U.S. patent application number 14/055771 was filed with the patent office on 2014-05-22 for system for extraction of tritium from liquid metal coolants.
This patent application is currently assigned to Lawrence Livermore National Security, LLC. The applicant listed for this patent is Lawrence Livermore National Security, LLC. Invention is credited to Daniel L. Flowers, Kevin J. Kramer, Jeffery Latkowski, Joel Martinez-Frias, Susana Reyes.
Application Number | 20140138257 14/055771 |
Document ID | / |
Family ID | 50726896 |
Filed Date | 2014-05-22 |
United States Patent
Application |
20140138257 |
Kind Code |
A1 |
Kramer; Kevin J. ; et
al. |
May 22, 2014 |
SYSTEM FOR EXTRACTION OF TRITIUM FROM LIQUID METAL COOLANTS
Abstract
A method for removing tritium from liquid lithium includes
mixing the liquid lithium containing trace amounts of tritium with
a molten salt and forming a salt of lithium and tritium. The method
also includes separating the liquid lithium from the salt of
lithium and tritium and circulating the molten salt in an
electrolyzer to form molecular tritium. The method further includes
bubbling an inert gas through the electrolyzer to remove the
molecular tritium and circulating the argon and removed molecular
tritium in a titanium getter to recover the tritium.
Inventors: |
Kramer; Kevin J.;
(Livermore, CA) ; Flowers; Daniel L.; (San
Leandro, CA) ; Latkowski; Jeffery; (Livermore,
CA) ; Martinez-Frias; Joel; (Redwood City, CA)
; Reyes; Susana; (San Francisco, CA) |
|
Applicant: |
Name |
City |
State |
Country |
Type |
Lawrence Livermore National Security, LLC |
Livermore |
CA |
US |
|
|
Assignee: |
Lawrence Livermore National
Security, LLC
Livermore
CA
|
Family ID: |
50726896 |
Appl. No.: |
14/055771 |
Filed: |
October 16, 2013 |
Related U.S. Patent Documents
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Application
Number |
Filing Date |
Patent Number |
|
|
61722569 |
Nov 5, 2012 |
|
|
|
Current U.S.
Class: |
205/705 ;
204/243.1 |
Current CPC
Class: |
C25B 1/02 20130101; C25B
15/08 20130101; C25B 9/00 20130101 |
Class at
Publication: |
205/705 ;
204/243.1 |
International
Class: |
C25C 3/34 20060101
C25C003/34 |
Goverment Interests
STATEMENT AS TO RIGHTS TO INVENTIONS MADE UNDER FEDERALLY SPONSORED
RESEARCH OR DEVELOPMENT
[0002] The United States Government has rights in this invention
pursuant to Contract No. DE-AC52-07NA27344 between the United
States Department of Energy and Lawrence Livermore National
Security, LLC for the operation of Lawrence Livermore National
Laboratory.
Claims
1. A method for removing tritium from liquid lithium, the method
comprising: mixing the liquid lithium containing trace amounts of
tritium with a molten salt; forming a salt of lithium and tritium;
separating the liquid lithium from the salt of lithium and tritium;
circulating the molten salt in an electrolyzer to form molecular
tritium; bubbling an inert gas through the electrolyzer to remove
the molecular tritium; and circulating the argon and removed
molecular tritium in a titanium getter to recover the tritium.
2. The method of claim 1 wherein the molten salt includes lithium
halides.
3. The method of claim 2 wherein the halides comprise at least one
of fluorine, chlorine, or bromine.
4. The method of claim 1 wherein the liquid lithium contains trace
amounts of tritium.
5. The method of claim 1 wherein mixing the liquid lithium with the
molten salt and separating the liquid lithium from the molten salt
are performed in a centrifugal contactor.
6. The method of claim 1 wherein the inert gas comprises at least
one of helium or argon.
