TRITIUM ISOTOPE SEPARATION
Aleea Tarnita, Nr. 7, Apt. 11
The substitution of an atom with an isotope of the same element in a molecular
species causes variation of most physical and chemical properties of the substances, as a
consequence of the corresponding mass variation. All the processes employed for isotope
separation purposes take advantage of differences in the behaviour of isotopic molecular
species. Under a generic term, all these differences are named isotopic effects.
A generalized enrichment element can be treated as a black box into which flows
material of a certain isotopic composition and out of which flow two streams, one
containing a higher percentage and the other a lower percentage of the desired isotope
than was present in the feed stream. The material with the higher percentage is generally
called product, and that with the lower percentage is called waste or tails. It should be
kept in mind that the words feed, product and tails can be used to refer to the inputs and
outputs of either a single element or a full cascade.
Obviously, the mass variety can be expressed, in any separation process, in an
individual behaviour diversity of the isotopic molecules, wich can lead to a nearly total
isotopic separation in a single separation unit. Frequently, the mass diversity can be
expressed in a diversity of statistical behaviour, materialized by a very small elementary
separation effect. Therefore the elementary process must be repeated several times to
obtain the desired isotopic concentration of the product.
Although isotope separation processes are based on fundamentally different
physical principles, the aspects connected with the multiplication of the elementary
separation effect can be tackled, ignoring the details of any particular process taken into
consideration. However, what is very important for any separation process is the
separation factor, a parameter which defines the variation of concentration of the
desired isotope after the separation was accomplished in a separation element, stage or
The separation and control of tritium is of increasing concern to the power reactor
industry. Leakage of fission tritium produced from chemical shims, from control rods,
from n-capture by water, from impurities in construction materials, etc., can bring the
oncentration up to the µCi/ml level in reactor coolants; this release to the environment is
a potentially serious adverse consequence of the extensive use of nuclear fission and
fusion power sources. The present trend appears to be toward zero release to the
environment. However, this approach substitutes an in-plant problem and only postpones
the environmental one: either an acceptable means must be found for releasing tritiated
water or a method for separation (concentration) and permanent storage must be
developed. The problem, of course, is also an important one for the fuel reprocessing
plants which handle 50 - 90% of the tritium produced in reactors.
The projected growth of thermal power reactor indicates that in about two decades
a production equal to the world's natural tritium inventory will be reached. In addition,
the development of fusion power will add greatly to the tritium inventory. Thus, it is
important to consider the development of simple, economic method to lower the tritium
concentration of process water to specified levels as well as to concentrate and store
highly enriched tritiated water .
To solve these problems, a series of separation processes have been studied and
applied for many years, but only some of these have been scaled at an industrial level.
At present, especially connected with the problem of recovery of tritium from
power reactors, in addition to active competition between some clasical processes,
advanced method based on laser techniques are being developed, with great hope in
This paper will present, shortly, the most significant processes and plants for
isotopic separation of tritium with particular emphasis on the analytical purposes
enrichment and recovery from nuclear and thermonuclear plants. A special section will
be assigned to laser isotope separation of tritium. More details can be found in our book
"TRITIUM ISOTOPE SEPARATION” published by CRC Press .
A. ENRICHMENT OF TRITIUM FOR ANALYTICAL PURPOSES
One possibility of measuring smaller tritium concentrations than direct counting
of the samples would allow, is artificially increasing tritium concentration of the samples
by increasing the specific activity through concentration of the water tritium content into
a smaller volume .
Processes used to achieve tritium isotopic enrichment, applied to water are:
electrolysis and distillation; those applied to hydrogen, prepared from water are:
thermal diffusion, distillation and gas-chromatography.
1. Water Electrolysis
The enrichment of tritium by electrolysis of water is a relatively simple method.
The process requires little sample handling and supervision during enrichment. The
simultaneous enrichment of a series of samples is easily feasible from a technical point of
view and has the advantage that standard and background samples can be run together
with the unknown samples under (almost) identical conditions.
The enriched effect of the process :
HTO HT + 1/2 O2
HT + H2O ↔ HTO + H2
is based on the higher precipitation velocity of the lighter hydrogen isotopes (protium and
deuterium) as compared to tritium which is reinforced by irreversible kinetic
effects with higher current densities (hydrogen overvoltage). With smaller current
densities the separation factor is determined solely by the equilibrium constant of the
exchange equilibrium, which adjusts itself to the catalytic electrode where the hydrogen
is in equilibrium with the surrounding water.
Hydrogen, produced by electrolysis of water, is depleted in the heavy isotopes (D
and T). Consequently, the tritium concentration in the residual water increases. If we are
able to express the ratio of final to initial tritium concentration, defined as enrichment
factor E, in measurable quantities, we can calculate the initial specific activity of the
sample by measuring the enriched water .
Apart from the enrichment factor, the recovery of tritium is of great importance.
The recovery and the required mass of the counting sample determine the initial mass of
the sample to be enriched, and thus the amount of water to be electrolysed.
A nominal enrichment run performed by Groeneveld is presented in Table 1 :
Table 1. Nominal enrichment run.
Sample charge: 246 g; NaOH solution added: 3 ml; water: 3 g and Na2O: 1 g; Current:
10 A; Time: 4,038 min; Cooling bath temperature: -2oC; Temperature of electrolyte: 0 8oC; Removed water: - by decomposition: 225 g; - by evaporation: 2.58 g; - by spray:
0.06 g; Volume concentration: 11.7; Enrichment factor for tritium: 10.7; Recovery for
Disadvantages of this process are the relatively high activity content in the
electrolyser outfit (furthermore occurs in the form of HTO, which is more radiotoxic than
HT), the high maintenance expenses, and high power consumption.
The process is the oldest large-scale industrially used process for hydrogen
isotopic separation. A series of promising research projects are being conducted with the
aim of reducing the mentioned disadvantages.
2. Water Distillation
Water distillation or water rectification exploits the vapour-liquid equilibrium ,
HTO(v) + H2O(l) ↔ H2O(v) + HTO(l)
by which tritium is enriched in the liquid phase. Normal operational parameters are 200
mbar low pressure and 60oC. Multiplication of the separation factor is carried out by a
counterflow of the steam or liquid phase in a rectifying column, the desired phase
reversal by condensation at the top, and revaporization at the bottom of the column.
Usually approximately 250 g of sample is distilled. The water recovery is 0.9992
and the loss of water 220 ± 140 mg. The loss in saturated steam at 100 oC in evaporation
flask and heated yoke would amount to 450 mg; the loss in saturated water vapour at
25oC amount to 15 mg .
The lower detection limit and the accuracy of tritium determinations have been
improved by electrolytic enrichment combined with distillation of the sample water.
Advantages of this process are robust construction o the equipment, manifold
experience gained in operation, high reliability and leakage safety caused by the vacuum.
In particular, there is no danger of a hydrogen explosion .
3. Thermal Diffusion
Besides electrolysis and distillation, thermal diffusion (TD) represents an
alternative technique for enrichment of tritium for analytical purposes. TD in isotopic
mixtures is of special interest from both a technical viewpoint because it is used for the
separation of isotopes [6-7] and from a theoretical viewpoint in view because of the close
correlation of this transport phenomenon with intermolecular forces . The essence of
this transport phenomenon is that in the presence of a temperature gradient in a binary or
multicomponent gaseous mixture, a transport of matter takes place thereby creating a
concentration gradient: the light component of the mixture concentrates in the region of a
higher temperature and the heavy component in the region of a lower temperature. The
magnitude of the elementary TD effect for a certain isotopic mixture is determined by the
value of the TD factor α T. The multiplication of the separation by TD is assured by
combining this effect with a convective curent in a TD column.
