• Nuclear Engineering and Technology
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• v.35 no.6
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• pp.515-522
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• 2003

Tritium behavior in the solid breeder blanket is one of the key factors in determining tritium self-sufficiency, as well as safety, of fusion reactors. Recently, a model has been developed to describe the tritium behavior in solid breeder material, which can predict the tritium release and inventory in the blanket. However, the model has limitation to account for tritium solubility effects, mainly existing as LiOT, especially inside the $Li_{2}O$ solid breeder. In order to improve the capability of predicting the LiOT precipitation in $Li_{2}O$ solid breeder, a new logic is developed and integrated in the existing tritium release and inventory calculation code. With the logic developed in this work, the code can have capabilities to analyze tritium release and inventories in $Li_{2}O$ under steady and transient conditions. It can be found that tritium inventory as LiOT is an important mechanism under pulsed operation, and the amount of inventory becomes higher as the tritium generation rate increases and the temperature decreases. Also, the temperature limits for the generation of LiOT precipitation are determined. Therefore the developed logic helps understand the tritium transport mechanism in $Li_{2}O$ solid breeder.

• Journal of Radiation Protection and Research
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• v.22 no.2
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• pp.119-125
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• 1997

PHWR achieves high neutron economy by adopting heavy water as its moderator and coolant. On the other hand it permits much tritium generation, compared to LWR, due to the neutron capture reaction of deuterium in heavy water. Meanwhile in the reactor core, $^3He formed as the result of-decay of tritium, captures a thermal neutron and transforms to tritium again. The existing calculation models on tritium generation in PHWR neglect the contribution of $^3He$ in both moderator and coolant due to its relatively low solubility. However the neutron capture cross-section of $^3He$ is almost $1.6{\times}10^7$ times as large as that of deuterium. That means that the dissolved amount of 0.03 ppm of $^3He$ in heavy water is enough to generate the same amount of tritium as that generated by the deuterium of total heavy water in the system. This study dealt with the contribution of $^3He$ to tritium generation. As a sample case, the contribution of $^3He$ to the tritium generation in Wolsong #1 was evaluated and compared to the measured values. According to the result of this study, it is concluded that $^3He$ in coolant contributes very much to the tritium generation but that in moderator shows negligible effects due to the low solubility and $^4He$ cover gas. At the beginning of the plant operation, the contribution of $^3He$ is slightly greater than the measured value but agrees well with the measured as the operating time increases.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.323-327
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• 2020

In this study, the neutronic calculation to obtain tritium breeding ratio (TBR) in a deuterium-tritium (D-T) fusion power reactor using Monte Carlo MCNPX is done. In addition, by using COMSOL software, an efficient cooling system is designed. In the proposed design, it is adequate to enrich up to 40% ⁶Li. Total tritium breeding ratio of 1.12 is achieved. The temperature of helium as coolant gas never exceed 687℃. As regards the tolerable temperature of beryllium (650℃), the design of blanket module is done in the way that beryllium temperature never exceed 600℃. The main feature of this design indicates the temperature of helium coolant is higher than other proposed models for blanket module, therefore power of electricity generation will increase.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.211-212
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• 2005

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.729-738
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• 2009

Given the evolution of High-Temperature Gas-cooled Reactor(HTGR) designs, the source terms for licensing must be developed. There are three potential source terms: fission products, actinides in the fuel and tritium in the coolant. It is necessary to provide first an inventory of the source terms under normal operations. An analysis of source terms has yet to be performed for HTGRs. The previous code, which can estimate the inventory of the source terms for LWRs, cannot be used for HTGRs because the general data of a typical neutron cross-section and flux has not been developed. Thus, this paper uses a combination of the MCNP, ORIGEN, and MONTETEBURNS codes for an estimation of the source terms. A method in which the HTR-10 core is constructed using the unit lattice of a body-centered cubic is developed for core modeling. Based on this modeling method by MCNP, the generation of fission products, actinides and tritium with an increase in the burnup ratio is simulated. The model developed by MCNP appears feasible through a comparison with models developed in previous studies. Continuous fuel management is divided into five periods for the feeding and discharging of fuel pebbles. This discrete fuel management scheme is employed using the MONTEBURNS code. Finally, the work is investigated for 22 isotope fission products of nuclides, 22 actinides in the core, and tritium in the coolant. The activities are mainly distributed within the range of $10^{15}{\sim}10^{17}$ Bq in the equilibrium core of HTR-10. The results appear to be highly probable, and they would be informative when the spent fuel of HTGRs is taken into account. The tritium inventory in the primary coolant is also taken into account without a helium purification system. This article can lay a foundation for future work on analyses of source terms as a platform for safety assessment in HTGRs.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.283.1-283
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• 1999

• Journal of the Korean Ceramic Society
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• v.54 no.1
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• pp.9-22
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• 2017

Many new functional materials have been studied for efficient production and storage of energy. Many new materials such as sodium-based and sulfide-based materials have been proposed for energy storage, but research on Li batteries is still dominant. Due to the influence of environmental concerns regarding nuclear energy, interest in and research on fusion power are steadily increasing. For the commercialization of nuclear fusion, a design standard based on a considerable level of physical analysis and modeling is proposed. Nevertheless, limitations of existing materials in nuclear fusion environments limit practical applications. Tritium propagation material for continuous fusion reaction is one of the core materials, and therefore research on this material is being carried out intermittently. The key material for Li-based energy storage and tritium generation is the functional material Li-M-O. In this review, a structural description of functional Li-M-O system materials and technical trends for its applications are introduced.

• Nuclear Engineering and Technology
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• v.24 no.2
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• pp.141-153
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• 1992

Excess tritium analysis was peformed to verify whether or not cold fusion occurs during electrolysis of heavy water in the current density range of 83~600 mA/$\textrm{cm}^2$ for a period of 24 ~ 48 hours with use of palladium electrodes of seven different processing treatments and geometries. The extent of recombination of D$_2$ and $O_2$gases in the electrolytic cell was measured for the calculation of accurate enthaplpy values. The behavior and interaction of hydrogen atoms with defects in Pd electrodes were examined using the Sieverts gas charging and the positron annihilation(PA) method. Slight enrichment of tritium observed was attributed to electrolytic enrichment but not to the formation of a by-product of cold fusion. The extent of recombination of D$_2$and $O_2$gases was 32%. Hence the excess heat measured during the electrolysis was considered to be due to the exothermic reaction of recombination but not to nuclear fusion. Lifetime results from the PA measurements on the Pd electrodes indicated that hydrogen atoms could be trapped at dislocations and vacancies in the electrodes and that dislocations were slightly more preferred sites than vacancies. It was also inferred from R parameters that the formation of hydrides was accompanied by generation of mostly dislocations. Doppler broadening results of the Pd electrodes indicated that lattiec defect sites where positrons were trapped first increased and then decreased, and this cycle was repeated as electrolysis continued. It can be inferred from PA measurements on the cold-rolled Pd and the isochronally annealed Pd hydride specimens that microvoid-type defects existed in the hydrogen-charged electrode specimen.