TRITIUM ISOTOPE SEPARATION

Gheorghe VASARUAleea Tarnita, Nr. 7, Apt. 11

Cluj-Napoca, ROMANIAE-mail:: gvasaru@hotmail.com

INTRODUCTION

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 cascade.

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 [1].

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 some laboratories,

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 [2].

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 [3].

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 [4]:

HTO HT + 1/2 O2

(1)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 kineticeffects 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 [5].

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 [3]:

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 tritium: 0.93

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 [4],

HTO(v) + H2O(l) H2O(v) + HTO(l) (2)

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 100oC in evaporation flask and heated yoke would amount to 450 mg; the loss in saturated water vapour at 25oC amount to 15 mg [3].

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 [4].

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 [8]. 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 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. [9], Boorman and Kronberger [10], Robinson et al., [11], Schirdewahn et al., [12], Gonsior [13], Verhagen and Sellschop [14-16], Buttlar and Wiik [17], Shimizu and Ravoire [18], Shimizu [19], Ravoire et al. [20], Verhagen [21], Lemarechal-Dupuis [22], Hugony et al., [23].

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 104 times as hazardeous as HT [24]. 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 104

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 [27]. The monitor was designed to measure tritium in air concentration up to 106

MBq/m3. In terms of derived air concentration, the range is up to 106 for HTO and 50 for HT .

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 H2 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 [1]. 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 [28]. Other packing materials used are activated alumina, silica and molecular sieves [29]. 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 H2-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 30oC. 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 moving- bed 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 [1].

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 [30]. 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/D2O 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) [31]. 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 [32]. 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 [1]. 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. ============================================================== No. Procedure Separation factor

1. Diffusion and membranes processes: ca. 1.6 2. 2. Ultracentrifuge and separation nozzle: ? 3. Laser excitation: 104

4. Adsorption processes: ca. 1.4 5. H2S, NH3 and methylamine exchange processes: 0.3 - 3 6. Rectification (distillation) as HTO: 1.05. 7. Catalytic exchange:

a. in the gas phase as HT: 4.50b. in the vapour phase as HTO: 0.40c. in the liquid phase as HTO: 0.20

2. Electrolysis as HTO: ca. 10 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), (2)

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 [1]

Table 3. Boiling point and vapour pressure of oxides of hydrogen isotopes.

H2O HDO D2O T2O HTOBoiling point (oC): 100.00 101.42 101.51 100.8Vapour pressure,: 23.76 22.12 20.6 19.8 21.7(mmHg), at 25oC

Water pressure ratios for isotopic water species P(H2O)/P(x) calculated using Van Hook's equation and data [33] for three temperatures are given in Table [4]:

Table 4. Vapour pressure ratios for isotopic water species P(H2O)/P(x).============================================================ Temperature,oC T2O DTO D2O HTO HDO 25 1.193 1.175 1.154 1.095 1.075 50 1.134 1.123 1.110 1.065 1.053 100 1.064 1.060 1.052 1.030 1.026

The measured liquid/vapour T ratios of Sepall and Mason [34] 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 = 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 = 1.5, only 50 stages would be necessary [1]. 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 France [35].

Sulzer Brothers Ltd. [36] has many years of experience in both H2O/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 H2/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 [36].

Table 5. Operating data of a Sulzer plant for tritium extraction from light water.

Quantity (l/h) Concentration (Ci/m3)

Feed: 190 1 Return: 189.875 0.5 Concentrate: 0.125 720 Operating pressure at top of column: 150 mbar Heating steam: 3,000 kg/h Cooling water: 160 m3/h Electric energy: 15 kW

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

under vacuumWhen 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 [37]:

- 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 water-

hydrogen isotope exchange process)- Absence of catalyst for isotope exchangeWD 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 fusion reactor.

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 [1]

Table 6. Physical properties of isotopic hydrogen species. ============================================================= H2* HD HT D2 DT T2

Molecular weight: 2.016 3.022 4.025 4.029 5.032 6.034 Boiling point (K): 20.39 22.14 22.92 23.67 24.38 25.04 Triple point (K): 13.96 16.60 17.62 18.73 19.71 20.62 Triple point pressure (mmHg): 54.0 92.8 109.5 128.6 145.7 162.0 Critical tempe rature (K): 33.24 35.91 37.13 38.35 39.42 40.44 Critical pressure (mmHg): 9,736 11,134 11,780 12,487 13,300 13,878Dissociation energy (eV): 4.476 4.511 4.524 4.553 4.588 Zero point energy (cm-1): 2,171.4 1,884.3 1,542.4

*/ 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 safety reasons.

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 [38].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 [30]

catalystDTO + D2 D2O + DT (3)

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) [39].

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 [39]:

- Vapour-phase catalytic exchange (VPCE) of tritiated D2O vapour with D2, producing a D2/DT mixture [38]

- Direct electrolysis (DE) of tritiated D2O, producing a D2/DT mixture- Combined electrolysis-chemical exchange CECE [40] of the tritiated D2O in

the liquid phase with D2/DT producing a D2/DT stream enriched in tritium- Liquid-phase catalytic exchange LPCE [39] of tritiated D2O with D2

producing a D2/DT mixtureEach 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 [1].

5. Chemical Exchange

Depending on the physicochemical form of tritium at the start of the chemical exchange reaction, three procedures are possible [4]:

HT(g) + H2O(l) H2(g) + HTO(l), (4)

HTO(v) + H2(g) H2O(v) + HT(g), (5)

HTO(l) + H2(g) H2O(l) + HT(g). (6)

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 300oC) has proved successful [4].

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 [1]. In this case the exchange is between water and HD or HT.