Separation of the tritium by TD represents one of the most important application of this
method. TD plants for analytical purposes have been operated by Almqvist et al. ,
Boorman and Kronberger , Robinson et al., , Schirdewahn et al., , Gonsior
, Verhagen and Sellschop [14-16], Buttlar and Wiik , Shimizu and Ravoire ,
Shimizu , Ravoire et al. , Verhagen , Lemarechal-Dupuis , Hugony et
4. Permeation Through Membranes
The monitoring of tritiated water vapour (HTD) in the presence of other airborne
radioactive species is difficult. A particularly important interfering species is tritiated
hydrogen gas (HT). The importance of differentiating between these two species is due to
their different radiological hazard. HTO is readily absorbed in the human body whereas
HT is not. A recent estimation of ICRP shows that HTO is 2.5 x 10 4 times as hazardeous
as HT . Both of these species could be present together within a plant for the removal
of tritium from a heavy water reactor or in the vicinity of a fusion reactor.
To provide adequate protection for workers and at the same time to avoid being
overly conservative, it is necessary to monitor separately the concentrations of these two
species in the air. Ideally, such a monitor should have a fast response with a signal in the
HTO channel because the presence of HT in the input being less than 1 part in 2.5 x 10 4
of the signal in the HT channel. The HTO contribution to the indicated HT concentration
is not important from the viewpoint.of a radiation protection. However, to assess the
prevailing conditions accurately, this contribution should also be small.
In order to meet these requirements, Osborne and McElroy [25-26] have studied
the possibility of using the permeation through membranes to separate and monitor
separately the HTO and HT species. Later, a prototype monitor that discriminates
between tritiated water (HTO) and tritiated hydrogen (HT) in the air was constructed
. The monitor was designed to measure tritium in air concentration up to 10 6
MBq/m3. In terms of derived air concentration, the range is up to 10 6 for HTO and 50 for
5. Adsorption and Chromatography
Vapor pressure differences can be realized and sometimes magnified by
adsorption of molecules on an active solid surface. Typical adsorbents are charcoal,
alumina, silica and molecular sieves. Each has some potential for use with H 2 at low
temperatures. In general the heavier isotope has a greater heat of adsorption and this can
utilize less active sieves. This should lead to an enhancement of separation factors. This
is confirmed by the data for H2O-HDO on charcoal . Although the method is simple
there are many difficulties in designing a large plant. The amount of water adsorbed per
gram of charcoal is small.
Adsorption is conveniently studied by gas-chromatography (GC) methods. One of
the first studies used palladium black . Other packing materials used are activated
alumina, silica and molecular sieves . Excellent separations can be achieved using
these packing materials because of a reasonably high separation factor value and the very
large number of theoretical plates.
The separation factors found on adsorption-desorption are significantly greater
than that given by vapor pressure ratios, thus implying an adsorption sieve selectivity.
Although adsorption-desorption processes has been shown to be successful, the
only extensive data appear to be on the systems H 2-palladium and water-charcoal. The
latter process would seem worthy of consideration for the tritium separation problem. A
vapour pressure ratio P(H2O)/P(HTO) of about 2 might be expected at 30 oC. It may turn
out, however, that the charcoal bed volume would be prohibitively large.
GC is usually considered an analytical type of operation and is essentially a batch
process. However, it can be made semi-continuous by cyclic operation or by a movingbed process.
B. RECOVERY AND ENRICHMENT OF TRITIUM FROM NUCLEAR PLANTS
The control of tritium is of increasing concern to the power reactor industry.
Leakage of fission tritium and production from chemical shims, from control rods, from
n-capture by water as well as from impurities in construction materials can bring the
concentration up to the µCi/ml level in reactor coolant; the release of this into the
environment is a potentially serious adverse consequence of the extensive use of nuclear
fission and fusion power sources. The present trend appears to be toward zero release into
the environment. However, this approach substitute an in-plant problem and only
postpones the environmental one; either an acceptable means must be found for releasing
tritiated water or a method for separation (concentration) and permanent storage must to
be developed. The problem, of course, is also an important one for the fuel reprocessing
plants which handle 50-90 % of the tritium produced in reactors.
The projected growth of thermal power reactors indicates that in the coming
decades a production equal to the world's natural tritium inventory will be reached. In
addition, the development of fusion power will add greatly to the tritium inventory. Thus,
it is important to consider the development of simple, economical methods to lower the
tritium concentration of process water to specified levels as well as to concentrate and
store highly enriched tritiated water .
An alternative method of control of tritium in reactor systems and reprocessing
plants is to separate the tritium and to concentrate it to a small volume that can be
economically stored or eliminated. In these systems the concentrations of tritium are very
low, with mole ratios of DTO/D2O of 10-5 in HWRs and HTO/H2O of 10-7 in reprocessing
plants . Thus, in the case of CANDU HWR tritium is produced in the moderator and
coolant circuits through neutron absorption by the deuterium atoms in heavy water. The
concentration of tritium, in the form of DTO molecules builds up slowly with time in
reactor operations. The mole ratio of DTO/D 2O after 12 years of operation of the
Pickering reactors is about 2.0 x 10-5 (30 Ci/kg) in the moderator and less than 1 x 10 -6 (2
Ci/kg) in the coolant. The tritium equilibrium concentration (at which the production rate
is balanced by the decay rate) is between 4 and 5 x 10 -5 mol of DTO per mol of D2O (50
to 75 Ci/kg) . Even though the quantities of water containing tritium may be large,
the quantities of tritium are very small.
The general objective of tritium separation and enrichment is to produce a stream
sufficiently enriched in tritium to be economical for storage or disposal and a stream
sufficiently depleted in tritium that it can be either recycled to the reactor system or
discharged directly into the environment.
A number of hydrogen isotope separation processes have been developed for the
production of heavy water . In general, these can be applied to tritium separation. The
tritium mass is higher than deuterium mass; the zero point energies are lower, leading to
greater separation factors. An additional factor is the radiation energy associated with the
decay which tends to promote chemical reactions . These processes include water
distillation, hydrogen distillation, electrolysis and chemical exchange of hydrogen
isotopes between hydrogen and water. The isotopic separation procedures presently
available for tritium are summarized in Table 2 [l]:
Table 2. Procedures for tritium separation.
Diffusion and membranes processes:
ca. 1.6 2.
Ultracentrifuge and separation nozzle:
H2S, NH3 and methylamine exchange processes:
0.3 - 3
Rectification (distillation) as HTO:
a. in the gas phase as HT:
b. in the vapour phase as HTO:
c. in the liquid phase as HTO:
2. Electrolysis as HTO:
3. Low-temperature rectification (cryogenic distillation) as HT: 1.80
The power consumption of all procedures is extremely high. Procedures 1 to 5 are
unacceptable from today's point of view since they are technically too complicated,
are unsuccessful for reasons of nuclear safety, are insufficiently developed, or cannot
economically be adapted in size. An exception is laser excitation which has the
greatest development potential of all methods.Procedures 6 to 9 are the most suitable
from today's standpoint. In general:
- The separation factor decreases with increasing temperature
- The tritium isotope preferably passes over into the liquid phase
- The lower the temperature, the greater the tendency of the tritium to pass over
into the liquid phase
1. Water Distillation
Water distillation, (WD), exploits the vapour-liquid equilibrium,
HTO(v) + H2O (l) ↔ H2O(v) + HTO (l),
by which tritium is enriched in the liquid phase. The separation factors for distillation
can be taken approximately as the ratio of the vapour pressures. Boiling point data
and vapor pressures ratio for isotopic species of interest are given in Table 3 
Table 3. Boiling point and vapour pressure of oxides of hydrogen isotopes.
Boiling point ( C):
(mmHg), at 25oC
Water pressure ratios for isotopic water species P(H2O)/P(x) calculated using
Van Hook's equation and data  for three temperatures are given in Table :
Table 4. Vapour pressure ratios for isotopic water species P(H2O)/P(x).
The measured liquid/vapour T ratios of Sepall and Mason  gave 1.1 at
25oC, 1.o6 at 50oC and 1.03 at 90oC.
Normal operational parameters are 200 mbar low pressure and 60oC.
Multiplication of the separation factor is carried out by a counterflow of the steam or
liquid phase in a rectifying column, desired phase reversal by condensation at the top
and revaporization at the bottom of the column.
Advantages of this process are the robust construction of the equipment, the
manifold experience gained in operation, high reliability, and leakage safety caused
by the vacuum. No danger of a hydrogen explosion.
