FUSION | MANAGING TRITIUM destructive than the immediate effects caused by the
radiation damage. In one such case, the tritium contamination normally present in heavy water up to several curies per litre was able to leach substantial amounts of chlorides out of the bodies of neoprene O-rings that were used for the seals”. The chlorides deposited into the stainless steel sealing surfaces above and below the trapped O-rings “led directly to the introduction of chloride-induced stress-crack corrosion in the stainless steel”.
Extracting tritium from the ‘blanket’ These examples of the challenging characteristics of tritium management show how important managing tritium is in the context of fusion. The US DOE’s Standard on Tritium Handling also includes
examples of situations where tritium’s combination of permeability and radioactivity affect storage decisions. It says design requirements for tritium are a function of the tritium form, quantity, concentration, pressure and period of storage. As a gas, high concentrations of tritium stored at high pressure (> 2,000 psia) are difficult to contain due to tritium and helium embrittlement of the container materials. As a liquid, tritiated water in the form of T2O is “somewhat corrosive”. This can be addressed with an overpressure of T2 gas, which suppresses formation of oxygen in the cover gas and peroxides in solution. In low concentrations, HTO recovered from tritium removal systems has been found to be corrosive when stored in liquid form in metallic containers and “resulted in the development of significant leaks in containers within days or weeks”. Instead, the Standard says, “Storage of this same water solidified on clay or on molecular sieve material, regardless of the quantity, is stable and noncorrosive and may be stored for many years in the container”. A paper presented at the 2022 IFE Community Workshop,
‘Efficient tritium extraction from PbLi: a potential IFE breeding material, by T.F. Fuerst, C.N. Taylor, and M. Shimada,’ discusses some blanket concepts. It says potential tritium breeders are broadly classified
as solid or liquid. In all solid breeder concepts, tritium diffuses out of a ceramic, is carried away by a helium sweep gas, and is harvested in a tritium extraction system. The ceramic pebbles could be pebbles of Li2TiO3 or Li4SiO4 or a ceramic foam with a continuous internal pore network.
Liquid breeder examples may be pure lithium, lead/lithium mixtures or fluoride molten salt. Liquid breeders are not susceptible to neutron-induced mechanical damage and because they flow they can be continuously processed and replenished. Pure lithium has good corrosion compatibility with
structural materials but reacts violently with air and water. The high tritium solubility reduces tritium loss but makes extraction more difficult. FLiBe has a low electrical conductivity and high heat capacity. But it has a high melting point and Be poses a health safety concern. PbLi has a low melting point, low viscosity, and a much lower chemical reactivity
but it has high density and corrosion issues with structural steels.
The US DOE Standard differentiates storage requirements over the short, medium and long term (given tritium’s decay into helium with a half life of around 12 years). It says that where tritium has to be readily available to the facility customers it can be stored in gaseous form. The storage container should be fabricated of all-metal, hydrogen- compatible materials including valves, valve seats and seals. Storage up to two years is similar, and “Experience has shown that tritium can be stored safely at near atmospheric pressure for long periods of time. If the buildup of helium in the supply does not impact the use, then storage as a gas is an acceptable alternative.” However, it says that tritide bed storage allows impurities such as nitrogen and oxygen to be removed from the gas stream as the bed is heated and cooled, along with helium from tritium decay. In addition, metal tritides significantly reduce the volume required to store tritium, without increasing the pressure of the gas during storage. However, long-term storage of hydrogen and tritium
in containers is well understood, in comparison to the understanding of long-term storage of metal tritides. Decades of experience of dealing with tritium is now
coming together with blanket design options as plans move from concept to physical investigation. The concept and operation of breeding blankets are
Above: Tritium, as an isotope of hydrogen, presents physical hazards as it causes embrittlement in container materials
38 | September 2025 |
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being tested at the ITER facility. More than a decade ago, in 2013, ITER announced plans to experiment with tritium production within the vacuum vessel by way of test blanket modules (TBMs). Four different test blanket module concepts based on similar technologies to those described above were developed by the ITER members: water-cooled lithium-lead (Europe); water-cooled ceramics breeder (Japan); helium-cooled ceramic breeder (China); and helium-cooled ceramic pebbles (Europe/Korea). These concepts will be simultaneously installed in two equatorial ports of the ITER machine (‘hot cells’) and operated to test the efficiency of tritium breeding and extraction systems. In addition to demonstrating the generation of tritium within a closed fuel cycle, the programme will also experiment with different coolants for the future power-to-electricity conversion cycle. That programme has moved forward. In 2020 ITER announced that teams had successfully demonstrated the remote handling replacement of ITER’s test blanket modules that will be performed in the ITER hot cell. The remote handling replacement of the TBM sets is scheduled during every long-term maintenance period during fusion operation at ITER, or approximately every two years. ■
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