S., Billone, M. C., Johnson, C. E., Liu, Y., Clemmer, R., Gohar, Y., Baker, C. C.   +9 more
core  

Research on the optimization method of tritium release efficiency in a water cooled blanket for CFETR

AIP Advances
The Chinese Fusion Engineering Test Reactor (CFETR) is a tokamak device located between ITER and DEMO, with the primary goal of generating useable fusion energy through D–T reactions.
Jie Li, Xianbing Xiang, Xiaoyi Zeng, Haiyan Yang, Jie Zhang   +4 more
doaj   +1 more source