Disadvantages are the size of the plant, which results from the necessary large
number of exchange plates, and the high power consumption. Thus, if the distillation
were carried out at about 60oC, for which the separation factor is α = 1.05, about 400
stages would be required to reduce tritium content from 10 µCi/ml to 3 x 10 -3 µCi/ml.
If the separation factor for the process were α = 1.5, only 50 stages would be
necessary . This constitutes the stripping, or decontamination section of the plant.
The concentrated tritium would then pass into the enrichment section. This section
rapidly becomes smaller in size and it is here that alternate methods might be utilized.
Current utilization is as a "finisher" in the final enrichment of heavy water and as an
"upgrader" to remove light water impurities from the moderator or heat exchange
system of HWR. It also enriches tritium in more toxic aqueous form compared with
the hydrogen form. It is a simple well- established technology and an installation built
for a detritiation plant based on WD has been considered for the Orphée reactor in
Sulzer Brothers Ltd.  has many years of experience in both H 2O/HTO vacuum
distillation of water and D2O/DTO low-temperature distillation of hydrogen. Since
1953, 35 plants have been supplied for the vacuum distillation of H2O/D2O and two
plants for low-temperature distillation of H 2/HD and HD/D2/DT. The size of a tritium
extraction plant from light water of a pressurized water reactor (PWR) by vacuum
rectification may be illustrated with the following example:
- Tritium production: 0.1 Ci/h (800 Ci/year)
- Tritium emission to environment: 0.01 Ci/h (80 Ci/year)
The tritium emission to the environment represents the total losses from the plant.
The operating data of a Sulzer plant for tritium extraction from light water for
720 Ci/year, based on 1 Ci/m3 in the reactor cooling water are given in Table 5 .
Table 5. Operating data of a Sulzer plant for tritium extraction from light water.
Operating pressure at top of column:
A tritium emission to the environment of 80 Ci/year with a tritium
concentration of 1 Ci/m3 in the reactor cooling water would allow an admissible
leakage rate of 80 m3/year (0.01 m3/h).
Features of Sulzer tritium extraction plant include
- Small dimensions
- Small hold-up allowing speedy restarting after interruptions
- Low steam consumption, only low-pressure steam needed
- Simple operation, little maintenance
Product-swept parts made of stainless steel, Sulzer packing of copper
No escape of tritium into building, because most part of the plant are
When designing a tritium extraction plant, it must be noted that the plant size
is proportional to the tritium production (i. e. to the number of Ci to be accumulated
nnually in the concentrate) and inversely proportional to the tritium concentration in
the feed (reactor water).
We can conclude that among various methods for separation of hydrogen
isotopes, WD has attractive possibilities in spite of its small separation factor because
the process has inherent advantages, such as :
- Simplicity of operation
- Absence of corrosive and toxic compound H2S (used in dual-temperature
water - H2S exchange process for heavy water production)
- Absence of water-hydrogen gas conversion by electrolysis (in the waterhydrogen isotope exchange process)
- Absence of catalyst for isotope exchange
WD has been, therefore, applied to the detritiation of drainage from a nuclear
fuel reprocessing plant, and will be adopted for the volume reduction of tritiated
water, based on tritium enrichment, from the cooling and safety systems of a nuclear
2. Hydrogen Distillation
Unlike the isotopes of other elements, the relatively large mass differences of H,
D, and T cause appreciable differences in properties of these isotopes and their
compounds. Table 6 lists some physical properties of the isotopic forms of hydrogen
Table 6. Physical properties of isotopic hydrogen species.
4.025 4.029 5.032 6.034
Boiling point (K):
22.92 23.67 24.38 25.04
Triple point (K):
17.62 18.73 19.71 20.62
Triple point pressure (mmHg): 54.0
109.5 128.6 145.7 162.0
Critical tempe rature (K):
37.13 38.35 39.42 40.44
Critical pressure (mmHg):
9,736 11,134 11,780 12,487 13,300 13,878
Dissociation energy (eV):
Zero point energy (cm ):
*/ The normal state (high temperature ortho-para composition) is used for this table.
The "equilibrium" state, the composition existing at the normal boiling point gives
slightly different properties.
From Table 6 it is seen that the boiling point of T2 is 25 K and of HT is 22.9 K
compared to 20.4 K for H2. Thus, as with deuterium,distillation of the elemental form
is a practical scheme. Similar to WD, the low-temperature rectification or cryogenic
distillation (CD) utilizes the liquid-gas equilibrium in the temperature range of liquid
nitrogen (20 to 30 K). A refrigeration machine with He circulation should be used for
Advantages of the procedure are the wide range of operating experiences and
the high plant availabilities. Hydrogen distillation has high separation factors and is
an established technology; and even though high refrigeration costs are incurred by
operation at liquid hydrogen temperatures, this is considered the best process for
tritium enrichment. It is used for the tritium recovery plant on the high flux reactor of
the Institut Max von Laue-Paul Langevin, Grenoble, France .The separation is
carried out in columns which are installed in cold boxes (Dewar vessels) with a
coresponding barrier effect.
Disadvantages are the need for intensive preliminary gas purification, i. e.
catalytic combustion of the residual oxygen and drying of the hydrogen flow, as well
as the retention of the contained nitrogen on regenerable molecular sieves.
To use hydrogen distillation to recover tritium from water streams, it is
necessary to couple it with a front-end transfer process that transfers the tritium from
the water molecule to the hydrogen molecule. For heavy water application, the
reaction is 
DTO + D2 D2O + DT
Tritium recovery for HWR generally would consist of two separate processes:
a transfer process that moves tritium from tritiated heavy water to tritiated hydrogen
and a concentration process that separates and enriches the tritium to high specific
activity tritiated hydrogen (T2) .
Since it is most economical to produce a small and highly concentrated tritium
stream for storage, CD of D2/DT/T2 is the most practical process available for the
final enrichment of tritium. The options of the transfer process are :
- Vapour-phase catalytic exchange (VPCE) of tritiated D2O vapour with D2,
producing a D2/DT mixture 
- Direct electrolysis (DE) of tritiated D2O, producing a D2/DT mixture
- Combined electrolysis-chemical exchange CECE  of the tritiated D2O in
the liquid phase with D2/DT producing a D2/DT stream enriched in tritium
- Liquid-phase catalytic exchange LPCE  of tritiated D 2O with D2
producing a D2/DT mixture
Each of these transfer processes may be coupled with CD of hydrogen
isotopes to provide a process for tritium recovery from heavy water.
3. Water Electrolysis
Water electrolysis, (WE), has long been used to concentrate small quantities
of tritium for analyses. Electrolysis has a high separation factor. The tritium is
concentrated in the electrolyte with the hydrogen depleted in the tritium. Several
stages of electrolysis are required to achieve the required tritium separation and
enrichment, and energy costs are high. There is also an absence of commercial cell
designs that are sufficiently leak-free to enable processing the tritium- enriched water.
The electrolysis technology is well developed, the process is probably the
simplest of all, the separation factor is the highest, the capital costs would be small,
and there is no doubt that an electrolytic separation plant would do the job .
5. Chemical Exchange
Depending on the physicochemical form of tritium at the start of the chemical
exchange reaction, three procedures are possible :
HT(g) + H2O(l) ↔ H2(g) + HTO(l),
HTO(v) + H2(g) ↔ H2O(v) + HT(g),
HTO(l) + H2(g) ↔ H2O(l) + HT(g).
In order to achieve the minimum velocities necessary to obtain the chemical
equilibria which are required for technical utilization, it is necessary to use
catalyzers.These have to be hydrophobic wherever an aqueous phase may prevent or
impair the catalytic process. Whereas suspension catalysers have proved impractical,
the use of fixed bed catalyzers (e.g., Pt/active carbon/polytetrafluoroethylene [PTFE],
in Raschig ring form, compressed at 2 bar and sintered at 300 oC) has proved
For exchange in the vapour phase, a "quasi-counterflow principle" can be
realized in each separation stage by renewed vaporization, exchange, and
condensation (phase separation).
Table 7 shows typical differences in equilibrium constants between deuterium
and tritium . In this case the exchange is between water and HD or HT.
