This web page describes an apparatus for concentrating the TRU in used nuclear power reactor fuel about 10X by extracting nearly pure UO2 from used fuel. This apparatus uses a closed recrystalization process to separate the used nuclear fuel bundle atoms into three molecular groups: pure UO2, Zr and everything else (UO2, TRU oxides, fission products).
THE FUEL PORTIONS:
The economic strategy is to minimize the used fuel reprocessing, shipping and storage costs by performing relatively simple preliminary operations on each power reactor site. These preliminary operations include:
a) Chopping up the used fuel;
b) Prebaking the chopped fuel to reject vapors and gases (Cs oxides + Kr + Ar + Xe);
c) Harvesting the used Zr hulls;
d) Dividing the remaining used fuel atoms into one large minimally radioactive UO2 portion (90%) and one smaller highly radioactive portion (UO2 + FP + TRU).
TRU is an acronym for TRans Uranium actinides, atoms with atomic numbers greater than 92 that are formed when the abundant uranium isotope U-238 is exposed to a thermal neutron flux. Typically in used CANDU reactor fuel about (1 / 2) of the TRU is Pu-239 and about (1 / 6) of the TRU is Pu-240.
TRU concentration is a physical process that selectively extracts uranium oxide from used CANDU reactor fuel, resulting in about a 10 fold increase in the TRU fraction in the remaining residue. The TRU concentration process is of fundamental economic importance in economic production of fuel for Fast Neutron Reactors (FNRs).
The motive for TRU Concentration is to mimimize the overall cost of producing initial FNR core fuel. This cost reduction is achieved by a reducing by 10 fold the used reactor fuel mass flow through the subsequent pyroprocess. This mass flow reduction realized by selective uranium oxide extraction which:
1) Increases the TRU/U ratio in the fuel concentrates as required for economic production of FNR core fuel;
2) Minimizes the mass of CANDU fuel concentrates that must be transported to a remote fuel reprocessing site;
3) Reduces the gamma emission from the extracted uranium oxide sufficiently to make that uranium oxide more economic to transport and store for future use as the main component of FNR blanket fuel.
4) Enables complete disposal of TRU from used water cooled reactor fuel, thus eliminating the necessity of DGRs for nuclear fuel waste disposal.
This used reactor fuel concentration process is of interest to parties that:
1) Have an inventory of used CANDU reactor fuel and who would like to use that used fuel to make FNR fuel;
2) Have an inventory of used nuclear fuel from water cooled reactors and who would like to mitigate the costs of transporting, storing and/or reprocessing of that used fuel;
3) Have an inventory of used nuclear fuel from water cooled reactors and who would like to rapidly convert that used fuel into stable elements that pose no risk to future human generations;
4) Seek to prevent nuclear weapon proliferation or further nuclear waste formation via use of closed system electrolytic fuel reprocessing but who want to minimize the overall cost.
About 2010 Peter Ottensmeyer realized that there is a way of converting used CANDU reactor fuel into FNR fuel with no spurious waste streams. The method, termed the Ottensmeyer Plan, has two major steps. The first major step involves cutting up the CANDU fuel to expose the UO2 pellet material and then baking the chopped up used fuel at 650 deg C in a vacuum oven to drive off Cs, other volitles and the inert gas fission products (FP) which must be captured.
The vacuum baking is followed by molecular component separation using a uranyl nitrate hexahydrate recrystallization cascade which divides the remaining used CANDU fuel into three portions:
a) zirconium hulls;
b) 90% of the pellet weight which is nearly pure uranium oxide;
c) 10% of the pellet weight which is residue containing UO2, FP and TRU.
In the dissolver upstream of the recrystallization cascade the irradiated zirconium hulls are captured in a basket for later use in FNR metallic fuel alloy.
The main reason for prebaking the uranium oxide is to exclude Cs-137 and inert gases from the recrystallization cascade.
For public safety reasons the resulting fuel concentrates should then be transported in shielded containers to a shared remote fuel pyroprocessing site.
Pyroprocessing involves electrolytic molten salt reprocessing of the residue to reduce the oxides to metals, to separate FNR core fuel components, to send fission products to 300 year isolated safe dry storage for natural decay and to fabricate suitably alloyed FNR fuel rods.
products are not waste. The fission products include rare earths that
are in high demand in the electrical industry. After 300 years in
storage the fission products will need further chemical processing to
extract the valuable elements. Most of this web page focuses on
practical implementation of the recrystalization cascade.
Cs isotope Cs-137 is strongly radioactive and should be trapped
during the prebaking and sent to 300 year storage. Also released
during prebaking are the inert gases krypton, xenon and argon that
were trapped within the used nuclear fuel. Since some of the inert
gas isotopes are radioactive these inert gases must be safely
captured and vented far from a metropolis.
These gases can be caught in an atmospheric pressure cold trap cooled by liquid nitrogen. The relevant boiling points are:
194.7 K sublimation
The neutron activated zirconium hulls were originally used in a CANDU reactor for enclosing the uranium oxide pellets used to fabricate CANDU reactor fuel bundles. This irradiated zirconium can be used in the FNR metallic fuel alloys to prevent the plutonium fraction of FNR core fuel from forming a low melting point Pu-Fe eutectic with the Fe fraction of the fuel tube alloy.
The largest residue portion, which comprises about 90% of the used nuclear fuel weight, consists of nearly pure uranium oxide. The ratios of the uranium isotopes in this portion are determined by the neutron irradiation history of the nuclear fuel. For used CANDU fuel this portion has a very low radioactivity permitting relatively easy and inexpensive handling, transportation, storage and reprocessing with minimal gamma ray shielding requirements. Nearly pure UO2 extracted from used CANDU fuel has a low radioactivity whereas UO2 extracted from used Light Water Reactor (LWR) fuel has a higher radioactivity due to the presence of a larger fraction of U-232. The required amount of UO2 shielding is set by the U-232, U-235 and Np-237 concentrations in the uranium oxide as well as impurities. These uranium isotopes may be present at low concentrations in the nearly pure UO2.
portion (10%) contains the balance of the used nuclear fuel weight.
For used CANDU fuel this portion is typically: oxides of U, TRU and
fission products. This portion is intensely radioactive and must be
handled and shipped in suitable shielded shipping containers that
have walls that have a gamma ray absorption thickness the equivalent
of a 30 cm thickness of lead and must be stored in dry shielded
containers or vaults. The cost of transporting this portion is
dominated by the cost of transporting the weight of the required
shielded shipping containers. A major safety concern with respect to
this portion is ensuring that the used fuel will not go critical if
water penetrates the used fuel container.
THE U-232 ISSUE:
One of the radioisotopes of concern is U-232 which can potentially occur in used CANDU fuel as a result of alpha particle capture and 4 n emission by impurity Th-232 atoms. Being an isotope of uranium it is not removed by the uranium selective recrystalization methodology used in this process. The U-232 has a half life of 72 years and its decay path involves a hard gamma emission. Since the Th-232 impurity content in the uranium used to produce CANDU fuel can potentially vary it may be necessary to measure the gamma ray output from the separated uranium oxide and to provide sufficient shielding to ensure safety compliance (Ref: Monica Regalbuto, firstname.lastname@example.org, Purex expert at INL).
THE NEPTUNIUM ISSUE:
For chemical reasons the neptunium is not excluded from UO2 during the uranyl nitrate hexahydrate recrystallization process.
For an overview of the Ottensmeyer Plan please review OTTENSMEYER PLAN.
overview of nuclear fuel waste processing see the paper:
Radioactive Waste Partitioning and Transmutation.
Reference:Â Japanese 2002 patent
overview of Uranyl nitrate hexahydrate [UO2(NO3)2.6H2O] solubility
see:Â Uranyl Nitrate
Uranyl nitrate hexahydrate solubility in nitric acid and its crystallization selectivity in the presence of nitrate salts
Uranyl nitrate hexahydrate solubility
A paper on decontamination factors actually realized is:Â Enhancement of Decontamination Performance of Impurities for Uranyl Nitrate Hexahydrate
relevant to the U-232 issue is:Â
Uranium-232 Production In Current Design LWRs
CONCENTRATION OPERATIONAL OBJECTIVE:
Selective extraction of uranium oxide should be done at existing CANDU reactor sites to realize about 90% of the spent CANDU fuel weight as nearly pure uranium oxide and the remaining 10% consisting of: (about 2.45% of spent CANDU fuel weight as a mixture of fission products + TRUs) + (remaining 7.55% of the CANDU fuel weight is uranium oxide).Â The uranium content of this mixture is used to meet the uranium metal content requirement of the FNR fuel. The extracted nearly pure uranium oxide must be sufficiently pure to reduce its radioactivity sufficiently to enable low cost transportation and storage.
that need to be performed at each major power reactor site
a) Chopping up the CANDU fuel bundles;
b) Prebaking the chopped pieces in a vacuum furnace to capture Cs, other volitiles and inert gases;
c) Dissolving the residue in nitric acid to make a warm saturated uranium nitrate hexahydrate [UO2(NO3)2.6H2O] solution;
d) Separating pure uranium nitrate hexahydrate from other substances using a 7 stage recrystallization cascade;
e) Nitric acid recovery leaving a 90% pure UO2 pile and a 10% residue pile.
