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Title: Liquid Scoping Study for Tritium-Lean, Fast Ignition Inertial Fusion Energy Power Plants

Abstract

In a thick-liquid protected chamber design, such as HYLIFE-II, a molten-salt is used to attenuate neutrons and protect the chamber structures from radiation damage. The molten-salt absorbs some of the material and energy given off by the target explosion. In the case of a fast ignition inertial fusion system, advanced targets have been proposed that may be Self-sufficient in the tritium breeding (i.e., the amount of tritium bred in target exceeds the amount burned). These ''tritium-lean'' targets contain approximately 0.5% tritium and 99.5% deuterium, but require a large pr of 10-20 g/cm{sup 2}. Although most of the yield is provided by D-T reactions, the majority of fusion reactions are D-D, which produces a net surplus of tritium. This aspect allows for greater freedom when selecting a liquid for the protective blanket (lithium-bearing compounds are not required). This study assesses characteristics of many single, binary, and ternary molten-salts. Using the NIST Properties of Molten Salts Database, approximately 4300 molten-salts were included in the study [1]. As an initial screening, salts were evaluated for their safety and environmental (S&E) characteristics, which included an assessment of waste disposal rating, contact dose, and radioactive afterheat. Salts that passed the S&E criteria were then evaluatedmore » for neutron shielding ability and pumping power. The pumping power was calculated using three components: velocity head losses, frictional losses, and lift. This assessment left us with 57 molten-salts to recommend for further analysis. Many of these molten-salts contain elements such as sodium, lithium, beryllium, boron, fluorine, and oxygen. Recommendations for further analysis are also made.« less

Authors:
; ; ; ;
Publication Date:
Research Org.:
Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
Sponsoring Org.:
USDOE
OSTI Identifier:
15013411
Report Number(s):
UCRL-ID-144899
TRN: US0600862
DOE Contract Number:  
W-7405-ENG-48
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
70 PLASMA PHYSICS AND FUSION TECHNOLOGY; 72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS; BERYLLIUM; BORON; DEUTERIUM; FLUORINE; IGNITION; LITHIUM; MOLTEN SALTS; NEUTRONS; OXYGEN; POWER PLANTS; SODIUM; THERMONUCLEAR REACTORS; TRITIUM; WASTE DISPOSAL

Citation Formats

Schmitt, R C, Latkowski, J F, Durbin, S G, Meier, W R, and Reyes, S. Liquid Scoping Study for Tritium-Lean, Fast Ignition Inertial Fusion Energy Power Plants. United States: N. p., 2001. Web. doi:10.2172/15013411.
Schmitt, R C, Latkowski, J F, Durbin, S G, Meier, W R, & Reyes, S. Liquid Scoping Study for Tritium-Lean, Fast Ignition Inertial Fusion Energy Power Plants. United States. doi:10.2172/15013411.
Schmitt, R C, Latkowski, J F, Durbin, S G, Meier, W R, and Reyes, S. Tue . "Liquid Scoping Study for Tritium-Lean, Fast Ignition Inertial Fusion Energy Power Plants". United States. doi:10.2172/15013411. https://www.osti.gov/servlets/purl/15013411.
@article{osti_15013411,
title = {Liquid Scoping Study for Tritium-Lean, Fast Ignition Inertial Fusion Energy Power Plants},
author = {Schmitt, R C and Latkowski, J F and Durbin, S G and Meier, W R and Reyes, S},
abstractNote = {In a thick-liquid protected chamber design, such as HYLIFE-II, a molten-salt is used to attenuate neutrons and protect the chamber structures from radiation damage. The molten-salt absorbs some of the material and energy given off by the target explosion. In the case of a fast ignition inertial fusion system, advanced targets have been proposed that may be Self-sufficient in the tritium breeding (i.e., the amount of tritium bred in target exceeds the amount burned). These ''tritium-lean'' targets contain approximately 0.5% tritium and 99.5% deuterium, but require a large pr of 10-20 g/cm{sup 2}. Although most of the yield is provided by D-T reactions, the majority of fusion reactions are D-D, which produces a net surplus of tritium. This aspect allows for greater freedom when selecting a liquid for the protective blanket (lithium-bearing compounds are not required). This study assesses characteristics of many single, binary, and ternary molten-salts. Using the NIST Properties of Molten Salts Database, approximately 4300 molten-salts were included in the study [1]. As an initial screening, salts were evaluated for their safety and environmental (S&E) characteristics, which included an assessment of waste disposal rating, contact dose, and radioactive afterheat. Salts that passed the S&E criteria were then evaluated for neutron shielding ability and pumping power. The pumping power was calculated using three components: velocity head losses, frictional losses, and lift. This assessment left us with 57 molten-salts to recommend for further analysis. Many of these molten-salts contain elements such as sodium, lithium, beryllium, boron, fluorine, and oxygen. Recommendations for further analysis are also made.},
doi = {10.2172/15013411},
journal = {},
number = ,
volume = ,
place = {United States},
year = {2001},
month = {8}
}

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