Table 7. Equilibrium constants for the hydrogen-water exchange reaction
HD + H2O
HT + H2O
Experimental data agree well with theory . Thus, those methods which
were shown to be best for deuterium will be excellent for tritium. At this point,
however, we will modify the criteria and make the assumptions that we are dealing
with large volumes of water containing a very low concentration of tritium, and that
we are interested primarily in water decontamination. A secondary goal will be to
concentrate the tritium removed.
Because of the large volumes of water to be treated and the low concentration
of tritium, about 10 µCi/ml will be assumed for reactor coolant. This level is about
3,000 times the maximum permissible concentration (MPC) for water. Consequently,
a high separation factor is desirable; and in this context, chemical exchange is a good
Advantages of catalytic exchange procedures are the small content, compact
construction and robust technology. A disadvantage is the as yet unproven long-term
performance of hydrophobic catalysts. Although at different levels of development,
all three procedures are already at the industrial stage .
Catalyzed chemical exchange always appear coupled with an another process.
The particular aspects connected with this method will be discussed below, in
connection with the description of each of the processes of major interest in the
5. Vapour-Phase CatalyticExchange(VPCE) and Cryogenic Distillation (CD)
This process was developed by the French Commisariat à l'Energie Atomique,
(CEA). Sulzer Brother Limited of Switzerland is licensed to market the process
commercially. The high flux reactor in Grenoble (France) which was built as a result
of Franco-German cooperation, is the first reactor installation in the world possessing
a plant in operation for the simultaneous extraction of hydrogen and tritium from
Tritium is formed here by neutron absorption into heavy water which
surrounds the fuel element. Without removal of tritium, the heavy water would reach
a tritium content of about 84,000 Ci/m3 (84 Ci/l), whereas with tritium extraction a
value of about 1,700 Ci/m3 (1.7 Ci/l) can be maintained. This requires an annual
extraction of 16,000 Ci, equivalent to about 60 nl of pure tritium.
The Grenoble tritium and hydrogen extraction plant is made up of two main
a).Catalytic exchange between heavy water vapor and deuterium gas at 200oC
and ca 1.2 bar, which allows the mass transfer of hydrogen and tritium from heavy
water to deuterium to reach equilibrium according to the following reactions ,
DTO(v) + D2(g) ↔ D2O(v) + DT(g)
HDO(v) + D2(g) ↔ D2O(v) + HD(g)
KT = 0.82
KH = 1.78
b).Low temperature rectification of the hydrogen isotopes HD/D2/DT/T2 at ca
1.5 bar in two column with extraction of HD at the top of the first column and pure
tritium at the bottom of the second column.
About 240 separation stages are present in the total height of the plant (15 m).
These separation stages concentrate the tritium to 99% at the bottom and the
hydrogen to 4o% at the head of the column. The product at the head is led off as gas
and burned in a burner to form water. The radioactive product at the bottom of the
column is led away in the gaseous form and stored in containers by means of a special
Operating data and operating media of the Grenoble plant are given in Table 8
Table 8. Operating data and operating media of the Grenoble tritium and
hydrogen extraction plant.
standard litres T2/year
Electric power input:
Cooling water consumption:
Compressed air consumption:
-----------------------------------------------------------------------------------------------------All the portions of the plant which come in contact with heavy water or heavy
water vapor are made from stainless steel.The evaporators are made from Monel alloy
in order to avoid the possibility of corrosion, since the heavy water can slightly
acidic. Welding of the stainless steel was carried out using the tungsten inert gas
[TIG], (i.e. argon arc welding protected from the air by means of a shielding inert
gas) process. The welded parts were subjected to surface crack detection and
radiographic testing. All components of the plant (vessels, valves) were pressure and
vacuum tested, both individually and after assembly. The number of flange
connections was kept small, in order to eliminate the danger of leaks as far as
possible. The seals used consist of two prestressed stainless steel annular discs,
between which an aluminium fillet is clamped. Lengthy and detailed tests, conducted
at extreme temperatures and temperature variations, have proved these seals to be
More stringent safety measures were necessary, because of radioactivity of the
heavy water and of the deuterium, together with the explosion danger associated with
The first tests with hydrogen began in August 1971. Start-up with deuterium and
nontritiated heavy water followed quickly and separation performances between
hydrogen and deuterium were controlled. It was only in August 1972 that the first
introduction of tritiated heavy water from the high flux reactor was treated. By July
1973 the accumulation of tritium only reached 5,000 Ci. By the end of 1976 there was
a tritium hold-up of 220,000 Ci in the plant and 230,000 litres of heavy water had
been treated. About 85,000 Ci of pure tritium with an average molar content of 98%
had been extracted in the gaseous phase .
Plant operation is semiautomatic. Personnel are required to carry out various
adjustments and maintenance concerned with the movement of heavy water and
regeneration of the adsorbers and carrying out various analyses. During working
hours, operation is ensured by four persons; outside working hours required
supervision is transferred to the reactor main control room. At a stop, plant goes
automatically into a safety position. Operation of the plant is continuous, start-up lasts
about 12 h and another 3 or 4 d are required to reach steady state conditions before
treating tritiated heavy water
The VPCE-CD process is not suitable for reconcentrating large quantities of
strong diluted D2O in one operation, because only a slight depletion of T 2 and H2 can
be achieved with three stage catalytic exchange. The size of a VPCE-CD plant will be
dictated by the tritium concentration to be maintained and the tritium production. The
lower the admissible tritium concentration and the bigger the tritium are, the bigger
the plant will be . With this process concentrated tritium is handled in its
elemental state and the maximum tritiated water concentration handled is that of the
heavy water feed .
The VPCE-CD process has been verified successfully, although is operated at
high temperature, each stage requiring the evaporation, superheating, vapour-phase
exchange, and condensation of the feedwater, it remained in attention as an important
process. Thus, a decision was reached around 1980 by Ontario Hydro to construct a
Tritium Removal Facility (TRF) in Canada, at the Darlington generating station, near
Toronto. This plant would reduce the tritium levels in the moderator and heat
transport systems of both the Pickering, Darlington and possibly the Bruce generating
stations. Detritiation at Pickering or Bruce would be accomplished by transporting
heavy water to and from the TRF .
The TRF is also based on a VPCE front-end process followed by
concentration through CD. The contract for TRF, consisting of the VPCE front-end,
feed treatment and CD, was awarded to Sulzer Canada Inc. The Ontario Hydro TRF
will produce 800 g/yr of tritium as T 2 at equilibrium, with a feed flow of 360 kg/h. In
comparison, the Grenoble plant has a feed rate of 25 kg/h.
6. Liquid-Phase Catalytic Exchange (LPCE) and Cryogenic Distillation
The discovery, at Chalk River Nuclear Laboratories (CRNL), Canada, of a
simple method of wetproofing platinum catalysts has stimulated an extensive research
program for the development of hydrogen isotope separation and hydrogen/oxygen
recombination processes. Over 14 years of study has resulted in highly efficient
catalysts which retain their activity while immersed in liquid water for periods of
more 3 years . Originally conceived for the separation of deuterium from ordinary
water, these catalysts now make feasible a wide range of processes in which water is
a reactant, coolant, product or a by- product, for which conventional catalysts are
The catalysts, for example, as platinum or palladium on carbon, silica, or
alumina are rendered "wet-proof" by a special process in which a Teflon or silicone
coating is applied to the surface. The Teflon provides a hydrophobic surface that
repels liquid water, but is permeable to gas and vapor, thus enabling isotope exchange
to take place in the presence of a liquid. Wetproofed catalysts are therefore capable of
achieving reaction rates comparable to those for conventional catalysts operating in
The current generation of catalysts which has been tested at temperatures up
to 250oC, have a demonstrated lifetime of at least three years, are easily regenerated
by heating in air at 150oC for a few hours and have now reached the stage of
The detritiation mechanism of the LPCE front end is similar to that of the
VPCE front end process except that the tritiated heavy water feed remains in the
liquid phase . In this process, during the first stage, tritium is transferred by
catalytic isotope exchange from the tritiated water into deuterium gas and the tritiumdepleted water is returned for reuse in the reactor. The deuterium gas containing
tritium then passes to a second stage, a CD system, where tritium is separated from
protium and deuterium. The tritium-depleted deuterium returns to the catalytic
exchange stage and tritium is collected in a separated vesel. The heart of the process
is a novel wet-proofed catalyst (of platinum crystallites deposited on carbon powder),
bounded to 6-mm diameter inert alumina spheres with polytetrafluoroethylene
The LPCE process is simpler and require less energy because the reaction is
conducted at ambient temperature in a single packed column . This proces has
therefore been chosen for CRNL's proposed demonstration plant for tritium recovery
from heavy water. The decision to build the CRNL tritium extraction plant (TEP) was
preceded by laboratory and pilot plant studies. A pilot plant was built at CRNL in
1979 to demonstrate LPCE process, to measure catalyst activity for hydrogen isotope
exchange and to demonstrate the lifetime performance of the catalyst. The tests were
carried out to determine catalyst activity under typical TEP conditions and provided
essential design data 
The CRNL TEP has been built to remove tritium from Atomic Energy of
Canada Ltd (AECL) heavy water, to reduce operator dose and to reduce emmisions.