SOLUBIlITY IN NITRIC ACID:
0 deg C = 98 g / 100 g H2O
20 deg C = 122 g / 100 g H2O
100 deg C = 474 g / 100 g H2O
of this data is that when a saturated uranyl nitrate hexahydrate
solution at 100 deg C is cooled to 20 degree C the weight of crystals
formed per 100 g H2O will be:
474 g - 122 g = 352 g
The initial solution weight
474 g + 100 g = 574 g
Thus more than 50% of the initial warm solution weight forms crystals.
Note that there is little merit in further increasing the weight fraction of crystals because, while in theory that reduces the requred number of stages, it sacrifices crystal face washing by the remaining liquid solution.
Off the top of my head uranium nitrate hexahydrate dissolves in water up to about 66 %, i.e. it is not solid UNH but diluted with the 34% water. When it crystallizes out it has a density as a solid of about 2.8 g/cc whereas the density of the solution is roughly
2.8 x .66 + 1.0 x .34 or about 2.2.
Therefore I expect the crystals that form spontaneously to sink.
However, if a cold surface is introduced into the hot solution, then I would expect the crystals to form on the cold surface and stick there (depending on the properties of that surface) until they are scraped off.
PURIFICATION OPERATING PRINCIPLE:
The dissolver temperature is kept at 100 degrees C, the dissolver solution is saturated by maintaining an excess of used CANDU fuel in the dissolver tank.
Assume that a
measured amount of solution from the dissolver tank is transferred
into Tank T1a.
Assume that we start with tank T1a holding a uniform warm saturated liquid HNO3 solution that contains multiple solutes. Assume that one solute So (uranyl nitrate hexahydrate) is dominant. The other non-dominant solutes are S1, S2, etc. From the perspective of the dominant solute So the fraction of each impurity Sn is:
(Sn / So).
If the solution is gradually cooled it will form dominant solute crystals surrounded by liquid solution containing the lesser solutes. We need the dominant solute crystal formation rate to be sufficiently small that the resulting dominant solute crystals are sufficiently large that their discharge from the tank can be mostly prevented by use of a simple grating or course filter. Now drain off the cool liquid solution portion in Tank T1a to Tank T1b. Use a course filter on the cool solution drain line to prevent the crystals flowing out the drain.
are highly regular atomic structures. During dominant solute crystal
formation within a saturated liquid solution mamy of the impurities
tend to be excluded from the dominant solute crystal and hence will
accumulate in the surrounding liquid and on the crystal faces. It is
important to maintain solution agitation sufficient to continuously
wash the impurities off the crystal faces.. If, after dominant solute
crystal formation, the surrounding liquid is drained off the excluded
impurity concentration in this drained off liquid is about:
(Sn / Sob)
where Sob is the portion of So in the drained off liquid and the excluded impurity concentration in the crystals is close to zero.
that it is physically practical to make the weight of dominant solute
crystals about equal to the weight of drained off liquid. Then:
Sob ~ (So / 2)
concentration in the drained off liquid is:
[Sn / Sob]
= (Sn / So)(So / Sob)
~ 2 (Sn / So)
Now warm up the crystals remaining in tank T1a. They are nearly pure with respect to exclusion of impurity Sn. Drain out this liquid to another tank T2a via a micron filter. This filtrate is very pure. The purpose of the micron filter is to try to trap in the tank T1a impurities that were incorporated into the solid crystal but not chemically bound. These impurities will tend to drain out to tank T1b via the course filter at the next opportunity.
description of the design and operation of the Uranyl Nitrate
Hexahydrate automatic crystal growing chamber is set out at:
Uranyl Nitrate Hexahydrate Crystal Growth
1) The process equipment must not be so large that a critical mass can accumulate anywhere in the apparatus. We must be careful in tank T1b where the impurity concentration is highest.
2) Nitric acid (4.5 M) acting on uranium oxide produces uranyl nitrate hexahydrate. The process of recrystalization of uranyl nitrate hexahydrate [UO2(NO3)2.6H2O] is used to selectively extract nearly pure UO2 from spent CANDU fuel. This process is ineffective at rejecting the elements Np and Cs. The Np simply stays mixed with the uranium. It is not a radioactivity problem unless it contains the isotope Np-237. Np-237 arises from fast neutron n > 2n reactions in U-238. In CANDU reactors the fast neutron flux is very low, so the Np-237 production is low so the contribution of Np-237 to spent fuel radioactivity is very low. This statement is not true for spent fuel from fast neutron reactors. Hence recrystalization of [UO2(NO3)2.6H2O] will not achieve comparable radioactivity reduction in spent fuel from fast neutron reactors.
3) The process of recrystalization of [UO2(NO3)2.6H2O] is also ineffective at rejection of CsNO3.2H2O from the crystals. Before the recrystalization steps Cs is rejected by heating the uranium oxide to over 650 degrees C at which point the isotopes Cs-133, Cs-135 and Cs-137 vaporize as the oxides Cs2O, Cs2O2, Cs2O3 leaving behind U3O8 with some contaminant Np. These vapors must be condensed in a cold trap. As indicated above the contaminant Np is not a problem unless it contains Np-237 from fast neutrons interacting with U-238.
4) During the initial prebaking most of the trapped radioactive Kr-81, Kr-85 and Ar-39 inert gas atoms are released in the vacuum furnace. This trapped inert gas mixture must be captured, condensed, stored, transported, separated from Cs, adequately mixed with the atmosphere and safely vented.
5) There must be a mechanism to safely prevent uncontrolled release of any remaining inert gases or Cs while new UO2 is being added to the dissolver tank.
U3O8 = UO2 + 2 UO3
UO2 + 2 HNO3 + 6 H2O = UO2(NO3)2.6H2O + H2 + heat
UO3 + 2 HNO3 + 5 H2O = UO2(NO3)2.6H2O
UO2(NO3)2.6H2O + heat = UO3 + 2 HNO3 + 5 H2O
The term "weak solution" refers to a UO2(NO3)2 4.5 M nitric acid solution which is saturated at 20 degrees C. The term "strong solution" refers to a UO2(NO3)2 4.5 M nitric acid solution which is saturated at 100 degrees C.
According to Wiki UO2(NO3)2.nH2O exists as a dihydrate, a trihydrate and a hexahydrate. It appears that unless concentrated nitric acid is used the result will be the hexahydrate. However, when the hexahydrate is gradually heated during UO2 recovery the hexahydrate may liberate water molecules to form the trihydrate and then dihydrate forms.
real life a single stage of separation gives:
A ~ 0.95
B ~ 0.05
Recrystalization separation factor = (A / B) ~ 19
As (I / U) rises the rejection of impurities on crystal formation becomes less ideal. Impurities that chemically bond with the uranium nitrate hexahydrate are not rejected.
P>RECRYSTALIZATION CASCADE DESCRIPTION:
The main apparatus used for selective uranium oxide extraction is a nitric acid dissolver followed by a seven stage uranyl nitrate hexahydrate recrystallization tank cascade.