7. A system for recovering tritium from a lithium coolant, the
system comprising: a lithium coolant circuit, wherein the lithium
coolant in the circuit includes trace amounts of tritium; a
plurality of contactors coupled to the lithium coolant circuit,
wherein the plurality of contactors are operable to mix the lithium
coolant with a molten salt and form a salt of lithium and tritium;
an electrolysis unit coupled to one or more of the plurality of
contactors; an inert gas source operable to bubble an inert gas
through the electrolysis unit; and a getter coupled to the
electrolysis unit and operable to recover the tritium.
8. The system of claim 7 wherein the molten salt comprises at least
one of LiF, LiCL, or LiBr.
9. The system of claim 7 wherein the plurality of contactors
comprise centrifugal contactors arranged in parallel.
10. The system of claim 7 wherein the getter comprises at least one
of a titanium or depleted uranium.
11. The system of claim 7 wherein the inert gas comprises at least
one of helium or argon.
Description
CROSS-REFERENCES TO RELATED APPLICATIONS
[0001] This application claims priority to U.S. Provisional Patent
Application No. 61/722,569, filed Nov. 5, 2012, and entitled:
"System for Extraction of Tritium from Liquid Metal Coolants," the
disclosure of which is hereby incorporated by reference in its
entirety for all purposes.
BACKGROUND OF THE INVENTION
[0003] The National Ignition Facility (NIF), the world's largest
and most energetic laser system, is operational at Lawrence
Livermore National Laboratory (LLNL) in Livermore, Calif. One goal
of operation of the NIF is to demonstrate fusion ignition for the
first time in the laboratory. Initial experiments are calculated to
produce yields of the order of 20 MJ from an ignited,
self-propagating fusion burn wave. The capability of the facility
is such that yields of up to 150-200 MJ could ultimately be
obtained. The NIF is designed as a research instrument, one in
which single "shots" on deuterium-tritium containing targets are
performed for research. A description of the NIF can be found in
Moses et al, Fusion Science and Technology, volume 60, pp 11-16
(2011) and references therein.
[0004] There is a rapidly growing need for power, and especially
for clean power. At LLNL a project known as Laser
Inertial-confinement Fusion Energy, (often referred to herein as
"LIFE") is working toward introduction of fusion based electric
power plants into the U.S. economy before 2030, and in a
pre-commercial plant format before that. LIFE technology offers a
pathway for the expansion of carbon-free power around the world. It
will provide clean carbon-free energy in a safe and sustainable
manner without risk of nuclear proliferation.
[0005] One challenge with respect to LIFE, as well as any
technology for generating electrical power to be distributed to
large numbers of consumers, is reliability. Consumers expect to
have extraordinarily high reliability in their electric power
supply. The result is that utilities that provide that electrical
power maintain their facilities to assure the required high
reliability. Thus, among the challenges with respect to fusion
power, is to provide mechanisms by which a reliable long-lived
fusion chamber can be provided in which the fusion reactions occur,
yet which can be maintained or replaced when necessary with minimal
downtime for the overall power plant.
[0006] Despite the progress made in the design of fusion based
electric power plants, there is a need in the art for improved
methods of extracting tritium from liquid coolants.
SUMMARY OF THE INVENTION
[0007] The present invention relates generally to methods and
systems for extracting tritium from lithium-based coolants. More
particularly, embodiments of the present invention provide methods
and systems for extracting tritium from lithium liquids that
includes trace amounts of lithium. Although embodiments of the
present invention are discussed in terms of lithium-based coolants,
other liquid metal coolants are included within the scope of the
present invention.
[0008] According to an embodiment of the present invention, a
method for removing tritium from liquid lithium is provided. The
method includes mixing the liquid lithium containing trace amounts
of tritium with a molten salt and forming a salt of lithium and
tritium. The method also includes separating the liquid lithium
from the salt of lithium and tritium and circulating the molten
salt in an electrolyzer to form molecular tritium. The method
further includes bubbling an inert gas through the electrolyzer to
remove the molecular tritium and circulating the argon and removed
molecular tritium in a titanium getter to recover the tritium.
[0009] According to another embodiment of the present invention, a
system for recovering tritium from a lithium coolant is provided.
