Kg Of Tritium Research Articles

Neutronics performance analysis on neutron consumption in a Fusion-Fission Hybrid System for tritium breeding

Comparison evaluation of system performance for Fusion-Fission Hybrid System as a tritium supply facility to perform D-T fusion researches was conducted. Performances for neutron consumption as well as k-eff variation, energy multiplication, tritium breeding and waste transmutation were investigated. All neutronics calculations were conducted by MCNPX2.6.0 and MCNP6.1 with ENDF/b-VII.0 neutron cross-section library.Analysis results for tritium breeding modules show that tritium with module using water or FLiNaBe coolant is effectively produced due to softening effect in spite of low 6Li enrichment. When designing hybrid system for tritium breeding, design of the fuel zone between tritium breeding zones is beneficial for tritium breeding. In this case, 7 kg of tritium is more produced without degradation of transmutation compared to the design of fuel zone in front of the tritium breeding zone. In the comparison evaluation of performance on coolants options; PbBi, LiPb, Na and FLiNaBe, tritium with Li containing coolant is produced in the fuel zone, however amount of produced tritium in the tritium breeding zone is reduced due to low neutron economy. Performance of tritium breeding with softening flux is degraded, since neutron economy is reduced by increasing the capture reaction.

Core configuration of a gas-cooled reactor as a tritium production device for fusion reactor

The performance of a high-temperature gas-cooled reactor as a tritium production device is examined, assuming the compound LiAlO2 as the tritium-producing material. A gas turbine high-temperature reactor of 300MWe nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations are carried out. To load sufficient Li into the core, LiAlO2 is loaded into the removable reflectors that surround the ring-shaped fuel blocks in addition to the burnable poison insertion holes. It is shown that module high-temperature gas-cooled reactors with a total thermal output power of 3GW can produce almost 8kg of tritium in a year.

TriToP – A compatibility experiment with turbomolecular pumps under tritium atmosphere

A turbo-molecular pump with magnetic bearing type Oerlikon Leybold MAG W 2800 has been tested under tritium atmosphere for one year with an overall throughput of about 1.3 kg of tritium (≈5 × 1018 Bq) in a closed loop. Evolution of gas composition has been regularly checked by in-line mass spectrometry. Experimental data of two long term tritium runs followed by a non-tritium run are shown. Results are discussed with regard to materials used inside the pump volume. Possible consequences for the use at the KATRIN experiment are given.

Effect of outside tritium source on tritium balance of a D-T fusion reactor

Attainable tritium breeding ration in the blanket system must be larger than the required breeding ratio when no effective tritium resources from outside are expected. It is revealed recently that a considerable amount of tritium can be trapped to the re-deposition layer of the first wall materials and that the time constant of this phenomenon is rather long. Then, the tritium breeding ratio around 1.1 is required in the blanket system when 3years is claimed for the tritium doubling time to prepare tritium for the initial inventory of a next reactor. Construction of an outside tritium supply is one of the possible ways to compensate the lack of tritium because it is generally considered that the attainable tritium breeding ratio in the solid breeder system is around 1.05. It is reported recently that a high-temperature gas-cooled reactor can produce 10kg of tritium per year. The preferable amount of tritium production rate of the outer tritium supply is discussed in this study from the viewpoint of tritium balance in a D-T power reactor.

Study of Tritium Production for Fusion Reactors Using High-Temperature Gas-Cooled Reactors

The performance of a high-temperature gas-cooled reactor as a tritium production device was examined. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) was assumed as the calculation target of a typical gas-cooled reactor, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the entire-core region of GTHTR300 were carried out considering its unique double heterogeneity structure. It was shown that gas-cooled reactors with thermal output power of 3 GW in all can produce 6~10 kg of tritium in a year.

Performance of high-temperature gas-cooled reactor as a tritium production device for fusion reactors

The performance of a high-temperature gas-cooled reactor as a tritium production device is examined. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) is assumed as the calculation target of a typical gas-cooled reactor, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the 3-dimensional entire-core region of GTHTR300 were carried out considering its unique double heterogeneity structure. It is shown that gas-cooled reactors with thermal output power of 3 GW in all can produce 5–8 kg of tritium in a year.