The mixed input to the cascade is (Um + Im) where:
Um = uranyl nitrate hexahydrate solution
Im = impurities (fission products and TRU)
In operation hot strong acid solution saturated with UO2(NO3)2 at 100 degrees C is fed from the dissolver into tank T1a and cool clean acid solution saturated with UO2(NO3)2.6H2O at 20 degrees C is later generated in tank T7a. As the dissolved used CANDU fuel moves from tank column 1 to tank column 7 it converts from being 97.55% UO2(NO3)2.6H2O to being nearly pure UO2(NO3)2.6H2O. The initial impurity fraction in used CANDU fuel is: 2.45% impurity to about 24.5% impurity.
consists of a dissolver, tanks Tnx where n = 1,2,3,4,5,6,7 and x =
a,b and two UO2 / acid recovery units. The dissolver contains
100degree C saturated solution with an excess of used CANDU fuel. The
system goes through a long series of temperature oscillations each
consisting of a slow cooling period from 80 degrees C down to 20
degrees C (320 minutes?) followed by a more rapid heating period from
20 degrees C back up to 80 degrees C (80 minutes?). After each
heating period clean hot solution is transferred one tank to the
right. During the subsequent cooling and crystal growth period TRUs +
FP are rejected from the crystals to the liquid. At the end of each
cooling period the remaining liquid solution flows down to tank Tnb
carrying the rejected impurities toward the impurity output. We can
refer to each complete module temperature oscillation as a
temperature cycle. Due to cleaner solution feedback the solution fed
back to tank Tna is cleaner than the solution that flowed into Tank
The cascade is designed so that the various tanks operate at progressively higher purities going to the right and at hi impuity concentrations going downward. At the end of each heating or cooling period acid solution is transferred from tank to tank nearly sequentially. While liquid transfers are occurring the liquid volume in the individual tanks fluctuates.
The tank temperatures are programmed to oscillate, typically between 20 degrees C and 90 degrees C. The circulated heating oil reaches up to 120 degrees C for warming and reaches down to 15 degrees C for cooling. Colder low end temperatures are possible with a suitable mechanical cooling equipment. The heating can be relatively fast (~ 1.0 degree C / minute) but the cooling must be slow (~ 0.25 deg C / minute) and carefully controlled. The required fine temperature control is achieved by controlling the circulated oil temperature. During a tank cooling periood UO2(NO3)2.6H2O crystals tend to grow on the textured surfaces.
Intertank liquid transfers are realized by applying air pressure over the source tank while venting the receiving tank. When the vent tube of the receiving tank fills with solution the liquid transfer is complete.
TANK Tx TO TANK Ty SOLUTION TRANSFER SUBROUTINE:
A liquid transfer from tank Tx to tank Ty is achieved by:
a) Stop solution agitation in tank Tx;
b) Open the NO top vent valve on the top of tank Ty;
c) Open desired NC liquid discharge route valve branch off the discharge tube of tank Tx;
d) Close the NO top vent valve on tank Tx;
e) Open the NC transfer air pressure valve on the top of tank Tx;
f) Wait for Tank Ty to indicate full or Tank Tx to indicate empty;
g) Close the NC transfer air pressure valve on the top of tank Tx;
h) Open the air vent valve on tank Tx;
i) Close the NC liquid discharge route valve branch off the teflon tube of tank Tx;
A cascade is formed by connecting multiple crystal growing stages in series with residue feedback. Tanks Tna, which each accommodate at least 2 U, oscillate in temperature. During a cooling period crystals form. During a warming period crystals melt.
The normal operation initial condition is that each of tanks Tna contain 2 U (nearly full) of hot saturated solution and each of tanks Tnb contain (U / 10).
During the period during which tanks Tna are cooling solution amount (U / 10) is shifted left one position in tanks Tnb starting with T1b and moving to the right to T8b so that T8b is emptied. Tank T1b discharges (U / 10) to the cascade residue output. The amount (U / 10) is indicated by total discharge of each upstream tank into the next downstream tank as indicated by air in the liquid transfer circuit. The amount U / 10 is set by a measured amount that enters tank T8b.
Wait until all tanks Tna reach the desired low temperature (20 deg C). At the end of the cooling period all of the remaining cool liquid in each tank Tna (approximate volume U) is transferred to the accompanying tank Tnb. Note that tanks Tnb must each accommodate more than 1.1 U. This transfer occurs until each upstream tank Tna is dry as indicated by air in the common liquid transfer circuit.
Then tanks Tna and Tnb are both warmed. During a warming period the solid crystals in tanks Tna melt. Wait until all tanks reach the desired high temperature (100 deg C). Then moving from right to left the warm saturated solution of volume U is completely discharged to the next a tank T(n+1)a as indicated by air in the liquid transfer circuit. The volume (U /10) of the liquid discharged from the last tank Tna is discharged to tank T8b as indicated by a liquid level sensor in the T8b vent line and the then the remaining (9 U /10) is discharged to the UO2(NO3)2.6H2O cascade output. New saturated solution measured amount U flows from tank T0a (volume U) into tank T1a. In each case air in the common liquid transfer line indicates complete transfer. The volume in Toa is indicated by a liquid sensor in its vent line.
Then the warm liquid in tanks Tnb for n = 1 to 7 is transferred into tanks Tna where it is mixed with the newly shifted in liquid so that tanks Tna all contain 2 U. Completion of this liquid transfer is indicated by tank full sensors in the vent lines of tanks Tna.
Note that the liquid insertion amount U flowing from tank T0a and recirculated amount (U / 10) flowing from tank T8b are accurately calibrated.
A practical cascade design is one where recrystallization at each stage is assumed to be imperfect. Assume that the weight of crystals = weight of solutution at the lowest temperature. Then the single stage transfer function is:
U + I = [(U / 2) + A I] + [(U / 2) + B I]
where cool output flow per thermal cycle is:
[(U / 2) + A I]
and where warm output flow per thermal cycle is:
[(U / 2) + B I]
A + B = 1
U = total uranium nitrate hexahydrate stage input
I = total impurity stage input
CASCADE BOUNDARY CONDITIONS:
Cascade input = (U + I)
Cascade main discharge = (9 / 10)(U) + k (I)
For each Tnb stage:
Main output = 10 X Aux output
For tank T1b:
Main output = 10 [(1 / 10)(U) + (1 - k)(I)]
= (U) + 10 (1 - k) I
Cascade Separation Factor:
[(I / U)] / [k (I) /(U)]
= [1 / k]
FIRST STAGE ANALYSIS:
Tank T1a Main input = (U + I)
Tank T1a Aux input = 10[Residue discharge]
= 10 [(1 / 10)(U) + (1 - k)(I)]
= (U) + (1 - k) (10 I)
T1a total input = 2 U + (11 I) - k (10 I)
T1a warm output = (U) + B I[(11) - 10 k]
T1a cool output = (U) + A I [(11) - 10 k]
Tank T1b Main input = (U) + A I [(11) - 10 k]
Tank T1b Aux input = (U / 10) + K2 I
Tank T1b Total input = (11 U / 10) + A I [(11) - 10 k] + K2 I
Tank T1b Aux input = (Tank T1b total input - Tank T1b main input)
= (11 U / 10) + A I [(11 ) -10 k] + K2 I - [(U) + A I [(11) - 10 k]]
= (U / 10) + K2 I
K2 ~ (1)
Tank T1b Total input = Tank Tib total output
(11 U / 10) + A I [(11) - 10 k] + K2 I
= (11 U / 10) + (1 - k) 11 I
A I [(11) - 10 k] + K2 I = ( 1 - k) 11 I
A [(11) - 10 k] + K2 = (1 - k) (11)
K2 = (1 - k)(11) - A [(11) - 10 k]
= 11 (1 - k - A) + 10 k A
= 11 (B - k) + 10 k A
= 11 B - 11 k + 10 k A
= 11 B - k - 10 k + 10 k A
= 11 B - k - 10 k (1 - A)
= 11 B - k - 10 k B
special case of many stages then:
k = 0:
K2 I = (11) I (1- A)
= (11 / 2) I B
which, as expected, is the same as the impurity carried forward via the T1a warm discharge.
The effective first stage separation factor is:
(I / U) / [B I ((11) - 10 k) / (U)]
= 1 / [B (11 - 10 k)] which for k = 0 becomes:
1 / [B (11))]
= 1 / 11 B
which for B ~ (1 / 20) becomes:
20 / 11
LAST STAGE ANALYSIS:
Tank Ta Main Input = Tank Ta Total Input - Tank Ta Aux Input
= Tank Ta Total Output - Tank Ta Aux Input
= 2 U + I - [U + (10 / 11) A I + (B I/ 11)]
= U + I - (10 / 11) A I - (B I/ 11)]
Tank Ta Aux Input = Tank Tb main output
= U + (10 / 11) A I + (B I/ 11)
Tank Ta warm output = U + B I
Tank Ta cool output = U + A I
Tank Ta total output = 2 U = Tank Ta total input
Tank Tb Main Input = U + A I
Tank Tb Aux input = (U + B I) / 10
Tank Tb total input = (11 / 10) U + A I + (B I / 10)
Tank Tb Main Output = U + (10 / 11)[A I + (B I / 10)]
= U + (10 / 11) A I + (B I/ 11)
Tank Tb Aux output = (U / 10) + (A I / 11) + (B I / 110)
Last stage separation factor:
[I - (10 / 11) A I - (B I / 11)] / (B I)
= (1 / B) - (10 / 11)(A / B) - (1 / 11)
= (1 / B) - (10 / 11)((1 - B) / B) - (1 / 11)
= (1 / 11 B) + (9 / 11)
For B = (1 / 20):
Last stage separation factor = (29 / 11)
Thus 10 or
more stages, each with two tanks, may be required to reduce the
impurity concentration by a factor of 1000.