A supplementary purpose of the plant is to demonstrate, in a full-scale environment,
LPCE-CD process technology . A future option to increase plant capacity is
conversion of the LPCE front end to a Combined Electrolysis Catalytic Exchange
7. Combined Electrolysis Catalytic Exchange (CECE) and Cryogenic
The CECE process has been developed by AECL at CRNL . It makes use
of a hydrophobic catalyst in a liquid phase exchange column combined with an
electrolytic cell to provide the hydrogen.
Tritiated heavy water from the reactor units is fed to the catalytic column and
allowed to flow downward countercurrently to a rising stream a D 2/DT gas generated
in electrolytic cells. Tritium moves from the gaseous D 2/DT stream to the liquid
D2O/DTO stream :
DT + D2O DTO + D2
The liquid stream enriched in tritium is collected in the electrolytic cells at the
bottom of the catalytic column. A portion of the concentrated DT/D 2 gas from the
electrolytic cells is directed to the CD systems for tritium extraction. Since deuterium
gas evolves preferentially to tritium in the electrolytic cell, the cell solution becomes
enriched with tritium.
The deuterium gas stream depleted in tritium from the CD is oxidised by the
oxygen stream from the electrolytic cells in a recombiner or burner to form detritiated
heavy water which is returned to the reactor units
This process is particularly attractive for the treatment of the very dilute
tritium streams in light water from reprocessing plant because of favorable separation
factors and its ability to concentrate the tritium .
It must be noted that CECE process is more complex than LPCE process and
requires handling of tritiated water at a concentration substantially higher than the
feed concentration from the reactor. In applications of tritium recovery from water
containing very low tritium concentrations this enriching feature of the CECE system
is considered very advantageous because of its large inherent H/T separation factors
and mild operating conditions. Till now, both the LPCE and CECE schemes have
been evaluated only on the laboratory scale . Development work is underway at
AECL to demonstrate these on a pilot plant scale.
At Chalk River is considered that for a given sized CD system, the tritium
extraction capacity may be increased by converting the LPCE process to the CECE
tritium recovery process, (CECE-TRP). Such a conversion is planned for CRNL
LPCE demonstration plant after the LPCE process has been demonstrated adequately.
For this application, the electrolysis cells must be designed to minimize leakage of
both tritiated heavy water and deuterium-tritium gas for occupational safety as well as
economic reasons .
8. Direct Electrolysis, (DE) and Cryogenic Distillation (CD)
Tritiated heavy water from the reactor units is fed to the electrolytic cells
where it is decomposed into an oxygen gas and a D2/DT gas stream 
2DTO 2DT + O2
2D2O 2D2 + O2
The D2/DT stream is purified and fed to the CD system, as in the VPCE
process. The deuterium gas is recombined with the oxygen generated in the
electrolytic cells to form detritiated heavy water which is then returned to the reactor
Electrolysis has a high separation factor. The tritium is concentrated in the
electrolyte. Several stages of electrolysis are required to achieve the required tritium
separation and enrichment, and energy costs are high. There is also an absence of
commercial cell designs that are sufficiently leak-free to enable processing of tritium
enriched water .
9. Combined Electrolysis Catalytic Exchange, (CECE)
Tritium recovery from light water streams from reactors and plants operated
by the Department of Energy contractors in the U.S. is being demonstrated under a
cooperative program  on a laboratory scale at the Mound Laboratory,
Miamisburg, OH. This facility incorporates the CECE-HWP (Heavy Water Process)
with the AECL hydrophobic platinum catalyst as the preconcentration step. Final
enrichment to pure tritium is by CD of hydrogen .
The Mound pilot CECE-TRP system will continue to be operated with goals
of obtaining operating experience, determining scale-up parameters and improving
system dependability. Finaly goal is to build a larger system which would be used to
treat aqueous tritiated waste for the Department of Energy (DOE).
The same CECE-TRP is also being developed independently in Belgium ,
Japan [50-52] and Germany , , to detritiate light water streams from nuclear
fuel reprocessing plants. Thus, the Belgians are developing the ELEX process (an
acronym for combined ELectrolysis chemical EXchange process) on behalf of the
European Economic Community. In Japan the proces is apparently referred to as
EXEL proces (an acronym for combined chemical Exchange Electrolysis process).
Both the EXEL and ELEX processes are based on the Canadian invention of the
hydrophobic platinum catalyst .
In Germany, the Dornier System GmbH studied the possibilty to utilize the
existing CECE-process, which is presently in pilot operation with tritium at
Kernforschungszentrum, Karlsruhe, in order to recover the tritium from aqueous
waste of nuclear fuel reprocessing plants, contaminated with tritium in the form of
In order to recover tritium from light water, research and development was
carried out at RIKEN, concerning a tritium separation process based on the principle
of hydrogen-water isotopic exchange reaction. The performance and durability of unit
operations for the process was studied. A pilot plant having a capacity of 1 ton/y (3.6
l/d) was designed and fabricated based on the results of the tests and studies. Using
this plant, tritiated water could be concentrated to the order of magnitude of 10 4.
Furthermore, the effect of the various operating conditions on the tritium
concentration factor was calculated by applying a data analysis program for the pilot
plant. This study offered prospect of practical application of the process by hydrogenvapour isotopic exchange reaction .
At the Tritium process laboratory in the JAERI (Japan) an apparatus for the
fuel clean-up process was developed. This was designed, fabricated and installed for
the experiments with up to 1 g of tritium. The function of the system is continuous
processing of a simulated plasma exhaust and separation of hydrogen isotopes and
impurity elements in it. Main components are palladium diffusers, catalytic reactors,
cold traps, an electrolysis cell and zirconium-cobalt beds. The apparatus was installed
in a glovebox and tested with hydrogen .
CECE provides a highly effective way to detritiate D2O. AECL has operatec a
prototype CECE plant to demonstrate detritiattion. Heavy water was detritiated up to
factors exceeding 50,000.
10. Thermal Diffusion
The possibilities of enrichment by thermal diffusion (TD) of the low
concentrations of tritium found in hydrogen or deuterium has been analyzed by
Takayasu et al. at Toyama University (Japan). The systems take into consideration
were both TH-H2 and TD-D2  and T2-H2, T2-He and TH-He .
In some recent papers, Yamamoto et al. [58-62] reported experimental and
analytical results for H2-HT isotope separation by TD.
In spite of the development of the TD column, isotope separation by TD is not
suitable for large scale because of extremely low thermodynamic efficiency. It,
however, possesses an advantage in small-scale operation in view of its simplicity of
apparatus and small inventory inherent in gas-phase operation. The process, therefore,
attracts attention to purification of tritium in a tritium production system and recovery
of tritium from used gas mixtures of hydrogen isotopes in fusion fuel cycle.
