U.S. patent application number 14/980617 was filed with the patent office on 2016-06-30 for nuclear materials processing.
This patent application is currently assigned to TerraPower, LLC. The applicant listed for this patent is TerraPower, LLC. Invention is credited to Ken Czerwinski, Joshua C. Walter.
Application Number | 20160189812 14/980617 |
Document ID | / |
Family ID | 55405444 |
Filed Date | 2016-06-30 |
United States Patent
Application |
20160189812 |
Kind Code |
A1 |
Czerwinski; Ken ; et
al. |
June 30, 2016 |
NUCLEAR MATERIALS PROCESSING
Abstract
This disclosure describes systems and methods for treating
nuclear fuel with supercritical fluids, such as supercritical
carbon dioxide. The addition of various ligands to the
supercritical fluids is disclosed, where one or more ligands can be
chosen to selectively remove one or more fission products from the
nuclear fuel. The nuclear fuel may be treated either within the
nuclear reactor or may be removed from the reactor before
treatment. This disclosure also presents methods and systems for
liquid nuclear fuel treatment with supercritical fluids in, for
example, a molten salt fast reactor, a traveling wave reactor, and
a containerized molten salt reactor.
Inventors: |
Czerwinski; Ken; (Seattle,
WA) ; Walter; Joshua C.; (Kirkland, WA) |
|
Applicant: |
Name |
City |
State |
Country |
Type |
TerraPower, LLC |
Bellevue |
WA |
US |
|
|
Assignee: |
TerraPower, LLC
Bellevue
WA
|
Family ID: |
55405444 |
Appl. No.: |
14/980617 |
Filed: |
December 28, 2015 |
Related U.S. Patent Documents
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Application
Number |
Filing Date |
Patent Number |
|
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62097235 |
Dec 29, 2014 |
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|
62098984 |
Dec 31, 2014 |
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62234889 |
Sep 30, 2015 |
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Current U.S.
Class: |
376/212 ;
376/311 |
Current CPC
Class: |
G21C 19/30 20130101;
G21C 1/22 20130101; G21C 19/50 20130101; G21C 1/03 20130101; G21C
3/24 20130101; Y02E 30/30 20130101; G21C 3/54 20130101; G21C 17/104
20130101; G21C 15/28 20130101; G21C 19/205 20130101; G21C 19/28
20130101; G21C 7/30 20130101; G21C 19/303 20130101; G21C 3/22
20130101 |
International
Class: |
G21C 19/30 20060101
G21C019/30; G21C 15/28 20060101 G21C015/28; G21C 19/50 20060101
G21C019/50; G21C 1/22 20060101 G21C001/22 |
Claims
1. A nuclear fission reactor comprising: a reactor core containing
a quantity of fuel salt including at least some fissionable
material and adapted to create a chain reaction in the fuel salt,
thereby generating heat and fission products in the fuel salt; at
least one heat exchanger adapted to transfer heat from the fuel
salt to a coolant; a power generation unit that converts heat in
the coolant into power; and a supercritical fluid separation system
adapted to remove at least some amount of fission products from the
fuel salt.
2. The nuclear fission reactor of claim 1 wherein the supercritical
fluid separation system comprises: a supercritical fluid contact
vessel which contacts fuel salt with a supercritical fluid; a fuel
salt transfer unit that circulates fuel salt between the reactor
core and the supercritical fluid contact vessel; a supercritical
fluid source fluidly connected to the supercritical fluid contact
vessel, the supercritical fluid including at least one ligand that
dissolves at least one fission product into the supercritical fluid
when the ligand contacts the at least one fission product; and a
controller controlling the transfer of fuel salt between the
reactor core and supercritical fluid contact vessel and controlling
the transfer of supercritical fluid from the supercritical fuel
source through the supercritical fluid contact vessel.
3. The nuclear fission reactor of claim 2 wherein the fuel salt
transfer unit maintains the fuel salt in a molten state when
transferring the fuel salt between the reactor core and the contact
vessel.
4. The nuclear fission reactor of claim 3 wherein the fuel salt
remains in a molten state while in the contact vessel during
contact with the supercritical fluid.
5. The nuclear fission reactor of claim 3 wherein the fuel salt
transfer unit maintains the fuel salt in a molten state when
transferring the fuel salt from the reactor core to the contact
vessel, and the contact vessel further comprises: a fuel salt
injector; a contact vessel environmental control system; and
wherein the controller operates the injector and environmental
control system so that the molten fuel salt solidifies into fuel
salt particles upon injection into the contact vessel.
6. The nuclear fission reactor of claim 5 wherein the fuel salt
transfer unit removes the fuel salt particles from the contact
vessel after the particles contact the supercritical fluid.
7. The nuclear fission reactor of claim 6 wherein the fuel salt
transfer unit melts the fuel salt particles to a molten fuel before
returning the molten fuel salt to the reactor core.
8. The nuclear fission reactor of claim 5 wherein the fuel salt
injector comprises: one or more nozzles adapted to disperse the
molten fuel salt into a spray of drops within the supercritical
fluid contained in the contact vessel, wherein the contact vessel
is maintained at a temperature and pressure by the environmental
control system that causes the dispersed molten fuel salt to
solidify into fuel salt particles.
9. The nuclear fission reactor of claim 5 wherein the contact
vessel environmental control system comprises one or more of: a
pressure sensor; a temperature sensor; a heater; a heat exchanger;
a fuel salt injection valve controlling the rate of flow of the
molten fuel salt into the contact vessel; a supercritical fluid
injection valve controlling the rate of flow of supercritical fluid
into the contact vessel; and a supercritical fluid extraction valve
controlling the rate of flow of supercritical fluid out of the
contact vessel.
10. The nuclear fission reactor of claim 2, wherein the
supercritical fluid flowing out of the contact vessel includes an
amount of fission products removed from the fuel salt and wherein
the supercritical fluid separation system further comprises: a
separation vessel fluidly connected to the contact vessel to
receive the supercritical fluid flowing out of the contact vessel
and separate at least some of the amount of fission products from
the supercritical fluid.
11. A nuclear fission reactor comprising: one or more fuel salt
containers, including a first fuel salt container containing a
quantity of fuel salt including at least some fissionable material;
a reactor core adapted to hold at least the first fuel salt
container and adapted to create a chain reaction in the fuel salt,
thereby generating heat in the fuel salt; at least one heat
exchanger adapted to transfer heat from the fuel salt to a coolant;
and a supercritical fluid extraction system adapted to pass a
supercritical fluid through the first fuel salt container, the
supercritical fluid extraction system including a supercritical
fluid source fluidly connectable to the first fuel salt container
that, when connected, delivers the supercritical fluid into the
first fuel salt container; a separation vessel fluidly connectable
to the first fuel salt container that receives the supercritical
fluid from the first fuel salt container; and a controller
controlling the transfer of supercritical fluid from the source
through the first fuel salt container into the separation
vessel.
12. A method of operating a nuclear fission reactor comprising:
charging a reactor core with an initial fuel containing fissionable
material; maintaining a first chain reaction in the reactor core
for a period of time, thereby generating a partially-reacted fuel
containing less fissionable material than in the initial fuel and
at least some amount of fission products greater than in the
initial fuel; contacting at least some of the partially-reacted
fuel with a supercritical fluid containing at least one ligand that
forms a metal complex with at least one fission product, thereby
creating a supercritical fluid and fission product mixture and a
regenerated fuel containing a lower amount of fission products than
in the partially-reacted fuel; separating the supercritical fluid
and fission product mixture from the regenerated fuel; initiating a
second chain reaction in the regenerated fuel; and extracting at
least some of the fission product from the supercritical fluid and
fission product mixture.
13. The method of claim 12 wherein the contacting is performed on
the at least some of the partially-reacted fuel without removing it
from the reactor core.
14. The method of claim 13 wherein the contacting is performed
without interrupting the first chain reaction in the reactor
core.
15. The method of claim 12 further comprising: removing the at
least some of the partially-reacted fuel from the reactor core
without interrupting the first chain reaction; performing the
contacting operation after removing the at least some of the
partially-reacted fuel from the reactor core; and returning the
regenerated fuel to the reactor core.
16. The method of claim 15, further comprising: returning the
regenerated fuel to the reactor without interrupting the first
chain reaction.
17. The method of claim 12 wherein the nuclear fission reactor is a
traveling wave reactor and the method further comprises: charging
the reactor core with a plurality of pins containing the initial
fuel containing fissionable material; maintaining the first chain
reaction in a breed-burn region of reactor core for a period of
time, thereby generating at least one pin containing
partially-reacted fuel; moving the at least one pin containing
partially-reacted fuel to a different position within the reactor;
and during or after the moving operation, contacting the
partially-reacted fuel in the at least one pin by passing the
supercritical fluid containing at least one ligand through the at
least one pin to obtain at least one pin containing regenerated
fuel.
18. The method of claim 17, wherein the moving operation is part of
a pin shuffling operation.
19. The method of claim 17 wherein initiating a second chain
reaction in the regenerated fuel is achieved by placing the at
least one pin containing the regenerated fuel at a location in the
breed-burn region of the reactor core.
20. The method of claim 12 wherein the nuclear fission reactor is a
molten salt reactor and wherein the contacting is performed on
partially-reacted molten salt fuel without removing it from the
reactor core.
21. The method of claim 20, further comprising: removing the at
least some of the partially-reacted molten salt fuel from the
reactor core without interrupting the first chain reaction;
performing the contacting operation after removing the at least
some of the partially-reacted molten salt fuel from the reactor
core; and returning regenerated molten salt fuel to the reactor
core.
22. The method of claim 21, further comprising: returning the
regenerated molten salt fuel to the reactor without interrupting
the first chain reaction.
Description
RELATED APPLICATIONS
[0001] This application claims the benefit of U.S. Provisional
Application Nos. 62/097,235, filed Dec. 29, 2014, 62/098,984, filed
Dec. 31, 2014, and 62/234,889, filed Sep. 30, 2015, which
applications are hereby incorporated by reference.
[0002] The present application is also related to U.S. patent
application Ser. No. ______ [Attorney Docket No.
14-001-UT1-US-INN], entitled "Targetry Coupled Separations" and
filed Dec. 23, 2015, which is specifically incorporated by
reference herein for all that it discloses and teaches.
INTRODUCTION
[0003] Several alternative designs for nuclear reactors have become
the subject of investigation. Two of these are molten salt reactors
and traveling wave reactors.
[0004] In a molten salt reactor, a radioactive fuel such as uranium
or thorium is dissolved into fluoride or chloride salts to form a
solution referred to as a "fuel salt." The fuel salt under normal
conditions is an immobile solid material, but when heated above
approximately 500.degree. C., it becomes a liquid. In a molten salt
reactor, the liquid fuel salt acts as both the heat source and a
heat transfer fluid that assists in removing heat from the reactor.
Tubes of fuel salt are deployed in a reactor core and, if the
concentration of the fissile material is high enough, a sustained
fission reaction may be created causing the fuel salt's temperature
to increase. In one design, the heated fuel salt may then be pumped
through a heat exchanger to transfer the heat to a different heat
transfer fluid (e.g., water or another molten salt). In an
alternative design, the second heat transfer fluid may be flowed
around stationary tubes of heated fuel salt. In either design, the
second heat transfer fluid is then used, directly or indirectly, to
generate power for beneficial use. Molten salt reactors are
considered safer than some other designs because, in the event of
an accident, the fuel salt will return to a solid, safe state. The
plant can operate near atmospheric pressure with a coolant that
returns to a solid form at ambient temperatures. This feature
simplifies the plant and enables safety systems that do not require
external electric power to safely shutdown, thereby assuring
greater safety for the public.
[0005] In a traveling wave reactor (TWR), sometimes also known as a
nuclear fission deflagration wave reactor or nuclear-burning-wave
reactor, the main reactor components are a reactor vessel filled
with a liquid sodium coolant and a reactor core. The reactor core
is submerged in the sodium pool in the reactor vessel. In the
center of the core are a few rods of enriched uranium (U-235),
surrounded by rods of depleted uranium (U-238). The U-235 serves as
an igniter, kick starting the traveling wave reaction--a
slow-moving chain reaction of parallel waves of fission traveling
through the uranium rods. These parallel waves initiate in the
center of the core, slowly consuming the fuel and generating heat
in the core. The sodium coolant is used to remove the heat from the
core. A containment vessel surrounds the reactor vessel to prevent
loss of sodium coolant in case of an unlikely leak from the reactor
vessel. The pumps circulate primary sodium coolant between the
reactor core and intermediate heat exchangers located in the pool.
