Fusion reactor designs that are fuelled by deuterium and tritium often include a lithium ‘blanket’ around the fusion site. This is where the necessary tritium fuel is ‘bred’: neutron bombardment causes the lithium to decay to tritium, which is extracted from the ‘blanket’ and ultimately recycled into the fusion reaction.
This cosy description masks a complex process in which the tritium being managed is not only a radiological hazard but a chemical hazard as well. Specific risks come with its nature as an isotope of hydrogen, but there are also physical hazards as it causes embrittlement in container materials, either due to radiological effects or due to the emergence of helium as a decay product.
Chemical interactions
The safe and secure management of tritium has been a nuclear industry focus for many decades. The US Atomic Energy Act of 1954, as amended (AEA) lists tritium as an “other” category of accountable nuclear material. Under this designation it must be controlled and accounted for financial and nuclear materials management purposes and protected in a manner consistent with its strategic and monetary importance. In the US the reportable quantity of tritium is one gramme, except for tritium contained in water used as a moderator in a nuclear reactor, which is not an accountable nuclear material. DOE requires tritium facilities with more than reportable 1 gramme quantity of tritium to establish material control and accountability systems to provide accurate nuclear materials inventory information while transactions exceeding the threshold must also be reported.
Hydrogen and tritium handling infrastructure typically uses Type 300-series austenitic stainless steels, due to their resistance to hydrogen isotope embrittlement. A 2021 paper in Fusion Engineering and Design: ‘Tritium embrittlement of austenitic stainless-steel tubing at low helium contents, by Timothy M. Krentz , Joseph A. Ronevich, Dorian K. Balch, Chris San Marchi’ discusses the effect of tritium and notes that there is some information on the performance of various alloys and some relationships have been established, such as correlation of embrittlement with hydrogen content and a positive effect of increasing nickel content. However, “there has been a lack of systematic studies focused on the effect that microstructural variation has on susceptibility to hydrogen embrittlement”.
Type 300-series austenitic stainless steels also have the advantage that they can be welded – the preferred joining strategy, as it is generally less leak prone than mechanical fittings. But the authors say stainless steels are not immune to embrittlement and exposure to tritium and its decay product (helium) reduces toughness. They say, “Tritium embrittlement is complicated by the combined effects of the hydrogen isotope and the aging effect of tritium’s helium decay product. Because helium is less soluble in the atomic lattice than hydrogen isotopes, mobile atomic helium ultimately becomes trapped as high-pressure nanoscale helium bubbles which are hypothesised to impede dislocation motion and thus alter plastic deformation”. They cite losses in fracture toughness of up to 95% for specimens with high helium buildup, while thermal history associated with welding tends to create microstructures and internal stresses that can be more sensitive to hydrogen-isotope embrittlement. Mechanisms of embrittlement in tritium environments are more complicated than for hydrogen, but “similar relationships have been established comparing composition and processing effects to mechanical performance in tritium environments”.
The US DOE’s Standard on Tritium Handling and Safe Storage, updated in 2015, includes experience with polymers, which all degrade when exposed to radiation. The standard says, “Both tritium and tritiated water permeate all polymers, and permeated tritium deposits the beta decay energy throughout the polymer bulk. Radiation-induced changes in polymer properties include softening or hardening, ductility loss, change in colour or dimensions, and gas evolution. It says, “Because of these effects, polymers should only be used in tritium systems where no metal alternatives exist”. In addition to polymer breakdown itself, products of degradation can form corrosive liquids or acids such as HF and HCl.
As a result, “The system should be designed to expose any polymers to as little tritium as possible” and the Standard notes that polymers are often used in gas-handling systems including in gaskets, O-rings, electrical cable insulation and valve parts, including seats, stem tips, and packing.
There are some flexibilities. Polymers that harden when exposed to radiation can be used, if they are replaced before they begin to deteriorate, but in this case all polymer parts have to be easily replaceable and regularly inspected and a replacement regime should be established. Protecting polymers from oxygen or air will help lengthen the lifetime of polymers exposed to tritium because many effects of radiation on polymers are accentuated by oxygen, and adding antioxidants may help. The temperature of any polymer parts should also be kept as low as possible as temperatures above about 120°C accelerate radiation effects. Finally, inert additives such as glass or graphite may enhance polymers’ resistance to radiation.
The US DOE Standard lists the effects on some common plastics.
- Ultra-high molecular weight polyethylene (UHMWPE) and high density polyethylene (HDPE) have been used for valve stem tips but “their performance has not reached the desired level”.
- Low-density polyethylene is “very permeable by tritium and tritiated water and should not be considered for use in tritium systems”.
- Teflon® (polytetrafluoroethylene) degrades and decomposes in tritium, forming hydrofluoric acid (HF). In humid air, both hydrochloric acid and HF are formed, which are both highly corrosive.
- Common chlorofluorocarbon polymers are incompatible with tritium. They, like Teflon®, degrade in tritium gas and should not be used.
- Elstomers with radiation damage harden and lose their sealing ability and, the Standard says, “tritium readily permeates into and diffuses through elastomeric materials and, depending upon thickness, begins appearing on the outside of the elastomeric seal within hours after exposure”.
The DOE Standard refers to decommissioning and dismantling of the Old Tritium Extraction Facility at Savannah River between 1995 and 1997 and says: “The damage done to organic materials by the presence of tritium in the internal structure of the material is not limited to the more obvious radiation damage effects”. Tritium, particularly in the form of T+, “has the insidious ability to leach impurities (and non-impurities) out of the body of the parent material”. It says, “In many cases, particularly where halogens are involved, the damage done by secondary effects such as leaching can be more 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 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.