C. RECOVERY AND ENRICHMENT
1. Fuel Cycle System for D-T Burning Fusion Reactor
Tritium technology is doubtlessly a most essential field for fusion energy
research and development, and a successful operation of a D-T burning fusion reactor
cannot be realized without the establishment of its fuel cycle. The fuel cycle for the
D-T burning fusion reactor contains five essential systems :
- The reactor cooling and tritium recovery system
- The mainstream fuel circulation system
- The tritium containment system
- The tritium recovery and waste treatment system
- The tritium production system.
a. Reactor Cooling and Tritium Recovery System
The primary fusion reaction in a D-T plasma is the T(D,n)4He reaction with an
energy release of w 17.5 MeV. The resulting neutron loses its energy in collisions
with the blanket material, thus heating the blanket. Since the blanket contains lithium,
the neutron can be captured by the lithium in an (n,α) reaction producing tritium.
Thus tritium is bred in the fusion reactor, and the bred tritium and the heat must be
efficiently removed from the blanket for use as fuel and for heat generation,
respectively. However, tritium is expected to be accompanied by significant amounts
of impurities such as H, N, C and O, and a series of chemical and physical processing
steps are needed for the recovery of tritium.
b. Mainstream Fuel Circulation System
The quantity of the fuel that must be supplied to the plasma is much larger
than the amount consumed by nuclear fusion, because only a fraction of the fuel
reacts during its residence time in plasma. Based on present forecasts of particle
confinement time, design studies on magnetically confined reactors project a burnup
fraction of 1 to 10 % of the injected fuel. Therefore, the unburned fuel needs to be
rapidly reprocessed and recirculated for reinjection into the plasma. However, the
exhausted plasma of the unburned fuel contains the helium ash produced by the
fusion reactions and various impurities such as H, C, N, O and Ar. As a consequence,
it is required that these impurities be continuously removed before refuelling the D-T
mixture. This fuel circulation requires a series of chemical and physical processing
steps including evacuation of the plasma chamber, removal of impurities, adjustment
of the D::T ratio, and reinjection of the fuel.
c. Tritium Containment System
The use of tritium leads to special concerns in the areas of personnel and
public safety. One of the most troublesome properties of tritium is that it readily
permeates through metals and other materials, with the results that the containment of
tritium is a difficult task. For this purpose, all the equipments containing tritium are
enclosed in glove boxes which have a controlled atmosphere. To maintain sufficiently
low tritium concentration in the glovebox atmospheres, the atmosphere is circulated
through a tritium removal system where in many cases tritium is removed by
oxidation by precious metal catalyst followed by adsorption of tritiated water on
molecular sieves. The tritium concentration in the glove box atmosphere is thus kept
below a specified level, and the rate of tritium release through the glove boxes to the
operation room air is minimized to assure the personal safety. Analyses of tritium
concentration or composition of the hydrogen isotopes in any gas or liquid are
unavoidably accompanied by the production of tritium waste. Before releasing these
waste gases to the stack, tritium must be removed to minimize environmental
impacts. Tritium can be removed by the oxidation/adsorption process.
d. Tritium Recovery and Waste Treatment System
During the operation of tritium systems a large amount of tritium waste
(usually in the form of tritiated water) is thus produced. However, if the tritium
concentration is adequately high, tritium must be recovered because it is a precious
fuel for the fusion reactors. On the contrary, the tritium recovery is not intended any
longer, but further waste treatment is needed because the concentration is too high to
be direct discarded to the environment.
e. Tritium Production System
Even if the breeding ratio of tritium exceeds unity during the fusion reactor
operation, tritium production is essential because a large amount of tritium is needed
in the initial period of the reactor operation. Lithium compounds are irradiated by a
nuclear reactor, and tritium thus produced is recovered by a series of chemical and
physical processing steps.
Another method of tritium production is recovery of tritium from the heavy
water used as a moderator for a nuclear fission reactor. Therefore, the heavy water
enrichment as well as the tritium recovery from the heavy water can be considered as
parts of the fuel cycle for the fusion reactor, to say nothing of the deuterium
2. Roles of Stage Processes in Fuel Cycle System
a. Cryogenic Distillation System
Although there are several methods which appear to be feasible for hydrogen
isotope separation, CD is highly atractive and promising, particularly in the
mainstream fuel circulation system [63-64]. Attractive features of CD are high
throughputs and high performance, comparatively small power consumption, easiness
of continuous operation, compact system scale, and negligible tritium permeation due
to the cryogenic temperature. CD is also a possible method for removal of H from HT or H-D-T mixture in other situations. Removal of H from these mixtures is needed
in the blanket loop, tritium production system, and recovery of tritium from glove box
atmospheres and other similar sources.
CD is the most attractive method even in cases where hydrogen isotopes to
process are in oxide forms, as long as tritium recovery is intended and high
performance is needed for enriching tritium (in these cases, the water/hydrogen
exchange method is used together with CD).
b. Cryogenic Falling Liquid Film Condenser (Helium Separator)
For removal of 3He from tritium (3He being produced by decay of tritium)
before feeding the isotope to the CD columns, a cryogenic falling liquid film helium
separator is used.
c. Water/Hydrogen Exchange Column
Although CD is highly attractive, it cannot be directly used in cases where
hydrogen isotopes to process are in oxide forms. These cases are found in the tritium
recovery from the heavy water (D2O, HDO and DTO), and tritium recovery from
glove box atmospheres and other similar sources (H 2O, HDO, HTO, D2O, DTO and
T2O). In these situations, the catalytic exchange columns are used together with CD
columns. Deuterium and tritium are enriched and transferred from the liquid water to
the hydrogen gas by water/hydrogen exchanges, and then hydrogen isotopes are
processed by CD columns for obtaining tritium free from protium.
On the other hand, if the tritium concentration is not adequately high, the
tritium recovery is not intended any longer. Since the amount of tritiated water
containing a low level of tritium is usually unacceptably large, its volume must be
greatly decreased before solidification and incapsulation. WD columns and
water/hydrogen exchange columns are very attractive candidates for these processes.
However, it should be noted that these two methods are hardly applicable to hydrogen
isotope separation in the situations in which hydrogen isotopes are in gas forms and
the tritium percentage is high (e.g in mainstream fuel circulation system), because
their system scales are expected to be much larger than those of CD columns, and
usage of tritium oxide in a large amount is somewhat dangerous.
d. Cascade Composed of Separate Elemental Stages
For hydrogen isotope separation there are several possible methods which
could be used, such as porous membrane method, Pd-alloy membrane method and
metal-hydride method, pending further R & D to prove the feasibility. To obtain
satisfactory separation performance, the separating elements must be connected in a
cascade. These methods can be unified in the category of hydrogen isotope separating
cascades, composed of separate elemental stages.
To establish design and operation methods for tritium separation processes,
development of mathematical simulation models, experimental works and computeraided simulation studies are needed.
3. Hydrogen Isotope Distillation
a. Tritium Systems Test Assembly (TSTA) [65-66]
In 1977 the Office of Fusion Energy, US Department of Energy funded the
LANL to design, construct, and operate the 135-136. The objectives of TSTA was to
develop those aspects of tritium technology related to the fuel cycle for fusion power
reactors and to develop the environmental and personal safety systems required for
such a tritium facility. The goal of the TSTA project was, also, to provide an
extensive data base which will be available to the designers of the first large D-T
burning fusion machine. This may be the engineering test facility (ETF) or the
international tokamak reactor (INTOR).
b. The Isotope Separation System (ISS) [64, 66-67].
At TSTA cryogenic fractional distillation is being used for hydrogen isotope
separation. The ISS is sized to handle the full flow appropiate to ETF or INTOR. It
has four principal duties. From a purified stream of 360 g mol/d of mixed hydrogen
isotopes, consisting nominally of equimolar quantities of deuterium and tritium with
low levels (1%) of hydrogen impurity, the ISS provides four product streams, namely
- an essentially tritium-free stream of H2 and HD wastes for disposal to the
- a high purity stream of D2, a stream needed by fusion reactors forrefuelling
and plasma heating by injection as a neutral beam
- a stream of basically pure DT for refuelling
- a high-purity stream of T2 for refuelling and for studies on properties of
tritium and effects of tritium on materials
Secondary duties of the ISS include: processing of the relative pure D 2 return
from a simulated neutral beam line (~275 g-mol of D2 per day)  and processing of
electrolytic hydrogens generated in the fuel clean-up system.
The ISS must perform all of the above separations all the above separation
continuously and reliably over the long period of time typical of commercially
operating power reactors.
c. Studies on Cryogenic Distillation Columns for Hydrogen Isotope
Separation in Japan .