The system includes a lithium coolant circuit. The lithium coolant
in the circuit includes trace amounts of tritium. The system also
includes a plurality of contactors coupled to the lithium coolant
circuit. The plurality of contactors are operable to mix the
lithium coolant with a molten salt and form a salt of lithium and
tritium. The system further includes an electrolysis unit coupled
to one or more of the plurality of contactors, an inert gas source
operable to bubble an inert gas through the electrolysis unit, and
a getter coupled to the electrolysis unit and operable to recover
the tritium.
[0010] Numerous benefits are achieved by way of the present
invention over conventional techniques. For example, embodiments of
the present invention provide methods and systems for extracting
tritium from lithium coolant fluids that are more efficient than
conventional techniques. These and other embodiments of the
invention along with many of its advantages and features are
described in more detail in conjunction with the text below and
attached figures.
BRIEF DESCRIPTION OF THE DRAWINGS
[0011] FIG. 1 is a simplified schematic diagram illustrating a 1.2
GW (thermal) power plant based on the LIFE engine.
[0012] FIG. 2 is a simplified schematic diagram of centrifugal
contactors connected in series.
[0013] FIG. 3 is a simplified plot of the liquid lithium flow
processed to recover tritium as a function of extraction efficiency
of centrifugal contactors.
[0014] FIG. 4 is a simplified plot showing the liquid lithium
fraction that is processed to maintain 0.1 wppm steady state
tritium concentration.
[0015] FIG. 5 is a plot showing the number of contactors as a
function of the centrifugal contactor extraction efficiency.
[0016] FIG. 6 is a simplified schematic diagram of a tritium
recovery system by molten salt extraction according to an
embodiment of the present invention.
[0017] FIG. 7 is a simplified schematic diagram of a tritium
recovery system utilizing redundancy.
[0018] FIG. 8 is a simplified flowchart illustrating a method of
extracting tritium according to an embodiment of the present
invention.
DETAILED DESCRIPTION OF SPECIFIC EMBODIMENTS
[0019] Liquid lithium is considered as the primary coolant and
breeder material in the design of the Laser Inertial Fusion Energy
(LIFE) engine. This work presents the analysis of a tritium
recovery system from the blanket liquid lithium for the LIFE engine
and is based on a molten salt extraction technology. The goal is to
size the primary components of the system. The dimensions of the
centrifugal contactors, electrolyzer, and getter are estimated as a
function of the desired tritium steady state inventory and the
component efficiencies. Results show that a relatively compact
system can recover tritium at the needed rate to maintain a low
tritium inventory.
[0020] The LIFE Engine is a laser-based energy system. Liquid
lithium is being considered as the blanket cooling and breeding
material for the LIFE engine due to its many virtues: low density,
high thermal capacity, good thermal conductivity, high potential to
breed tritium due to its high neutron capture cross-section, and
high affinity for hydrogen isotopes that minimizes tritium
permeability. High hydrogen affinity, however, complicates tritium
recovery while operational safety demands a low tritium inventory.
Contact with water is avoided at all times.
[0021] The tritium recovery system is designed to keep the tritium
steady state inventory of the LIFE engine power plant at less than
about 50 g. The amount of liquid lithium in LIFE pilot plant is
.about.1000 tons. If we have .about.1000 m.sup.3 of liquid lithium
in our 1.2 GW(th) LIFE plant we need a steady sate tritium
concentration of 0.1 wppm. If our tritium recovery system has 90%
efficiency then we need a system that can process .about.1% of the
lithium flow to decrease tritium concentration from 0.1 wppm to
0.01 wppm.
[0022] FIG. 1 is a simplified schematic diagram illustrating a 1.2
GW (thermal) power plant based on the LIFE engine. The design of
the pilot LIFE engine power plant includes four loops, each with an
intermediate heat exchanger, steam generator, and reheater. Thus
one tritium recovery system processes .about.7 kg/s of lithium flow
to reduce tritium concentration from 0.1 wppm to 0.01 wppm.
