Fuel cycle and trends in tritium processing

Fusion Engineering and Design 18 (1991) 3-13 North-Holland

3

Fuel cycle and trends in tritium processing W.T. S h m a y d a Ontario Hydro, Research Division, 800 Kipling Ave., Toronto, Ontario, M8Z5S4, Canada

This paper reviews the development of tritium handling systems which have been designed to process torus effluent from fusion reactors. Several subsystems: torus evacuation, impurity stripping and processing, and tritium storage are reaching a level of maturity which permit realistic extrapolation to reactor scale. Long term tests using tritium are required to shakedown these processes. Their status are reviewed and areas which require additional research are discussed.

1. Background The first significant symposium on tritium was hosted at the University of Nevada in May 1973 [1]. While the conference covered a broad range of topics: tritium production in fission reactors, tritium detection and measurement, chemical, environmental and biological effects, and health physics issues, there was no discussion of fusion fuel systems. In fact, fusion itself only received a passing reference. It was recognized at this conference however, that the use, control, and processing of tritium in the nuclear power industry was of growing importance. Projected growth of nuclear power suggested that by year 2000, tritium production in fission reactors would match the global inventory produced by cosmic radiation. If fusion became practical, tritium production would significantly increase the world inventory of tritium. To minimize an undue burden on our ecosystem, tritium handling practices which restricted tritium emission to the environment would be required. As a first step in this direction, the Atomic Energy Commission through Monsanto Research Corporation initiated an extensive review [2] of tritium handling techniques in use within the industry and government laboratories in the United States. The Controlled Thermonuclear Research community meanwhile had made significant progress towards breakeven during the early seventies. It appeared reasonable that fusion power reactors based on a deuterium-tritium fuel cycle could be realized by the turn of the century. If the endeavour was to be successful, the data base developed by the tritium community needed to be integrated into the fusion program.

The first information exchange between these two communities was implemented through a symposium entitled "Tritium Technology Related to Fusion Power Reactors" held at Mound in October 1974. Several papers discussed fusion cycle relevant topics and identified areas which required additional research. It is of passing interest from a historical perspective, that tritium recovery from blankets was considered to be the most challenging task which would face machine designers. The uncertainty rested in several factors: materials selection for the blanket, the coolant type, the structure of the blanket, and its operating temperature. Equipment associated with plasma exhaust processing, on the other hand, appeared to be available or readily attainable by modest extension of existing technology and consequently the difficulties of exhaust processing appeared to be more tractable. Of the papers delivered at the 1974 symposium, two represent significant milestones in the development of fuel systems. 1.1. The ORMAK F / BX The paper entitled "Engineering Studies of Tritium Recovery from CTR Blankets and Plasma Exhaust" [3] focussed on the assessment of tritium systems for a large divertor pumped tokamak, the ORMAK F / B X . Four subsystems were foreseen as necessary to recycle tritium from the plasma exhaust to the tokamak fuelling port: gas removal from the plasma, removal of hydrogen-free species from the effluent stream, isotope separation, and fuel injection. Gas removal from the plasma would be effected with a divertor, even though the need, functionality, or design of divertors was not entirely clear at this stage

0 9 2 0 - 3 7 9 6 / 9 1 / $ 0 3 . 5 0 © 1991 - Elsevier Science Publishers B.V. All rights reserved

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vK l~ Shmayda / Fuel cycle attd trends in tritium processing