Safe handling experience of a tritium storage bed

In an ITER facility, about 3 kg of tritium will be stored in about 50 hydride beds. The safe design and operation of tritium storage beds will be one of the most important points for enhancing overall safety in the facility. In the Tritium Process Laboratory at the Japan Atomic Energy Agency, many tritium storage beds with zirconium–cobalt (ZrCo) and uranium (U) have been used with/without self-accountancy measures, and 20 years of safe handling experience has been accumulated. From this experience, the key issues to be considered for the safety design are the effects of tritium decay, such as decay heat transfer and 3He behavior, which can cause an increase in temperature and pressure in the tritide vessel, along with the normal protection against over-temperature, over-pressure and leaks in the metal-hydride bed. Concerning the safety operation, the key issues are the hydrogenation–dehydrogenation cycle sequence under the requirements of the storage system and the emergency action, such as rapid hydrogen recovery and loss of normal cooling function.

Inertial Fusion Energy Power Reactor Fuel Recovery System

A conceptual design is proposed to support the recovery of un-expended fuel, ash, and associated postdetonation products resident in plasma exhaust from a ~ 2 GW IFE direct drive power reactor. The design includes systems for the safe and efficient collection, processing, and purification of plasma exhaust fuel components. The system has been conceptually designed and sized such that tritium bred within blankets, lining the reactor target chamber, can also be collected, processed, and introduced into the fuel cycle. The system will nominally be sized to process ~ 2 kg of tritium per day and is designed to link directly to the target chamber vacuum pumping system. An effort to model the fuel recovery system (FRS) using the Aspen Plus engineering code has commenced. The system design supports processing effluent gases from the reactor directly from the exhaust of the vacuum pumping system or in batch mode, via a buffer vessel in the Receiving and Analysis System. Emphasis is on nuclear safety, reliability, and redundancy as to maximize availability. The primary goal of the fuel recovery system design is to economically recycle components of direct drive IFE fuel. The FRS design is presented as a facility sub-system in the context of supporting the larger goal of producing safe and economical IFE power.

04/01515 Preparation and crystallization of forsterite fibrous gels: Tsai, M. T. Journal of the European Ceramic Society, 2003, 23, (8), 1283–1291

The performance of a high-temperature gas-cooled reactor as a tritium production device is examined. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) is assumed as the calculation target of a typical gas-cooled reactor, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the 3-dimensional entire-core region of GTHTR300 were carried out considering its unique double heterogeneity structure. It is shown that gas-cooled reactors with thermal output power of 3 GW in all can produce 5–8 kg of tritium in a year.

Oxidative Decontamination of Tritiated Materials Employing Ozone Gas

ABSTRACTThe Princeton Plasma Physics Laboratory has developed a process by which to significantly reduce surface and near surface tritium contamination from various materials. The Oxidative Tritium Decontamination System (OTDS) reacts gaseous state ozone (accelerated by presence of catalyst), with tritium entrained/deposited on the surface of components (stainless steel, copper, plastics, ceramics, etc.) for the purpose of activity reduction by means of oxidation-reduction chemistry.1 In addition to removing surface and near surface tritium contamination from (high monetary value) components for re-use in non-tritium environments, the OTDS has the capability of removing tritium from the surfaces of expendable items, which can then be disposed of in a less expensive fashion. The OTDS can be operated in a batch mode by which up to approximately 20kg of tritium contaminated (expendable) items can be processed and decontaminated to levels permissible for free release (< 16.66Bq/100cm2). This paper will discuss the OTDS process, the level of tritium surface contamination removed from various materials, and a technique for “deep scrubbing” tritium from sub-surface layers.

Tritium inventory of the Black Sea

Tritium and helium-3 measurements from the southern half of the Black Sea are described. A tritium maximum centered at approximately 1200 m and accompanied by a radiogenic helium-3 maximum suggests deep injection of near surface water in the eastern part of the basin. The results are combined with three other tritium data sets to discern the temporal variation in the past three decades. The observed temporal variation is slower than would be expected from radioactive decay alone in the upper few hundred meters where the major proportion of tritium resides. This trend suggests that there must have been a tritium source(s) other than that produced by the atmospheric nuclear tests. It is estimated that approximately 0.1 kg of tritium of non-atmospheric origin had entered the Black Sea since 1969.

