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Title: Normal operation and maintenance safety lessons from the ITER US PbLi test blanket module program for a US FNSF and DEMO

A leading power reactor breeding blanket candidate for a fusion demonstration power plant (DEMO) being pursued by the US Fusion Community is the Dual Coolant Lead Lithium (DCLL) concept. The safety hazards associated with the DCLL concept as a reactor blanket have been examined in several US design studies. These studies identify the largest radiological hazards as those associated with the dust generation by plasma erosion of plasma blanket module first walls, oxidation of blanket structures at high temperature in air or steam, inventories of tritium bred in or permeating through the ferritic steel structures of the blanket module and blanket support systems, and the 210Po and 203Hg produced in the PbLi breeder/coolant. What these studies lack is the scrutiny associated with a licensing review of the DCLL concept. An insight into this process was gained during the US participation in the International Thermonuclear Experimental Reactor (ITER) Test Blanket Module (TBM) Program. In this paper we discuss the lessons learned during this activity and make safety proposals for the design of a Fusion Nuclear Science Facility (FNSF) or a DEMO that employs a lead lithium breeding blanket.
Authors:
; ; ; ;
Publication Date:
OSTI Identifier:
1162197
Report Number(s):
INL/JOU-13-30100
DOE Contract Number:
DE-AC07-05ID14517
Resource Type:
Journal Article
Resource Relation:
Journal Name: Fusion and Engineering Design; Journal Volume: 89; Journal Issue: 9-10
Research Org:
Idaho National Laboratory (INL)
Sponsoring Org:
DOE - SC
Country of Publication:
United States
Language:
English
Subject:
70 PLASMA PHYSICS AND FUSION TECHNOLOGY fusion, safety, DCLL, DEMO, PbLi, blanket.