Apparatus and methods for extraction of trapped helium-3

Abstract

Helium-3 (He-3) trapped in containers and tritium trapped within titanium material are recovered. A glove box system has a leak-tight chamber filled with inert gas, connected to an inert gas tank, and an exhaust port connected to a ventilation system and a He-3 recovery system within the chamber. A heater heats the container to immobilize tritium to release He-3 and tritium trapped in titanium within the container. A vacuum system configured to recover He-3 and tritium released from the container. A getter bed purification system purifies, separates, and recovers He-3 and tritium released from the container. Also taught are evacuating the He-3 and tritium released by the heating into a recovery vessel, circulating the He-3 and tritium through one or more getter beds to remove impurities from He-3, including at least one uranium getter bed removing tritium from He-3 gas, storing tritium and He-3 gas separately.

Description

APPARATUS AND METHODS FOR EXTRACTION OF TRAPPED HELIUM-3CROSS-REFERENCE

[0001] This application claims priority from United States Application No. 63 / 749,979 filed January 27, 2025, the contents of which are incorporated herein by reference.TECHNICAL FIELD

[0002] The present invention relates to apparatus and methods for recovering helium-3 trapped in containers adapted to immobilize tritium.BACKGROUND

[0003] In nuclear power reactors that use heavy water as a coolant and a moderator, the heavy water gradually becomes tritiated. Tritiated heavy water is produced from neutron capture from the deuterium molecule of heavy water. In There is an existing TRF (Tritium Removal Facility) to remove tritium from heavy water (S.K. Sood et.al. (1985) Removal and Immobilization of Tritium from Ontario Hydro’s Nuclear Generating Stations, Fusion Technology, 8:2P2, 2478-2485). The tritium can be extracted from the heavy water in the reactor and stored in containers adapted to immobilize tritium, called Immobilized Tritium Containers (ITCs) in the TRF (N.P. Kherani & W.T. Shmayda (1985) Gas Handling Systems Using TitaniumSponge and Uranium Bulk Getters, Fusion Technology, 8:2P2, 2399-2406). The ITCs contain titanium sponges that capture tritium within the containers. Over time, tritium decays into helium-3, a stable helium isotope, which increases pressure within the ITCs. The ITCs from the TRF can be safely depressurized using a tool to extract free He-3 gas within the ITCs and compress the He-3 into bottles which can be sold using a specialized tool. The tool is not able to extract all of the He-3 as some if it remains trapped in the titanium sponge in the ITCs.

[0004] There is a need for apparatus and methods that can extract He-3 trapped in the ITCs.

[0005] Heung, LK (Titanium for long-term tritium storage (U), WestinghouseSavannah River Company, WSRC-TR-94-0596, 1994) teaches that when the He-3 to titanium mole ratio is higher than 0.4 He-3:Ti, the evolution rate of free He-3 becomes significant at temperatures as low as 200°C. Analysis of the Phase-1 tool shows a similar result, namely there is insignificant free He-3 in ITCs with a He-3:Ti mole ratio lower than 0.36. The Heung study showed that a minimum temperature of 700°C is required to remove He-3 when the He-3:Ti mole ratio is about 0.3 and below. A further study by Zhou et.al. investigated He-3 release from titanium tritide film (Zhou X, Peng S, Long X and Luo S. Progress of helium evolution in aging titanium tritide film. Fusion Science and Technology, vol.60 P905-909, 2011.) In the Zhou study, the maximum temperature required to release He-3 from the titanium tritide films having a 0.325 He-3:Ti mole ratio was between 447°C-607°C. Therefore, there is a need in the art for apparatus and methods to extract He-3 trapped in titanium within ITCs. There is also a need to separate and capture the tritium that will be simultaneously released from the ITCs with He-3 .

[0006] The use of vacuum to recover gases is known, for example (see, e.g. US2022037047 paragraph

[0013] , EP3053638

[0004] ). The use of heat to recover gases is known, for example (e.g. CN104538074A, abstract; EP3053638

[0037] ). Methods to recover gases, including He-3, are also known in the art (see for example US2022065760AA, US2002066288AA, DE19927773A1 and EP3742456A1). However, these methods do not address recovering He-3 trapped in titanium with tritium nor do they address methods of separating the He-3 and tritium.

[0007] Each gas has its own unique properties and challenges in determining a useful method for recovery. Furthermore, the context in which the gas is found can create significant challenges for recovery.

[0008] Methods of recovering He-3 are known, for example US2002066288AA teaches a system for recovering He-3 from titanium dioxide, including a solar collector on a satellite suitable for receiving ilmenite ore (FeTiO3) containing titanium dioxide with He-3 adsorbed therein, said solar collector being adapted to heat the titanium dioxide. KR20120086544A teaches that He-3 can be separated from stored tritium in heavy waterthrough a membrane. CN112331366 teaches a deuterium and tritium fuel storage system that includes a He-3 recovery unit. There is still a need for apparatus and methods for removal of residual He-3 from trapped in titanium within containers that immobilize tritium.

[0009] It is also known in the art that tritium will contaminate the extracted He-3 rendering it unsuitable for further use. CN111495177A discloses a system and method for tritium removal and purification. The patent teaches the release of tritium. The system is operated within a sealed container (glove box) to contain the release of tritium and removes impurities including using adsorption beds and drying using a molecular sieve. This application does not contemplate recovery and separation of He-3 from the tritium from recovered He-3.

[0010] Gas purification apparatus are known in the art, for example, US20150004053A1 , which discloses a glovebox system for removing contaminants from an inert atmosphere. US20150004053A1 does not contemplate separation of tritium or recovery of He-3. Glovebox cleanup systems are also used in TRFs to remove tritium, moisture, oxygen and excessive heat from the inert gas environment inside the glovebox and to provide a leak tight system for working with isotopes. The glovebox cleanup system can comprise a blower, chiller, flow sensor and individual getter beds to purify gases. (Heics et.al. (1988) Ontario Hydro Research Division Tritium Laboratory: Design and Operating Experience, Fusion Technology, 14:2P2B, 1277-1281).