of development. Cryosorption pumps operating at liquid helium temperatures would pump all the effluent species from the divertor throat. Perceived advantages favouring cryosorption pumping were: scalability to reactor size, reliability since there were no moving parts, simplicity of operation, low cost relative to the alternate options, minimal contamination of the effluent stream, and finally no reliance on oil scaled backing pumps. The cyclic nature of cryopump operation was considered to be the pump's chief disadvantage. Regeneration frequency would depend on two competing factors. If too infrequent, tritium inventory in the pumps would be excessive. If too often, regeneration costs would be unacceptably high. Two approaches were promulgated for separating the hydrogen-free spccics from the effluent stream: chemical cracking over uranium beds and permeation. The former technique which had been proven in tritium handling facilities as a gas purification technique albeit for lower flow rates and gas quantities could be implemented with minimal extension of existing technology. A second promising approach with the potential for near term implementation was permeation through palladium/silver membranes. In this case however, compressors were foreseen on the upstream side to achieve acceptable throughputs. While not discussed in any detail, the hydrogen waste stream was expected to contain tritium quantities which could not be vented directly to the environment. A separate tritium removal system would oxidize the hydrogen in the effluent stream. Long term burial of the resultant tritiated waste water was proposed. The need to enclose process equipment, for building atmosphere detritiation, monitoring systems, tritium receiving, storage and dispensing facilities was recognized. It was felt that experience and technology under development in other programs would be directly transferrable. 1.2. The Los Alamos Test Reactor

The second paper entitled "Preliminary Design of the Tritium Handling Facilities for the Los Alamos Fusion Test Reactor" [4] described a tritium handling system for a theta pinch. The paper is particularly noteworthy for its richness of design detail. It anticipated design practices adopted in current tokamak systems. The authors strove to manage the entire tritium cycle, accounting for the tritium distributed throughout the fuel cycle and anticipating, monitoring and controlling, both chronic and accidental tritium releases to the reactor hall and to the environment. A

tritium emission limit of 7 kBq/m ~ leaving the facility was proposed. The proposed facility embodied four subsystems: moderate term storage, a fuel cycle loop, tritium waste treatment, and reactor hall atmosphere processing. A site inventory of 2.3 grams was selected as the balance between maintaining a reliable tritium supply on demand for torus operation and not compromising control over the majority of the inventory. 15 TBq of tritium would be processed through the reactor each day. The moderate-term storage facility was a separate building. Tritium gas would be received on site in metal bottles and stored in a vault. When required, a portion of the gas would be transferred to uranium storage beds which formed part of the tritium injcction system. The vault would provide security against intrusion and fire and minimize the potential for accidental release of the entire site inventory. Tritium recovery from the vault atmosphere was proposed. The fuel cycle loop would consist of three subsystems: an injection loop, a vacuum loop, and a tritium recovery loop. Tritium gas would be dcsorbed from uranium storage beds into calibrated temperature-regulated volumes, assayed, mixed with deuterium and then injected into the torus by volume expansion. Particular attention was given to the leaktightness and mechanical integrity of this loop. The vacuum pump loop was entirely cryogenic. Roughing from atmosphere or during bakeouts would be accomplished by cryosorption on molecular sieve at liquid nitrogen temperatures. Effluent would be removed after the burn by cryopumping at liquid helium temperatures with two pumps sets, one intended primarily for DT and the other for helium. DT gas regenerated from the pump sets would be routed to uranium storage beds. As water was expected to be the dominant contaminant in the effluent stream, it would be frozcn out before entering the eryopumps on traps maintained at 77 K. Gas regenerated from each trap would bc routed to the tritium recovery loop. Several interesting design features were included in the vacuum loop. The pump stations were modular to reduce downtime during repairs. The most critical component, the pump head, could be replaced without breaking vacuum and as a result minimize contamination of the torus and minimize tritium release to the reactor hall. Local ventilation was introduced in the vicinity of the pump stations to collect tritium release due to chronic leakage or during routine maintenance of the valves. The plasma effluent which eventually arrived at the recovery uranium bed was expected to bc essentially