[0023] FIG. 1 illustrates two of the four loops. First, liquid
lithium transfers energy from the fusion chamber 110 to the
intermediate heat exchanger 112 where the thermal energy of the
liquid lithium is transferred to the molten salt. The molten salt
in the secondary loop (also referred to as an intermediate loop)
carries thermal energy from the intermediate heat exchanger 112 to
the steam generator and reheater 114. The steam then goes to the
Rankin cycle turbines. Each of the four loops will also have a
tritium recovery system by molten salt extraction (labeled "Lithium
Processing" in FIG. 1).
[0024] Several options have been proposed for tritium recovery from
liquid lithium. These include use of a permeable window, a
gettering process, a cold trap, distillation, molten salt, and
gettering and molten salt combined. Because the molten salt method
is based on liquid diffusion at relatively low temperature
(.about.500.degree. C.) it results in a relatively compact system
with relatively low energy consumption. For that reason, and others
discussed below, the molten salt extraction is utilized in the
embodiments described herein.
[0025] FIG. 8 is a simplified flowchart illustrating a method of
extracting tritium according to an embodiment of the present
invention. As illustrated in FIG. 8, the molten salt extraction
method includes four steps in this embodiment. The method includes
mixing a lithium fluid containing trace amounts of tritium with a
molten salt (e.g., lithium halides, for example, including LiF,
LiCl, LiBr, or combinations thereof) in a centrifugal contactor
(810). In other embodiments, one or more sodium salts can be
utilized as the molten salt. The intimate contact between the
molten salt and the trace amounts of tritium in the lithium fluid
preferentially extracts the LiT into the salt phase (i.e., a salt
of lithium and tritium). The lithium fluid and the salt of lithium
and tritium are separated in the centrifugal contactor (812). The
salt of lithium and tritium is circulated in an electrolyzer and
the LiT is oxidized to form T2 (814), which is swept from the salt
of lithium and tritium by bubbling an inert gas (e.g., argon)
through the salt (816). Finally, the inert gas is circulated in a
titanium getter which recovers the tritium from the inert gas
(818).
[0026] The most important characteristic of the molten salt
extraction method is being free from solid-state diffusion. Liquid
phase diffusion is often faster than solid-state diffusion.
However, distillation requires very high temperatures. Maroni et
al. were the first to propose the use of molten salt extraction for
tritium recovery from liquid lithium. In previous experiments
molten salt had been used to remove impurities from liquid metals.
They observed that some experiments with lithium/hydrogen galvanic
cells showed that the lithium hydride formed during discharge was
preferentially extracted into the electrolyte.
[0027] As described above, the liquid lithium (also referred to as
lithium liquid) that contains tritium is mixed with lithium halide
salts in a centrifugal contactor/separator. It has been shown that
the volumetric distribution coefficient, Dv (defined as the ratio
of tritium content per unit volume in the salt to tritium content
per unit volume in the lithium) was between 2 and 4, verifying that
the LiT moves preferentially from the liquid lithium to the salt
when they are mixed. Then, the salt of lithium and tritium is
circulated to an electrolyzer where the LiT is oxidized to form T2,
which is swept from the salt phase by bubbling an inert (i.e.,
noble) gas. The noble gas circulates through a getter that recovers
the tritium. It has been estimated that a 20% electrolyzer recovery
efficiency is used to capture tritium from the molten salt by
measuring the hydrogen isotopes added to the molten salt and the
hydrogen concentration in the noble gas coming out of the
electrolyzer.
[0028] Embodiments of the present invention provide a molten salt
extraction method that can be used in a full-scale fusion system.
Important issues that are addressed are: 1) The need to reduce the
halide impurities in the primary lithium blanket to safe levels
from an activation perspective. This depends on how well the
centrifugal contactors separate the lithium and molten salt after
LiT is transferred. 2) The need to minimize impurities in the salt
that can decrease electrolyzer efficiency. Once again, this depends
on how well the contactor-separator works.
[0029] The following provides an analysis of a complete molten salt
extraction system to recover tritium from the liquid lithium of the
LIFE engine power plant. We estimate the size of the major
components and calculate system energy consumption.