[0011] Methods and systems are known for tritium storage. GB2097991 A1 discloses a container for immobilizing tritium for long term storage. CN112331366A discloses a system to store deuterium-tritium fuel. Methods and systems are also known for immobilizing or trapping tritium (see e.g. WO19078758A2, WO16034745A2, DE102015016873A1 and US2017169907AA). Methods are also known in the art for use of titanium for gas storage (CN104538074A, US5602886A, US2022037047AA and GB2097991A1). However, there remains a need for apparatus and methods to remove He-3 trapped in titanium matrices within tritium storage vessels.

[0012] Methods are known for separation or recovery of He-3 or tritium. US2016019993AA discloses a means and method to separate a collect tritium from molten salt reactors. CN117819497A discloses a multi-membrane coupling process for extracting He-3 from a high tritium heavy water tank headspace mixture. CN116913558A discloses a gas cooled reactor and fusion reactor to obtain tritium and / or He-3. EP3053638A1 discloses a process for detritiation of tritiated water. CN117623241A discloses a method and system to extract He-3 from high-tritium heavy water mixed gas. KR20120086544A discloses separating tritium and He-3 through a membrane. JP7151879A2 discloses a He-3 gas recovery system and He-3 gas circulating system for He-3 generated in tritium waste. CN115631877A discloses producing He-3 from the heavy water moderator of a heavy water reactor. US3794875A discloses a pump that can remove He-3 from the deuterium and tritium. US2022037047AA discloses a container for tritium with a He-3 diffusing window. Foster et.al. discloses that He-3 accumulated in hydride beds from tritium decay can be removed through baking the bed beyond its normal operating temperature (P. J. Foster et.al. (2021) Hydride Bed Helium-3 Recovery and Partial Regeneration, Fusion Science and Technology, 77:3, 195-198). Jung et.al. discloses He-3 produced by tritium decay in storage beds can separated from the beds by a gas circulation process (Pil-KapJung et.al. Helium blanketing and gas circulation effect of hydriding reaction with ZrCo, Fusion Engineering and Design (2020) 157:111626). However, these methods do not address recovering He-3 trapped in titanium with tritium and / or the separation of tritium from He-3.SUMMARY

[0013] Example embodiments relate to apparatus and methods for recovering helium-3 (He-3) trapped in containers adapted to immobilize tritium. In some aspects the present invention relates to an apparatus and method for use in a Tritium Recovery Facility (TRF) for recovering helium (He-3) trapped in containers adapted to immobilize tritium.

[0014] An example embodiment of the present invention is an apparatus for recovering helium-3 (He-3) trapped in containers adapted to immobilize tritium containing a titanium material, tritium trapped within the titanium material and He-3 trapped within the titanium material, the apparatus comprising:a glove box system comprising at least one leak-tight chamber filled with inert gas, a port connected to an inert gas tank, an exhaust port connected to a ventilation system and a He-3 recovery system configured to operate within the chamber filled with inert gas, the He-3 recovery system comprising:a heater for heating a container adapted to immobilize tritium to release He- 3 and tritium trapped in titanium within the container;a vacuum system configured to recover He-3 and tritium released from a container adapted to immobilize tritium; anda getter bed purification system for purifying, separating, and recovering He-3 and tritium released from a container adapted to immobilize tritium.

[0015] In one aspect, the container adapted to immobilize tritium further contains free helium-3 gas.

[0016] In one aspect, in the glove box system further comprises an antechamber configured to be filled with inert gas for transferring the container adapted to immobilize tritium into the leak-tight chamber filled with inert gas.

[0017] In one aspect the heater is configured to heat the container adapted to immobilize tritium to a temperature providing maximal release of helium-3 from titanium inside the container and not damaging the glove box system.

[0018] In one aspect the heater is configured to heat the container adapted to immobilize tritium to a temperature between 600-750°C.

[0019] In one aspect the vacuum system comprises a vacuum pump; a recovery vessel for capturing the helium-3 and tritium gas released from the container adapted toimmobilize tritium; lines connected to the container to the vacuum pump and connected to the vacuum pump to the recovery vessel; and valves for controlling the flow of gas through the vacuum system.

[0020] In one aspect the getter bed purification system comprises:at least one nickel getter bed for removing oxygen and nitrogen impurities; at least one molecular sieve for removing moisture;at least one uranium storage getter bed for capturing and storing tritium; at least one zirconium-iron getter bed for removing any remaining tritium not captured by the uranium getter beds;at least one storage container for recovering helium-3 gas separated from tritium;lines connected to the recovery vessel of the vacuum system, the getter beds and the storage container;at least one pump for circulating gas through the getter bed purification system;and valves for controlling the flow of gas through the getter bed purification system.

[0021] In one aspect the helium-3 recovery system further comprises a gas analyzer configured to monitor the level of tritium, helium-3 and impurities in gas circulating within the helium-3 recovery system.

[0022] In one aspect the gas analyzer is a mass spectrometer or a gas chromatograph.

[0023] In one aspect the getter bed purification system further comprises at least one secondary clean-up uranium getter bed for removing excess tritium when the tritium levels flowing through the glove box exhaust port are greater than 100 pCi / m3.

[0024] In one aspect the glove box system further comprises a second chamber filled with air and the helium-3 recovery system further comprises a secondary cleanup systemadapted to operate in the second chamber for maintaining the leak-tight chamber filled with inert gas free of gas impurities.