W.T. Shmayda / Fuel cycle and trends in tritium processing

pure hydrogen. Any remaining impurities would be absorbed by the recovery uranium beds. The major loads on the tritium recovery loop would be tritiated water from the torus which had been collected by the traps and tritium gas discharged from the mass spectrometric analysis system. These effluent streams however, were expected to be small, marginally accountable and of insignificant economic value. If necessary, the water could be reduced on copper/copper-oxide beds. A more significant tritium bearing stream could arise from the desorption of helium from the cryopumps. In this case, the selective removal of tritium with the aid of uranium scavenger beds was proposed. The Tritiated Waste Treatment loop was expected to process two types of emissions: chronic releases arising from one of several sources and cleanup of large accidental releases into the reactor hall. The former required a small scale, 7 liters/s, facility while the latter dictated high throughputs, 5 m3/s. Two loops which differed in scale only were proposed. All hydrogen isotopes would be converted to water over a precious metal catalyst. Volatile organic compounds would be reduced to water and CO z. The tritiated water would be collected on molecular sieve dryers, which when spent would be packaged in storage vessels, and sent to disposal sites. 2. The

defining

years

The following ten years marked intense national and international activity to integrate the tritium fuel cycle into fusion machines and to manage the tritium effluent streams according to established nuclear principles. The design work was biased towards tokamaks, in particular the Engineering Test Facility (ETF) and the International Tokamak Reactor (INTOR) [5,6,7]. As the size and operating conditions of reactor relevant machines were refined, tritium process designers provided more detailed fuel cycles. By 1981, a complete deuterium-tritium-lithium fuel cycle had been developed for INTOR [8]. Flow rates of the order of 1 to 2 k g / d a y through the impurity removal subsystems were envisioned. The resident tritium inventory within the impurity processing subsystem was estimated at 200 grams. A target release rate of 7.6 × 1011 B q / d a y was set: 1 X 1011 B q / d a y from the coolant, 9 x 10 l° B q / d a y from the building ventilation, 6 x 1011 B q / d a y from process loops, and 3 x 10 l° B q / d a y as solid waste. The primary tenants intended to guide the development of future fuel cycle systems were presented: - maintain as low a working tritium inventory in the process systems as possible,

5

minimize tritium release to the environment, minimize tritiated solid waste production, and - maintain worker exposure levels as low as realistically practical. Researchers at Mound [9,10] and Sandia [11,12,13] National Laboratories collated and expanded their experience on reducing the tritium content in airborne effluent streams before discharging to the environment. These streams could include various complex compounds; tritiated organic volatile compounds, volatile tritiated solvents, tritium gas emission from process loops or tritium entrained in inert gases. The systems which could reduce complex compounds to water over hot catalytic beds and subsequently dry the carrier stream in molecular sieve beds were presented as mature technology which was being upgraded to meet fusion requirements and could be transferred to industry as the need arose. Los Alamos National Laboratory launched the design and construction of a new facility, Tritium Systems Test Assembly (TSTA) with the intent of demonstrating all key components relevant to the DT fuel cycle [14,15,16]. The major DT processing systems envisioned included: a continuous tritium waste treatment system to strip residual tritium from all effluent streams, a torus evacuation system based on cryosorption pumping, a fuel cleanup system that relied on chemical reactions over hot metal getters, mainly uranium, and a cryogenic distillation system for hydrogen isotope separation. The key task for this new facility would be to demonstrate the continuous operation of a fuel cycle which could process typical fusion reactor exhaust at the rate of 1.5 to 2 kg/day. Programmes aimed at the development of tritium technology for controlled fusion reactors were also launched in Europe, Canada [17] and Japan [18]. In Europe, the Programme Definition Group on Tritium and Blanket Technology recommended the implementation of a tritium technology programme within the European Community on the assumption that the Community would build and operate a DT fuelled machine as the next step [19]. The objectives were to develop and test the key tritium handling systems. A tritium handling system was designed to accommodate experiments planned for Z E P H Y R [20]. Researchers at JET submitted the first of several tritium handling schemes for a DT cycle [21]. -

-

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The key elements of a fuel loop based on current wisdom is presented in fig. 1. For the N E T / I T E R