Tritium activities in Canada

Canadian tritium activites comprise three major interests: utilites, light manufacturers, and fusion. There are 21 operating CANDU reactors in Canada; 19 with Ontario Hydro and one each with Hydro Quebec and New Brunswick Power. There are two light manufacturers, two primary tritium research facilities (at AECL Chalk River and Ontario Hydro Technologies), and a number of industry and universities involved in design, construction, and general support of the other tritium activities. The largest tritum program is in support of the CANDU reactors, which generate tritium in the heavy water as a by-product of normal operation. Currently, there are about 12 kg of tritium locked up in the heavy water coolant and moderator of these reactors. The fusion work is complementary to the light manufacturing, and is concerned with tritium handling for the ITER program. This included design, development and application of technologies related to Isotope Separation, tritium handling, (tritiated) gas separation, tritium-materials interaction, and plasma fueling.

Ventilation of North Pacific intermediate waters: The role of the Alaskan Gyre

Hydrographic data, tritium data, and potential vorticity calculations suggest that although North Pacific Intermediate Water is formed in the northwest, the Alaskan Gyre might be an additional ventilation site. The proposed ventilation is quantified by a vertical column tritium inventory, which indicates an excess of 0.08 kg of tritium in the Alaskan Gyre. An evaluation of the energy stored in the water column and of wind and buoyancy forcing shows that during winter conditions enough energy can be pumped into the system to force 26.80 σθ to outcrop in the Alaskan Gyre. Model results suggest that relatively limited outcrops in time and space (tens of days and several hundred kilometers in diameter) can account for the excess tritium.

Feasibility study of a magnetic fusion production reactor

A magnetic fusion reactor can produce 10.8 kg of tritium at a fusion power of only 400 MW —an order of magnitude lower power than that of a fission production reactor. Alternatively, the same fusion reactor can produce 995 kg of plutonium. Either a tokamak or a tandem mirror production plant can be used for this purpose; the cost is estimated at about $1.4 billion (1982 dollars) in either case. (The direct costs are estimated at $1.1 billion.) The production cost is calculated to be $22,000/g for tritium and $260/g for plutonium of quite high purity (1%240Pu). Because of the lack of demonstrated technology, such a plant could not be constructed today without significant risk. However, good progress is being made in fusion technology and, although success in magnetic fusion science and engineering is hard to predict with assurance, it seems possible that the physics basis and much of the needed technology could be demonstrated in facilities now under construction. Most of the remaining technology could be demonstrated in the early 1990s in a fusion test reactor of a few tens of megawatts. If the Magnetic Fusion Energy Program constructs a fusion test reactor of approximately 400 MW of fusion power as a next step in fusion power development, such a facility could be used later as a production reactor in a spinoff application. A construction decision in the late 1980s could result in an operating production reactor in the late 1990s. A magnetic fusion production reactor (MFPR) has four potential advantages over a fission production reactor: (1) no fissile material input is needed; (2) no fissioning exists in the tritium mode and very low fissioning exists in the plutonium mode thus avoiding the meltdown hazard; (3) the cost will probably be lower because of the smaller thermal power required; (4) and no reprocessing plant is needed in the tritium mode. The MFPR also has two disadvantages: (1) it will be more costly to operate because it consumes rather than sells electricity, and (2) there is a risk of not meeting the design goals.

Tritium inventories of the world oceans and their implications

Tritium inventories are given for the Pacific Ocean from 51°N to 60°S. As expected, the highest inventories are found north of 20°N. Inventories drop rapidly from 20°N to 10°N. From 10°N southwards, inventories in the top 500 m are essentially constant. From these data, it is calculated that 107±21 kg of tritium were present in the Pacific Ocean in 1970. A total of ∼ 300±80 kg of tritium was present on the Earth's surface in 1970. This indicates that 550±160 kg of tritium have been produced by nuclear detonations. It is estimated that 62% of the tritium in the Pacific Ocean was added through molecular exchange, 5% through runoff from the continents, and 33% as precipitation.