[0025] In one aspect the secondary cleanup system comprises,at least one tritium monitor, at least one oxygen monitor and at least one moisture monitor connected to the leak-tight chamber filled with inert gas for monitoring tritium, oxygen and moisture content within said chamber;a secondary purification system comprising, at least one uranium getter bed for capturing tritium; at least one nickel getter bed for removing oxygen; at least one molecular sieve for removing moisture; and at least on zirconium-iron getter bed for removing any trace tritium not captured by the one or more uranium getter beds.a vacuum system configured to receive gas from the helium-3 recovery system and the antechamber and configured to circulate the gas through the secondary purification system;lines connected to the antechamber, the leak-tight chamber filled with inert gas and the helium-3 recovery system to the secondary clean-up system;at least one pump for circulating gas through the secondary clean-up system getter beds and lines;valves for controlling the flow of gas through the secondary clean-up system; andan exhaust port connecting the secondary clean-up system to the ventilation system.

[0026] In one aspect the apparatus further comprisesat least one line to connect the getter bed purification system to a new, tritium-free container adapted to immobilize tritium for capturing tritium released from the one or more uranium storage getter beds when they are heated; and valves for controlling the flow of the tritium into the new, tritium-free container adapted to immobilize tritium.

[0027] In one aspect the helium-3 recovery system has a capacity to capture over 600,000 Ci of discharged tritium.

[0028] In one aspect the apparatus is configured to operate automatically.

[0029] In one aspect the inert gas is argon gas.

[0030] In one aspect the apparatus further comprises an electric hoist attached to the top of the glove box system for placing the container adapted to immobilize tritium into the heater.

[0031] Another example embodiment of this invention is a method of recovering helium-3 (He-3) trapped in containers adapted to immobilize tritium containing a titanium material, tritium trapped within the titanium material and He-3 trapped within the titanium material, the method comprising:moving a container adapted to immobilize tritium into a glove box chamber filled with inert gas;placing the container adapted to immobilize tritium into a heater; energizing the heater;heating the container adapted to immobilize tritium;evacuating the He-3 and tritium released by the heating into a recovery vessel;circulating the He-3 and tritium through one or more getter beds to remove impurities from He-3, the one or more getter beds including at least one uranium getter bed capable of removing tritium from He-3 gas and storing tritium; and transferring the He-3 gas to a storage vessel.

[0032] In one aspect the step for circulating the helum-3 (He-3) and tritium through one or more getter beds to remove impurities from the He-3 further comprises,circulating the He-3 and tritium through one or more nickel getter beds to remove oxygen impurities;circulating the He-3 and tritium through one or more molecular sieve getter beds to remove moisture; andcirculating the He-3 through one or more zirconium-iron getter beds to remove any traces of tritium not capture by the one or more uranium getter beds.

[0033] In one aspect the container adapted to immobilize tritium further contains free helium-3 (He-3) gas and following the step of placing the container into a heater the method further comprises,vacuuming any He-3 gas from the container into a recovery vessel; circulating the He-3 gas from the recovery vessel through one or more getter beds to remove impurities from the helium-3 gas;capturing tritium by circulating the gas through at least one uranium storage getter bed; andtransferring the He-3 gas to an outside storage vessel.

[0034] In one aspect the method further comprises,heating the one or more getter beds adapted to remove tritium from He-3 to a temperature sufficient to release tritium from the uranium storage getter bed; andrecovering tritium gas released from the heated uranium storage getter bed in a new, tritium-free container adopted to immobilize tritium.

[0035] In one aspect, the uranium getter beds are heated to 350-450°C to release tritium.

[0036] In one aspect the method further comprises performing an isotope exchange with deuterium or protium gas to remove any residual tritium remaining in the one or more getter beds and disposing of the gas used in the isotope exchange by treating the gas to remove tritium or capturing and storing the gas used in the isotope exchange in containers adapted to prevent tritium release.BRIEF DESCRIPTION OF THE DRAWINGS

[0037] Reference will now be made, by way of example, to the accompanying drawings which show example embodiments, and in which:

[0038] Fig. 1 is a schematic diagram of an example embodiment of an apparatus for recovering He-3 trapped in containers adapted to immobilize tritium.

[0039] Fig. 2 is a process flow diagram for an example embodiment of a method of recovering He-3 trapped in containers adapted to immobilize tritium.DETAILED DESCRIPTION

[0040] Throughout the following, specific details are set forth in order to provide a more thorough understanding to persons skilled in the art. However, well-known elements may not have been shown or described in detail. Accordingly, the description and drawings are to be regarded in an illustrative, rather than a restrictive, sense.

[0041] Reference will be made below in detail to exemplary embodiments of the invention, examples of which are illustrated in the accompanying drawings.

[0042] Figure 1 illustrates, according to one example embodiment, an apparatus for recovering helium-3 (He-3) 100 trapped in containers adapted to immobilize tritium 101.

[0043] The containers adapted to immobilize tritium 101 contain a titanium material, tritium trapped within the titanium material and He-3 trapped within the titanium material (not shown). The container adapted to immobilize tritium 101 may further contain free He-3 gas (not shown).

[0044] At the Tritium Removal Facility (TRF) designed to extract tritium from heavy water, the tritium is stored in containers adapted to immobilize tritium called Immobilized Tritium Containers (ITCs). Over-time tritium decays into He-3, a stable helium isotope, increasing the pressure within the ITCs. The half-life of tritium is 12.323 years. The ITCs can besafely depressurized, and the free He-3 gas within the ITCs recovered through a Phase I tool available in the TRF that can extract up to 10,000 liters of He-3 annually. The ITCs, as an example of containers adapted to immobilize tritium, are composed of a 316 stainless steel exterior and are designed to load 500,000 Ci of tritium. The container is designed to accommodate the loading of tritium, the degassing of helium and the recovery of tritium with heating. The tritium is loaded into approximately 841 grams of titanium sponge material at 50% loading capacity. Tritium activity in the ITCs can vary form 1000 - 500, 000 Ci with gas purity from 10-99.9%. The ITCs can hold He-3 for up to 50 years with a maximum pressure capacity of 86000 kPa. The Phase I tool allows the ITCs to be degassed to a pressure of 100 kPa and returned to a storage vault maintained at a temperature below 40°C. However, there is still a significant amount of tritium left to decay and He-3 trapped within the titanium sponge. Effective recovery of He-3 using the Phase I tool requires typically 20 years of aging. Once helium and tritium are extracted to the maximum extent possible, the ITCs can be disposed of as Low Level or Intermediate Level Waste. He-3 recovered from a container adapted to immobilize tritium can be resold to provide a source of He-3. Further, ITCs with greater than 185,000 Ci of tritium activity have effective He-3 recovery even under heat and vacuum.