6

W.T. Shmayda / Fuel cycle and trends in tritium proces'sing

design, effluent will be removed from the torus at 75 moles D T / h o u r . The probable gas composition in mole percent is 94% DT, 1% H2, 3% He and 2% impurities during the 'Burn and Dwell' mode [22]. The anticipated impurity composition is 1% C , Q .... 0.1% CO, CO 2, Ar, and NQ 3, and 0.2% N 2, O> and Q 2 0 where Q is used to represent one of the three hydrogen isotopes (H, D, T). The effluent is separated into two streams in the Impurity Stripping subsystem. The impurity free stream is isotopically adjusted to meet torus fuelling requirements and stored or returned to the reactor via the Fuel Blending and Delivery subsystem. The effluent stream leaving the Impurity Processing subsystem is discharged via the Waste Treatment subsystem to the environment.

hausts without excluding lorus access to other subsystems. Research effort is directed towards the development of two alternative types of pumps; turbomolccular pumps and cryopumps. 4.1.1. Turbomolecular pumps

Turbomolecular pumps mechanically compress gases entering the inlet by means of a series of bladed rotor and stator sets and exhausts them continuously via an outlet port. These pumps offer two advantages. They are inherently low tritium inventory devices and their operating cycle can be readily tailored to the application. Oil lubricated pumps with capacities in the range of 10 000 liters/s are commercially available although a data base for tritium service is non-existent. Scalability to larger capacities is technically feasible. However particular attention is required to dcvelop units which can function in the proximity of strong and varying magnetic fields, can endure the mechanical shocks o1 torus disruptions and can withstand the high thrust loads on the bearings that arise during an accidental air ingress into the torus. Magnetic bearings, if thcy can operate in the vicinity of the torus and are not degraded by tritium gas absorption, could replace existing oil lubricated bearings or greasc lubricated ceramic

4.1 Torus el,acuation

The function of the Torus Evacuation Subsystem is to remove gas from the reaction volumc and to deliver it to the Impurity Stripping Subsystem. This subsystem consists of valving, tubulation and pumps. Careful design of the tubulation is required to attain high conductance pathways between the torus and the pump ex-

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W.T. Shmayda / Fuel cycle and trends in tritium processing bearings and could circumvent the need to handle and dispose of tritiated lubricant. 4.1.2. Cryopumps Cryopumping removes molecules from the gas phase by physisorbing the gas molecules on cold surfaces; cryosorption pumping relies on chilled sorbants such as charcoal or molecular sieve whereas cryocondensation pumping utilizes bare metallic surfaces. Two cryopump options which maintain high specific pumping speeds for both helium and DT are under investigation, staged pumping and co-pumping. Staged pumps intercept the gas flow with cryocondensing panels held at 4.2 K to preferentially pump DT before the helium reaches the cryosorption panels [23,24,25]. Co-pumping relies on the simultaneous cryosorption of helium and DT on a 4.2 K surface usually by depositing a frost layer to enhance the helium pumping speed and pump capacity [26]. For example, the utility of argon spray to enhance the helium pumping speed in the presence of DT has been demonstrated. The complexity of the argon injection and removal subsystems needs to be weighted against the advantage of simplifying the cryopump design [27]. Several concerns are usually associated with staged pumps: lengthy regeneration times, the compatibility of the sorbent binder for tritium service and the inherent working tritium inventory of the pump. Results from a test programme aimed at developing cryosorption panels for the ITER application are reported at this conference [28]. Concerns regarding pump inventory can be reduced through frequent regeneration cycles. Near continuous pumping, from the evacuation subsystem point of view, can be achieved through the use of dual cryopump circuits which are alternated, one on-line while the second is regenerated. The challenges which faces cryopump designers are the development of staged pumps with rapid heat-up times and low DT transmittance to the cryosorbing surface intended to pump helium. Current staged pump designs accommodate approximately 10% of the DT stream along with the helium and impose the need for a subsystem which separates the DT from the helium downstream of the pumps during regeneration. An interesting operating regimes of co-pumping which depends on the fraction of the panel's storage capacity utilized has been observed for 4.2 K charcoal cryosorption panels. Experiments at TSTA have demonstrated that the co-pumping of helium and hydrogen isotopes can proceed with a progressive but tolerable reduction in the helium pumping speed at a low panel capacity [29,30]. At capacities exceeding 10%