These heat exchangers have non-radioactive intermediate sodium
coolant on the other side of the heat exchanger. Heated
intermediate sodium coolant is circulated to steam generators that
generate steam to drive turbines of electrical generators.
[0006] In theory, TWRs require no fuel reprocessing, use depleted
or natural uranium as their primary fuel, require only a small
amount of enriched uranium at start-up, and never need refueling.
This core longevity depends on the size of the initial charge of
the uranium and on the fuel burn-up achieved during reactor
operation.
Nuclear Materials Processing
[0007] This disclosure describes systems and methods for treating
nuclear fuel with supercritical fluids, such as supercritical
carbon dioxide. The addition of various ligands to the
supercritical fluids is disclosed, where one or more ligands can be
chosen to selectively remove one or more fission products from the
nuclear fuel. The nuclear fuel may be treated either within the
nuclear reactor or may be removed from the reactor before
treatment. This disclosure also presents methods and systems for
liquid nuclear fuel treatment with supercritical fluids in, for
example, a molten salt fast reactor, a traveling wave reactor, and
a containerized molten salt reactor.
[0008] An aspect of the present disclosure is a nuclear fission
reactor that includes a reactor core containing a quantity of fuel
salt including at least some fissionable material. The reactor is
adapted to create a chain reaction in the fuel salt, thereby
generating heat and fission products in the fuel salt. At least one
heat exchanger is provided to transfer heat from the fuel salt to a
coolant and a power generation unit is also provided that converts
heat in the coolant into power. The reactor is further provided
with a supercritical fluid separation system that is adapted to
remove at least some amount of fission products from the fuel salt.
In an ex situ treatment embodiment, the supercritical fluid
separation system includes: a supercritical fluid contact vessel
which contacts fuel salt with a supercritical fluid; a fuel salt
transfer unit that circulates fuel salt between the reactor core
and the supercritical fluid contact vessel; a supercritical fluid
source fluidly connected to the supercritical fluid contact vessel;
and a controller controlling the transfer of fuel salt between the
reactor core and supercritical fluid contact vessel and controlling
the transfer of supercritical fluid from the supercritical fuel
source through the supercritical fluid contact vessel. The
supercritical fluid may include at least one ligand that dissolves
at least one fission product into the supercritical fluid when the
ligand contacts the at least one fission product in the fuel salt.
The supercritical fluid separation system may further include a
separation vessel fluidly connected to the contact vessel to
receive the supercritical fluid and fission product mixture flowing
out of the contact vessel and separate at least some of the amount
of fission products from the supercritical fluid. In an embodiment,
the supercritical fluid is supercritical carbon dioxide containing
a ligand that dissolves at least one fission product.
[0009] The fuel salt transfer unit may or may not maintain the fuel
salt in a molten state when transferring the fuel salt from the
reactor core to the contact vessel or from the contact vessel to
the reactor core depending on the embodiment. Likewise, the fuel
salt may or may not remain in a molten state while in the contact
vessel during contact with the supercritical fluid. When the fuel
salt transfer unit maintains the fuel salt in a molten state when
transferring the fuel salt from the reactor core to the contact
vessel, the contact vessel may include a molten fuel salt injector
and a contact vessel environmental control system. The controller
may furthermore operate the injector and environmental control
system in a manner that causes the molten fuel salt to solidify
into fuel salt particles upon injection into the contact vessel.
The fuel salt transfer unit may also remove the solid fuel salt
particles from the contact vessel after the particles contact the
supercritical fluid. The fuel salt transfer unit may then melt the
fuel salt particles to a molten state before returning the molten
fuel salt to the reactor core. Alternatively, the transfer unit may
return the solid fuel salt particles to the reactor core. In an
embodiment, the fuel salt injector may include one or more nozzles
adapted to disperse the molten fuel salt into a spray of drops
within the supercritical fluid contained in the contact vessel,
wherein the contact vessel is maintained at a temperature and
pressure by the environmental control system that causes the
dispersed molten fuel salt to solidify into fuel salt particles. In
an embodiment, the contact vessel environmental control system may
include one or more of: a pressure sensor; a temperature sensor; a
heater; a heat exchanger; a fuel salt injection valve controlling
the rate of flow of the molten fuel salt into the contact vessel; a
supercritical fluid injection valve controlling the rate of flow of
supercritical fluid into the contact vessel; and a supercritical
fluid extraction valve controlling the rate of flow of
supercritical fluid out of the contact vessel.
[0010] Another aspect of this disclosure is a nuclear fission
reactor having a supercritical fluid extraction system adapted to
pass a supercritical fluid through a fuel salt container. The
nuclear fission reactor includes: one or more fuel salt containers,
including a first fuel salt container containing a quantity of fuel
salt including at least some fissionable material; a reactor core
adapted to hold at least the first fuel salt container and adapted
to create a chain reaction in the fuel salt, thereby generating
heat in the fuel salt; at least one heat exchanger adapted to
transfer heat from the fuel salt to a coolant; and the
supercritical fluid extraction system. The supercritical fluid
extraction system further includes: a supercritical fluid source
fluidly connectable to the first fuel salt container that, when
connected, can deliver the supercritical fluid into the first fuel
salt container; a separation vessel fluidly connectable to the
first fuel salt container that receives the supercritical fluid
from the first fuel salt container; and a controller controlling
the transfer of supercritical fluid from the source through the
first fuel salt container into the separation vessel. The nuclear
fission reactor may further include a container transfer unit that
transfers the fuel salt container between the reactor core and the
separation system. The separation system may further be adapted to
pass the supercritical fluid through the fuel salt container while
the fuel salt container is in the reactor core.
[0011] In an embodiment, the supercritical fluid is supercritical
carbon dioxide containing a ligand that dissolves at least one
fission product. The ligand may be selected from cupferron,
chloroanillic acid, .beta.-diketone,
N-benzoyl-N-phenylhydroxylamine, .alpha.-dioximines
diaminobenzidine, a porphyrine compound such as porphine,
8-hydroxyquinoline, nitrosonapthols, nitrosophenols,
ethylenediaminetetraacetic acid, diphenylcarbazide,
diphenylcarbazone, Azoazoxy BN, sodium diethlydithiocarbamate,
dithizone, bismuthiol II, thiothenoyltrifluoracetone, thioxine,
thiophosphinic acid, phosphine sulfide, phosphorothioic acid, and
tributylphoshpate.
[0012] Yet another aspect of this disclosure is a traveling wave
reactor that includes a supercritical fluid separation system. The
traveling wave reactor includes: a reactor core containing a first
fuel assembly having at least one fuel pin containing a fuel
material including at least some fissionable material and fission
products; a reactor vessel containing a primary sodium coolant,
wherein the reactor core is within the reactor vessel and in
contact with the primary sodium coolant; an assembly shuffling
system adapted to move the first fuel assembly from a first
location within the reactor core to a second location within the
reactor core; and a supercritical fluid separation system that
removes fission products from the fuel material in the first fuel
assembly. The supercritical fluid separation system may further
include: a supercritical fluid source fluidly connectable to the
first fuel assembly that delivers the supercritical fluid into the
first fuel assembly; a separation vessel fluidly connectable to the
first fuel assembly that receives the supercritical fluid from the
first fuel assembly; and a controller controlling the transfer of
supercritical fluid from the source through the first fuel assembly
into the separation vessel. The supercritical fluid source may be
fluidly connectable to at least one pin in the first fuel assembly
and may deliver the supercritical fluid into the at least one pin.
The supercritical fluid separation system may further include a
separation vessel fluidly connectable to at least one pin in the
first fuel assembly that receives the supercritical fluid from the
at least one pin in the first fuel assembly and a controller that
controls the transfer of supercritical fluid from the source
through the at least one pin of the first fuel assembly into the
separation vessel. The assembly shuffling system may be adapted to
remove the first assembly from the reactor core and the
supercritical fluid separation system may remove fission products
from the first assembly when the first assembly is outside of the
reactor core. In an alternative embodiment, the assembly shuffling
system may be further adapted to move the first assembly from the
first location to an intermediate location within the reactor core
before moving the first assembly to the second location and the
supercritical fluid separation system may remove fission products
from the first assembly when the first assembly is in the
intermediate location within the reactor core. In yet another
embodiment, the supercritical fluid separation system may remove
fission products from the first assembly when the first assembly is
in the first location within the reactor core.
[0013] The traveling wave reactor may include a fission product
handling system that receives fission products from the
supercritical fluid separation system. The traveling wave reactor
may also include a swelling monitoring device that monitors
expansion of the fuel material during operation of the traveling
wave reactor and the controller may control the transfer of
supercritical fluid based on the expansion of the fuel material.
The traveling wave reactor may further include a coolant monitoring
device that monitors a concentration of fission products in the
coolant and the controller may control the transfer of
supercritical fluid based on the concentration of fission products
in the coolant. The supercritical fluid separation system may be
further adapted to remove fission products from the primary sodium
coolant. The traveling wave reactor may further include a transfer
vessel adapted to hold the first fuel assembly in argon and the
supercritical fluid separation system may be further adapted to
remove fission products from argon that has been exposed to the
first fuel assembly. The traveling wave reactor may further include
a coolant cleaning system including an absorber that removes
fission products from the primary sodium coolant and the
supercritical fluid separation system may be adapted to remove
fission products from the absorber.
[0014] In an embodiment, the supercritical fluid is supercritical
carbon dioxide containing a ligand that dissolves at least one
fission product. The ligand may be selected from cupferron,
chloroanillic acid, .beta.-diketone,
N-benzoyl-N-phenylhydroxylamine, .alpha.-dioximines
diaminobenzidine, a porphyrine compound such as porphine,
8-hydroxyquinoline, nitrosonapthols, nitrosophenols,
ethylenediaminetetraacetic acid, diphenylcarbazide,
diphenylcarbazone, Azoazoxy BN, sodium diethlydithiocarbamate,
dithizone, bismuthiol II, thiothenoyltrifluoracetone, thioxine,
thiophosphinic acid, phosphine sulfide, phosphorothioic acid, and
tributylphoshpate.
[0015] Yet another aspect of this disclosure is a method of
operating a nuclear fission reactor. In this aspect, the method
includes: charging a reactor core with an initial fuel containing
fissionable material; maintaining a first chain reaction in the
reactor core for a period of time, thereby generating a
partially-reacted fuel containing less fissionable material than in
the initial fuel and at least some amount of fission products
greater than in the initial fuel; contacting at least some of the
partially-reacted fuel with a supercritical fluid containing at
least one ligand that forms a metal complex with at least one
fission product, thereby creating a supercritical fluid and fission
product mixture and a regenerated fuel containing a lower amount of
fission products than in the partially-reacted fuel; separating at
least some of the supercritical fluid and fission product mixture
from the regenerated fuel; initiating a second chain reaction in
the regenerated fuel; and extracting at least some of the fission
product from the supercritical fluid and fission product mixture.
In the method the contacting may be performed on the at least some
of the partially-reacted fuel without removing it from the reactor
core. In the method the contacting operation may be performed
without interrupting the first chain reaction in the reactor core.
The method may further include: removing the at least some of the
partially-reacted fuel from the reactor core without interrupting
the first chain reaction; performing the contacting operation after
removing the at least some of the partially-reacted fuel from the
reactor core; and returning the regenerated fuel to the reactor
core. The method may include returning the regenerated fuel to the
reactor without interrupting the first chain reaction.
[0016] In the method, the nuclear fission reactor may be a
traveling wave reactor and the method may further include: charging
the reactor core with a plurality of pins containing the initial
fuel containing fissionable material; maintaining the first chain
reaction in a breed-burn region of reactor core for a period of
time, thereby generating at least one pin containing
partially-reacted fuel; moving the at least one pin containing
partially-reacted fuel to a different position within the reactor;
and during or after the moving operation, contacting the
partially-reacted fuel in the at least one pin by passing the
supercritical fluid containing at least one ligand through the at
least one pin to obtain at least one pin containing regenerated
fuel. The moving operation may be part of a pin shuffling
operation. In the method, initiating a second chain reaction in the
regenerated fuel may be achieved by placing the at least one pin
containing the regenerated fuel at a location in the breed-burn
region of the reactor core.
[0017] In the method, the nuclear fission reactor may be a molten
salt reactor and the contacting operation may be performed on
partially-reacted molten salt fuel without removing it from the
reactor core. This method may further include: removing the at
least some of the partially-reacted molten salt fuel from the
reactor core without interrupting the first chain reaction;
performing the contacting operation after removing the at least
some of the partially-reacted molten salt fuel from the reactor
core; and returning regenerated molten salt fuel to the reactor
core. The method may further include returning the regenerated
molten salt fuel to the reactor without interrupting the first
chain reaction.