[0045] Tritium is a beta emitter that decays into He-3 by emitting a beta particle, also known as an electron, and an antineutrino from a neutron in its nucleus. The tritium decay and helium formation over time can be estimated using the following equations:Mole of He- 3 (rriHe3 at t years) where mT2(initiai) is the initial amount of tritium:&Moles of tritium at t years (rriT2 at t years) where rriT2(initiai) is the initial amount of tritium:

[0046] Using these equations, the decay of tritium and the formation of He-3 can be calculated and expressed as a percentage over ten half-lives in containers adapted toimmobilize tritium 101. Also relevant to the equation are the following parameters known to the person of skill in the art. Standard Temperature and Pressure (STP) Conditions are 760 torr, 0°C (1 atm and 273.15 K) and 1 mole of ideal gas at STP = 22.414 L and for Tritium activity, 1 Ci = 3.7x1O10disintegrations / second = 3.7x1O10Bq = 37 GBq, Half-life of tritium = 12.323±0.004 years, Tritium (T2) gas = 9619 Ci / g and Tritium gas contains 58023 Ci / mole. Table 1 below shows the calculated percent tritium decay and helium formation with age. By measuring the initial tritium activity data when containers adapted to immobilize tritium 101 are loaded the number of moles of tritium in the container can be calculated. The initial moles of tritium in the above equations can be used calculate the tritium activity over time and the formation of He-3 as expressed as the percentage formation of He-3 and tritium decay. These percentages are critical for selecting containers for maximum He-3 recovery and determine the optimal age to dispose of the containers.Table 1: Percent tritium decay and helium formation with age

[0047] After 40 years, tritium activity is reduced to 10.54% of its initial value, corresponding to an 89.47% formation of He-3. The rate of reduction is not significant after 80 years. The residual tritium activity is a crucial factor for maximum helium recovery and the disposal of containers adapted to immobilize tritium 101.

[0048] Titanium metal, including that within containers adapted to immobilize tritium 101 or ITCs, is saturated with tritium and stored at room temperature. The titanium metal will retain decaying He-3 until the He-3 content reaches 0.3 moles of He-3 per mole of tritium at which point the He-3 released by tritium decay enters the void space in the ITC. Beyond this point, the He-3 release rate equals the production rate by tritium decay. For example, an ITC contains 841 grams (17.57 moles) of titanium, and the timelines for helium release are 6.5 years for an initial tritium activity of 500,000 Ci and over 20 years for 200,000 Ci at 0.3 He / Ti mole ratio, extending to 8.13 years and over 30 years, respectively, at 0.36 He / Ti mole ratio. To analyze the extraction of free He-3 using only a tool to depressurize the ITCs, the free He-3 gas was recovered in the TRF. The He-3 recovery data was analyzed for He-3 retention in the titanium matrix, considering initial tritium activity, gas pressure in the ITCs before and after extraction, and average He-3 recovery using the specialized He-3 tool. Taking into account the ITC container's free volume (6.035 liters), titanium metal content (841g), and pressure exerted by the free He-3 gas (initial pressure), the average mole ratio of He-3 to Ti moles retained in the titanium metal wasestimated. The corresponding average He-3 to titanium mole ratio is 0.36. He-3 gas above this ratio is extractable using the He-3 tool, while helium below this ratio requires heat and vacuum for recovery. The volume left in each ITC container, maintaining positive pressure to reduce air ingress, is around 21.3 L (STP), which can be extracted without heat. The minimum, maximum, and average values of the ITC extracted data are provided below in Table 2.Table 2 Helium Recovery data using Helium-3 specialized tool

[0049] Based on the analysis of the free He-3 extraction data and operational practice, the following assumptions can be made for estimating the free He-3 gas for extraction without heat and vacuum: the average mole ratio of free He-3 to titanium sponge metal is 0.36, the helium gas formed due to tritium decay will be trapped inside the titanium metal until reaching a ratio of 0.36, the gas formed will exert pressure, and the trapped gas inside the titanium metal at 0.36 He / Ti is 141.77 L (STP). The average gas pressure after extraction using the specialized tool to depressurize the ITC is 57.88 psia, corresponding to 21.3 L (STP) of free He-3 remaining in the ITC. ITCs aged 20 years and above would provide an extractable gas volume of over 100 L (STP) using Phase-1 tool.However, heat and vacuum are required to extract the remaining He-3 trapped in titanium within ITCs. Heating the titanium will also release tritium which requires the tritium to be safely captured and stored as well as a the He-3 extraction to be performed in a sealed environment such as inside a glove box.

[0050] The apparatus 100 illustrated in Figure 1 comprises a glove box system 104 and a He-3 recovery system 103 configured to operate within the chamber filled with inert gas 102. The glove box comprises at least one leak-tight chamber filled with inert gas 102, a port for connecting to an inert gas tank 106, an exhaust port for connecting to a ventilation system 124. In some embodiments the inert gas is argon gas. The glove box system 104 further contains in some embodiments, an antechamber 127 configured to be filled with inert gas for transferring the container adapted to immobilize tritium 101 into the leak-tight chamber filled with inert gas 102. The apparatus in an example embodiment further comprises an electric hoist (not shown) attached to the top of the glove box system 104 for placing the container adapted to immobilize tritium 101 into the heater 105.