In operation hot strong acid solution saturated with UO2(NO3)2.6H2O at 100 degrees C is fed from the dissolver into tank T1a main inlet. Tank T1a is then cooled to 20 degrees C. The remaining cool liquid is transferred from tank row a to tank row b. Thr remaining solid is then warmed. It converts from being 97.55% UO2(NO3)2.6H2O to being nearly pure UO2(NO3)2.6H2O. The initial impurity fraction in used CANDU fuel is: 2.45%.
The pure liquid is transferred from tank T1a to Tank T2a.
The contents of Tank T1b are heated and recycled back into tank T1a.
The contents of tank T0a are transferred into T1a.
Tank T1a then starts a new cooling cycle.
The contents of tank T1b should be heated to 100 degrees C before being transferred back into tank T1a.
The impurity fraction at the cascade residue discharge is ~ 24.5%.
The cascade is fed by a closed dissolver. The dissolver has a heating coil and a removable fuel basket to enable removal of undissolved fuel and zirconium hulls. In normal operation the dissolver is always hot at 100 degrees C. The dissolver solution is saturated by maintaining an excess of used CANDU fuel. The dissolver receives HNO3 recovered from the two cascade outputs.
The cascade consists of a series of nitric acid resistant tanks designated by Tnx. n = 1, 2, 3, 4, 5, 6, 7 x = a, b The detail of the connections to tank T7x and T1b differs from the other tanks to enable the solution discharged by these tanks to feed the HNO3 recovery apparatus.
The tank discharge tubes are fed from the very bottom of each tank. The discharge connections to the adjacent downstream tanks are to the space above the liquid levels in the downstream tanks to ensure that there are liquid breaks in the connections between adjacent tanks. Each tank overhead gas space is vented to a common overhead vent pipe. This vent is sometimes closed to permit easy liquid transfer between tanks while safely containing hot HNO3 and related gases.
The tanks Tna are weakly agitated during crystal growth by circulating solution so that there is rising solution convection within the UO2(NO3)2.6H2O solution. The purpose of the weak agitation is to minimize impurity accumulation on the crystal faces during crystal growth and to prevent drain blockage by small crystals.
Each tank has temperature sensors which are used to regulate the heating/cooling rates and to indicate when a heating or cooling cycle is complete.
During the heating period the temperature of the oil is about 20 degrees C warmer than the temperature of the solution. During the cooling period the temperature of the oil is about 5 degees C cooler than the solution.
At the end of the heating and cooling periods the solution in each tank is transferrred into the next downstream tank.
Absent U-232 in the usedfuel this cascade should reduce the uranium oxide gamma emission per kg down to the level of new CANDU fuel formed from natural uranium_______??. The amount of U-232 gamma emission will be a strong function of the original Th-232 impurity concentration in CANDU fuel and a weak function of the age of the used CANDU fuel.
During an intertank solution transfer each transfer normally runs until there is no liquid in the upstream tank. The exceptions to this rule are for strong hot solution transfers from the dissolver into tank T1a. In these cases the solution transfer stops when the desired liquid level in tank T1a is attained.
The tanks all have liquid level overfill sensors.
control logic which will stop the liquid transfer sequence and alarm
if any tank liquid level exceeds its design maximum. The floor under
the tanks is covered with an acid resistant stainless steel sheet and
is sloped to a common drain. The drain goes to the basement level
dirty acid drain down tank. Thus any acid leak anywhere in the system
gravity drains into this dirty acid drain down tank.
To enable system maintenance the dissolver basket containing remaining fuel and zirconium hulls is removed and placed behind a shielded barrier, all the tanks are heated to 100 degrees C to fully dissolve the remaining solids and then the entire system solution volume is drained down into the shielded below grade acid drain down tanks. Below the drain down tanks are pumps which can be used to transfer the acid solution back into the cascade after the service work is complete.
requires drain and rinse valves to enable cascade draining and
Â Â SYSTEM OPERATION: The equipment consists of a dissolver, 28 tanks Tnx where n = 1,2,3,4,5,6,7 and x = a,b,c and two UO2 / acid recovery units. The dissolver contains impurities at the equilibrium concentration in the used CANDU fuel. The system goes through a long series of temperature oscillations each consisting of a slow cooling period from 80 degrees C down to 20 degrees C (320 minutes?) followed by a more rapid heating period from 20 degrees C back up to 80 degrees C (80 minutes?). After each heating period hot solution is transferred down one tank. During the subsequent cooling and crystal growth period TRUs + FP are rejected from the crystals to the solution. At the end of each cooling period the remaining liquid solution flows forward one tank to the right carrying the rejected impurities toward the impurity output. We can refer to each complete module temperature oscillation as a temperature cycle.
The tanks oscillate in temperature. The cascade operates by taking
advantage of the temperature dependence of UO2(NO3)2 solubility in
4.5 M HNO3. Lowering the temperature of a saturated solution triggers
recrystallization. During slow recrystallization impurity atoms are
excluded from the UO2(NO3)2.6H2O crystals and concentrate in the
surrounding liquid. In order to realize high purity UO2(NO3)2.6H2O a
cascade containing multiple successive recrystalization steps is
used. Â While a tank is being heated or cooled there is
no solution flow in or out of that tank. After a cooling period there
is a cool solution discharge period. After a warming period there is
a warm solution discharge period. During the cooling period crystals
form which increase the residue concentration in the liquid
surrounding the crystals. At the end of the cooling period crystals
containing nearly pure uranium nitrate hexahydrate are retained by a
screen and the surrounding cool liquid is transferred one tank
position toward the residue discharge. Then the tank is heated. The
UO2(NO3)2.6H2O crystals in tanks re-dissolve in the cleaner
surrounding warm acid solution. The solution resulting from melting
the remaining crystals in tank T7x is transferred into the UO2 / HNO3
recovery system. Â Then there is a warming period during
which the crystals melt. Then the warm liquid flows one tank position
downward toward the nearly pure uranium discharge. Then the tank is
reloaded from its two upstream tanks. Then another cooling period
commences. Â During steady state operation the fraction:
(TRU + FP) / U at the cascade residue discharge builds up to be about
10X the (TRU + FP) / U fraction in the used CANDU fuel, which
fraction is several orders of magnitude higher than the corresponding
value of (TRU + FP) / U at the nearly pure uranium oxide discharge.
There are seven_____ columns of tanks in four rows designated n = 1, 2, 3,
4, 5, 6,7 and x = a, b. The tank at column n, row x is Tnx. The
seven_____ columns and two rows are required to hit two targets. As
uranyl nitrate hexahydrate moves from column 1 to column 7____ the ratio of I /
U in the nearly pure uranium output must drop at least 1000 fold. As
uranyl nitrate hexahydrate moves from row a to row b the ratio
of I / U in the FNR fuel feedstock must increase at least 10 fold. The used CANDU fuel is dissolved in warm HNO3. Then on row a the Nth
tank of the cascade behaves as follows: PROCESS
ECONOMICS: With careful
adjustment one system will process UO2 atÂ 83 kg / day
yielding a TRU output of The ultimate
value of fissile is $7600 / kg Thus the gross
value of this fissile fuel after pyroprocessing is: Assume that
the TRU concentration apparatus costs $500,000 and needs to earn a 2
year simple payback on capital plus $100,000 / year for cost of
operation. Gross income
is $2523 / day X 365 days / year = $920,968 year. Hence,TRU
concentration costs $350,000 / year leaving: One CANDU
reactor produces 4 g TRU / kg UO2 X 100,000 kg UO2 / 1.5 years = 267
kg TRU / year One TRU
concentrator produces: Need 2 systems
at $500,000 each to balance one CANDU reactor. Â
a) Transfer uranium nitrate hexahydrate solution from two upstream tanks to tank Tna until both upstream tanks are empty;
b) Cool tank Tna to form uranyl nitrate hexahydrate crystals + excess solution. The crystals must form about half the total weight.
c) Transfer the remaining cool liquid from tank Tna to tank Tnb until there is no more liquid in Tank Tna;
d) Warm tank Tna to melt the crystals remaining in tank Tna;
e) Transfer the entire warm solution from tank Tna to tank T(n+1)a.