7

of the design value, deuterium pumping occurs as a surface frost and the helium pumping speed appears to remain constant but at the reduced level. These tests suggest that staged pumping is not required. Furthermore the reduction in helium pump speed can be partially recovered by removal of the condensing chevrons required in the staged pump. While co-pumping simplifies pump design and operation a DT-helium separation stage is required between the pump exhaust and the impurity removal subsytem. 4.2. Impurity stripping The function of the Impurity Stripping Subsystem is to receive the torus exhaust, separate the gaseous impurities, including helium, from the main stream and to deliver the impurity free stream to the Isotope Separation Subsystem. The residual impurity stream is sent to the Impurity Processing Subsystem. Each subsystem may contain several ancillary components such as surge tanks, " D E O X " catalyst to remove any free oxygen from the effluent stream, circulation pumps, and tritium process monitors. The need for these components is generic and is not discussed here except to mention that the performance of large scroll pumps [31] and the utility of a 2 cm 3 ionization chamber for process monitoring [32] are reported at this conference. Several approaches to separate the impurity stream from the main stream have been developed: impurity reduction over hot metal beds, cryopurification and separation by permeation. 4.2.1. Hot metal beds The use of hot metals, uranium in particular, to purify hydrogen isotope fusion reactor exhaust streams is based on operating experience established in the tritium community and predates its proposed role in fusion [4,16]. When a hydrogen stream containing impurities comes in contact with hot uranium, the impurities are converted directly to oxides, nitrides, and carbides. Little hydrogen is retained since the solubility of hydrogen in uranium is negligible above 900 K. This process has an inherently low residual tritium working inventory and is simple to operate. These attributes however must be balanced against several drawbacks. Significant quantities of low level tritiated uranium waste can be expected to arise particularly if the internal reactors components are carbon based. Container designs must minimize tritium permeation and ensure integrity over a broad temperature range. This approach, has fallen out of favour in recent years as a method for processing the main stream, and appears to

W. T. Shmayda / Fuel ¢Tcle and trends in tritium processhzg be restricted to processing effluent streams which have been stripped of DT. 4.2.2. Cryopurification The continuous separation of impurities from the effluent stream by condensation on a sorbent is a maturing technology. A design capable of processing 1 kg/day, 35% tritium, and using a working inventory of 110 g is under evaluation at TSTA. This scale approaches ITER sizing requirements [29,30]. The loop consists of a catalytic reactor, a train of two molecular sieve beds in series and a regeneration loop. The catalytic reactor which contains 387 g of precious metal catalyst is maintained at 450 K to remove any free oxygen from the stream. The two molccular sieve beds containing 1.6 kg of Linde 5 A sieve are maintained at 77 K to purify the effluent stream of all expected compounds except hydrogen and helium. Several operational factors are being assessed: continuous removal of impurities, on-line regeneration, and the process monitoring schemes. Cryopurification offers several advantages: low tritium permeation to the surroundings, low solid waste generation, and robustness. The trapping efficiencies of 5 A sieve operating below 93 K for impurities typical of the torus effluent are presented at this conference

impurity concentrations 0.2 to 9.5 vol% [43]. This work concluded that neither CO~ nor CH~ had any practical effect on the permeation performance of Pd-Ag membranes. Degradation in the performance of permeation due to CO could be mitigated by operating the membrane above 300 ° C. Operating above 375 °C resulted in objectionable production of methane from the presence of CO and hydrogen. More recently concerns regarding the tritium compatibility of palladium diffusers have driven the need for long term tests using tritium. During the permeation process a significant quantity of tritium is contained within the P d - A g bulk. The helium-3 content arising from tritium decay is expected to increase within the metal matrix with operating time. Changes in the permeability and mechanical properties of the membrane can be expected. Tests spanning 450 days of operation with tritium permeating at the approximate rate of 3.6 cm-~/cm 2 (STP) through Pd-25(~i Ag membranes held al 280 ° C under an upstream pressure ot 3.3 bar demonstrated an 8~~ increase in the permcaut flux. While not considered significant to the operation of the diffuser, the effect has been attributed to helium-3 incorporation in the metal lattice [39]. 4. 3. Impurity processing