[0051] The glove box system 104 can have glass windows on the front, permits gloved access to handling areas, and provides easy access to valves and instrumentation. Controls located inside the glove box system 104 may be fully visible on an Human Machine Interface (HMI) display monitor located outside the glove box system. The glove box system 104 can also further comprise a fan 135 and chiller 136 to regulate gas flow and temperature within the glove box system 104.

[0052] The He-3 recovery system 103 comprises a heater 105 for heating a container adapted to immobilize tritium 101. The heater causes the release of He-3 and tritium trapped in titanium within the container 101. The He-3 recovery system also comprises a vacuum system 108 configured to recover He-3 and tritium released from a container adapted to immobilize tritium 101 and a getter bed purification system 113 for purifying, separating, and recovering He-3 and tritium released from a container adapted to immobilize tritium 101. The He-3 recovery system 108 may also be designed to allow forevacuation of all process lines and purging and flushing of all process lines with helium or argon.

[0053] The heater 105 in some embodiments heats the container adapted to immobilize tritium 101 to a temperature providing maximal release of He-3 from titanium inside the container 101 while also not damaging the glove box system 104. The heater 105 can be configured to heat the container adapted to immobilize tritium 101 to a temperature between 600-750°C. As an example of containers adapted to immobilize tritium 101 , ITCs are made of 316 stainless steel which has a melting point of 1370-1398°C. This stainless steel has good oxidation resistance in intermittent service up to 870°C, however continuous use of stainless steel 316 is not recommended in the range of 425-860°C. The maximum recommended design temperature of ITCs is 550°C under vacuum (10-5to 10’6Pa(a)) therefore, ITCs exposed to this temperature should be disposed of afterwards. In an example embodiment, the temperatures between 650-700°C best preserve glove box system 104 integrity. The ITCs can be heated through surface heating with heat transported to the interior through conduction. Table 4 below shows He-3 recovery with time using various temperatures to heat ITCs.Table 4 He-3 recovery from ITCs using heat with timeTemperature Helium Release (L, He / Ti Remarks References (°C) STP) MoleRatio600 15.75 0.325 Feasible 8700 25.6 0.3 Limited by 7Glove boxoperatingconditions1000 64.97 0.2 Not feasible 11400 104.35 0.1 Not feasible 1

[0054] Since the ITC heaters operate between 650°C and 700°C, the titanium metal sponge temperature can reach around 600°C but requires a longer time to achieve 700°C. Thus to estimate He-3 recovery with heat, a He-3 recovery of 15.75 L (STP) per ITC container is assumed, with the potential to achieve up to 25.6 L (STP) by reaching 700°C.

[0055] he vacuum system 108 in some embodiments comprises a vacuum pump 109, a recovery vessel 111 for capturing the helium-3 and tritium gas released from the container adapted to immobilize tritium 101 , lines for connecting the container to the vacuum pump 112 and for connecting the vacuum pump to the recovery vessel 110 and valves 110 for controlling the flow of gas through the vacuum system. Using combined heating and vacuum on the example of ITCs that had first been depressurized of free He-3 gas was recovered with heating and vacuum.

[0056] The getter bed purification system 113 in one embodiment comprises at least one nickel getter bed 114 for removing oxygen and nitrogen impurities, at least one molecular sieve 115 for removing moisture, at least one uranium storage getter bed 116 for capturing and storing tritium and at least one zirconium-iron getter bed 118 for removing any remaining tritium not captured by the uranium getter beds 116. The getter bed purification system 113 also comprises at least one storage container 126 for recovering helium-3 gas separated from tritium, lines 120 for connecting the recovery vessel of the vacuum system, the getter beds and the storage container, at least one pump 121 for circulating gas through the getter bed purification system 113 and valves 119 for controlling the flow of gas through the getter bed purification system. The operating temperatures for the nickel 114 and zirconium-iron 118 getter beds are 400°C and 350°C, respectively. The uranium getter beds 116 are operated at ambient temperature to capture and store tritium. The getter beds annuluses (not shown) and all lines in the He-3 recovery system may be vacuumed using a vacuum pump 109 in the system prior to using the apparatus 100 to extract He-3. The system lines can also be purged with helium when the He-3 extraction is complete.

[0057] Tritium immobilized in titanium within containers adapted to immobilize tritium 101 will be discharged from the container into the He-3 recovery system 103 once heat and vacuum is applied to the container 101. Based on the equilibrium isotherms, the percentage of hydrogen isotopes that can be directly removed by heat and vacuum is a function of temperature and pressure. A discharge temperature of 600°C and a vacuumof 1 torr are required to recover 95% of the hydrogen isotopes in a reasonable amount of time. The recovery is also sensitive to pressure, with a slight increase in operating pressure resulting in a drop in recovery efficiency. Table 5 shows that at 1 torr and 530°C, a 95% recovery rate can be achieved. However, with a slight increase in pressure to 2.2 torr, the required temperature increases to 590°C. This indicates that maintaining the required vacuum levels is crucial for maximum tritium recovery. If the required vacuum levels are not maintained, the efficiency of tritium recovery will be significantly impacted.Table 5 Tritium recovery with temperature and pressure

[0058] An analysis using the ITCs located in the TNF facility calculated the average initial tritium activity with gas purity of ITCs (Table 6). The average maximum initial tritium activity is 488,889 Ci for ITCs with 99% and above purity, while the minimum average value is 356,644 Ci for ITCs with gas purity less than 90%. The average initial tritium activity of 488,889 Ci for gas purity 99% and above is considered for estimating tritium activity with age. Therefore, in ITCs 95% of tritium is recoverable tritium and the residual tritium activity remaining will be 5% after applying heat and vacuum to the ITC.Table 6 The minimum, maximum and average tritium initial activity of ITCs with gas purity> > < > <<

[0059] The He-3 recovery system 103 in some embodiments further comprises a gas analyzer 137 configured to monitor the level of tritium, He-3 and impurities in gas circulating within the He-3 recovery system 103. The gas analyzer 137 can be a mass spectrometer or a gas chromatograph. The gas analyzer 137 monitors tritium content in the gas recovered in the recovery vessel 111 prior to extracting He-3 to monitor for impurities, after He-3 and tritium are released into the recovery vessel 111 and after the He-3 gas is purified through getter beds. A line with a valve connects the recovery vessel 111 to the gas analyzer 137.