f) Repeat the above sequence subject to readiness of both upstream and down stream tanks. Total mass flow into tank T1a is (U + I) where U = uranium nitrate hexahydrate and I = impurity Cold liquid discharge from top of T1a is (U / 2) + A I Hot liquid discharge from main output of T1a is (U / 2) + B I A + B = 1.000
Due to crystal exclusion of impurity I material and careful mechanical design: A ~ 0.99 B ~ 0.01 Note that A and B must be experimentally determined. With reference to the diagram: __________ indicating that the impurity concentration in the UO2 is reduced more than 100 fold. Thus a separation factor of 19 gives an impurity enhancement of 10 fold and and impurity reduction of more than 100 fold in UO2. Recall that: A + B = 1 and Separation factor = (A / B) ~ 19 Calculation check: B = (1 - A) B^2 = (1 -2 A + A^2) B^3 = (1 - 2A + A^2 - A + 2 A^2 - A^3) = (1 - 3A + 3 A^2 - A^3) Substitue for B in the above equations and check that the sum of the total outputs = (U + I). Â SOLID RECOVERY: The U + Pu + Fp impurities are harvested from the liquid solution in the dissolver after the dissolver has reached its equilibrium impurity concentration. The extracted dissolver solution is heated to evaporate the nitric acid which is recovered and recycled. A significant amount of energy is required. The resulting dry residue is the feedstock for successive operations to make FNR core fuel. Very strong UO2(NO3)2.nH2O from tank T5 is transferred to the UO2 /acid recovery unit. Heating this very strong solution to recover the UO2 and the contained acid and the cesium oxide compounds requires a substantial amount of energy. The very strong solution in the UO2 /acid recovery unit is mildly heated to drive off the nitric acid which is recycled. The dry residue is then heated to 650 degrees C to drive off radioactive cesium oxides which are caught in a cold trap and sent to 300 year storage. The remaining high purity depleted uranium oxide has a very low radioactivity and should be stored for future use as FNR blanket rod material. Â Note that over time the quantity of Zr hulls in the dissolver will gradually increase. From time to time the process must be stopped to allow Zr extraction. Â SOLUBIlITY: 0 deg C = 98 g / 100 g H2O 20 deg C = 122 g / 100 g H2O 100 deg C = 474 g / 100 g H2O Solute mass per 100 g H2O that moves one tank forward with each thermal cycle: (474 g - 122 g) / 100 g = 352 g UO2(NO3)2 / 100 g H2O Solute mass per 100 g H2O that moves two tanks backward with each thermal cycle: = 122 g / 100 g H2O Net solute that moves forward one tank in one thermal cycle is: 352 g / 100 g H2O - 2(122 g / 100 g H2O) =Â 108 g / 100 g H2O The amount of CANDU fuel processed per thermal cycle is: [UO2 / UO2(NO3)2] X 108 g / 100 g H2O = [(238 + 32) / (238 + 32 + 2 (14 + 48))] X 108 g / 100 g H2O = [(270) / (270 + 124)] X 108 g / 100 g H2O =Â 74.01 g / 100 g H2O Mass per tank per 100 g H2O that must be heated and cooled in each thermal cycle is: 100 g H2O + 474 g UO2(NO3)2 Â TANK VOLUMES: The volume of a basement drain down tank sufficient to absorb the entire volume in the dissolver, Tnx is: ________ We may need to be concerned about accumulating a critical mass in the dissolver. Hence, we may need to rethink these tank sizes. Note that the submerged tube surface area must be consistent with the assumed heat transfer rate. Â HEAT REQUIRED TO DRIVE ONE THERMAL CYCLE OF OPERATION: ____________________ Estimate the heat required to swing all 21 tanks through 105 degrees C: _________ 5 tanks X 574 g / stage/ 100 g H2O X 80 deg C X 1 cal / g-deg C X 4.18 J / cal = 959,728 J / 100 g H2O = 959,728 J / 100 g H2O X 1 W-s / J X 1 kWt / 1000 W X 1 h / 3600 s =Â 0.2666 kWht / 100 g H2O Note that this figure does not include the heat capacity of the tank metal. Heat of formation of UO2(NO3)2.6H2O: = -2739.5 Btu/lb of UO2(NO3)2 This is the heat that must be supplied to release UO2 from UO2(NO3)2.6H2O crystals. 1 Btu = heat required to raise 1 lb of H2O 1 deg F = heat required to raise 454 g of H2O (1 / 1.8) deg C = heat required to raise 252.2 g of H2O 1 deg C = 252.2 cal 1 lb = 454 g Hence: (1 Btu / lb) = 252.2 cal / 454 g = 0.5555 cal / g Thus the heat required to recover UO2 from UO2(NO3)2.6H2O is: 2739.5 Btu/lb of UO2(NO3)2 =2739.5 Btu/lb X (0.5555 cal / g) / (1 Btu / lb) = 1522 cal / g of UO2(NO3)2.6H2O Each thermal cycle produces 108 g of UO2(NO3)2.6H2O. Thus the heat required for UO2 recovery per thermal cycle is: 108 g X 1522 cal / g = 164,376 cal = 164,376 cal X 4.18 J / cal = 687,092 J = 687,092 J X 1 W-s / J X 1 kWt / 1000 W X 1 h / 3600 s =Â 0.1908 kWht Thus the energy required per thermal cycle / 100 g H2O / tank is: 0.1908 kWht + 0.2666 kWht =Â 0.45746 kWht / 100 g H2O / tank From above, each thermal cycle produces: 74.01 g UO2 / 100 g H2O Thus at a minimum a facility processing 5000 kg / day of used CANDU fuel has an average thermal power consumption of: 5000 kg / day X 1 day / 24 h X 0.47546 kWht/74.01 g X 1000 g / kg =Â 1338.38 kWt By the time fan and pump loads for heat removal are added and the heat capacity of the tanks is included this will be about 2.0 MWt. Note that this is an average heating power. In reality the system heats for 80 minutes followed by a cooling period of 320 minutes. Thus during the heating periods the peak power is 5X the average or: 5 X 2 MWt =Â 10 MWt. Note that the UO2 / HNO3 recovery unit can be run continuously. Hence about 800 kWt is needed continuously and about 6 MWt are needed with a 20% duty cycle. Â SOLVE FOR ACTUAL HEATING AND COOLING TIMES TANK REQUIREMENT: Each temperature cycle = 400 minutes. Number of temperature cycles per day = [24 h X 60 m / h] / [400 min / cycle] = 3.6 cycles / day In order to produce 5000 kg / day of UO2, each cycle must produce: 5000 kg / 3.6 = 1389 kg UO2 Each 100 g of acid produces: 74.01 g UO2. Hence the required number of 100 g units of acid per tank is: 1389 kg / 74.01 g = 18.77 X 10^3 Hence the required amount of acid / tank is: 18.77 X 10^3 X 100 g = 1877 kg Hence it appears that the estimated tank volumes are 2 X larger than necessary. However, that extra volume allows for the tube and manifold volume in a shell and tube heat excahnge configuration with water-glycol in the tubes and volume expanding crystals on the shell side. Wrapped around the outside of the vertical tubes must be a perforated sheet to stop crystals falling into the shell side drain. Â MECHANICAL USED CANDU FUEL FEED: a) From the used CANDU fuel inventory that has been out of a CANDU reactor for at least 10 years withdraw used CANDU fuel bundles as required. b) Mechanically shear the used CANDU fuel bundles into small pieces, each about 3 cm long. c) Feed these pieces into the dissolver at a controlled rate sufficient to keep the dissolver solution saturated at 100 degrees C. Â ZIRCONIUM RECOVERY: Over sufficient time the dissolver will fill up with Zr hulls and must be stopped to remove these hulls. The dissolver contains a large basket to expedite Zr hull recovery. Â DISSOLVER OPERATION: a) The dissolver is a nitric acid resistant tank with a removeable top. The dissolver tank contains a nitric acid resistant basket to allow convenient zirconium hull recovery. An empty basket is lowered into the dry dissolver tank, the dissolver tank top is replaced and the dissolver tank is evacuated. This evacuated air is exhausted to the atmosphere. The evacuation valve is then closed. b) Then hot nitric acid temporarily stored in the drain down tank is pumped into the dissolver. c) Due to the low overhead pressure in the dissolver tank nitric acid containing a low UO2(NO3)2 concentration flows from the drain down tank into the dissolver where it is heated to 100 degrees C. Used CANDU fuel is added to the dissolver via a feed tube. After some time in the nitric acid at 100 degrees C in the dissolver solution becomes saturated with UO2(NO3)2. Undissolved zirconium pieces collect in the dissolver's bottom basket. Radioactive inert gas fission products such as krypton bubble up through the liquid acid and collect in the sealed space above the acid. d) Then a controlled volume of the hot nitric acid solution is pumped from the dissolver tank into tank T1 via the port at the bottom of the dissolver. Sufficient nitric acid solution remains in the dissolver tank over this port to prevent the radioactive inert gas fission products on top of the dissolver solution from exiting the dissolver via its bottom port. e) Eventually when the inert gas pressure over the acid in the dissolver tank becomes too high or when the dissolver tank full of Zr hulls the dissolver is cooled to 20 degrees C to reduce the partial pressure of remaining HNO3 gas in the dissolver head space. f) The inert gas plus some HNO3 gas in the dissolver head space are evacuated via a cold trap. The HNO3 vapor is caught in the cold trap. The radioactive inert gases are sent either to a high stack or to a pressure tank for later safe release to the atmosphere at a remote location. Ideally the inert gas should be stored to allow it to naturally decay. After a