[0030] Centrifugal Contactors
[0031] LiF--LiCl--LiBr (22-31-47 mol %) is the primary choice for
molten salt in an embodiment although other mixtures can be
utilized. The selection of this primary choice was made primarily
on melting point consideration: i.e., <450.degree. C. is
required. LiF--LiCl--LiBr (22-31-47 mol %) also showed good
stability during testing, demonstrating that the particular ratio
discussed as the primary choice is not required by the present
invention. We consider that due to similar properties of the molten
halides the analysis presented here will be valid if another halide
is utilized. Most research has been done with 1:1 lithium to salt
volume ratio and no issues have been reported, thus the 1:1 ratio
is recommended in the interest of reducing overall salt inventory
in the plant.
[0032] FIG. 2 is a simplified schematic diagram of centrifugal
contactors connected in series. Lithium and salt flow
counter-current, mixing and separating in every centrifugal
contactor. Lithium and molten salt flows are in
counter-current.
[0033] The centrifugal contactor 1 210 receives salt with no
tritium and lithium with low tritium concentration (most tritium in
liquid lithium is found as LiT) because tritium is previously
extracted in centrifugal contactor 2 220 and contactor 3 230. The
two flows mix and separate with the lithium leaving with less
tritium and the salt leaving contactor 1 210 with a small amount of
tritium and entering contactor 2 220 where it mixes with lithium
with higher tritium concentration, and so on. Increasing liquid
lithium purity utilizes more contactors in series. It is also
possible to connect the centrifugal contactors in parallel. In that
case, the fraction of lithium flow that needs to be processed
increases because the concentration of tritium will be higher after
the single contactor stage.
[0034] There is a trade-off between the contacting step
(centrifugal contactors) and the salt processing step
(electrolyzer). If the contacting step is larger with high energy
consumption then the electrolyzer can be smaller and use less
energy because less salt with higher tritium concentration will
flow from the centrifugal contactors to the electrolyzer. Recovery
efficiency determines the number of contactors, and electrolyzer
volume is determined by the average current density. The more
contactors in series, the less salt volume flows into the
electrolyzer, enabling compact electrolyzer design. Some
experiments demonstrated 100% hydrogen recuperation efficiency from
molten salt electrolysis. More experimental work is needed, but
with the available information it seems better to maximize
electrolysis because the centrifugal contactors are more complex
and demand more energy. Thus, in some embodiments, parallel
centrifugal contactors are utilized, for example, for the pilot
LIFE engine plant.
[0035] Electrolyzer
[0036] The electrolysis unit is the least studied component of the
system. Recovery efficiency, defined as the weight of the hydrogen
recovered from the salt divided by the weight of hydrogen added to
the salt, is believed to be the most important electrolyzer
parameter. Electrolysis experiments have been conducted to separate
hydrogen from molten salt, reporting recovery efficiency ranging
from 20% to 100%. The higher efficiency is explained by the use of
an argon bubbler that was also the anode of the cell that removed
the hydrogen before back mixing occurred. The kinetics of tritium
recovery in the contactor units has been studied but did not
include a study of electrolyzer kinetics. Typically, an assumption
of 90% electrolysis process efficiency is used. It has been shown
that applying less than 0.6 V does not release any tritium.
Recovery efficiency subsequently increased until reaching a maximum
at 0.9 V. After that, recovery efficiency decreases for higher
voltage. Voltage should not be increased beyond 1.5 V to avoid
molten salt decomposition that occurs at .about.2 V.
[0037] In the electrolyzer the LiT in the salt phase is oxidized to
become molecular tritium (T2). The T2 collection electrode must be
designed to recover tritium before it can react back into the salt
phase. In an particular embodiment, a porous electrode enabling
surface oxidation of T2 with simultaneous argon flow through the
small electrode orifices is utilized to carry away the tritium
before back reaction can take place.
[0038] Getter
[0039] The noble gas (e.g., Ar)-T2 mixture flows from the
electrolyzer to a getter. Even though there might not be many
getters specifically designed to extract tritium from argon, this
component is more conventional because there are many hydrogen
getters, and argon/helium is unreactive. Several getter materials
can be used: titanium, yttrium, and depleted uranium, which is
likely the best option.