[0060] In one aspect the getter bed purification system 113 further comprises at least one secondary clean-up uranium getter bed 117 for removing excess tritium when the tritium levels flowing through the glove box exhaust port is greater than 100 pCi / m3. The apparatus can further comprise a tritium monitor 138 to detect tritium in the He-3 recovery system 103. The secondary clean-up uranium bed 117 is used to detect excessive levels of tritium in the exhaust before— and after using the apparatus of any of the example embodiments. If the tritium level is greater than 100 pCi / m3the glove box system 104 should alarm, shut the exhaust port valves and redirect the gas to the one or more uranium clean-up beds 117 to remove excess tritium before directing the exhaust. Line 143 is used to clean up the beds and the lines before and after using the apparatus. If the tritium level exceeds the specified limit, T2 monitor 170 should trigger an alarm and shut the exhaust port in the vent line (not shown), redirecting the gas to one or more uranium cleanup beds (117) to remove excess tritium before directing the exhaust. The exhaust line 132 connecting the exhaust port of the apparatus to the ventilation system 133 and can further comprise a tritium monitor and fan.

[0061] According to some embodiments, the glove box system 104 further comprises a second chamber filled with air 141 and the helium-3 recovery system 103 further comprises a secondary cleanup system 140 adapted to operate in the second chamber for maintaining the leak-tight chamber filled with inert gas 102 free of gas impurities.

[0062] The secondary cleanup system 140 in some embodiments comprises at least one tritium monitor 151 and 167, at least one oxygen monitor 151 and 165 and at least onemoisture 151 and 166 monitor, a secondary purification system 153 comprising, a vacuum system 142, a pump 150, an exhaust port 172 and lines with one or more valves 155 and 160 connecting the components of the secondary clean up system 140. The monitors 151 are connected through lines 149 to the leak-tight chamber filled with inert gas 102 for monitoring tritium, oxygen and moisture content within said chamber 102. The secondary purification system 140 comprises in some embodiments at least one uranium getter bed 158 for capturing tritium; at least one nickel getter 156 bed for removing oxygen; at least one molecular sieve 157 for removing moisture and at least one zirconium zinc getter bed 159 to capture trace amounts of tritium not captured by the one or more uranium getter beds 158. The vacuum system 142 is configured to receive gas from the He-3 recovery system 103 and the antechamber 127 and configured to circulate the gas through the secondary purification system 140. The line 145149 and 143 connects the antechamber 127, the leak-tight chamber filled with inert gas 102 to the glove box inert gas environment. The metal bellows pump 150 and vacuum pumps 144 and 146 circulate gas through the secondary clean-up system 140 getter beds and lines. Valves along the lines allow for flow of gas to be controlled through the secondary clean up system 140. The exhaust port 172 connects the secondary-clean up system 140 to the ventilation system 133. A tritium monitor 170 and a fan are located along the lines connecting the exhaust port 172 to the ventilation system 133. The metal bellows pump 150 draws gas from the inert gas environment in the upper glove box 102 to monitor oxygen, moisture, and tritium levels in the upper glove box's inert gas environment and circulates it back. Vacuum lines 143 and 144 are used for vacuuming the lines in the helium recovery process line and equipment (101, 106, 105, 110, 111, 114 to 119, 121, and 138). Vacuum lines 145 and 146 are used for vacuuming the antechamber 127 and the annulus of getter beds (114-118) in the process system.

[0063] The secondary clean up system 140 monitors and removes any gas impurities such as oxygen, nitrogen, and moisture that may enter the glove box chamber filled with inert gas 102 due to air leaks or tritium leaks from the He-3 recovery system 103 process lines and components. If impurities such as oxygen and moisture are detected without tritium presence, the system discharges the contaminated gas into the exhaust 172 andreplaces it with fresh argon to maintain positive pressure. If tritium is detected, the secondary clean up system removes it by circulating the gas through secondary clean up system getter beds, including uranium getter beds 158 for tritium storage, and finally through zirconium-iron beds 159 for trace gas removal.

[0064] In one aspect the apparatus 100 further comprises at least one line with a valve 122 to connect the helium-3 purification system to a new, tritium-free container adapted to immobilize tritium (not shown). The new, tritium-free container is for capturing tritium released from the uranium storage getter bed 116 when it is heated. The apparatus 100 also comprises a valve for controlling the flow of the tritium into the new, tritium-free container adapted to immobilize tritium. The containers can be calibrated to enable the loading of tritium onto the new container and to assay of tritium.

[0065] The He-3 recovery system 103 in some example embodiments has a capacity to capture over 600,000 Ci of discharged tritium. The apparatus 100 can be configured to operate automatically.

[0066] Another example embodiment of this invention is illustrated by Figure 2. Figure 2 shows a method of recovering helium-3 (He-3) trapped in containers adapted to immobilize tritium containing a titanium material, tritium trapped within the titanium material and He-3 trapped within the titanium material. The method comprises steps of moving a container adapted to immobilize tritium into a glove box chamber filled with inert gas 202, placing the container adapted to immobilize tritium into a heater 204, energizing the heater 206, heating the container adapted to immobilize tritium 208, evacuating the He-3 and tritium released by the heating into a recovery vessel 210, circulating the He-3 and tritium through one or more getter beds to remove impurities from He-3 212 and transferring the He-3 gas to a storage vessel 214. The one or more getter beds include at least one uranium getter bed capable of removing tritium from He-3 gas and storing tritium 214.

