search query: @keyword tungsten / total: 2
reference: 1 / 2
« previous | next »
Author: | Fernández Navarro, Alejandro |
Title: | Impact of fusion neutrons on helium production in beryllium and tungsten, and tritium breeding in ITER and DEMO |
Publication type: | Final Project work |
Publication year: | 2014 |
Pages: | viii + 75 Language: eng |
Department/School: | Perustieteiden korkeakoulu |
Main subject: | Ydin- ja energiatekniikka (Tfy-56) |
Supervisor: | Groth, Mathias |
Instructor: | Airila, Markus ; Rintala, Lauri |
Electronic version URL: | http://urn.fi/URN:NBN:fi:aalto-201411123014 |
OEVS: | Electronic archive copy is available via Aalto Thesis Database.
Instructions Reading digital theses in the closed network of the Aalto University Harald Herlin Learning CentreIn the closed network of Learning Centre you can read digital and digitized theses not available in the open network. The Learning Centre contact details and opening hours: https://learningcentre.aalto.fi/en/harald-herlin-learning-centre/ You can read theses on the Learning Centre customer computers, which are available on all floors.
Logging on to the customer computers
Opening a thesis
Reading the thesis
Printing the thesis
|
Location: | P1 Ark Aalto 2005 | Archive |
Keywords: | blanket technology tritium breeding beryllium tungsten lithium fusion |
Abstract (eng): | The project studies blanket designs of ITER and DEMO for neutron shielding, helium production and tritium breeding. On the one hand, a comparison has been made between beryllium and tungsten as first wall materials. On the other hand, tritium breeding blanket models have been studied, focus on the European test blanket module (TBM) concepts, the helium-cooled pebble bed (HCPB) and the helium-cooled lithium-lead (HCLL). The choice of plasma facing materials and the tritium breeding technology are key issues in the technological development of future fusion power plants. Whereas the ITER design includes beryllium as the first wall material of the blanket and tungsten in the divertor, DEMO will possibly use tungsten for both surfaces, due to beneficial characteristics of this material related to lower tritium retention and lower erosion rates. As future DEMO-type reactors are intended to be tritium self-sufficient, the reactors would dedicate most of the blanket to tritium breeding. Both analytical (multigroup diffusion theory) and Monte-Carlo methods were utilized to calculate the neutron fluxes and neutron induced reactions. The Serpent code is used to run Monte-Carlo simulations. The results for the Be-W comparison indicate that W is a better first wall material in terms of blanket shielding capability for high-energy neutrons and showing lower helium production in the first wall. However, the simulations for the HCPB and HCLL models show that the use of a Be first wall instead of W leads to a substantial increment of the tritium breeding ratio (TBR), allowing the use of lithium with lower enrichment. The assessment of the European tritium breeding blanket concepts indicated that HCPB models have a higher TBR and better shielding capability than HCLL models, being the HCPB with Be as first wall the most efficient breeding blanket. Finally, lithium depletion simulations for the HCPB and HCLL models showed that these blankets can be easily designed to work without recharging lithium during their estimated lifespan of 5 years. |
ED: | 2014-11-16 |
INSSI record number: 50040
+ add basket
« previous | next »
INSSI