[331. 4.2.3. Permeation An elegant approach for separating impurities from the torus exhaust relies on hydrogen permeation through a palladium-silver alloy held at an elevated temperature, usually on the order of 600 K. Permeation as an impurity stripping process offers continuous operation capability, does not generate solid waste products, permits low inventory operation, and provides a pure, helium-free product stream. Several research groups have addressed the objections raised by early investigators [16]: embrittlement due to temperature cycling, high operating upstream pressures, and poisoning by ammonia and methane. Embrittlement arising from phase changes of the palladium when hydrogen is incorporated in its bulk was circumvented through the alloying of palladium with 20 to 25 weight wt% silver [34]. Reduction in the upstream pressure has been realized through the use of multiple tubular permeation cells connected to a common header [35,36,37]. Parametric studies on the poisoning effects by the most likely impurities to be present in the exhaust stream, CO, CO z, CH 4, have been carried out as a function of membrane temperature over the range 100°C to 450°C, hydrogen pressures 0.3 to 14 kPa and

The function of the Impurity Processing Subsystem is to remove chemically bound tritium from the impurity stream received from the Impurity Stripping subsystem. This stream is expected to contain CQ~, NH3, modest amounts of T~ and DT depending on the efficiency of the Impurity Stripping subsystem, traces of more complex tritiated organic vapours and several tritium free gaseous species. Several options are under development. 4.3.1. Catalytic oxidation The first detailed scheme proposed for separating chemically bound tritium from the torus exhaust [16] relied on experience developed in glovebox cleanup systems [40,41,42]. In this approach, chemically bound tritium is oxidized over a precious metal catalyst dispersed on a high surface area alumina base at 800 K, to form tritiated water vapour and tritium free gaseous compounds. Subsequently the water vapour is separated from the latter by cyrotrapping on molecular sieve at 160 K. Upon regeneration, the water is reduced to DT over uranium held at 750 K. This concept has been successfully tested at the I kg tritium/day level [29]. The advantages and shortcomings of hot uranium beds have been discussed in see-

W.T. Shmayda / Fuel cycle and trends in tritium processing tion 4.2.1. However operation in a DT-starved effluent stream dramatically reduces tritium permeation losses from the reactor vessel and renders the use of hot metal beds less objectionable. These tests have uncovered several areas which need further attention: accountability of tritium resident in the cryotrap, water hang-up in the loop between the trap and the reducer, and the low efficiency of hot uranium beds for reducing water [26]. Electrolysis has been envisioned as an alternative to hot uranium beds for reducing water to DT from the onset of fuel cycle development. Its chief advantage stems from the dramatic reduction in waste generation in comparison to that produced by hot metal beds. Its chief disadvantage arises from the need to handle very toxic water. Implementation in fuel loops has awaited the development of tritium compatible electrolysers. Three cell configurations have evolved in the past twelve years: the vapour phase ceramic cell, the capillary electrolyser, and the membrane cathodic electrolyser. The ceramic electrolysis cell utilizes yttrium stabilized zirconia tubes which act as electrolysers to decompose water vapour. This operating mode provides the cell with several advantages: low working inventory relative to liquid cell designs, reduced radiation damage by comparison to polymer membranes, and negligible production of spent electrolyte. A cell operating at 900 K has processed helium which contained 3% TeO for seven months and has demonstrated the viability of this concept by converting 7.4 × 1014 Bq of tritiated water per day to T 2 with a conversion efficiency of 99.9% Currently, ten cells are under long term evaluation for a broader range of operating conditions to determine system throughputs and cell compatibility with tritium [43,44]. The membrane cathode electrolyser consists of two concentric tubes separated by a 300 /zm thick pallad i u m / silver membrane. Tritiated electrolyte is fed into the outer volume while the inner volume is evacuated. The device combines two processes: cathodic charging of the membrane with tritium ions drawn from the electrolyte and vacuum extraction of the permeant hydrogen flux on the downstream side. This device, particularly suited to reducing highly concentrated tritiated water, is attractive because of its simplicity of design in comparison to wet cells and because of the absence of pressure regulating devices. All materials in contact with tritiated water are metallic. Evaluation of an industrial model operating at 80 °C with a current density of 75 m A / c m 3 in a p H 10, NaOH solution has demonstrated the device's ability to process 0.8 g / h of