It is well known that with the growth of tritium engineering technologies in
relation to the research and development program of a fusion reactor, it is necessary
to provide a safe facility in which workers are protected from radiation exposure of
the tritium activity and tritium is prevented from being released into the environment.
For this purpose, detailed design studies and safety analyses of a tritium processing
laboratory in which grams levels of tritium are to be handled, were conducted at
Although exists a great experience in a specific situation of tritium recovery
from HWR, it is still inadequate to be applied to other situations because of its rather
weak theoretical basis. Column cascades handle a variety of feeds, and the input and
output specifications for the cascades vary greatly from situation to situation. In
addition, in the mainstream fuel circulation system for the fusion reactor, the feed
condition may be greatly variable during the operation. Another fact to be noted is
that the radiological standards for tritium lost to the environment are becoming
increasingly restricted. Thus to establish the design, construction, operation and
control methods for column cascades for all the situation, systematic and extensive
studies are further needed.
In those studies, both theoretical and experimental approaches are imperative
on various subjects :
- Physicochemical studies on cryogenic properties of hydrogen isotopes
- Development of a set of computer simulation procedures and programs
- Detailed analyses on steady state and dynamic column behaviour
- Laboratory scale experiments measuring fundamental parameters and
verifying some of the significant computer predictions
- Development of analysis techniques for measuring mole fractions of
molecular species with high accuracy, adquately short time and moderate
- Operational tests or engineering development for a single, practical scale
column and multiple interlinked column.
Cryogenic hydrogen properties pertinent to fusion technology have been
studied mainly by Souers . He has published a very useful report presenting a
significant amount of information on cryogenic hydrogen data. However, properties
of HT, DT and T2 are uncertain and should be further studied.
Development of simulation procedures and computer programs has been made
mostly by Kinoshita [63, 71-75]. Although a great progress has been made and a
number of efficient computer codes are available, the information on dynamics and
control for column cascades is not yet adequate.
As principal part of the TSTA project, a column cascade of four interlinked
columns was under development at LANL. During this development, at JAERI,
Kinoshita et al., has performed the most of computer simulation work, particularly for
the start-up of the cascade and the design of more reliable control system.
The researches performed at JAERI includes:
- CD columns in the mainstream fuel circulation system; for removal of H
from H-T or H-D-T mixtures (needed in the blanket loop and tritium
production system) and for recovery of tritium from glove box atmospheres
and other similar sources [63, 69, 73-74]
- Cryogenic falling liquid film condenser (He separator) [76-78]
- WD columns [75, 79]
- Mathematical simulation of a multistage-type water/hydrogen exchange
column for recovery of tritium from heavy water and for decrease in the
volume of tritiated water 
- Mathematical simulation of hydrogen isotope separating cascades by the
porous membrane method 
Because of recentness of fusion energy research as well as special feature of
hydrogen isotope separation in tritium systems previously described, computer
simulation study for those stage processes is still in a very early stage. Rather
simplified simulation models have been used to avoid mathematical complication.
The simplified models are in some cases satisfactory but in other cases unacceptable
even in approximate calculations. In cases where simplified models are unacceptable,
rigorous models must be developed.
d. A New System for Complete Separation of 3He and T2 Composed of a
Falling Liquid Film Condenser and a Cryogenic Distillation Column with a
Feedback Stream [76-79]
It is well known that tritium has a relative short half-life time (12.3 years) and
that by decay, 3He is produced. If tritium has been stored for a significant period, a
large percentage of 3He is expected to be contained in the tritium. For this reason,
removal of He from tritium is needed in some situations for use of tritium.
From the possible methods applicable to the present case, Kinoshita et al.  have analyzed separation characteristics of a falling liquid film condenser and
have proposed this cryogenic process as the most attractive one for removing He from
hydrogen isotopes. It has two sections, a cooled section and a packed section. The gas
mixture of He and hydrogen isotopes is continuosly fed to the part of the column
between the two sections. Heat substraction is made from the column wall by the
refrigerant (He gas) which flows through the outer shell. There is a heater at the
bottom which boils up the liquid. As the gas flows up within the column,
condensation of hydrogen isotopes preceeds and the liquid film formed by
condensation falls down along the column wall. Mass transfer phenomena occur
between the gas and the liquid film which flow counter-currently, and the penomena
promote separation of helium and hydrogen isotopes. It is obvious that liquid flows
increase as the liquid film falls down toward the bottom. A packed section in which
there are adequate liquid flows is located near the bottom to greatly increase the
vapour-liquid interface area. An essentially He-free stream of hydrogen isotopes is
withdrawn from the bottom of the column. The gas which leaves the top of the
column is transferred to another process because it unavoidable contains hydrogen
isotopes. This unit will be incorporated between the fuel clean-up unit (FCU) and the
isotope separation system (ISS) by CD in the TSTA . Similar condensers is
chosen for both the fusion engineering device (FED) and the INTOR .
The new system composed of a falling liquid film condenser and a CD
column with a feedback stream presents a significantly high tritium recovery
4. New Concepts for the Recovery and Isotopic Separation of Tritium in
This section introduces new AECL (Canada) concepts for the recovery of
tritium from. light water coolant of LiPb blankets and high pressure He-coolant of Liceramic blankets. Applications of these concepts to fusion reactors is illustrated by
Dombra et al.  with conceptual system design for the anticipated next European
torus (NET) blanket requirements.
a. Tritium Recovery and Concentration Process for LiPb Blanket
Options for detritiation of light water .
CD is the only viable option for the final separation of elemental tritium for
the fusion reactor scale, as is apparent from the selection of this process for the
TSTA, Grenoble, CRNL and Darlington tritium separation plants.
For tritium from heavy water extraction, the selected processes are: LPCE at
CRNL and VPCE at Grenoble and Darlington. As the separation factor for stripping
of tritium from light water is much lower, these processes are unattractive for light
water applications, except in combination with a tritium enrichment process.
WD is used exclusively on a large scale in all CANDU reactors for upgrading
of heavy water (H/D separation) for which the separation factor is lower than for H/T
separation. Since steam is much cheaper than electric energy, this process is a leading
candidate for large scale enrichment and stripping of tritium in light water. Due to the
large electrolytic tritium enrichment factor in light water, electrolysis is not wellsuited as a partner for DW. For the state-of-the-art processes, the best alternative
apparently is a DW and LPCE or VPCE combination. In addition, a CECE can be
added for a separate and smaller scale detritiation of wastewater. Electrolysis and
CECE are largely stand-alone alternatives. Because electrolysis provides no feed gas
enrichment, it requires a larger CD unit with a difficult HT to DT conversion step.
Also, the relatively small size of electrolysis available for tritiated water service
increases cost and complexity of large electrolytic plants. Both electrolysis and CECE
processes therefore become increasing unattractive with increase in size.
c. Research and Development of Tritium Technology at the Karlsruhe
Research Centre (Germany) .
The Karlsruhe Research Centre (KfK) has carried out research and
development activities on tritium for many years within the frame work of nuclear
fuel reprocessing and radioecology. These activities have been greatly expanded as a
result of participation of KfK in the European fusion technology program. The KfK
program includes experiments in the field of the plasma exhaust process, tritium
extraction from ceramic and liquid metal breeder materials and tritium removal from
waste streams and ecology. The final goal of this work is the development of fuel
cycle components and systems for the Next European Torus (NET).
d. Recent Results and Developments in Tritium Processing
- The Canadian fusion fuels technology project - This was formed in 1982
to undertake research and engineering in tritium technology and robotics for fusion
applications. The total current program is about 10 M$ per annum including
contributions from subcontractors and cost sharing with external projects .
- Fusion technology development program for tritium in United States This is centered around TSTA at LANL. Objectives of this project are to develop and
demonstrate the fuel cycle for processing the reactor exhaust gas (unburned
deuterium and tritium plus impurities), and the necessary personnel and
environmental protection systems for the next generation of fusion devices. The
TSTA is a full-scale system for an INTOR/ITER sized machine. That is, TSTA has
the capacity to process tritium in a closed loop mode at a rate of 1 kg/d, requiring a
tritium inventory of about 100 g. The TSTA program also interact with all other
tritium-related fusion technology programs in the United States and all major
program abroad .