suitable decay period vent the residual inert gas to the atmosphere. Note that radioactive Kr-81, Kr-85 and Ar-39 must be well mixed with the atmosphere. g) The nitric acid in the cold trap is isolated and is recycled back to the disssolver. h) When the dissolver is full of Zr hulls the dissolver is drained into the drain down tank. i) The dissolver tank top is removed. The basket containing zirconium pieces is removed from the dissolver and is air dried. The neutron activated zirconium is harvested from the dissolver basket for future use as a component of FNR fuel. j) Transport the neutron activated zirconium and the CANDU fuel concentrates to the remote irradiated zirconium and fuel stores. k) The now empty dissolver tank basket is replaced and the dissolver batch cycle repeats. Note that the dissolver tank is sufficiently large that one dissolver batch will serve many temperature cycles. Note that as the system operates the amount of UO2 in the dissolver gradually diminishes but the acid liquid level in all the tanks remains almost constant and the UO2(NO3)2 concentration in the dissolver solution remains almost constant. From time to time new used CANDU fuel is added via the feed tube. Â MISCELLANEOUS CASCADE ISSUES: 4) Provide perforated sheet cylinders in each tank to provide crystal growth surfaces and to prevent crystals being sucked into the inlet of a backward pump. These must be perforated all the way to their bottom edges to ensure complete drainage. 5) It is necessary to carefully control the liquid transfers to realize the optimum acid volume in each tank at a particular time in the operating cycle. If there is too much acid in a tank the system will be energy inefficient on thermal cycling and the forward solute propagation will be poor. If there is too little acid in a tank not all the crystals will be dissolved during a tank heating cycle, leading to insufficient crystal growth in the next tank during its cooling cycle. Thus the liquid levels must be carefully controlled. Each tank should have a precise liquid level sensing device and a pressure sensor at the tank discharge. The level control in tank T1 is particularly important as it sets the levels in the other tanks if the tank flows are all properly balanced. Thus the tube sheets and end manifolds must have provisions for the required shell side liquid level sensors. Â UO2 RECOVERY AND ACID RECOVERY 1) Chop up the used fuel bundles. 2) Move the [U3O8 + contamination Cs2O,Cs2O2, Cs2O3] from #1 above to a suitable furnace. 3) Heat the [U3O8 + contamination Cs2O,Cs2O2, Cs2O3] residue above 650 C to drive off the Cs2O, Cs2O2, Cs2O3 as a condensable gas. 4) Condense and collect the radioactive Cs2O, Cs2O2, Cs2O3. 5) Send the radioactive Cs2O, Cs2O2, Cs2O3 to fission product 300 year storage. After 30 years in storage the radioactivity should be dominated by Cs-137 (30 year half life) and Cs-135 (3.0 X 10^6 year half life). Use a segragated store in case this material contains other contaminants. 6) The remaining solution dry residue is CANDU fuel concentrates for further reprocessing (fission products + TRUs + uranium oxide). 7) Apply the cascade. 8) Heat the cascade discharge products to evaporate, condense, recover and recycle the HNO3. 8) Transport the CANDU fuel concentrates (fission products + TRUs + uranium oxide) to the remote site for electrolytic reprocessing. Â MINIMIZASTION OF CANDU FUEL REPROCESSING COST: 1) The purification of U is not a matter of choice. The subsequent pyroprocess has to extract enough U in total first so that the cadmium cathode can start working on the transuranics. Otherwise it extracts the U into the molten Cd until the transuranics start coming out. That step is done cleaner on the iron cathode. It just happens that the iron cathode produces pure U as it extracts it without the transuranics. 2) To make the overall process as inexpensive and short as possible, the pre-extraction of U should ideally go to as high a percentage of U as possible without introducing unacceptable levels of impurities. Â SYSTEM SERVICE: The envisaged cascade is orders of magnitude more mechanically simple than the earlier student design summarized below. From time to time the system will likely need mechanical service. To safely enable such service both the HNO3 and the radioactive species must be completely drained. To service the system heat the entire system to 100 degrees C to dissolve all of the UO2(NO3)2. Lift out the dissolver basket with a gantry crane and place it in a shielded enclosure and then drain all the fluids from the drain valves below the bottoms of the tanks to shielded below grade drain down tanks. If the radiation from residue remaining in the system is too high flush the system with clean 4.5 M nitric acid. Â UO2(NO3)2 DATA: MP = 60.2 deg C Dissociation at 118 deg C UO2(NO3)2 is hygroscopic forming 6(H2O), 3(H2O), 2(H2O) TOXICITY OF UO2(NO3)2: 12 mg / kg (dog) Natural Uranium: U-238 99.27%, 4.47 X 10^9 Y U-235 0.711% 700 X 10^6 Y U-234 ~ 0.019% Price of uranium oxide ~ $28 / lb Â MECHANICAL DESIGN: One of the important design constraints is set by tube sheet flexing. During the heating period the heat exchange tubes are about 20 degrees C hotter than the shell wall. During the cooling period the tubes are about 5 degrees C cooler than the shell wall. Hence even if the tubes and shell wall are perfectly TCE matched the tube sheets will still flex back and forth with each temperature cycle. Hence the heat exchange tubes must be centrally located in the tube sheets and the width of the non-tubed ring around them in combination with the tube sheet material thickness must accommodate continuous tube sheet flexing. Thus the tank height, the temperature coefficient of expansion, the Young's modulus, the yield stress, the tube wall thickness and the shell wall thickness will all be important. Keeping the tubes centrally located will reduce the heat transfer area. In order to move this design forward we need the physical properties of the material, which will be determined in part by the corrosion resistance. It may prove necesssary to make the top tube sheet thinner than the bottom tube sheet which must support the fluid head. Â CASCADE MATERIAL SELECTION: Peter Ottensmeyer: The only plastic that is HNO3 resistant is Teflon (PTFE). WHAT ABOUT EPOXYS? The preferred materials are 440 stainless steel, ceramic, and titanium Â John Rudesill: The concept Charles has described has merit. Counter current separations are ideally more efficient than single stage separations and are a preferred design practice in chemical engineering when practical. The link I sent to Charles earlier https://www.rolledalloys.com/technical-resources/environments/nitric-acid/ indicates that various SS alloys are used in contact with 60% HNO3 at temperatures well in excess of 110 C. Plastic could work, but is both structurally and thermally inferior to SS alloys, The heat conduction coefficient is also much lower than metals. I will dig a little deeper for approved metals for this service. We have to be aware that the FP's contain halides which can make HNO3 far more corrosive. I need to see the expected solution analysis of the product coming from the fuel element dissolution step to account for the halide content. I am aware than jacketed tanks are made of even carbon steel and lined with teflon to make them near impervious to corrosion until the eventual pin hole forms leading to liner failure and jacket penetration. Heat transfer is impeded, but is still workable. Teflon can be used over 200 C as long as it is not used in a structural role. Similarly metal tubing can be teflon coated. For some services polyvinyl difluoride (F analog of Saran) can be used in place of teflon and it is somewhat more structurally robust. I will read over the procedure Charles has provided and see if I can make a hand sketch of this process to study. It makes sense to scope this concept out at the kg/d scale or even smaller before making any commitments to a detailed design for a ~5mt/d facility. Materials handling will be a dominant aspect of the design given the radio hazard content. A 5 mt /d process rate may seem like a lot, but is a very modest pilot plant. There is little incentive to try to save on equipment costs as they will be at large variance with usual engineering economy of scale cost calculations. The design must prioritize absolute containment and separation completeness over almost all other considerations. We want to do each step only once--no do overs at least in a continuous flow mode. In a batch mode, a step can be repeated if necessary. Also, batch mode can enable single shift 5 day week operation, initially. Continuous operation must be 24/7. As I write this, I think batch mode is the wiser choice. Once it is up and running and understood, a prudent consideration of continuous 24/7 operation is reasonable. Â
Assume that we can process 14,000 ml of solution per temperature cycle. That is a solution mass of:
(2.2 kg / 1000 ml) X 14,000 ml / cycle = 30.8 kg / cycle.