[0018] These and various other features as well as advantages which
characterize the systems and methods described herein will be
apparent from a reading of the following detailed description and a
review of the associated drawings. Additional features are set
forth in the description which follows, and in part will be
apparent from the description, or may be learned by practice of the
technology. The benefits and features of the technology will be
realized and attained by the structure particularly pointed out in
the written description and claims hereof as well as the appended
drawings.
[0019] It is to be understood that both the foregoing general
description and the following detailed description are explanatory
and are intended to provide further explanation of the invention as
claimed
BRIEF DESCRIPTION OF THE DRAWINGS
[0020] The following drawing figures, which form a part of this
application, are illustrative of described technology and are not
meant to limit the scope of the invention as claimed in any manner,
which scope shall be based on the claims appended hereto.
[0021] FIG. 1 illustrates a simplified schematic view of a molten
salt fast spectrum nuclear reactor, in accordance with one or more
embodiments of the present disclosure.
[0022] FIGS. 2A and 2B illustrate a simplified schematic view of a
molten salt fast spectrum nuclear reactor with a protective layer
disposed on one or more internal surfaces of the reactor, in
accordance with one or more embodiments of the present
disclosure.
[0023] FIG. 3 illustrates an embodiment of a nuclear power plant
for generating power from a nuclear reaction using a molten
chloride fast reactor.
[0024] FIG. 4 illustrates another embodiment of a simplified
schematic view of a molten salt nuclear reactor.
[0025] FIG. 5 is a block flow diagram of an embodiment of an
example supercritical fluid treatment system.
[0026] FIG. 6 is an embodiment of a method for treating a fuel salt
reactor with a supercritical fluid treatment system.
[0027] FIG. 7 is a block flow diagram of the example fuel salt
reactor used with the example supercritical fluid treatment
components.
[0028] FIG. 8 is a block flow diagram of the example containerized
fuel salt reactor used with the example supercritical fluid
treatment components.
[0029] FIG. 9 is a block flow diagram of the example traveling wave
reactor used with the example supercritical fluid treatment
components of FIG. 5.
[0030] FIG. 10 is an embodiment of a method for operating a reactor
with supercritical fluid separation.
DETAILED DESCRIPTION
[0031] This disclosure describes systems and methods for treating
nuclear fuel with supercritical fluids, such as supercritical
carbon dioxide. The addition of various ligands to the
supercritical fluids is disclosed, where one or more ligands can be
chosen to selectively remove one or more fission products from the
nuclear fuel. The nuclear fuel may be treated either within the
nuclear reactor or may be removed from the reactor before
treatment. This disclosure also presents methods and systems for
liquid nuclear fuel treatment with supercritical fluids in, for
example, a molten salt fast reactor, a traveling wave reactor, and
a containerized molten salt reactor.
[0032] As used herein, fissionable material includes any nuclide
capable of undergoing fission when exposed to low-energy thermal
neutrons or high-energy neutrons. Furthermore, for the purposes of
this disclosure, fissionable material includes any fissile
material, any fertile material or combination of fissile and
fertile materials. As used herein, a direct fission product is the
atom that remains after fission of a fissile atom. As used herein,
an indirect fission product is a decay daughter, grand-daughter,
etc., that results from the radioactive decay of a direct fission
product. However, at any given point in time, some quantity of a
particular species fission product compound, such as .sup.99Mo,
will be a direct product and the remaining quantity will be
indirect products, as there can be multiple decay chains at work.
As used herein, a fuel salt can include target material (material
that can undergo fission) as well as ancillary material (including
post-radiation fission products and non-fissile salts in the
fuel).
[0033] The disclosed treatment systems and methods can be used with
molten salt reactor designs and related systems, where the molten
salt includes molten fluoride fuel salt, molten chloride fuel salt,
fuel salts of UCl.sub.xF.sub.y variety, as well as bromide fuel
salts. Binary, ternary and quaternary chloride fuel salts of
uranium, as well as other fissionable elements, are contemplated.
This disclosure also presents methods and systems for manufacturing
such fuel salts, for creating salts that reduce corrosion of the
reactor components and for creating fuel salts that are not
suitable for weapons applications.
[0034] Metalloids are elements with both metallic and non-metallic
properties, and include arsenic, selenium and tellurium. A metal is
an element that forms positive ions in solution, and produces
oxides that form hydroxides, rather than acids, with water. Metals
include alkali metals, alkali-earth metals, transition metals,
noble metals (including the precious metals gold, platinum and
silver), rare metals, rare-earth metals (lanthanides), actinides
(including the transuranic metals), light metals, heavy metals,
synthetic metals and radioactive metals. Specific examples are
given herein of extraction methods for extracting lanthanides and
actinides (collectively referred to as the f-group elements from
the filling of their 4f and 5f orbitals). The f-group elements are
commonly produced by nuclear fission reactions, and the actinides
are radioactive. Transition metals are commonly used or produced in
many industrial processes and products, such as mineral production
or fly ash.
[0035] Suitable fluids and/or supercritical fluids for use in the
disclosed embodiments include carbon dioxide, nitrogen, nitrous
oxide, methane, ethylene, propane and propylene. (See Table I)
Carbon dioxide is a particularly suitable fluid for both
subcritical and supercritical fluid extractions because of its
moderate chemical constants (T.sub.C=31.degree. C., P.sub.C=73 atm)
and its inertness (i.e. it is non-explosive and thoroughly safe for
extractions, even extractions performed at supercritical
conditions). Carbon dioxide also is abundantly available and
relatively inexpensive. Virtually any condition above the critical
temperature and pressure for carbon dioxide is acceptable for
producing a supercritical carbon dioxide fluid solvent useful for
practicing the extraction process as described herein.
[0036] The fluids listed in Table I may be used either individually
or in combination, as mixed fluids or supercritical fluid
solvents.
TABLE-US-00001 TABLE I PHYSICAL PARAMETERS OF SELECTED
SUPERCRITICAL FLUIDS* Molecular T.sub.c P.sub.c .rho..sub.c Fluid
Formula (.degree. C.) (atm) (g/ml) .rho.400.sub.atm** Carbon
dioxide CO.sub.2 31.1 72.9 0.47 0.96 Nitrous oxide N.sub.2O 36.5
71.7 0.45 0.94 Ammonia NH.sub.3 132.5 112.5 0.24 0.40 .eta.-Pentane
C.sub.5H.sub.12 196.6 33.3 0.23 0.51 .eta.-Butane C.sub.4H.sub.10
152.0 37.5 0.23 0.50 .eta.-Propane C.sub.3H.sub.6 96.8 42.0 0.22 --
Sulfur hexafluoride SF.sub.6 45.5 37.1 0.74 1.61 Xenon Xe 16.6 58.4
1.10 2.30 Dichlorodifluoromethane CCl.sub.2F.sub.2 111.8 40.7 0.56
1.12 Trifluoromethane CHF.sub.3 25.9 46.9 0.52 -- Methanol
CH.sub.3OH 240.5 78.9 0.27 -- Ethanol C.sub.2H.sub.5OH 243.4 63.0
0.28 -- Isopropanol C.sub.3H.sub.7OH 235.3 47.0 0.27 -- Diethyl
ether (C.sub.2H.sub.25).sub.2O 193.6 36.3 0.27 -- Water H.sub.2O
374.1 218.3 *data from Matheson Gas Data Book (1980) and CRC
Handbook of Chemistry and Physics (CRC Press. Boca Raton. Florida
1984). **T.sub.r = 1.03
[0037] In addition, a modifying solvent (also referred to as an
extractant) may be added to the fluid, including supercritical
fluids, to improve the solvent characteristics thereof. Currently,
useful modifying solvents include water, organic solvents, such as
low to medium boiling point alcohols and esters, particularly the
lower alkyl alcohols and esters, such as methanol, ethanol, ethyl
acetate and the like; and phosphate esters, particularly lower
alkyl phosphate esters, such as tributyl phosphate. With more
specificity, but without limitation, the modifiers are usually
added to the fluids at proportions of between about 0.1% and 20.0%
by weight. The extractants contemplated for use herein may not be
supercritical fluids at the disclosed operating conditions. Such
extractants may be simply dissolved in the fluid solvents,
including the supercritical fluid solvents, to improve the solvent
properties.
[0038] In one embodiment, the chosen extractant is combined with a
supercritical fluid at the described proportions prior to feeding
the supercritical fluid to the extraction vessel. Alternatively,
the supercritical fluid is fed to the extraction vessel without the
extractant. The extractant is then introduced into the extraction
vessel and thereby combined with the supercritical fluid.
[0039] Extractants also include chelating agents. Chelating agents
useful for solubilizing metals and metalloids in supercritical
fluids are listed in Table II. The list of chelating agents is not
exhaustive and for illustration only. Many other chelating agents,
now known or hereafter discovered that are useful for forming metal
and metalloid chelates, also may be used.
TABLE-US-00002 TABLE II COMMONLY USED METAL CHELATING AGENTS Oxygen
Donating Chelating Agents Cupferron Chloroanillic acid and related
reagents .beta.-diketones and related reagents
N-Benzoyl-N-phenylhydroxylamine and related reagents Macrocyclic
compounds Nitrogen Donating Chelating Agents .alpha.-dioximines
Diaminobenzidine and related reagents Porphyrines and related
reagents Oxygen and Nitrogen Donating Chelating Agents 8-Hydroxy
quinoline Nitrosonapthols and nitrosophenols EDTA and other
complexionates Diphenylcarbazide and diphenylcarbazone Azoazoxy BN
Sulfur or Phosphorus Donating Chelating Agents Sodium
diethlydithiocarbamate and related reagents Dithizone and related
reagents Bismuthiol II Thiothenoyltrifluoracetone Thioxine
Thiophosphinic acids Phosphine Sulfides Phosphorothioic acids
Tributylphosphate and related reagents
[0040] Prior to discussing treating fuel salt with supercritical
fluid in greater detail, a brief discussion of the general
components of various nuclear reactors will be discussed. FIGS. 1-7
generally describe systems and methods of operating a molten salt
nuclear reactor, a traveling wave reactor, and a containerized
molten salt reactor. For instance, FIGS. 1-3 depict various
embodiments of a molten salt nuclear reactor 100 for operating in a
fast spectrum breed-and-burn mode. These are just examples to
provide context for discussion of treating fuel salt with
supercritical fluid and the reader should understand that
potentially any molten fuel nuclear reactor could be adapted to use
the fuel embodiments described below. While various fluoride salts
may be utilized in molten salt nuclear reactors as described below,
fluoride-based fuel salts typically display heavy metal
concentrations significantly below that which may be achieved with
chloride-based and chloride-fluoride-based fuel salts described in
the present disclosure.
[0041] FIG. 1 illustrates a simplified schematic view of a molten
salt fast spectrum nuclear reactor 100, in accordance with one or
more embodiments of the present disclosure. In one embodiment, the
reactor 100 includes a reactor core section 102. The reactor core
section 102 (which may also be referred to as the "reactor vessel")
includes a fuel input 104 and a fuel output 106. The fuel input 104
and the fuel output 106 are arranged such that during operation a
flow of the molten fuel salt 108 is established through the reactor
core section 102. For example, the fuel input 104 and/or the fuel
output 106 may consist of conical sections acting as converging and
diverging nozzles respectively. In this regard, the molten fuel 108
is fluidically transported through the volume of the reactor core
section 102 from the input 104 to the output 106 of the reactor
core section 102. Although FIG. 1 shows fluid fuel flow with
arrows, it is to be appreciated that the direction of flow may be
modified as appropriate for different reactor and plant
configurations. Specifically, FIG. 1 shows fluid fuel 108 flow from
the `bottom` to the `top` in the central core region, and
alternative apparatuses may create and/or maintain a fluid fuel 108
flow from the top towards the bottom in the central core
region.
[0042] The reactor core section 108 may take on any shape suitable
for establishing criticality within the molten fuel salt 108 within
the reactor core section 102. By way of non-limiting example, the
reactor 100 may include, but is not limited to, an elongated core
section, as depicted in FIG. 1. In addition, the reactor core
section 108 may take on any cross-sectional shape. By way of
non-limiting example, the reactor core section 108 may have, but is
not required to have, a circular cross-section, an ellipsoidal
cross-section or a polygonal cross-section.