[0067] The step for circulating the He-3 and tritium through one or more getter beds to remove impurities from He-3 further comprises in one embodiment, circulating the He-3 and tritium through one or more nickel getter beds to remove oxygen impurities, circulating the He-3 and tritium through one or more molecular sieve getter beds to remove moisture and circulating the He-3 through one or more zirconium-iron getter beds to remove any trace tritium remaining that was not captured by the uranium getter beds. The gas flows through the getter beds in the order of nickel getter beds, molecular sieves, uranium storage getter beds and zirconium-iron getter beds.

[0068] If the container adapted to immobilize tritium contains free He-3 gas then after placing the container into a heater the method further comprises, vacuuming any He-3 gas from the container adapted to immobilize tritium into a recovery vessel, circulating the He-3 gas from the recovery vessel through one or more getter beds to remove any impurities from the He-3 gas, capturing tritium by circulating the gas through at least one uranium storage getter bed and transferring the He-3 gas to an outside storage vessel. The remaining steps of the method illustrated in Figure 2 are then followed.

[0069] The method in an example embodiment may further comprise heating the one or more uranium getter beds adapted to remove tritium from He-3 to a temperature sufficient to release tritium from the uranium getter bed and recovering tritium gas released from the heated uranium getter bed in a new, tritium-free container adopted to immobilize tritium. The uranium getter beds in one embodiment are heated to 350-450°C to release tritium.

[0070] In one aspect the method further comprises performing an isotope exchange with deuterium or protium gas to remove any residual tritium remaining in the one or more getter beds and disposing of the gas used in the isotope exchange by treating the gas to remove tritium or capturing and storing the gas used in the isotope exchange in containers adapted to prevent tritium release. Using ITCs contained in the TRF facility as an example, 5% of the tritium within an ITC, the residual tritium, that cannot be directly recovered by heat and evacuation can be recovered through isotope exchange at about 500°C. The number of exchanges required depends on several variables, including theamount of tritium to be exchanged, the amount of deuterium (or protium) used in each exchange, and the desired completeness of recovery. Typically, five to ten exchanges with about 20% of the capacity gas will reduce the residual tritium concentration by a factor of ten. To estimate the amount of hydrogen or deuterium required for this process, the initial amount of residual tritium, the efficiency of each exchange, and the capacity of the gas used in each exchange were considered. The hydrogen or deuterium used in isotope exchange will be tritium-contaminated and must be disposed of properly, either by treating the gas to remove tritium or by capturing and storing the contaminated gas in secure containers designed to prevent tritium release.

[0071] The proposed He-3 recovery apparatus and method embodiments of the present invention are capable of performing the isotope exchange process to reduce residual tritium levels in containers adapted to immobilize tritium. After reducing the residual tritium to an acceptable level through isotope exchange, the ITC will be disposed of as low-level solid or intermediate-level solid waste following established procedures. ITCs are designed for one-time use because heating to 600°C in the presence of tritium saturates the vessel walls with tritium at equilibrium levels, causing surface contamination and potentially weakening the vessel walls. This approach ensures the safe and efficient recovery of residual tritium, proper disposal of contaminated hydrogen or deuterium, and compliance with waste disposal regulations.

[0072] The methods of the example embodiments can further comprise monitoring the tritium, oxygen and moisture impurities in a glove box chamber filled with inert gas. The methods may comprise a further step of monitoring impurities in a second glove box chamber containing a secondary cleanup system wherein the system may alarm, shut off flow to the exhaust and scavenge tritium in one or more uranium getter beds when tritium levels in the glove box chamber filled with inert gas exceed 100 pCi / m3

[0073] All patent and other publications referenced herein are incorporated by reference into this specification as if the entire contents were set out in this specification. The embodiments illustrated and described herein are intended to be examples only.Those of skill in the art may effect alternations, modifications and variations to the embodiments without departing from the intended scope of the present invention. In particular, features from one or more of the embodiments may be selected to create alterative embodiments comprised of a subcombination of features not described above. In addition, features from one or more of the described embodiments may be selected or combined to create alternate embodiments. Any dimensions provided in the drawings are for illustrative purposes only. Although the foregoing invention has been described in some detail by way of illustration and example for purposes of clarity of understanding, the description is not intended as a limitation on the scope of the present disclosure but is instead provided as a description of exemplary embodiments.

Claims

1. WHAT IS CLAIMED IS:

1. An apparatus for recovering helium-3 (He-3) trapped in a container adapted to immobilize tritium, said container containing (a) a titanium material, (b) tritium trapped within the titanium material and (c) He-3 trapped within the titanium material, the apparatus comprising:a glove box system comprising at least one leak-tight chamber filled with inert gas, a port connected to an inert gas tank, an exhaust port connected to a ventilation system and a He-3 recovery system configured to operate within the chamber filled with inert gas, the He-3 recovery system comprising:a heater for heating the container tritium to release He-3 and tritium trapped in titanium within the container;a vacuum system configured to recover He-3 and tritium released from the container; anda getter bed purification system for purifying, separating, and recovering He-3 and tritium released from the container.

2. The apparatus as claimed in claim 1 , wherein the container further contains free He-3 gas.

3. The apparatus as claimed in claims 1-2, wherein in the glove box system further comprises an antechamber configured to be filled with inert gas for transferring the container into the leak-tight chamber filled with inert gas.

4. The apparatus as claimed in claims 1-3, wherein the heater is configured to heat the container to a temperature providing maximal release of He-3 from titanium inside the container and to a temperature not damaging to the glove box system.