The weight of uranyl nitrate hexahydrate UO2(NO3)2.6H2O processed per cycle is:
[(2.8 X 0.66) / 2.2] X 30.8 kg / cycle = 25.872 kg / cycle
Assume 6 temperature cycles per day:
6 cycles / day X 25.872 kg / cycle = 155.23 kg / day uranyl nitrate hexahydrate
Weight Fraction UO2 = UO2 / [UO2(NO3)2.6H2O]
= (238 + 32) / [238 + 32 + 2(62) + 6(18)]
= 270 / [270 + 124 + 108]
= 270 / 502
Hence UO2 throughput = 155.23 kg /day X 0.5378
= 83.483 kg /day
83 kg / day X 4 g / kg = 332 g / day TRU.
0.332 kg/ day X $7600 / kg = $2523 / day
which must include the cost of pyroprocessing.
920,868 - 350,000 = $570,560 / year
for pyroprocessing 332 gm / day TRU or 1.66 kg / day of FNR core fuel.
0.332 kg TRU / day X 365 day / year = 121 kg / year
Thus we need 2.2 TRU concentrators for every CANDU reactor.
There are seven_____ columns of tanks in four rows designated n = 1, 2, 3, 4, 5, 6,7 and x = a, b. The tank at column n, row x is Tnx. The seven_____ columns and two rows are required to hit two targets. As uranyl nitrate hexahydrate moves from column 1 to column 7____ the ratio of I / U in the nearly pure uranium output must drop at least 1000 fold. As uranyl nitrate hexahydrate moves from row a to row b the ratio of I / U in the FNR fuel feedstock must increase at least 10 fold.
The used CANDU fuel is dissolved in warm HNO3. Then on row a the Nth
tank of the cascade behaves as follows:
adjustment one system will process UO2 atÂ 83 kg / day
yielding a TRU output of
The ultimate value of fissile is $7600 / kg
Thus the gross
value of this fissile fuel after pyroprocessing is:
Assume that the TRU concentration apparatus costs $500,000 and needs to earn a 2 year simple payback on capital plus $100,000 / year for cost of operation.
is $2523 / day X 365 days / year = $920,968 year. Hence,TRU
concentration costs $350,000 / year leaving:
One CANDU reactor produces 4 g TRU / kg UO2 X 100,000 kg UO2 / 1.5 years = 267 kg TRU / year
Need 2 systems
at $500,000 each to balance one CANDU reactor.
ÂPRIOR STUDENT WORK The Ottensmeyer Plan originated in student work at the University of Toronto circa 2010. The following notes are an edited version of student work. There are various claims made by the students based on references that should be confirmed before large sums of money are committed to implementation of the process. 2.1.1 Fuel Bundle Shearing CANDU nuclear fuel bundles are made from fuel pellets, inserted into zircaloy cladded fuel rods and loaded into channels of a cylindrical metal assembly (See Appendix B for diagram and composition). Thus, the first step of reprocessing begins with the removal of these long, narrow fuel rods from the bundle and liberating the spent fuel. The proposed process achieves this by mechanically shearing the rods into small segments, approximately 3cm , and dissolving the exposed fuel in a hot nitric acid dissolver solution. Other alternatives for fuel liberation were considered, such as chemical decladding, mechanical decladding and perforating the cladding. Decladding proved uneconomical due to excess waste and high losses, while perforating the cladding alone does not provide enough exposed fuel surface area to dissolve it in a timely fashion . Sawing as an alternative to mechanical shearing was also considered; however, it is less consistent and produces more metal fines . 2.1.2 Dissolution The dissolver selected is a countercurrent flow, multistage, rotary dissolver. After shearing, pellets are fed into one end of the dissovler, while hot nitric acid is fed into the other end at 4M and 95Â°C . The resulting solution, after dissolution of the spent fuel, is approximately 300g/L (using U as a basis). The concentration and temperature of nitric acid was found by analysis of calculated rate data and experimental data (see Appendix C). While higher concentrations of nitric acid yield higher rates of dissolution (75 minutes versus approximately 3 hours), it also introduces further hazards with regards to acidity and corrosion. The dissolver stage was not rate limiting in the process and thus the lower, less hazardous concentration could be applied. As the dissolver drum rotates, the pellets in the initial stage are transferred along its length, while nitric acid flows counter-currently dissolving the fuel within. The drum is rotated to propel the solids forward and moves in a rocking motion to establish appropriate agitation for speedier dissolution. It also allows for better mass transfer and more efficient dissolution, as the least soluble fuel particles are contacted with the strongest acid . See Appendix E-1 for schematic. At the last stage of the dissolver, the solid cladding is ejected from the bottom, while the loaded nitric acid, containing uranium nitrate, plutonium nitrate and other dissolved fission products and actinides, flows out from the opposite end. The main dissolution reactions (Rxn. 1 and 2) are the dissolution of uranium dioxide (UO2) in nitric acid. Rxn. 2 is dominant when nitric acid concentration is less than 10M . The other actinides dissolution reactions proceed similarly. UO2+ 4HNO3 = UO2(NO3)2+ 2NO2+ 2H2O Î”HrÂ°= -74.9 kJ/mol [Rx.1] 3UO2 + 8HNO3 = 3UO2(NO3)2 + 2NO + 4H2O Î”HrÂ°= -367.6 kJ/mol [Rx.2] During dissolution, NOx gases and radioactive iodine vapours from the fuel are emitted and sent for gas treatment (see section 6.2) while the concentrated solution is sent for crystallization. The undissolved cladding is rinsed, monitored for fissile material, packaged and transferred to the solid waste storage area for disposal. This dissolver design is an update on more traditional spent fuel dissolvers that involve placing pellets into perforated baskets, followed by immersion in hot nitric acid. There are a number of operating limitations associated with this method, including the requirement of batch processing. Furthermore, the use of a highly corrosive solvent, off-gas emissions, as well as the potential for criticality, leads to challenges and limitations on the amount processed per batch. The equipment also decreases any criticality risk due to the processing of less mass in one location and its long, narrow geometry . 2.1.3 Crystallization and Clarification The dissolver solution contains both long lived actinides used for FNR fuel and shorter-lived fission products. Before separation of actinides from fission products, the majority of U can be crystallized out of solution directly into Uranyl Nitrate Hexahydrate (UNH), a yellow green crystal. Due to its high concentration of U in solution (300g/L present as uranyl ions, UO22+), as well as its relative insolubility, a decrease in solution temperature to 10-20Â°C is capable of crystalizing approximately 70-80 wt% of the U .This was determined based on solubility-temperature data, as well as experimental results (see Appendix D). Co-crystallization of smaller amounts of Pu(VI) and Np(VI) is possible and favourable, as this would further diminish the amount of actinides requiring later processing. The crystallization reaction proceeds as follows: UO22++ 2NO3- + 6H2O = UO2(NO3)2â€¢6H2O Î”HrÂ°= -20 kJ/mol [Rx.3] One issue with crystallization is the contamination of U with fission products. Testing has shown that there is negligible occlusion of fission products within the crystals, and crystals can be washed to achieve a decontamination factor of about 100 for fission products (after 3 wash cycles) . Washing is done with cooled water in the last stages of a multistage crystallizer. Although crystallization prior to extraction is not carried out in conventional PUREX, this modified process uses it to reduce the amount of aqueous and organic solution being processed. This addition provides process and economic benefits (e.g. less process fluid), as well as environmental benefits as the corresponding liquid waste requiring treatment and disposal is reduced. While it may be possible to achieve greater amounts of U crystallization with lower temperatures (i.e. ~99% at -30Â°C ), liquid-liquid extraction equipment (see Section: 2.1.4 Main Liquid-Liquid Extraction and Stripping) would still be required to achieve separation of the remaining actinides. The smaller temperature reduction for approximately 70% U crystallization maximizes the benefit of crystallization, without incurring