[0043] The dimensions of the reactor core section 102 are selected
such that criticality is achieved within the molten fuel salt 108
when flowing through the reactor core section 102. Criticality
refers to a state of operation in which the nuclear fuel sustains a
fission chain reaction, i.e., the rate of production of neutrons in
the fuel is at least equal to rate at which neutrons are consumed
(or lost). For example, in the case of an elongated core section,
the length and cross-sectional area of the elongated core section
may be selected in order to establish criticality within the
reactor core section 102. It is noted that the specific dimensions
necessary to establish criticality are at least a function of the
type of fissile material, fertile material and/or carrier salt
contained within the reactor 100. Principles of molten fuel nuclear
reactors are described in U.S. patent application Ser. No.
12/118,118 to Leblanc, filed on May 9, 2008, which is incorporated
herein in the entirety.
[0044] The reactor core section 102 is formed from any material
suitable for use in molten salt nuclear reactors. For example, the
bulk portion of the reactor core section 102 may be formed, but is
not required to be formed, from one or more molybdenum alloy, one
or more zirconium alloys (e.g., ZIRCALOY.TM.), one or more niobium
alloys, one or more nickel alloys (e.g., HASTELLOY.TM. N) or high
temperature ferritic, martensitic, or stainless steel and the like.
It is further noted that the internal surface may coated, plated or
lined with one or more additional materials in order to provide
resistance to corrosion and/or radiation damage, as discussed in
additional detail further herein.
[0045] In the embodiment shown, the reactor 100 includes a primary
coolant system 110 that takes heat from the reactor core 102 and
transfers that heat to the secondary coolant 126 in the secondary
coolant system 120 via the heat exchanger 119. In the embodiment
illustrated in FIG. 1, the molten fuel salt 108 is used as the
primary coolant 118. Cooling is achieved by flowing molten fuel
salt 108 heated by the ongoing chain reaction from the reactor core
102, and flowing cooler molten fuel salt 108 into the reactor core
102, at the rate necessary to maintain the temperature of the
reactor core 102 within its operational range. In this embodiment,
the primary coolant system 110 is adapted and sized to maintain the
molten fuel salt 108 in a subcritical condition when outside of the
reactor core 102.
[0046] The primary coolant system 110 may include one or more
primary coolant loops 112 formed from piping 114. The primary
coolant system 110 may include any primary coolant system
arrangement known in the art suitable for implementation in a
molten fuel salt context. The primary coolant system 110 may
circulate fuel 108 through one or more pipes 114 and/or fluid
transfer assemblies of the one or more primary coolant loops 112 in
order to transfer heat generated by the reactor core section 102 to
downstream thermally driven electrical generation devices and
systems. For purposes of simplicity, a single primary coolant loop
112 is depicted in FIG. 1. It is recognized herein, however, that
the primary coolant system 110 may include multiple parallel
primary coolant loops (e.g., 2-5 parallel loops), each carrying a
selected portion of the molten fuel salt inventory through the
primary coolant circuit.
[0047] In an alternative embodiment (an example of which is shown
in FIG. 2A), the primary coolant system 110 may be configured such
that a primary coolant 118 (different than the molten fuel salt
108) enters the reactor core section (e.g., main vessel). In this
embodiment, the fuel salt 108 does not leave the reactor core
section, or main vessel, but rather the primary coolant 118 is
flowed into the reactor core 102 to maintain the temperature of the
core within the desired range. It is noted that in this embodiment
the reactor 100 may include an additional heat exchanger (not
shown) in the reactor core section 102, or main vessel. In this
embodiment, the secondary coolant system 120 may be optional; the
usable power can be derived directly from the primary coolant
system 110. In this embodiment, the primary coolant may be a
chloride salt with a suitable melting point. For example, the salt
may be a mixture of sodium chloride and magnesium chloride.
[0048] In the embodiment shown in FIG. 1, the primary coolant
system 110 includes one or more pumps 116. For example, one or more
pumps 116 may be fluidically coupled to the primary coolant system
110 such that the one or more pumps 116 drive the primary coolant
118, in this case the molten fuel 108, through the primary
coolant/reactor core section circuit. The one or more pumps 116 may
include any coolant/fuel pump known in the art. For example, the
one or more fluid pumps 116 may include, but are not limited to,
one or more mechanical pumps fluidically coupled to the primary
coolant loop 112. By way of another example, the one or more fluid
pumps 116 may include, but are not limited to, one or more
electromagnetic (EM) pumps fluidically coupled to the primary
coolant loop 112.
[0049] FIG. 1 further illustrates that the reactor 100 includes a
secondary coolant system 120 thermally coupled to the primary
coolant system 110 via one or more heat exchangers 119. The
secondary coolant system 120 may include one or more secondary
coolant loops 122 formed from piping 124. The secondary coolant
system 120 may include any secondary coolant system arrangement
known in the art suitable for implementation in a molten fuel salt
context. The secondary coolant system 120 may circulate a secondary
coolant 126 through one or more pipes 124 and/or fluid transfer
assemblies of the one or more secondary coolant loops 122 in order
to transfer heat generated by the reactor core section 102 and
received via the primary heat exchanger 119 to downstream thermally
driven electrical generation devices and systems. For purposes of
simplicity, a single secondary coolant loop 124 is depicted in FIG.
1. It is recognized herein, however, that the secondary coolant
system 120 may include multiple parallel secondary coolant loops
(e.g., 2-5 parallel loops), each carrying a selected portion of the
secondary coolant through the secondary coolant circuit. It is
noted that the secondary coolant may include any second coolant
known in the art. By way of example, the secondary coolant may
include, but is not limited to, liquid sodium.
[0050] It is further noted that, while not depicted in FIG. 1, the
reactor 100 may include any number of additional or intermediate
heating/cooling systems and/or heat transfer circuits. Such
additional heating/cooling systems may be provided for various
purposes in addition to maintaining the reactor core 102 within its
operational temperature range. For example, a tertiary heating
system may be provided for the reactor core 102 and primary coolant
system 110 to allow a cold reactor containing solidified fuel salt
to be heated to an operational temperature in which the salt is
molten and flowable.
[0051] Other ancillary components 127 may also be utilized, as
illustrated, in the primary coolant loop 112. Such ancillary
components 127 may be include one or more filters or drop out boxes
for removing particulates that precipitate from the primary coolant
118 during operation. To remove unwanted liquids from the primary
coolant 118, the ancillary components 127 may include any suitable
liquid-liquid extraction system such as one or more co-current or
counter-current mixer/settler stages, an ion exchange technology,
or a gas absorption system. For gas removal, the ancillary
components 127 may include any suitable gas-liquid extraction
technology such as a flash vaporization chamber, distillation
system, or a gas stripper. Some additional embodiments of ancillary
components 127 are discussed in greater detail below.
[0052] It is noted herein that the utilization of various metal
salts, such as metal chloride salts, in reactor 100 may cause
corrosion and/or radiation degradation over time. A variety of
measures may be taken in order to mitigate the impact of corrosion
and/or radiation degradation on the integrity of the various
salt-facing components (e.g., reactor core section 102, primary
coolant piping 114, heat exchanger 119 and the like) of the reactor
100 that come into direct or indirect contact with the fuel salt or
its radiation.
[0053] In one embodiment, the velocity of fuel flow through one or
more components of the reactor 100 is limited to a selected fuel
salt velocity. For example, the one or more pumps 116 may drive the
molten fuel 108 through the primary coolant loop 112 of the reactor
100 at a selected fuel salt velocity. It is noted that in some
instances a flow velocity below a certain level may have a
detrimental impact on reactor performance, including the breeding
process and reactor control. By way of non-limiting example, the
total fuel salt inventory in the primary loop 112 (and other
portions of the primary coolant system 110) may exceed desirable
levels in the case of lower velocity limits since the
cross-sectional area of the corresponding piping of the primary
loop 112 scales upward as flow velocity is reduced in order to
maintain adequate volumetric flow through the primary loop 112. As
such, very low velocity limits (e.g., 1 m/s) result in large
out-of-core volumes of fuel salt and can negatively impact the
breeding process of the reactor 100 and reactor control. In
addition, a flow velocity above a certain level may detrimentally
impact reactor performance and longevity due to erosion and/or
corrosion of the internal surfaces of the primary loop 112 and/or
reactor core section 102. As such, suitable operational fuel salt
velocities may provide a balance between velocity limits required
to minimize erosion/corrosion and velocity limits required to
manage out-of-core fuel salt inventory. For example, in the case of
a molten chloride fuel salt, the fuel salt velocity may be
controlled from 2-20 m/s, such as, but not limited to, 7 m/s.
[0054] FIGS. 2A and 2B illustrate a simplified schematic view of a
molten salt fast spectrum nuclear reactor 100 with a protective
layer 128 disposed on one or more internal surfaces of the reactor
100, in accordance with one or more embodiments of the present
disclosure.
[0055] In one embodiment, the protective layer 128 is disposed on
one or more surfaces of the reactor 100 facing the fuel salt 108 of
the reactor 100. The protective layer 128 may provide resistance to
corrosion and/or radiation degradation of one or more reactor
salt-facing surfaces of the reactor 100. For the purposes of the
present disclosure, a material resistant to corrosion and/or
radiation degradation is interpreted as any material displaying
resistance to corrosion and/or radiation degradation superior to
the underlying bare surface of the reactor 100.
[0056] The protective layer 128 may include any material known in
the art suitable for providing an internal surface of a reactor
with corrosion and/or radiation resistance to a corresponding
nuclear fuel salt. Thus, the material of the protective layer 128
may vary depending on the salt 108 used. In one embodiment, the
protective layer 128 includes one or more refractory metals. For
example, the protective layer 128 may include, but is not limited
to, one or more of niobium, molybdenum, tantalum, tungsten or
rhenium. In another embodiment, the protective layer 128 includes
one or more refractory alloys. For example, the protective layer
128 may include, but is not limited to, one or more of a molybdenum
alloy (e.g., titanium-zirconium-molybdenum (TZM) alloy), a tungsten
alloy, tantalum, a niobium or a rhenium. In another embodiment, the
protective layer 128 includes one or more nickel alloys. In another
embodiment, the protective layer 128 includes a carbide, such as,
but not limited to, silicon carbide.
[0057] In an embodiment, the protective layer 128 is formed by
plating the internal surface of the one or more portions (e.g.,
piping 114 or primary loop 112) of the reactor 100 with the
selected protective material. In another embodiment, the protective
layer 128 includes one or more coatings of the selected protective
material disposed on the internal surface of one or more portions
of the reactor 100. In yet another embodiment, the bulk material of
the various components may be formed with one or more of the
protective materials described above. For instance, the piping 114
of the primary coolant loop 112 may include, but is not limited to,
TZM piping.
[0058] In one embodiment, as shown in FIG. 2A, the internal
salt-facing surface of the reactor core section 102 includes a
protective layer 128. For example, the vessel of the reactor core
section 102 may be formed from steel or a zirconium alloy, with
refractory alloy or nickel alloy plating disposed on the internal
salt-facing surface of the reactor core section 102 to form the
protective layer 128. For instance, the reactor core section 102
may include, but is not limited to, a molybdenum-based protective
layer 128 having a thickness from approximately 5-7 mm, with the
vessel of the reactor core section 102 having a wall thickness of
approximately 9-11 cm thick.
[0059] Similarly, as shown in FIG. 2B, the salt-facing surface of
the piping 114 of the primary coolant loop 112 (which may be the
internal and/or external surface of piping or other components)
includes a protective layer 128. For example, refractory alloy or
nickel alloy plating may be disposed on the salt-facing surface of
the piping 114 of the primary coolant loop 112 to form the
protective layer 128.
[0060] FIG. 3 illustrates an embodiment of a nuclear power plant
for generating power from a nuclear reaction using a molten fuel
salt, in this case a molten chloride fast reactor (MCFR). For a
power plant application, the heat generated by the MCFR will be
converted into electrical power by power conversion hardware. In
the embodiment shown, Rankine cycle power conversion hardware was
used with water (steam) as the working fluid. The conversion
efficiency of a Rankine cycle plant is in large part determined by
the temperature (and pressure) of the steam entering the turbines,
where higher temperatures correlate to higher efficiency.
Performance is coupled to steam pressure as well as temperature and
the highest efficiency Rankine cycle plants use supercritical and
ultra-supercritical steam. Although a Rankine cycle steam turbine
was used for illustration purposes, heat engines based on other
cycles are also feasible such as closed-cycle gas turbines (e.g.,
air, helium, or CO.sub.2) based on the Brayton cycle.
[0061] The power conversion system encompasses all systems that
come into contact with the power conversion system working fluid.
