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XYLENE POWER LTD.

TRU CONCENTRATION

By Charles Rhodes, P.Eng., Ph.D.

INTRODUCTION:
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 DEFINITION:
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:
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).
 

MOTIVATION:
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.
 

APPLICATION:
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.
 

OVERVIEW:
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.

The fission 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.
 

The 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:

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.

The smaller 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, monica.regalbuto@inl.gov, Purex expert at INL).
 

THE NEPTUNIUM ISSUE:
For chemical reasons the neptunium is not excluded from UO2 during the uranyl nitrate hexahydrate recrystallization process.
 

REFERENCES:
For an overview of the Ottensmeyer Plan please review OTTENSMEYER PLAN.

For an overview of nuclear fuel waste processing see the paper:
Radioactive Waste Partitioning and Transmutation.
Reference: Japanese 2002 patent

For an overview of Uranyl nitrate hexahydrate [UO2(NO3)2.6H2O] solubility see: Uranyl Nitrate Solubility
and
Uranyl nitrate hexahydrate solubility in nitric acid and its crystallization selectivity in the presence of nitrate salts
and
Uranyl nitrate hexahydrate solubility

A paper on decontamination factors actually realized is:  Enhancement of Decontamination Performance of Impurities for Uranyl Nitrate Hexahydrate

A paper relevant to the U-232 issue is:  Uranium-232 Production In Current Design LWRs
 

 

FUEL 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). 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.

 

The operations that need to be performed at each major power reactor site include:
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.
 

 

UO2(NO3)2 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

The imporance 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.

Email from Peter:
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.
 

 

RECRYSTALLIZATION 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.

The crystals 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.

Now assume 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)

The impurity 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.

 

A detailed description of the design and operation of the Uranyl Nitrate Hexahydrate automatic crystal growing chamber is set out at:
Uranyl Nitrate Hexahydrate Crystal Growth
 

RECRYSTALLIZATION PROCESS CONSTRAINTS:
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.

 

CHEMISTRY:
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
 

 

DEFINITIONS:
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.
 

 

HYDRATION OF UO2(NO3)2:
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.  

 

Typically in real life a single stage of separation gives:
A ~ 0.95
and
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)
 

 

CASCADE OVERVIEW:
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.
 

The equipment 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 Tnb.
 

CASCADE DESIGN:
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:
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;
 

CASCADE FORMATION:

CASCADE OPERATION:
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.
 

CASCADE MATHEMATICS:
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]
where:
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]

FLOWS
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