9.5 x 1015 B q / l i t e r tritiated water for 60 hours [45]. The electrolyser cell volume is 69 cm 3. Preparations are under way to process 3.7 x 1016 B q / l i t e r water. Qualification tests at these concentrations will address concerns regarding the integrity of the membrane in the radiation enhanced corrosion environment and demonstrate the strength of this approach. The capillary electrolyser is a wet cell designed specifically for tritium applications [46]. Two platinum electrodes sandwich a porous vitreous silica layer. When wetted, the silica layer divides the chamber into two gas impermeable compartments which prevents the electrolytically produced tritium and oxygen from mixing. Tritiated sulfuric acid is fed to the silica layer by capillary action through a tiny feed channel. The cell operates at 5 °C to minimize radiolytic decomposition of the electrolyte, to retard metal corrosion and to reduce the quantity of tritiated water entrained in the product stream. It features a cell volume of 20 cm 3 with a total volume of 40 cm 3 including the ancillary feed tanks and feed lines. The design throughput for the celt is approximately 4 g / h of tritiated water. Preparations are underway to test the cell with tritiated water activities as high 3.7 x 1016 Bq/liter. These tests will address several concerns: tritium loss in the spent electrolyte, cell corrosion, the radiation stability of the electrolyte, and the degradation in performance due to bubble formation in the capillaries.

4.3.2. Catalytic decomposition Reclamation of chemically bound tritium through catalytic decomposition of methane and ammonia has undergone testing with hydrogen [47,48]. The primary effluent stream is separated into an impurity free hydrogen stream and an effluent stream by permeation at 650 K. Water in the effluent stream is converted to hydrogen by the water gas shift reaction over a zinc stabilized copper chromite catalyst held at 450 K. Hydrocarbons, methane in particular, and ammonia are decomposed over a nickel catalyst at 780 K. A P d / A g permeator operating at 650 K separates the newly formed hydrogen stream from the effluent gases. This impurity processing concept offers several advantages over that discussed in the previous section. Efficient cracking of ammonia and methane are possible at a lower temperature, 700 K rather than 900 K. As a result, tritium permeation losses and solubilization in the metal container walls are reduced. In comparison to chemical beds, less solid waste is generated in catalytic beds since few reaction products are re-

t0

W.T. Shmayda / Fuel cycle and trends" m tritium processing

tained by the catalyst. Eliminating direct oxidation of the impurities also significantly reduces the need for high activity electrolysis cells and ancillary subsystems capable of handling the high activity tritiated water. A data base elucidating the performance of this system for tritiated effluent streams is required. Experiments suggest that 0.2 moles of nickel catalyst is consumed per mole of methane decomposed without degradation in the catalyst's performance [48]. While catalyst consumption can be reduced further through regeneration with deuterium, tests are needed to measure the tritium content in the regeneration stream and if necessary to establish a process for the ultimate disposition of that tritium.

4.3.3. Isotopic exchange o~'er a catalyst

Isotopic exchange activated by contact with a hot catalytic surface has been proposed as a method for extracting chemically bound tritium from the effluent. This concept promulgates swamping the effluent stream with hydrogen and equilibrating the mixture on a 1000 K platinum surface to exchange the chemically bound tritium with hydrogen [49]. The process envisioned is a batch process. If successful, the initial tritium content is expected to be reduced by approximately nine orders of magnitude and to permit discharge of the processed effluent stream directly to the environment. Experiments are under preparation to validate this model.