In the case of a steam Rankine cycle plant as illustrated, this
includes a steam generator 152, a turbine system 154, water
circulation loop 162 including one or more water circulation pumps
156 and a cooling tower 158, electrical generation equipment 160
and a control system 162. In addition, a fuel storage system 166
for storing new fuel salt and a reaction product storage system 168
to receive and safely contain used fuel salt are illustrated. As
illustrated in FIG. 3, the power conversion system starts with a
primary coolant transferring heat to the power cycle working fluid
through a heat exchanger (e.g. steam generator 152).
[0062] The power conversion system receives thermal power from the
reactor 100 and converts that heat into mechanical and then
electrical power. The analysis specifically focused on using
conventional steam Rankine cycle hardware for power conversion. The
analyzed configuration has three turbines, with a high pressure
turbine (HPT), intermediate pressure turbine (IPT), and low
pressure turbine (LPT), illustrated simply as the turbine system
154.
[0063] The model in FIG. 3 is simplified in that it shows only the
major components of the power plant. In the model used, the HPT
receives steam from the "main steam" generation system that is
heated by the primary cooling fluid carrying thermal energy from
the reactor. Exhaust from HPT is sent to the reheat steam
generation system, where the primary cooling fluid transfers heat
to the exhaust from the HPT, and that heated steam is fed to the
IPT. The exhaust from the IPT is fed to directly to the LPT to
extract additional enthalpy. There are often multiple turbines in
parallel, particularly for the LPT. In the model used, there are
twin LPTs that are used for the final expansion step. In the model,
all turbines are on a common shaft and direct coupled to an
electrical generator 160. The outlet of the LPT flows to a
condenser that cools the steam to near ambient temperature. For
this analysis, the LPT is assumed to be a once-through condenser
that receives heat from a large body of water, like a large lake or
river. After the condenser, the water is pumped and sent through
several feedwater heaters. The feedwater heaters preheat the
feedwater by mixing with steam extracted from various points on the
turbines. The preheated fluid from the feedwater heaters is then
fed to the steam generator, where it is heated to temperature for
the main turbine.
[0064] FIG. 4 illustrates another embodiment of a simplified
schematic view of a molten salt nuclear reactor 200. The reactor
200 is a pool-type reactor in which in some examples the fuel salt
108 may be flowing through the pool, such as through piping. In
other examples, the fuel salt 108 is circulating within the pool
but is contained within and is never removed from the core.
[0065] In the example shown in FIG. 4, the fuel salt is contained
in tubes 204 that are located at the center of a pool 210 of
coolant 202 in a closed reactor vessel 206. One or more tubes 204
that contain nuclear fuel 108 may be mechanically treated as unit
and referred to as a "fuel assembly". The top portion of the
reactor vessel 206 may be filled with some inert gas 218 such as
argon. The fuel tubes 204 are arranged in an array similar to
conventional solid fuel arrays in a light water reactor. The
coolant 202 transfers heat from the center of the pool 210 to heat
exchangers 208 located on the periphery of the pool 210. In the
embodiment shown, the circulation of the coolant 202 (illustrated
by the dashed arrows 212) within the pool 210, which may be natural
or induced by an impeller or other mechanism (not shown), convects
heat away from the fuel tubes 204 to be removed by the heat
exchangers 208.
[0066] The heat exchangers 208 transfer heat from the pool 210 to a
secondary coolant system 214. In an embodiment, the secondary
coolant is water that is boiled in the heat exchangers and the
resulting steam 216 is used to drive turbines (not shown) for the
generation of power.
[0067] An optional set of reflector modules 232 may be provided
around the array of fuel tubes either within the reactor vessel as
shown in FIG. 4, and/or external to the reactor vessel, to increase
the efficiency of the reactor. Optional shutdown rods may be
provided to maintain the reactor subcritical when needed.
[0068] Following its initial start-up with enriched (.about.12%
.sup.235U) fuel, an MCFR may not require the ongoing feed of
enriched fissile material. Instead, an MCFR can be fed depleted or
natural uranium, among other fertile materials. During normal
operations, modelling shows that the reactor slowly breeds up in
reactivity. To counter this increase in reactivity, a small
quantity of fully mixed fuel salt may be removed and replaced with
fertile feed salt. The addition of fertile material is, in effect,
a control rod that reduces reactivity.
[0069] Rather than going to disposal, used MCFR fuel can be
collected until a sufficient amount is available to start a new
reactor. Such a daughter reactor contains a chemically identical
fuel salt, and thus, is able to be started without any new
enrichment. By transferring used fuel, in total, to a daughter
plant for use as the initial fuel for that plant, creation of a
fleet of MCFRs significantly reduces the use of actinides and
defers the vast majority of radioactive waste until the end of
fleet build-out. For ultimate disposal of actinide-bearing fuel
salt, prior work found that the salt itself could be packaged,
without chemical conversion, into a suitable waste form.
[0070] In an embodiment, the reactor 200 in FIG. 4 is operated as a
traveling wave reactor (TWR). Such a traveling wave reactor is
disclosed in more detail in U.S. patent application Ser. No.
11/605,943 filed Nov. 28, 2006 in the names of Roderick A. Hyde, et
al. and titled "Automated Nuclear Power Reactor For Long-Term
Operation", which application is assigned to the assignee of the
present application, the entire disclosure of which is hereby
incorporated by reference.
[0071] In operation, the fuel in the TWR may remain solid, such
solid fuel often referred to as a fuel pin 204 because of its
cylindrical shape. In this embodiment, pins 204 may be in the form
of a solid cylinder of material, which may or may not be porous and
may or may not be enclosed in a tube. In an alternative embodiment,
the fuel is in the form of particulate material loosely packed and
contained by the tubes 204, thus the fuel pin 204 in this
embodiment includes the particulate fuel and tube 204.
[0072] In an embodiment of a TWR, fuel assemblies of one or more
pins may be moved during operation in order to maintain the
burnfront in a static location within the array of pins. For
example, as the fuel in a particular assembly in a central location
is breed up and then fissioned to the point that it is no longer
contributing to the overall criticality of the reactor, that
assembly may be moved to a location at the periphery of the tubes
and replaced with a fresh assembly. In this manner, the traveling
burnfront may be maintained at a static location within the
reactor, with spent assemblies being replaced by fresh assemblies
from within the reactor as needed to maintain criticality.
[0073] TWRs were generally described above in the Introduction. In
an embodiment, the TWR 200 has a cylindrical reactor core submerged
in a large sodium pool 202 in the reactor vessel 206. In an
embodiment, the reactor vessel and fuel assemblies are completely
surrounded on the sides and below by a contiguous containment
vessel (not shown) that does not have any penetrating components so
that in the event that sodium coolant leaks from the reactor
vessel, there can be no release of coolant to the environment
unless both the reactor vessel and the containment vessel are
breached.
[0074] In operation, pumps circulate the primary sodium coolant 202
within the pool so that it passes up, through the reactor core
toward the surface of the coolant pool. In the embodiment shown,
intermediate heat exchangers are submerged in the coolant pool
above the reactor core. Through the intermediate heat exchangers,
heat is transferred from the primary sodium coolant being
circulated in the pool to the intermediate sodium cooling system.
Heated intermediate sodium coolant is circulated to the steam
generators (not shown) that generate steam to drive turbine and
electrical generators. In an embodiment in which the primary sodium
coolant may be exposed to fission products, this design prevents
radioactive materials from being removed from the reactor vessel as
part of the cooling and electricity generation process. Reactor
containment is completed using an upper steel dome (not shown) that
engages with the containment vessel to completely enclose the
reactor core and reactor vessel.
[0075] During periods of reactor shut down, the plant electrical
loads are provided by the grid and decay heat removal is provided
by pony motors on the coolant pumps delivering reduced flow through
the heat transport systems. In the event that grid power is not
available, decay heat is removed using two dedicated safety class
decay heat removal systems: the Reactor Vessel Air Cooling System
(RVACS) and the Auxiliary Cooling System (ACS), which operate
entirely by natural circulation with no need for electrical
power.
[0076] In an embodiment, the TWR 200 may have a cylindrical
geometry composed of hexagonally shaped fuel pin bundles, or
assemblies. The assemblies contain a combination of enriched and
depleted uranium metal alloy fuel pins. In an embodiment, the fuel
pins are clad in ferritic-martensitic steel tubes.
[0077] In the center of the core are a few pins, or assemblies of
pins, containing enriched uranium (.sup.235U), surrounded by pins
or assemblies of depleted uranium (.sup.238U). The .sup.235U serves
as an igniter, kick starting the traveling wave reaction--a
slow-moving chain reaction of parallel waves of fission traveling
through the uranium rods. These parallel waves initiate in the
center of the core, slowly consuming the fuel and generating heat
in the core. In operation, the chain reaction creates a breed-burn
zone in the core that does not move through fixed core material.
Instead, a "standing" wave of breeding and burning is established
by periodically moving core material in and out of the breed-burn
region around the center of the core. This movement of fuel
assemblies is referred to as "fuel shuffling".
[0078] In the embodiment, metal fuel is used because it takes
advantage of the high heavy metal loadings and excellent neutron
economy, which allows an effective breed and burn process in TWRs.
The uranium metal may be alloyed with 5 to 8% zirconium to
dimensionally stabilize the alloy during irradiation and to inhibit
low-temperature eutectic and corrosion damage of the cladding. A
sodium thermal bond may be used to fill the gap that exists between
the uranium alloy fuel and the inner wall of the clad tube to allow
for fuel swelling and to provide efficient heat transfer which
keeps the fuel temperatures low. Individual fuel pins may have a
thin wire from 0.8 to about 1.6 mm diameter helically wrapped
around the circumference of the clad tubing to provide coolant
space and mechanical separation of individual pins within the
hexagonal fuel assembly housing that also serves as the coolant
duct. The cladding, wire wrap and housing may be fabricated from
ferritic-martensitic steel to take advantage of its superior
irradiation performance.
Treatment with Supercritical Fluids
[0079] The embodiments shown in FIGS. 5-10 will be discussed with
supercritical carbon dioxide (sCO.sub.2) as the supercritical
fluid. However, various other supercritical fluids, such as those
discussed above, can also be used, and the application of the
following embodiments is in no way limited to the application of
sCO.sub.2.
[0080] Supercritical carbon dioxide (sCO.sub.2) has been examined
for extraction on metals and metalloids from both aqueous and solid
solutions. Accordingly, sCO.sub.2 combined with various ionic
liquids (ILs) can be utilized as ligands to extract metal ions from
solutions. Similar methods may be used to extract metals or
metalloids from solid materials, such as contaminated paper,
fabrics, or even soils. Extraction of bulk materials requires the
material to be in an ionic form, such as a uranyl, lanthanide, or
actinide ions in solution. Thus, current bulk fissionable material
recycling techniques using sCO.sub.2 solutions require dissolution
of the bulk material into a solution. It may be possible to treat
used fuel (including nuclear fuels considered for molten-salt
reactors) with sCO.sub.2 in a manner which does not require
dissolution. As an example, metal fuel from a breed and burn
reactor, such as a TWR, can be treated with an sCO.sub.2 system
that does not dissolve the U metal but does remove selected fission
products (with high cross sections for parasitic absorption). A
sCO.sub.2 system may be capable of selectively removing these
elements and their corresponding isotopes. A list of elements
soluble in ILs is shown in Table III.
TABLE-US-00003 TABLE III Occurrence of selected elements in TWR
spent fuel and IL solubility. Fractional Fractional Element
Absorption Element Absorption Pd 2.38% Ru101 1.18% Ru 1.95% Pd105
1.13% Sm 1.25% Tc99 1.02% Mo 1.21% Rh103 1.02% Cs 1.16% Pd46 0.73%
Tc 1.02% Cs133 0.73% Rh 1.02% Mo97 0.45% Nd 0.85% Sm149 0.43% Xe
0.41% Ru102 0.41% Eu 0.30% Mo95 0.41%
[0081] For ILs, the sCO.sub.2 may be useful as a means of
introducing uranium into the IL. In other cases, it may be
appropriate to have direct dissolution of oxides into an IL.