[0040] System Analysis.
[0041] For the 1.2 GW (thermal) LIFE engine pilot power plant,
tritium breeding is estimated at Rb=200 g/d. The total volume of
liquid lithium (blanket, pipes, pumps, and intermediate heat
exchangers) is estimated to be 100 m.sup.3. At density pLi=480
kg/m.sup.3 at T=500.degree. C., the total lithium mass is 480 tons.
The LIFE engine power plant has the goal of keeping the steady
state tritium inventory <50 g. Tritium steady state
concentration in the lithium loop is therefore .about.0.1 wppm. If
we used the values of DV=2.9 (measured) and .epsilon.=90%, we can
plot the liquid lithium flow that needs to be processed to recover
the tritium as a function of the extraction efficiency of the
centrifugal contactors (ii). FIG. 3 is a simplified plot of the
liquid lithium flow processed to recover tritium as a function of
extraction efficiency of centrifugal contactors. Extraction
efficiency is difficult to calculate or estimate because it depends
on specific geometry and operating conditions of the centrifugal
contactors.
[0042] In one implementation, one volume of liquid lithium is mixed
with three volumes of molten salt, resulting in a total mixed flow
of 210 m.sub.3/h. If we use the 45 m3/h (200 gpm) centrifugal
contactor we need 5 sets of three contactor per loop, or 15
contactor per loop, 60 contactors total.
[0043] FIG. 4 is a simplified plot showing the liquid lithium
fraction that is processed to maintain 0.1 wppm steady state
tritium concentration. To estimate the number of centrifugal
contactors we selected a commercial product that has a maximum
throughput of 757 liters per minute (200 gpm), with a footprint of
152 cm.times.152 cm (60''.times.60'') and height 163 cm (64'').
Using an estimate for the extraction efficiency, it is possible to
calculate the number of contactors needed for the whole LIFE engine
plant as a function of the centrifugal contactor extraction
efficiency. FIG. 5 is a plot showing the number of contactors as a
function of the centrifugal contactor extraction efficiency.
[0044] If we conservatively assume n=0.4, the total number of
centrifugal contactors is 16, or 4 contactors in parallel per loop.
FIG. 6 is a simplified schematic diagram of a tritium recovery
system by molten salt extraction according to an embodiment of the
present invention. As illustrated in FIG. 6, with four centrifugal
contactors (Centrifugal Contactors 1-4) in parallel, the
electrolyzer 612, the getter 614, and the lithium fluid circuit 620
and the molten salt circuit 622. In this example, a depleted
uranium getter is utilized. Additionally, although argon is used as
the noble gas, this is not required by the present invention. As
previously described, the molten salt flows from each of the
centrifugal contactors to an electrolyzer 612 where the T- is
oxidized to form T2, and a stream of argon sweeps the molecules of
T2 before back reaction occurs. The argon stream with tritium (and
possibly other impurities) goes through a getter unit 614 where
depleted uranium absorbs the tritium, which is later released by
heating during the regeneration stage.
[0045] The tritium recovery system by molten salt extraction
illustrated in FIG. 6 is suitable for use with the 1.2 GW (thermal)
LIFE engine power plant or other suitable sources that generate
tritium.
[0046] FIG. 7 is a simplified schematic diagram of a tritium
recovery system utilizing redundancy. Examples of the redundant
systems include the redundant contactors (Centrifugal Contactor 1
Redundant, Centrifugal Contactor mostly salt Redundant, and
Centrifugal Contactor mostly lithium Redundant), redundant
electrolysis tank (Electrolysis Tank Redundant) and the redundant
getters (Getters Redundant). One of ordinary skill in the art would
recognize many variations, modifications, and alternatives.
[0047] It is also understood that the examples and embodiments
described herein are for illustrative purposes only and that
various modifications or changes in light thereof will be suggested
to persons skilled in the art and are to be included within the
spirit and purview of this application and scope of the appended
claims.
* * * * *