- Japan’s two major nuclear fusion research and development plans for
the next stage, post JT-60, of the Nuclear Fusion Council (AEC). - These plans
consists of the development of a tokamak-type large facility and the comprehensive
development of fusion reactor technology. The latter means to promote the reactor
technologies which will be essential in the future to construct, not only a D/T burning,
but also a DEMO reactor .
- Tritium technology activities carried out in EEC laboratories - These are
associated with the European fusion program and the European Joint Research
Center, and deal with the design priorities for NET .
- Tritium studies within the frame work of the NET recently undertaken
in the Bruyeres-le-Chatel Center (CEA France), inside the European fusion
technology program - The 25 years of tritium experience refers to purification of DT
mixtures by cryosorption on molecular sieves, purification of DT on uranium beds,
storage of DT in T-He implanted materials, and corrosion of materials by tritiated
waters species .
- Hydrogen Isotope Separation System (HISS) at Mound Laboratory,
Miamisburg, Ohio – This is a general-purpose tritium recovery and enrichment
processor that uses low-temperature distillation as the separation process. The HISS
processes feed mixtures containing all three isotopes of hydrogen (H, D, T) and yields
an enriched tritium product as high as 99.95% tritium while producing a discardable
- A tritium laboratory has been designed and developed at Ontario Hydro
Research Division -. Its mandate responds to the three corporate needs in which in
order of priority are: to provide a service facility in support of the tritium removal
facility (TRF) to provide a research laboratory for CANDU tritium related work and
to develop a tritium handling data base. The facility is currently licensed to a
maximum tritium inventory of 3000 Ci. The license restricts the total emission from
the laboratory to a maximum of 1 µCi/m 3 of tritium when averaged over a 1 week
period. Current handling practices strive to and have successfully restricted emissions
below 1% of the emission limit .
- New tritium laboratory and process systems has been completed at the
LANL. The Weapon Engineering Tritium Facility (WETF), approximately 8000
square feet of floor space and highly automated processing equipment, replaces an
aging tritium laboratory. It provides improved protection to personnel involved in
tritium handling operations, reduced routine release of tritium to the atmosphere, and
reduced potential for a major tritium release that might result from an accident to the
building or human error .
- Data collection system for obtaining information which can be used to help
determine the reliability and availability of future fusion power plants has been
installed at the LANL's TSTA - Failure and maintenance data on components of
TSTA tritium system have been collected since 1984. The focus of the data collection
has been the TSTA tritium waste treatment system (TWT) which has maintained high
availability since it became operational in 1982. Data collection is still in progress
- New effluent cleanup system for tritium recovery (abbreviated VERS) has
been designed by the LLNL – The purpose of this is for use at their tritium facility
in cleaning Ar box atmospheres and pump exhaust. It is the standard catalystmolecular sieve (zeolite) system, and is directly descended from the cleanup system
at Mound Laboratory .
- Major new tritium handling facility, devoted to the investigation of
problems related to the operational safety and fusion power plants is constructed
at Joint Research Center (JRC), Ispra (Italy). The facility will provide the
necessary safe operating conditions and tritium infrastructure in which both Ispra
personnel and those of other institutions can conduct a wide range of experimental
- Comparison of the processes for recovering tritium from an aqueous
lithium salt blanket on the NET fusion device – This comparison recommended a
system which uses acombination of WD, VPCE and CD. The WD unit is comparable
in size to the D2O upgrader at the Darlington nuclear generating station, and the CD
unit is similar to that provided for the Chalk River tritium extraction plant .
D. LASER ISOTOPE SEPARATION OF TRITIUM
Among many techniques for tritium isotope separation, the one utilizing infra
read laser- induced multiphoton dissociation (IRMPD) has been extensively
investigated because of its possible high separation factors. This method is based on
laser-induced dissociation of a tritiated molecule that yields easily separated products.
For this is was necessary to find a suitable tritium-substituted molecule (working
molecule or working substance) that has selective absorption bands in the range of IR
wavelength. The energy used for laser isotope separation (LIS) is deposited
selectively into the intended isotopes, while in other statistical methods the energy is
distributed randomly among all kinds of components. Thus, this new technique uses
energy highly efficiently, and will be especially useful for low abundance isotopes.
From a practical point of view, it is necessary to optimize the gas temperature
and irradiation wavelength, in order to obtain the best performance of this working
molecule. The major criteria of such optimization are operating pressure, selectivity,
and critical fluency. While the selectivity should satisfy a certain requirement,
increase of the operating pressure and decrease of the critical fluence are strongly
required for increase of the pressure treatment rate.
Tritium isotope separation using laser method has been intensively
investigated since 1979, especially by two groups: at the Institute of Physical and
Chemical Research, (RIKEN) Wako, Saitama, Japan; Lawrence Livermore National
Laboratory (LLNL), Livermore, CA, US and at Ontario Hydro Research Division
Principle of the method
This method is based on isotopic selective laser-induced photochemical
reaction of a working substance. The working substance is tritiated in an isotopeexchange tower, after contact with tritium-containing effluent water. No enrichment
is required in this stage.
In a photochemical reactor, selective laser-induced dissociation of a tritiated
working molecule occurs while the protonated (or deuterated) working substance
reacts to a far lesser extent. The product enriched in tritium is recovered and stored,
and the tritium depleted working substance is recirculated and retritiated.
For the development of this new technique, the following three steps are
necessary: In the first stage, the most suitable working substance must be surveyed. In
the second stage, the reaction kinetics must be fully studied in batch irradiation
experiments. Without such study, the third stage or the design of a continuous reactor
The working substances surveyed for tritium laser isotope separation are:
- Halogenated methanes (particularly CTF3) [96-107]
- Halogenated ethanes (CF3CTClF) , (C2TF5) , [109-110]
- Chloroform (CTCl3) [111-112]
- Halogenated propanes (i-C3TF7) 
From the certain halomethanes whose IRMPD have been extensively studied
for tritium LIS the best results have been obtained for trifluoromethane-T (CTF 3),
with selectivity of over 104 at 9P(8) line (1,057 cm-1) at -78oC for 85 to 205 Torr
CHF3 and 1µTorr CTF3. By addition of 100 Torr argon, as buffer gas, critical
fluency of CTF3 was decreased from 136 J/cm2 to 54 J/cm2 .
Among halogenated ethanes, for which far lower threshold fluency for MPD
is expected, tetrafluoromethane-T (CF3CTClF) and pentafluoroethane-T (C2TF5) has
been found to give both high selectivity (~400) and moderated fluency 8 J/cm 2 for
CF3CTClF at 973 cm-1. and 20 to 30 J/cm2 for C2TF5 .
The T/H separation using pentafluoroethane-T (C 2TF5) has been performed
with a selectivity > 500 for P(20) line at 944.2 cm-1 and later, T/D separation with
single step separation factors exceeding 3,000 for 10P(34) line at 931 cm -1 and –
The T/D separation in which MPD of chloroform (CTCl3 in a back-ground of
CDCl3) was achieved using a CO2-laser pumped NH3 laser at 12.08 µm (828 cm-1)
with a lower limit single step separation factor exceeding 15,000 at room temperature
Finally, from the investigations on IRMPD of heptafluoropropanes it was
found that the irradiation of an i-C3HF7 – i-C3TF7 mixture ( 2 Torr) with a CO2-laser at
10R(30) at 982.1 cm-1, gave an appreciably low critical fluency (~9 J/cm 2) for iC3TF7 dissociation and an extremely high selectivity exceeding 1,400, . These
values may be further improve by using a 13CO2 laser to irradiate at the MPD peak of
i-C3TF7 or by utilizing a CO2 laser with short pulse duration to overcome the collision
de-excitation at higher sample pressures. The i-C3HF7 – i-C3TF7 system showed an
extremely large hydrogen isotope exchange rate with HTO, indicating an important
practical advantage in LIS cycle for tritium removal from water. These results
indicate that i-C3TF7 can be used as one of the most promising molecules in a LIS
cycle for tritium removal from water .
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