[0082] Metals of interest to nuclear waste processing, such as
actinides, lanthanides, and transition metals, have been
characterized chemically using highly soluble fluorinated
(3-diketones in sCO.sub.2. Extraction can be accomplished by using
appropriate chelating agents as extractants. For example, La and Eu
extraction with greater than 90% effectiveness has been
demonstrated using fluorinated diketones combined with
tri-butylphosphate (TBP). In this process, a room temperature ionic
liquid, an imidazolium-based 1-butyl-3-methylimidazolium (BMIM)
with bis(trifluoromethylsulfonyl)-imide (also known as Tf2N--,
which is properly described as (CF3SO2)2N--) was used as a
complexing agent because of the complexing agent's ability to
solubilize CO.sub.2. In this manner, a full water/RTIL/sCO.sub.2
system is developed. A similar process with other ionic liquids and
metal chelating agents (extraction agents) was conducted and is
summarized in Table IV. Note that Eu and La are both extracted with
all systems except when using thenoyl tri-fluoroacetone (TTA)
without TBP. The latter only extracted La while not separating
(extracting) Eu.
[0083] Using sCO.sub.2 separation on target material and/or during
the post-processing of nuclear fuel, .beta.-diketones can be used
to selectively bind with oxides or metal in the presence of uranium
species. The extractions performed in Table IV were carried out
with the extractant/sCO.sub.2 mixture at 150 atm for one hour at
50.degree. C. Based on this information, it is anticipated that
.beta.-diketones can be used to selectively bind with radioisotope
oxides or metals while not substantially dissolving fissionable
material.
TABLE-US-00004 TABLE IV Degree of extraction (%) of EUIII and LaIII
from BMIMTf2N with different beta-diketones (with or without TBP).
Eu3+ La3+ HFA w/o TBP 90.5 90.4 HFA w/TBP 99.9 92.6 TTA w/o TBP --
87.1 TTA w/TBP 95.5 90.5 HFA = hexafluoroacetylacetone, TTA =
4,4,4-trifluoro-1-(2-thienyl)-1,3-butanedione
[0084] In general, an obstacle to CO.sub.2 solvation is low solvent
power of CO.sub.2 (non-polar). Metals and metal chelates have low
solubility in sCO.sub.2 with CO.sub.2 solubility parameters in the
range of 4-5 cal/cm.sup.3. This can be overcome by adding
CO.sub.2-philic functional groups such as fluoroethers,
fluoroacrylates, fluoroalkyls, silicones, and certain phosphazenes.
Fluorinated beta-diketones (with and without tributyl phosphate)
have been demonstrated in current techniques to extract a variety
of metals. Bis(trifluoroethyl) dithiocarbamate exhibits higher
solubility than non-fluorinated counterparts; 10-4 mol/L for
fluorinated vs. 10-6 to 10-7 mol/L for non-fluorinated. As another
example, Diethyldithiocarbamate (DDC) can be 3-800 times less
soluble in sCO.sub.2 at 100 atm than
bis(trifluoroethyl)dithiocarbamate (FDDC). Since sCO.sub.2 density
change is nearly linear with pressure, the solubility also changes
nearly linearly with solubility increasing with increasing
pressure. Lanthanides, actinides, copper, arsenic, and antimony
(and other products of irradiated targets) can have concentrations
on the order of lemon CO.sub.2. Water and soil extraction has been
demonstrated in current techniques with 1000-10000 molar ratio of
chelate to metal in solution.
[0085] A system for removing fission products from salt-based fuels
may be chemically similar to the process developed for metallic
fuels. This is because salts, by themselves, are insoluble in
sCO.sub.2. Extraction agents, such as diketones, may be used to
draw select metals into the sCO.sub.2 phase as described, above.
Physically, the clean-up system may be made to avoid pressurization
of the reactor vessel during a leak in the sCO.sub.2 clean-up
system. Additionally, the salts in their liquid states may be at
temperatures high enough to dissociate the diketones. To avoid both
of these obstacles, a system may be designed such that the
molten-salt is pumped external to the reactor vessel and injected
into a vessel containing the sCO.sub.2. The sCO.sub.2 system may be
maintained at a temperature low enough to solidify the molten-salt,
resulting in a high surface area solid. Provided the sCO.sub.2 can
be maintained at a sufficiently low temperature, the beta-diketones
or other appropriate separation agent may be co-mixed with the
sCO.sub.2 during salt injection, avoiding dissociation.
[0086] Alternatively, the separation agent may be injected into the
system in a batch-wise fashion following salt injection. In either
case, the result is a solution of (selected) metal-complexes
solvated in the sCO.sub.2 diketone solution. The solution may then
be pumped to a secondary system where temperature or pressure is
adjusted to remove the metal complexes (product) from the solution
without substantial effect on the target molten salt fuel. Again,
it is likely that the metal complex is removable form the target
solution without dropping the CO.sub.2 to a gaseous state (below
the critical point) via heating, cooling, or both. Heat may be used
to volatilize the metal complexes so that a separate gas phase
occurs within the sCO.sub.2 solution. The sCO.sub.2 may
alternatively be cooled or heated near and above the critical point
where its solubility typically changes significantly with changes
in temperature and pressure, resulting in a separate, liquid metal
complex phase which was forced out of solution due changes in
thermodynamic condition. This phase can then be transferred, such
as by way of pumping, from the extraction system to system designed
for interim or long term storage. Whether further heating or
cooling is used to separate the metal complex or other product,
ultimately further heating can be used to thermally decompose the
diketones, leaving behind the metal fission product(s).
[0087] FIG. 5 is a block flow diagram of an embodiment of an
example supercritical fluid treatment system 500. The example
system 500 includes a reactor 502 and supercritical fluid treatment
components 504, including fluid storage 506, supercritical fluid
container 508, supercritical fluid control 510, extractant(s)
storage 512, contact vessel 514, separation unit 516, transfer unit
518, and reprocessing/waste 520. Ancillary components, such as
pumps, valves, sensors, etc., are not shown. Other embodiments can
include more or fewer components.
[0088] The reactor 502 is a nuclear fuel reactor. For example,
reactor 502 is a fuel salt reactor, a containerized fuel salt
reactor, or a traveling wave reactor, as those reactors, their
contents, and methods of operation, are described above. Reactor
502 can be a different type of reactor in other examples.
[0089] The supercritical fluid separation components 504 bring a
supercritical fluid into contact with irradiated nuclear fuel
and/or the reactor 502. Generally, supercritical fluids have
properties that can permit selective separation of produced
radionuclides from a target material, such as uranium, and/or
remove residues containing fission products from equipment in or
related to the reactor 502. Supercritical fluids retain both gas
and liquid properties. They can have viscosities that resemble gas
and diffusion properties between a gas and liquid. Supercritical
fluids can effectively penetrate solid materials. Various
supercritical fluids can be used for material processing and
chemical reactions owing to their unique physico-chemical
properties. Examples of supercritical fluids and some of their
properties are discussed above.
[0090] Fluid CO.sub.2 is stored in fluid storage 506. There, the
fluid CO.sub.2 is stored at a temperature and pressure such that
the CO.sub.2 is in non-supercritical form. Fluid storage 506 can be
a gas reservoir that is in fluid communication with the sCO.sub.2
container 508. However, as noted below with reference to separation
unit 516, economics and other factors might dictate that sCO.sub.2
is not transitioned to the gas phase, in which case separation
components 504 does not include fluid storage 506.
[0091] The sCO.sub.2 container 508 is pressurized and has a
temperature such that the CO.sub.2 is in supercritical form. For
example, sCO.sub.2 container 508 operates at a temperature greater
than 32.degree. C. and at a pressure greater than 73 atm.
[0092] The supercritical fluids used in system 500 can be combined
with specific extractants to selectively remove or separate
radionuclides. These extractants are stored in extractant storage
512 and mixed with the sCO.sub.2. Example extractants are discussed
above, such as, for example, cupferron, chloroanillic acid,
.beta.-diketone, N-benzoyl-N-phenylhydroxylamine,
.alpha.-dioximines diaminobenzidine, a porphyrine such as porphine,
8-hydroxyquinoline, nitrosonapthols, nitrosophenols,
ethylenediaminetetraacetic acid, diphenylcarbazide,
diphenylcarbazone, Azoazoxy BN, sodium diethlydithiocarbamate,
dithizone, bismuthiol II, thiothenoyltrifluoracetone, thioxine,
thiophosphinic acid, phosphine sulfide, phosphorothioic acid, and
tributylphoshpate. In some embodiments, sCO.sub.2 is used without
being mixed with any extractants.
[0093] As shown in FIG. 5, the extractants are mixed with the
sCO.sub.2 within the supercritical fluid container 508, at a
supercritical fluid control 510, or at some point before the
sCO.sub.2 contacts the nuclear fuel. For example, the sCO.sub.2 is
passed through a column containing the extractants to dissolve the
extractants into the sCO.sub.2 stream. If the selected extractant
is not particularly soluble in sCO.sub.2, this operation may also
include modifying the extractant to make it more soluble, such as
by adding CO.sub.2-philic functional groups such as fluoroethers,
fluoroacrylates, fluoroalkyls, silicones, and certain phosphazenes
to a selected ligand. In an embodiment, the extractant may be a
fluorinated .beta.-diketone and a trialkyl phosphate, or a
fluorinated .beta.-diketone and a trialkylphosphine oxide. In
another embodiment, the ligand may be selected from
dithiocarbamates, thiocarbazones, .beta.-diketones and crown
ethers.
[0094] The sCO.sub.2 and irradiated nuclear fuel from reactor 502
are introduced to the contact vessel 514. In embodiments, the
sCO.sub.2 is introduced directly into the reactor 502. The nuclear
fuel includes a plurality of radioisotopes and radionuclides, such
as, for example, .sup.99Mo, .sup.238U, .sup.131I, .sup.51Cr,
.sup.225Ra, and .sup.225Ac.
[0095] In embodiments, irradiated nuclear fuel is injected into the
contact vessel 514 via a fuel salt injector. The fuel salt injector
includes one or more nozzles that disperse the molten fuel salt
into a spray, mist, or fog of droplets. The contact vessel 514 can
be operated as a batch or continuous process, or a combination of
both.
[0096] Also, the contact vessel 514 can include an environmental
control system. The environmental control system is capable of
monitoring and regulating the temperature and pressure within the
contact vessel 514, as well as the flow rates of the sCO.sub.2 and
nuclear fuel. In embodiments, the environmental control system
includes a pressure sensor, a temperature sensor, a heater, such as
a heating jacket, and a heat exchanger. The environmental control
system can also include the fuel salt injector, thereby controlling
the flow rate of nuclear fuel into the contact vessel and a
supercritical fluid injection valve that controls the flow rate of
supercritical fluid into the contact vessel 514. In embodiments,
the environmental control system can maintain the contact vessel
514 at a temperature and pressure that causes the dispersed molten
fuel salt to solidify into fuel salt particles.
[0097] The environmental control system can also regulate the flow
of supercritical fluid and nuclear fuel out of the contact vessel
514. For example, one or more extraction valves, controlled by the
environmental control system, can regulate the flow rate of
supercritical fluid out of the contact vessel 514. In embodiments,
the fluid flow out of the contact vessel is a combination of
sCO.sub.2 and nuclear fuel. In other embodiments, the nuclear fuel
exits contact vessel 514 in one stream (and is routed to transfer
unit 518), and the sCO.sub.2 is routed to separation unit 516.
[0098] After contacting the nuclear fuel and/or reactor 502
components, the sCO.sub.2 will have one or more fission products
dissolved therein. Separation unit 516 separates the sCO.sub.2 from
the chelate and/or waste, and in embodiments, from the nuclear
material. The sCO.sub.2 exits the separation unit 516 and is routed
back to fluid storage 506 or supercritical fluid container 508. The
nuclear material exits the separation unit 516 and is routed to the
transfer unit 518, and the chelate and/or waste exits the
separation unit 516 and is routed to reprocessing/waste 520.
[0099] It may be impractical to transition sCO.sub.2 to the gas
phase in separation unit 516 economically because that transition
would require either recompression of the CO.sub.2 to the
supercritical state or a steady supply of high pressure CO.sub.2.
Additionally, there is a potential safety risk inherent to
confining a high pressure solution of a highly compressible fluid.
Furthermore, the off-gas CO.sub.2 would need to be collected in a
container capable of further decontamination or disposal, because
some residual radioactive materials or decay products might remain
in the carbon dioxide gas.
[0100] In embodiments, the separation unit 516 includes a `back
extraction` process which does not require gasification of the
sCO.sub.2 as part of the separation of the radioisotopes from the
sCO.sub.2. In this type of process, metal or metalloid species are
removed from solid or liquid solutions by using supercritical
fluids to form a metal or metalloid chelate. The supercritical
fluid will typically contain a solvent extractant, such as a few
percent H.sub.2O or MeOH. The metals or metalloids are then
back-extracted from the sCO.sub.2 solution by using an acidic
solution, one of which may be halogenated. By back extracting to
another (aqueous) solution, decompression of the sCO.sub.2 is
avoided. What is left is the other solution bearing the selected
radioisotopes, which is routed to reprocessing/waste 520, and the
sCO.sub.2 that can be readily reused by routing it to the
supercritical fluid container 508.