unnecessary costs and complexities. Before proceeding to the main extraction stage the exiting dissolver solution must be clarified, also by centrifugation, to remove any suspended particles that may interfere in extraction. The collected particles are recycled for further dissolution. The dissolver solution then proceeds to a mixing tank for acidity and concentration adjustment, mainly to maintain the high concentration of nitric acid (approximately 5-6M) required for high efficiency separation in the main extraction stage . The UNH crystals which exit the crystallizer as a slurry, are gravity fed into a mixing tank for further processing (see section 2.1.7 Metal Formation for more details). The crystallizer selected is an annular, rotary, screw-type, continuous crystallization device, with a cooling jacket for temperature control. This design was selected due to its continuous operation and its similarity in functioning to the other screw-type equipment used in the overall process. The crystallizer was designed to operate at an incline to aid in nucleation and growth of crystals. This also facilitates easier removal of crystals without damage. As the internal spiral blades rotate they dislodge the crystals growing on the wall and move them upward to a centrifugal basket for washing and filtration, while the remaining dissolver solution flows downward. The rotary screw design, in a similar way to the rotary dissolver, also provides compartments or stages that allow for crystal washing to be done internally. See Appendix E-1 for schematic. One operational disadvantage of the crystallizer design is the potential for crystal accumulation and blockage of discharge streams. This can be mitigated by appropriate operation. For example, the screw rotation, dissolution feed and coolant feed can be adjusted to restrict new crystal growth or break down crystal agglomerates. Scrub acid may also be fed to decompose any crystal blockages  (see Section 3. for more details). 2.1.4 Main Liquid-Liquid Extraction and Stripping After crystallization, clarification and concentration, the dissolver solution is ready for the main liquid-liquid extraction and stripping. The solution is fed into the middle of a multistage centrifugal contactor to begin the separation of the remaining actinides from fission products.  M. Nakahara, Y. Sani, Y. Koma, M. Kamiya, A. Shibata, T. Koizumi and T. Koyama, "Separation of Actinide Elements by Solvent Extraction Using Centrifugal Contactors in the NEXT Process," J Nucl Sci Technol, vol. 44, no. 3, pp. 373-381, 2007.  J. D. Law, T. G. Garn, D. H. Meikrantz and J. Warburton, "Pilot-Scale TRUEX Flowsheet Testing For Separation of Actinides and Lanthanides from used Nuclear Fuel," Separation Science and Technology, vol. 45, pp. 1769-1775, 2010.  H. M. Mineo, H. Isogai, Y. Morita and G. Uchiyama, "An Investigation into Dissolution Rate of Spent Nuclear Fuel in Aqueous Reprocessing," J. Nucl. Sci. Technol., vol. 41, no. 2, pp. 126-134, 2004.  R. Jubin, "Spent Fuel Reprocessing," in Introduction to Nuclear Chemistry and Fuel Cycle Separations Course, Consortium for Risk Evaluation With Stakeholder Participation, Nashville, 2008.  T. Takata, Y. Koma, Sato, Koji, Kamiya, Masayoshi, A. Shibata, K. Nomura, H. Ogino, Koyama, Tomozo and S.-i. Aose, "Conceptual Design Study on Advanced Aqueous Reprocessing System for Fast Reactor Fuel Cycle," J Nucl Sci Technol, vol. 41, no. 3, pp. 307-314, 2004.  K. Ohyama and K. Nomura, "Development of Uranium Crystallization System in â€œNEXTâ€ﾝ Reprocessing Process," in Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems, Boise, 2007. Â Appendix D: Crystallization of Uranium in Nitric Acid Data The following graph depicts the solubility of uranium in nitric acid, as a function of uranium and nitric acid concentration, plus temperature. Moving from right to left, the uranium concentration decreases with temperature before reaching a minimum point, where water and nitric acid crystallize. To avoid this co-crystallization, a crystallizer must operate to the right of this point. The graph shows that as the concentration of nitric acid increases, the minimum point shifts from right to left and a higher UNH crystallization is possible. To maximize crystallization, the feed uranium and nitric acid concentration should be maximized as well . Figure 3 Solubility Curves of Uranyl Nitrate The above equilibrium data is supported by experimental data. Crystallization tests at high uranium concentration (300-600 g/l) and high nitric acid concentration (4-6M) to simulate spent fuel dissolver solution have yielded 70-80% uranium recovery at 10-20Â°C, and up to 95% at -10Â°C . The results for a run with 500g U/l and 5M nitric acid are summarized below: Figure 4 Uranium and nitric acid concentration in mother liquor (left) and recovery of uranium (right) Appendix E-1: Equipment Diagrams The following appendix features schematic diagrams of certain non-standard key equipment to aid in the visualization and understanding of their functioning, as described in Section 2.1 of the Process Description, and also as discussed in the non-key unit sizing section. Figure 5 Fuel Bundle Shearer  Figure 6 Continuous Rotary Dissolver  Figure 7 Rotary Crystallizer Crystallizer is scaled down from a tested device, having a feed rate of 1380L/hr and a residence time of 1 hour . The dimensions of internal rotary cylinder with the blades of the tested crystallizer device are 11 centimeters of internal diameter and 50 centimeters by length . Since the amount of uranium crystals to be dealt with is much less, the dimensions are rescaled by reducing the original volume of 0.00475m3 by half to 0.002375m3 in order to design for smaller blades. Cost Table Appendix T: Works Cited for Appendices  Advamacs, "Concentration Calculator," [Online]. Available: http://www.trimen.pl/witek/calculators/stezenia.html. [Accessed November 2012].  D. P. Jackson, "NWMO Background Papers: Technical Methods; Status of Nuclear Fuel Reprocessing, Partitioning, and Transmutation," NWMO, 2003.  D. Hart and D. Lush, "THE CHEMICAL TOXICITY POTENTIAL OF CANDU SPENT FUEL," NWMO, 2004.  L. Johnson and J. Tait, "Source terms for 36Cl in the assessment of used fuel disposal," Atomic Energy of Canada Limited Technical Record, 1992.  J. Tait, I. Gauld and G. Wilkin, "Derivation of Initial Radionuclide Inventories for the Saftey Assessment of the Disposal of Used CANDU Fuel," Atomic Energy of Canada Limited Report, 1989.  G. Faure, Principles and Applications of Geochemistry, Upper Saddle River: Prentic Hall Inc., 1998.  R. Hart and G. Morris, "Crystallization temperatures of uranyl ntirate-nitric acid solutions," Prog. Nucl. Energy III, p. 544, 1958.  T. Chikazawa, T. Kikuchi, A. Shibata, T. Koyama and S. Homma, "Batch Crystallization of Uranyl Nitrate," J. Nucl. Sci. Technol., vol. 45, no. 6, pp. 582-587, 2008.  T. Todd, "Spent Nuclear Fuel Reprocessing," in Nuclear Regulatory Commission Seminar, Rockville, 2008.  R. Jubin, "Used Fuel Reprocessing," in CRESP Nuclear Fuel Cycle Course, 2011.  Rousselet-Robatel, "ROUSSELET ROBATEL MODEL LX MULTISTAGE CENTRIFUGAL EXTRACTOR," [Online]. Available: http://www.rousselet-robatel.com/products/pdfs/multistage-centrifugal-extractor-operatingprinciple. pdf. [Accessed 2012].  P. A. Haas, R. D. Arthur and S. W.B., "Development of Thermal Denitration to Prepare Uranium Oxide and Mixed Oxides for Nuclear Fuel Fabrication," Oak Ridge National Laboratory, Oak Ridge, 1981.  S. Jeong, S. Park, S. Honge, S. C.S. and S. Park, "Electrolytic production of metallic uranium from U3O8 in a 20kg batch scale reactor," J. Radioanal. Nucl. Chem., vol. 268, no. 2, pp. 349-356, 2006.  T. Takata, Y. Koma, Sato, Koji, Kamiya, Masayoshi, A. Shibata, K. Nomura, H. Ogino, Koyama, Tomozo and S.-i. Aose, "Conceptual Design Study on Advanced Aqueous Reprocessing System for Fast Reactor Fuel Cycle," J Nucl Sci Technol, vol. 41, no. 3, pp. 307-314, 2004.  R. Herbst and M. Nilsson, "Standard and advanced separation: PUREX processes for nuclear fuel reprocessing," in Advanced Separation Techniques for Nuclear Fuel Reprocessing and Radioactive Waste Treatment, 2011, pp. 141-173. This web page last updated June 21, 2021 A^4 B^2 I Tsors. Â
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