5. The apparatus as claimed in claim 4, wherein the temperature is 600-750°C.

6. The apparatus as claimed in claims 1-5, wherein the vacuum system comprises a vacuum pump; a recovery vessel for capturing the He-3 and tritium gas released from the container; lines connected to the container to the vacuum pump and connected to the vacuum pump to the recovery vessel; and valves controlling a flow of gas through the vacuum system.

7. The apparatus as claimed in claims 1 to 6, wherein the getter bed purification system comprises one or more of:at least one nickel getter bed for removing oxygen and nitrogen impurities; at least one molecular sieve for removing moisture;at least one uranium storage getter bed for capturing and storing tritium; andat least one zirconium-iron getter bed for removing any remaining tritium not captured by the uranium getter beds;and wherein the getter bed system comprises at least one storage container for recovering He-3 gas separated from tritium;lines connected to the recovery vessel of the vacuum system, the getter beds and the storage container;at least one pump for circulating gas through the getter bed purification system; andvalves controlling the flow of gas through the getter bed purification system.

8. The apparatus as claimed in claims 1 -7, wherein the He-3 recovery system further comprises a gas analyzer configured to monitor one or more of the level of tritium, He-3 and impurities in gas circulating within the He-3 recovery system.

9. The apparatus as claimed in claim 8, wherein the gas analyzer is a mass spectrometer or a gas chromatograph.

10. The apparatus as claimed in claims 1 -9, wherein the getter bed purification system further comprises at least one secondary clean-up uranium getter bed for removingexcess tritium when the tritium levels flowing through the glove box exhaust port are greater than 100 pCi / m311. The apparatus as claimed in claims 1 to 10, wherein the glove box system further comprises a second chamber filled with air and the He-3 recovery system further comprises a secondary cleanup system adapted to operate in the second chamber for maintaining the leak-tight chamber filled with inert gas free of gas impurities.

12. The apparatus as claimed in claims 10-11, wherein the secondary cleanup system comprisesone or more of at least one tritium monitor, at least one oxygen monitor and at least one moisture monitor connected to the leak-tight chamber filled with inert gas for monitoring tritium, oxygen and moisture content within said chamber;a secondary purification system comprising one or more of at least one uranium getter bed for capturing tritium; at least one nickel getter bed for removing oxygen; at least one molecular sieve for removing moisture; and at least one zirconium-iron getter bed for removing any trace tritium not captured by the one or more uranium getter beds;a vacuum system configured to receive gas from the He-3 recovery system and the antechamber and configured to circulate the gas through the secondary purification system;lines connected to the antechamber, the leak-tight chamber filled with inert gas and the He-3 recovery system to the secondary clean-up system;at least one pump for circulating gas through the secondary clean-up system getter beds and lines;valves for controlling the flow of gas through the secondary clean-up system; andan exhaust port connecting the secondary clean-up system to the ventilation system.

13. The apparatus as claimed in one of claims 1 -12, further comprising at least one line connecting the getter bed purification system to a second container adapted to immobilize tritium thus capturing tritium released from the uranium storage getter bed when it is heated; andvalves for controlling the flow of the tritium into the second container adapted to immobilize tritium.

14. The apparatus as claimed in claims 1-13, wherein the He-3 recovery system has a capacity to capture over 600,000 Ci of discharged tritium.

15. The apparatus as claimed in claims 1-14, wherein the apparatus is configured to operate automatically.

16. The apparatus as claimed in claims 1-15, wherein the inert gas is argon or helium gas.

17. The apparatus as claimed in claims 3-16, wherein the apparatus further comprises an electric hoist attached to the top of the glove box system for placing the container into the heater.

18. A method of recovering helium-3 (He-3) trapped in trapped in a container adapted to immobilize tritium containing (a) a titanium material, (b) tritium trapped within the titanium material and (c) He-3 trapped within the titanium material, the method comprising:moving the container into a glove box chamber filled with inert gas; placing a container adapted to immobilize tritium into a heater; activating the heater;heating the container;evacuating the He-3 and tritium released by the heating into a recovery vessel;circulating the He-3 and tritium through one or more uranium storage getter beds to remove impurities from He-3, the one or more getter beds including at least one uranium getter bed capable of removing tritium from He-3 gas and storing tritium; andtransferring the He-3 gas to a storage vessel.

19. The method of claim 18 wherein the step for circulating the helum-3 (He-3) and tritium through one or more getter beds to remove impurities from He-3 further comprises,circulating the He-3 and tritium through one or more nickel getter beds to remove oxygen impurities;circulating the He-3 and tritium through one or more molecular sieve getter beds to remove moisture andand circulating the He-3 through at least one zirconium-iron getter bed for removing any trace tritium not captured by the one or more uranium storage getter beds.

20. The method of claims 18-19 wherein the container further contains free (He-3) gas and following the step of placing the container into a heater the method further comprises,vacuuming any He-3 gas from the container into a recovery vessel; circulating the He-3 gas from the recovery vessel through one or more getter beds to remove any impurities from the helium-3 gas;capturing tritium by circulating the gas through at least one uranium storage getter bed; andtransferring the He-3 gas to an outside storage vessel.

21. The method as claimed in claims 18-20, further comprising,heating the one or more getter beds adapted to remove tritium from He-3 to a temperature sufficient to release tritium from the one or more uranium storage getter beds;recovering tritium gas released from the heated at one or more uranium storage getter beds in a new, tritium-free container adopted to immobilize tritium.

22. The method as claimed in claims 18-20, wherein the one or more uranium storage getter beds are heated to 350-450°C to release tritium.

23. The method as claimed in claims 18-22, further comprising performing an isotope exchange with deuterium or protium gas to remove any residual tritium remaining in the one or more getter beds and disposing of the gas used in the isotope exchange by treating the gas to remove tritium or capturing and storing the gas used in the isotope exchange in containers adapted to prevent tritium release.

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