[0101] Using a back extraction process can be advantageous in an
automated system and in a continuous treatment, although even in a
semi-automated, batch treatment system the ability to recycle
sCO.sub.2 without the added step of repressurization would be
cost-advantageous. Back extraction may, or may not, remove the
extractant with the radioisotope. In an embodiment, fresh
extractant may need to be added to the sCO.sub.2 before it can be
reused as an extraction compound. It should be noted that ILs could
also be used for the back extraction process.
[0102] Alternatively, in separation unit 516 the solution removed
from the contact vessel 514, containing the sCO.sub.2, elements and
isotopes removed from the used fuel, can be brought to below the
critical point and converted to the gaseous phase. This conversion
leaves behind the extractant ligand and the separated elements or
isotopes. The extractant can then be brought to above its
volatilization temperature and converted to a vapor phase, leaving
behind the selected element or isotopes. Variations of this scheme
may be used as appropriate. For example, lowering the solution to
below the liquidus point of the carbon dioxide may be preferred if
the chosen extractant and liquid CO.sub.2 are insoluble.
[0103] Another alternative is to, in separation unit 516, raise the
temperature of the supercritical solution to above the
volatilization point of the extractant (e.g. greater than
100.degree. C. to 200.degree. C.) or to above the decomposition
temperature (e.g. greater than 200.degree. C. to 300.degree. C.).
In either case, the metal may substantially or partially
precipitate from the sCO.sub.2 once the extractant is lost. Removal
of the extractant vapor or decomposition product can be
accomplished by a gas phase separation or, as above, by converting
the CO.sub.2 to a liquid phase. Furthermore, the solution may
change temperature or pressure from a supercritical condition to a
second supercritical condition, the second condition having a
solubility of the extractant lower than the solubility of the first
condition. By this process, all or a portion of the extractant may
be recovered without leaving the supercritical state.
[0104] Current techniques use sCO.sub.2-ionic liquid processes to
remove both lanthanides and actinides from aqueous solutions, and
some current modified versions have been proposed. One such process
is called super-DIREX, which is short for `supercritical direct
extraction`. The super-DIREX process is expected to minimize the
cost of reprocessing because the heavy metals (U, Pu, Np, Am and
Cm) are directly extracted from a spent fuel powder in a column
covering the dissolution. One experiment claimed up to 30%
reduction in waste stream volumes if sCO.sub.2 methods were
utilized. However, other work has cited other critical issues, such
as counteracting the acidity of TBP-HNO.sub.3 solutions and the
build-up of .sup.14C.
[0105] Metallic fuel, including those metal fuels appropriate for
vented pin configurations and/or a traveling wave reactor,
typically includes metal fuel capable of high burn-up contained
within vented, ferritic martensitic stainless steel cladding. At
the end of life, the fuel generally has a highly porous matrix of
metallic form fuel and solid fission products which precipitated
from the fuel during the burn cycle.
[0106] The transfer unit 518 is an optional component that
circulates the liquid nuclear fuel between the reactor 502 and the
contact vessel 514. The transfer unit 518 can include one or more
pumps, valves, sensors, and flow meters. Additionally, the transfer
unit 518 can include a holding vessel for holding the nuclear
material and raising the temperature and/or pressure of the nuclear
material.
[0107] As shown, the transfer unit 518 directs nuclear material
back to the reactor 502. Alternatively, some or all of the nuclear
material can be directed to waste processing. Another alternative
is to re-use the nuclear material, once the fissions products are
removed in contact vessel 514 and/or separation unit 516, in
another facility similar to or different from the reactor 502. For
example, the reactor 502 is a breed and burn type reactor such as a
TWR. It may be practical to remove the fission products and then
perform a thermo-mechanical treatment within the contact vessel 514
and/or separation unit 516 used for solvation in order to modify
the structural material for continued in-reactor use. Once the
fission products are removed, the nuclear material may be brought
to significantly higher temperatures (which could be made to exceed
the fuel melting point) and pressures (10's of MPa's).
[0108] Reprocessing/waste system 520 includes one or more processes
and/or storage facilities. As shown, the extractants, such as
chelates and ligands, as well as fission products, are routed to
reprocessing/waste system 520. Through one or more processes, the
ligands can be separated from the fission products. In some
instances, the separated ligands can be reused in the system
500.
[0109] FIG. 6 is an embodiment of a method 600 for treating a fuel
salt reactor with a supercritical fluid treatment system. The
example method 600 includes generating a supercritical fluid
(operation 602), introducing an extractant into the supercritical
fluid (operation 604), contacting a volume of the supercritical
fluid with a molten fuel salt (operation 606), separating the
supercritical fluid from the fuel salt (operation 608), directing
contacted fuel salt to a reactor core (operation 610), and
separating fission products form the supercritical fluid (operation
612). The supercritical fluid treatment system 504 shown and
described with reference to FIG. 5, and any one of the reactors
described above with reference to FIGS. 1-5, can be used to
implement example method 600. Other embodiments can include more or
fewer operations.
[0110] The example method 600 begins by generating a supercritical
fluid in a fluid providing operation 602. As discussed above, the
supercritical fluid can be supercritical carbon dioxide (sCO.sub.2)
which can be generated at using known techniques. Once the
supercritical fluid is generated, an extractant is introduced into
the supercritical fluid such that the ligand is dissolved into the
supercritical fluid in an extractant adding operation. As discussed
above the extractant adding operation may include modifying or
preparing the selected extractant for effective use with the chosen
supercritical fluid.
[0111] Next, a volume of irradiated, molten fuel salt is contacted
with a volume of the supercritical fluid in a contacting operation
606. This contacting can occur within the reactor core or in a
separate contact vessel, as described above with reference to FIG.
5.
[0112] During contact, the supercritical fluid removes one or more
types of fission products from the irradiated fuel salt. The
extractant in the supercritical fluid forms a complex with one or
more fission products, resulting in a sCO.sub.2-radioisotope
complex solution.
[0113] The supercritical fluid is then separated from the fuel salt
in a separation operation 608. In embodiments, the contacted fuel
salt is directed back to the reactor core in a fuel return
operation 610. As discussed above with reference to FIG. 5, the
temperature and pressure of the fuel salt may need to be increased
before re-introduction into the reactor core.
[0114] The separated supercritical fluid is also subjected to one
or more separation processes to extract the supercritical fluid
from the fission products in a separation operation 612. Techniques
and conditions for this separation are discussed above with
reference to FIG. 5.
[0115] FIG. 7 is a block flow diagram of the example fuel salt
reactor 100 used with the example supercritical fluid treatment
components 504. The example fuel salt reactor 100 is shown and
described above with reference to at least FIGS. 1-3. The example
supercritical fluid treatment components are also shown and
described above with reference to FIG. 5. Other embodiments can
include more or fewer components.
[0116] In the embodiment shown, molten fuel salt is fed into the
contact vessel 514 from reactor 100. The fuel salt returns to the
reactor 100 via transfer unit 518 and the supercritical fluid,
including the chelates in solution, are sent to the separation unit
516 for additional processing.
[0117] Alternatively, the supercritical fluid is fed directly into
the reactor 100 during reactor operation. In another embodiment,
the supercritical fluid is fed into the reactor 100 during a
reactor shut-down period where little to no fuel salt is
present.
[0118] FIG. 8 is a block flow diagram of the example containerized
fuel salt reactor 400 used with the example supercritical fluid
treatment components 504. The example fuel salt reactor 400 is
shown and described above with reference to at least FIGS. 1-4. The
example supercritical fluid treatment components are also shown and
described above with reference to FIG. 5. Other embodiments can
include more or fewer components.
[0119] The reactor 400 includes one or more fuel salt containers
490, each of which containing fuel salt that includes at least some
fissionable material. The containers 490 may be storage containers,
or, alternatively may be containers for use within the reactor such
as fuel tubes or assemblies.
[0120] In the embodiment shown, supercritical fluid is directed
into one or more fuel containers 490. The containers 490 may be
located in the reactor core or may be removed for treatment. After
treatment, the containers 490 may be returned to the reactor core
or stored. The containers 490 may hold a fuel salt which may be
molten or solid, or may hold another nuclear fuel such as loose
particulate fuel or porous matrix fuel as described with reference
to the TWR above. The flow rate of supercritical fluid into the
fuel container 490 is regulated by the supercritical fluid control
510. Inside the fuel container 490, the supercritical fluid
contacts the fuel. As discussed above, one or more fission products
are extracted by the supercritical fluid during this contact by
dissolution. Then, the supercritical fluid exits the fuel container
490 and is directed to the separation unit 516. In embodiments, the
flow rate of supercritical fluid out of the fuel container 490 is
regulated by a controller in communication with a sensor and
valve.
[0121] In embodiments, the supercritical fluid is used to prepare
the fuel container 490 for shut-down. Removing fission products
from the fuel container 490 may greatly enhance the disposability
of the fuel container 490, as >90% of targeted fission products
may be removed using sCO.sub.2 treatment, and even more can be
removed with multiple sCO.sub.2 solution treatments. In some cases,
it may be advantageous to apply multiple cycles such as repeated
treatments or multiple different treatments to increase the removal
of fission products. For example, in some cases, two treatments
could give 99% removal of accessible fission products. Three would
give 99.9% and so forth.
[0122] Any appropriate factors may be used to determine the number
and/or type of processing treatments and may be based on fission
products dissolved or stuck inside the solid fuel matrix where
sCO.sub.2 solution cannot penetrate. It should be noted, however,
that it may be possible to operate at temperature and timescales
which would allow for diffusion of solution soluble metals out of
the bulk fuel matrix and into solution. This may lower the short
term heat load of the spent fuel assembly, decrease the dangers of
handling and transporting the assembly, and make it more suitable
for long-term disposal.
[0123] FIG. 9 is a block flow diagram of the example TWR 900 used
with the example supercritical fluid treatment components of FIG.
5. The example TWR 900 is shown and described above with reference
to at least FIG. 4. The example supercritical fluid treatment
components are also shown and described above with reference to
FIG. 5. Other embodiments can include more or fewer components.
[0124] The TWR 900 includes a reactor core and a reactor vessel
containing a primary sodium coolant. The reactor core is submerged
with the primary sodium coolant. The TWR 900 also includes at least
one assembly 904 in the reactor core that includes one or more
solid fuel pins that contain fissionable material and fission
products. Also, the TWR 900 includes an assembly shuffling system
that is configured to move the assemblies between various positions
within the reactor core.
[0125] The supercritical fluid can contact nuclear fuel from the
TWR 900 in a contact vessel 514, as shown, where the nuclear fuel
is directed away from the reactor core. Alternatively, the
supercritical fluid can be contacted with an assembly when the
shuffling system has removed, temporarily, the assembly from the
reactor core. Still another alternative is that the supercritical
fluid can contact nuclear fuel directly within the reactor
core.
[0126] The supercritical fluid control 510 can control the transfer
of supercritical fluid to the TWR 900 into the fuel assembly 904
and/or contact vessel 514. The control can be based on the
expansion of the nuclear fuel and/or based on the concentration of
fission products in a coolant used in the TWR 900.
[0127] In embodiments, the TWR 900 includes a transfer vessel that
is configured to hold a fuel assembly 904 in argon. The
supercritical fluid control 510 can direct supercritical fluid to
contact a fuel assembly 904 that is held in argon, not shown in
FIG. 9. There, the supercritical fluid can remove fission products
from the argon that is exposed to the fuel assembly 904. The
supercritical fluid, now a mixture with fission products, is
returned to the separation unit 516 as discussed above.
[0128] In embodiments, the TWR 900 includes a coolant cleaning
system that includes an absorber 902. The absorber, such as a
packed bed or adsorption membrane, removes fission products from
the primary sodium coolant used in the TWR 900. The supercritical
fluid control 510 can direct supercritical fluid to the absorber
902 to dissolve and remove fission products from the absorber 902.
The supercritical fluid, now a mixture with fission products, is
returned to the separation unit 516 as discussed above.