4.4. Tritium storage

The function of the Tritium Storage subsystem is to accept purified tritium or DT gas from the various sources and to secure it against accidental release. The storage system should be capable of readily returning the gas upon demand. While many storage options involving gas, liquid, and chemical storage have been entertained over the past two decades, only the last option appears viable for fusion reactor applications. While dispensing tritium from gas containers is convenient, tritium contained in gas bottles is vulnerable to accidental release. In addition gas storage inherently imposes an inconveniently low storage density, < 100 B q / c m 3 of storage volume, particularly if subatmospheric conditions are adopted for safety reasons. Chemical beds, uranium storage beds for example, can safely operate at storage densities in the range of 10 T B q / c m 3 of uranium powder. Tritium storage in liquid form, primarily as water, increases storage density rela-

tive to gaseous storage but introduces the hazard of toxic water handling and complicates tritium gas extraction. Several tritide forming materials are under evaluation as storage media in chemical beds. Uranium has provided a long standing and respected service as a storage medium in the tritium community [50,51]. As a result, uranium represents the storage medium against which alternates are compared. Uranium offers a sound tritium data base, good loading kinetics over a broad storage capacity range characterized by the tritium/metal ratio (T/M), T / M 0.2 to T / M / ~ 2.9, and a broad equilibrium dissociation pressure range, approximately 10 ~' mbar at room temperature and 1 bar at 690 K. Bed operation at storage capacities exceeding T / M ~- 1.5 is not common for safety and kinetic reasons. Two underlining concerns drive efforts to develop alternative storage media. Firstly, uranium is a controlled strategic material and therefore an attendant paper trail monitored by government regulatory, agencies is required. Secondly. uranium tends to decrepitate. Without preventative measures, fines which distribute through the process loop can accelerate component wear and can bc inhaled by operators during routine loop maintenance. Several alternatives have been proposed. Three have undergone limited testing: Zr-Co, LaNi3Mn z and LaNi 5 ~AI, with tritium. The applicability of Z r - N i based alloys for hydrogen storage are discussed in a paper submitted at this conference [52]. Zirconium-Cobalt [5354,55,56] forms a pseudo-binary tritide. While the equilibrium dissociation pressure is similar to that of uranium, the isochorcs arc inclined, increasing with increasing hydrogen content in the host material. Hydrogen tests suggest that Z r - C o may be an acceptable alternative to uranium. To this end, long term tests with tritium are underway in several laboratories to determine the impact of radiation damage and helium incorporation in thc Z r - C o matrix on storage capacity and loading kinetics. Two ternary alloys, LaNi3Mn z and L a N i s . , A I , which are under evaluation [57,58] differ from both uranium and Z r - C o in that they retain decay helium within the metal lattice for extended periods of time. The attractive feature of these metal tritidcs is that the regenerated tritium is helium free. Long term testing of the ternary alloys with tritium is particularly critical. Radiation decay can scramble the alloy mixture. Decay helium within the alloy will stress the matrix and may adversely affect the pressure isochores and loading kinetics. As with all ternary alloys, tests are necessary to measure disproportioning of the alloy as a result of temperature cycling.

W. 7". Shmayda / Fuel cycle and trends in tritium processing

5. Summary The status of tritium processing loops for controlled thermonuclear devices has been reviewed. Several options are available for the next generation fusion machine. At least one of these, impurity separation by molecular sieves maintained at 77 K followed by tritium reclamation over hot uranium, can be extrapolated to reactor scale with confidence. A second option based on separating the effluent stream into a pure hydrogen stream by permeation and subsequently burning the effluent to reclaim the tritium in a vapour phase electrolysis cell has undergone extensive tritium testing. Evaluation of an I T E R scale system would be prudent. A third promising option which also relies on permeation as the first step and has undergone extensive hydrogen testing, extracts the chemically bound tritium from the effluent by catalytic decomposition. Performance evaluations using tritium are planned and necessary if this option is to be a serious fuel clean-up contender. However each of these options suffer drawbacks and underscore the need for developing more benign, both to operators and to the environment, simpler and consequently more reliable systems.

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