Table 7. Equilibrium constants for the hydrogen-water exchange reaction (theoretical).

============================================================= Temperature (oC) HD + H2O HT + H2O 20 3.78 6.24 100 2.65 3.83 200 2.09 2.66 300 1.75 2.09 400 1.54 1.77 500 1.37 1.52

Experimental data agree well with theory [41]. 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 candidate.

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 [4].

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 detritiation operations.

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 heavy water.

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 parts [38]:

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 [38], [42]:

DTO(v) + D2(g) D2O(v) + DT(g) KT = 0.82 (7) HDO(v) + D2(g) D2O(v) + HD(g) KH = 1.78 (8)

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 filling device.

Operating data and operating media of the Grenoble plant are given in Table 8 [42]:

Table 8. Operating data and operating media of the Grenoble tritium and hydrogen extraction plant.

============================================================= Intake: 16.7 litres/h Head product: 60 litres H2O/year Bottom product: 60 standard litres T2/year (160000 Ci/year)D2 consumption: 60 standard litres/h O2 consumption: 35 standard litres/h Electric power input: 800 kW Cooling water consumption: 25 m3/h Compressed air consumption: 50 standard m3/h D2O hold-up: 1,000 litres D2 hold-up: 85 standard m3 T2 hold-up: 60 standard lHe hold-up: 35 standard m3

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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 completely leak-free.

More stringent safety measures were necessary, because of radioactivity of the heavy water and of the deuterium, together with the explosion danger associated with deuterium.

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 [38].

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 T2 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 [36]. 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 [39].

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 [43].

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 T2 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 (CD)

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 [44]. 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 completely ineffective.

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 dry environments.

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 commercial exploitation.

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 [39]. In this process, during the first stage, tritium is transferred by catalytic isotope exchange from the tritiated water into deuterium gas and the tritium-depleted 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 (Teflon)) [45].

The LPCE process is simpler and require less energy because the reaction is conducted at ambient temperature in a single packed column [44]. 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 [45]

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 [46]. A future option to increase plant capacity is conversion of the LPCE front end to a Combined Electrolysis Catalytic Exchange (CECE) process.

7. Combined Electrolysis Catalytic Exchange (CECE) and Cryogenic Distillation (CD)

The CECE process has been developed by AECL at CRNL [40]. 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 D2/DT gas generated in electrolytic cells. Tritium moves from the gaseous D2/DT stream to the liquid D2O/DTO stream [39]:

DT + D2O DTO + D2 (9)

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/D2 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 [30].

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 [39]. 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 [47].

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 [39]

electrolysis2DTO 2DT + O2 (10) electrolysis2D2O 2D2 + O2 (11)

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 units.

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 [30].

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 [48] 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 [47].

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 [49], Japan [50-52] and Germany [4], [53], 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 [47].

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 HTO.

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 104. 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 hydrogen-vapour isotopic exchange reaction [54].

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 [55].

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 [56] and T2-H2, T2-He and TH-He [57].

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 OF TRITIUM FROM THERMONUCLEAR PLANTS

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 [63]:

- 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 SystemThe 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,) 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 SystemThe 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 SystemDuring 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 SystemEven 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 production.

2. Roles of Stage Processes in Fuel Cycle System

a. Cryogenic Distillation SystemAlthough 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 H-T 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 ColumnAlthough 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 (H2O, 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 StagesFor 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 computer-aided 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 [64]:

- an essentially tritium-free stream of H2 and HD wastes for disposal to the atmosphere

- 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 materialsSecondary duties of the ISS include: processing of the relative pure D2 return

from a simulated neutral beam line (~275 g-mol of D2 per day) [66] 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 [69].

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 JAERI.

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 [69]:

- 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 cost

- 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 [70]. 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 [80]

- Mathematical simulation of hydrogen isotope separating cascades by the porous membrane method [81]

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. [76-79] 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 [77]. Similar condensers is chosen for both the fusion engineering device (FED) and the INTOR [76].

The new system composed of a falling liquid film condenser and a CD column with a feedback stream presents a significantly high tritium recovery percentage 100%)

4. New Concepts for the Recovery and Isotopic Separation of Tritium in Fusion Reactors

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 Li-ceramic blankets. Applications of these concepts to fusion reactors is illustrated by Dombra et al. [82] with conceptual system design for the anticipated next European torus (NET) blanket requirements.

a. Tritium Recovery and Concentration Process for LiPb BlanketOptions for detritiation of light water [82].

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 well-suited 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) [83].

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 [84].

- 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 [85].

- 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 [86].

- 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 [87].

- 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 [88].

- 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 raffinate [89].

- 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/m3 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 [90].

- 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 [91].

- 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 [92].

- 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 catalyst-molecular sieve (zeolite) system, and is directly descended from the cleanup system at Mound Laboratory [93].

- 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 studies [94]

- 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 [95].

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 (OHRD), Canada.

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 isotope-exchange 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 is impossible

The working substances surveyed for tritium laser isotope separation are:- Halogenated methanes (particularly CTF3) [96-107] - Halogenated ethanes (CF3CTClF) [108], (C2TF5) [97], [109-110]- Chloroform (CTCl3) [111-112]- Halogenated propanes (i-C3TF7) [113]

Results 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 [100].

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/cm2 for CF3CTClF at 973 cm-1.[108] and 20 to 30 J/cm2 for C2TF5 [107].

The T/H separation using pentafluoroethane-T (C2TF5) 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 – 78oC [106].

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 [108-109].

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/cm2) for i-C3TF7 dissociation and an extremely high selectivity exceeding 1,400, [114]. 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 [110].

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