[0129] A system using supercritical fluid treatments may remove
fission products prior to the end of life by incorporating the
separation process such as sCO.sub.2 process into the fuel
management or `shuffling` cycle to remove fission products
periodically during irradiation (operation of the reactor). For
example, a TWR re-fueling system may incorporate a sealed enclosure
for raising the assembly out of the coolant pool. The TWR
containment enclosure may be provided with sufficient cooling
capability to manage assembly decay heat during the treatment
process while the assembly 904 is out of the coolant pool. In an
alternative embodiment, the system may be designed that one or more
assembly positions around the periphery of the core are treatment
positions where the assemblies may be connected to the treatment
system 504. The system may be made more robust such that fission
products may be removed, in containment, with minimal system
modifications. Such a system would not require large vessels and
piping, due to the high density of sCO.sub.2. Concentrations of
greater than 10-4 kg metal/kg solution are possible. At end of
life, each assembly contains the maximum amount of fission
products, on the order of 50 kg. The solution density is on the
order of 1000 kg/m3. Therefore only 5 m3 of sCO.sub.2 solution
would be needed in some cases to contain all the fission products
in a single assembly. Treating the assembly at more frequent
intervals would obviously reduce this maximum volume. Furthermore,
since the CO.sub.2 may be separated from the fission products and
re-entered into the system, the inventory can be additionally
reduced.
[0130] FIG. 10 is an embodiment of a method 700 for operating a
reactor with supercritical fluid separation. The example method 700
includes charging the reactor core (operation 702), maintaining a
chain reaction (operation 704), contacting a volume of reacted fuel
with a supercritical fluid (operation 706), initiating a chain
reaction in regenerated fuel (operation 708), and separating
fission products from the supercritical fluid (operation 710). The
supercritical fluid treatment system 504 shown and described with
reference to FIG. 5, and the reactor 100 described above at least
with reference to FIGS. 1-3, can be used to implement example
method 700. Other embodiments can include more or fewer
operations.
[0131] The example method 700 begins by charging the reactor core
with nuclear fuel in a charging operation 702. The charging
operation 702 varies depending upon the type of nuclear reactor
employed in the example method 700. Examples of charging a molten
salt reactor, a traveling wave reactor, and a containerized molten
salt reactor are described above.
[0132] After charging the reactor core in charging operation 702, a
chain reaction is maintained within the reactor core at or above
criticality in a first fission operation 704. Maintaining the chain
reaction varies depending upon the type of nuclear reactor employed
in the example method 700.
[0133] At some point after the chain reaction is started and
maintained, a volume of supercritical fluid is contacted with a
volume of irradiated nuclear fuel. As discussed above,
supercritical carbon dioxide (sCO.sub.2) can be used as a
supercritical fluid. Contacting the supercritical fluid and nuclear
fuel varies depending on the specific type of reactor used, with
some examples discussed above.
[0134] As an example, and as discussed above, the supercritical
fluid can be contacted with at least some of the partially-reacted
nuclear fuel without removing the nuclear fuel from the reactor
core. This introduction of supercritical fluid can be performed
without interrupting one or more chain reactions within the reactor
core.
[0135] As another example, and as discussed above, some of the
partially-reacted fuel can be removed from the reactor core and
then contacted with the supercritical fluid, for example in a
contact vessel. The removal of the partially-reacted fuel can be
performed without interrupting one or more chain reactions within
the reactor core. After contacting with the supercritical fluid,
the partially-reacted fuel can be returned to the reactor core
without interrupting the one or more chain reactions within the
reactor core.
[0136] As discussed above with reference to the TWR, the nuclear
fuel can be contained within one or more fuel pins or pin
assemblies. The pins can be moved within the reactor core to
different positions throughout operation. During or after movement
of a pin, the supercritical fluid can be contacted with the fuel
within the pin.
[0137] After the contacting the supercritical fluid and
partially-reacted nuclear fuel in contacting operation 706, a chain
reaction is initiated in the now-treated nuclear fuel in a second
fission operation 708. As discussed above, in embodiments the
partially reacted nuclear fuel can be contacted with supercritical
fluid within the reactor. In those embodiments, the reactor
operation can continue as before. In other embodiments, the
partially-reacted nuclear fuel is directed out of the reactor core
and then is contacted with the supercritical fluid. In those
embodiments, the partially-reacted nuclear fuel can be redirected
back to the same reactor core or directed to a different, perhaps
newly starting, reactor core.
[0138] The method 700 also includes separating the fission products
from the supercritical fluid in a separation operation 710.
Embodiments of systems and methods for separating fission products
from the supercritical fluid are discussed at least with reference
to FIG. 5.
[0139] Yet another possible application of supercritical fluids is
reformation of spent or previously-irradiated fuel. Reformation of
fuel after irradiation generally may be designed to allow treatment
of the entire fuel assembly for fission product, lanthanide, or
actinide removal treatments without modification of the nuclear
fuel assembly or fuel pins contained within. In one example of a
sealed vessel and supercritical fluid treatment, a previously
burned nuclear fuel assembly may be placed into a sealable pressure
vessel. The vessel may then be filled with pressurized sCO.sub.2
and one or more extraction agent (such as diketones, or any other
appropriate agent) is added to create an extracting solution in the
absence of an IL or aqueous component. Because of the presence of a
vent in the existing fuel assembly for fission gas venting, and the
nature of supercritical fluids, the sCO.sub.2-extractant solution
will work to fill the fuel pin and the matrix of porous fuel (i.e.
supercritical fluids behave as low surface tension, low viscosity
fluids which fill the volume they are contained within). The
sCO.sub.2 solution will begin to solvate targeted fission products
(or other materials, if so desired and a proper ligand chosen),
leaving the uranium metal matrix unaffected. The fission products
will then begin to diffuse out of the fuel pin such that the
concentration of the overall system tends toward equilibrium. The
solution containing the dissolved metal can then be slowly released
from the pressure vessel. New, clean solution may be re-added to
the pressure vessel. Agitation, heat and/or continued
pressurization and depressurization may be applied to the system to
enhance the solvation rate. For example, the system may operate at
greater than 7.5 MPa (approximate critical point at 51.degree. C.)
and be oscillated by +/-0.1 MPa to enhance `pumping` of sCO.sub.2
solution in and out of the porous fuel.
CONCLUSION
[0140] While particular aspects of the present subject matter
described herein have been shown and described, it will be apparent
to those skilled in the art that, based upon the teachings herein,
changes and modifications may be made without departing from the
subject matter described herein and its broader aspects and,
therefore, the appended claims are to encompass within their scope
all such changes and modifications as are within the true spirit
and scope of the subject matter described herein. It will be
understood by those within the art that, in general, terms used
herein, and especially in the appended claims (e.g., bodies of the
appended claims) are generally intended as "open" terms (e.g., the
term "including" should be interpreted as "including but not
limited to," the term "having" should be interpreted as "having at
least," the term "includes" should be interpreted as "includes but
is not limited to," etc.). It will be further understood by those
within the art that if a specific number of an introduced claim
recitation is intended, such an intent will be explicitly recited
in the claim, and in the absence of such recitation no such intent
is present.
[0141] For example, as an aid to understanding, the following
appended claims may contain usage of the introductory phrases "at
least one" and "one or more" to introduce claim recitations.
However, the use of such phrases should not be construed to imply
that the introduction of a claim recitation by the indefinite
articles "a" or "an" limits any particular claim containing such
introduced claim recitation to claims containing only one such
recitation, even when the same claim includes the introductory
phrases "one or more" or "at least one" and indefinite articles
such as "a" or "an" (e.g., "a" and/or "an" should typically be
interpreted to mean "at least one" or "one or more"); the same
holds true for the use of definite articles used to introduce claim
recitations. In addition, even if a specific number of an
introduced claim recitation is explicitly recited, those skilled in
the art will recognize that such recitation should typically be
interpreted to mean at least the recited number (e.g., the bare
recitation of "two recitations," without other modifiers, typically
means at least two recitations, or two or more recitations).
[0142] Unless otherwise indicated, all numbers expressing
quantities of ingredients, properties such as molecular weight,
reaction conditions, and so forth used in the specification and
claims are to be understood as being modified in all instances by
the term "about." The term "about" is not intended to either expand
or limit the degree of equivalents which may otherwise be afforded
a particular value. Further, unless otherwise stated, the term
"about" shall expressly include "exactly," consistent with the
discussions regarding ranges and numerical data. The term "about"
in the context of the present disclosure means a value within 15%
(.+-.15%) of the value recited immediately after the term "about,"
including any numeric value within this range, the value equal to
the upper limit (i.e., +15%) and the value equal to the lower limit
(i.e., -15%) of this range. For example, the value "100"
encompasses any numeric value that is between 85 and 115, including
85 and 115 (with the exception of "100%", which always has an upper
limit of 100%).
[0143] Concentrations, amounts, and other numerical data may be
expressed or presented herein in a range format. It is to be
understood that such a range format is used merely for convenience
and brevity and thus should be interpreted flexibly to include not
only the numerical values explicitly recited as the limits of the
range, but also to include all the individual numerical values or
sub-ranges encompassed within that range as if each numerical value
and sub-range is explicitly recited. As an illustration, a
numerical range of "4% to 7%" should be interpreted to include not
only the explicitly recited values of about 4 percent to about 7
percent, but also include individual values and sub-ranges within
the indicated range. Thus, included in this numerical range are
individual values such as 4.5, 5.25 and 6 and sub-ranges such as
from 4-5, from 5-7, and from 5.5-6.5; etc. This same principle
applies to ranges reciting only one numerical value. Furthermore,
such an interpretation should apply regardless of the breadth of
the range or the characteristics being described.
[0144] Notwithstanding that the numerical ranges and parameters
setting forth the broad scope of the invention are approximations,
the numerical values set forth in the specific examples are
reported as precisely as possible. Any numerical value, however,
inherently contain certain errors necessarily resulting from the
standard deviation found in their respective testing
measurements.
[0145] Furthermore, in those instances where a convention analogous
to "at least one of A, B, and C, etc." is used, in general such a
construction is intended in the sense one having skill in the art
would understand the convention (e.g., "a system having at least
one of A, B, and C" would include but not be limited to systems
that have A alone, B alone, C alone, A and B together, A and C
together, B and C together, and/or A, B, and C together, etc.). In
those instances where a convention analogous to "at least one of A,
B, or C, etc." is used, in general such a construction is intended
in the sense one having skill in the art would understand the
convention (e.g., "a system having at least one of A, B, or C"
would include but not be limited to systems that have A alone, B
alone, C alone, A and B together, A and C together, B and C
together, and/or A, B, and C together, etc.). It will be further
understood by those within the art that typically a disjunctive
word and/or phrase presenting two or more alternative terms,
whether in the description, claims, or drawings, should be
understood to contemplate the possibilities of including one of the
terms, either of the terms, or both terms unless context dictates
otherwise. For example, the phrase "A or B" will be typically
understood to include the possibilities of "A" or "B" or "A and
B."
[0146] In some instances, one or more components may be referred to
herein as "configured to," "configurable to," "operable/operative
to," "adapted/adaptable," "able to," "conformable/conformed to,"
etc. Those skilled in the art will recognize that such terms (e.g.,
"configured to") can generally encompass active-state components
and/or inactive-state components and/or standby-state components,
unless context requires otherwise.
[0147] With respect to the appended claims, those skilled in the
art will appreciate that recited operations therein may generally
be performed in any order. Also, although various operational flows
are presented in a sequence(s), it should be understood that the
various operations may be performed in other orders than those
which are illustrated, or may be performed concurrently. Examples
of such alternate orderings may include overlapping, interleaved,
interrupted, reordered, incremental, preparatory, supplemental,
simultaneous, reverse, or other variant orderings, unless context
dictates otherwise. Furthermore, terms like "responsive to,"
"related to," or other past-tense adjectives are generally not
intended to exclude such variants, unless context dictates
otherwise.
[0148] It will be clear that the systems and methods described
herein are well adapted to attain the ends and advantages mentioned
as well as those inherent therein. Those skilled in the art will
recognize that the methods and systems within this specification
may be implemented in many manners and as such is not to be limited
by the foregoing exemplified embodiments and examples. In this
regard, any number of the features of the different embodiments
described herein may be combined into one single embodiment and
alternate embodiments having fewer than or more than all of the
features herein described are possible.
[0149] While various embodiments have been described for purposes
of this disclosure, various changes and modifications may be made
which are well within the scope of the technology described herein.
For example, although not explicitly stated Raman spectroscopy may
be but one of many techniques used to monitor fuel salt quality
during operation of a molten salt reactor and, likewise, multiple
Raman probes may be used in order to get an understanding of the
variations in fuel salt quality at different locations within the
reactor. Numerous other changes may be made which will readily
suggest themselves to those skilled in the art and which are
encompassed in the spirit of the disclosure and as defined in the
appended claims.
* * * * *