• 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.

• 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.

• Transactions of the Korean hydrogen and new energy society
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• v.23 no.1
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• pp.49-55
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• 2012

Hydrogen isotopes accountancy and storage are important functions in a nuclear fusion fuel cycle. The hydrogen isotopes are safely stored in metal hydride beds. The tritium inventory of the bed is determined from the decay heat of tritium. The decay heat is measured by circulating helium through the metal hydride bed and measuring the resultant temperature increase of the helium flow. We are reporting our preliminary experimental results on the hydrogen isotopes accountancy and storage performance in a metal hydride bed.

• Journal of Korean Society for Quality Management
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• v.19 no.2
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• pp.96-107
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• 1991

At the design stage of a plant, the plausible causes and pathways of release of hazardous materials are not clearly known. Thus there exist large amount of uncertainties on the consequences resulting from the operation of a fusion plant. In order to better handle such uncertain circumstances, we utilize the Probabilistic Risk Assessment(PRA) for the safety analyses on fusion power plant. In this paper, we concentrate on the tritium release accident. We develop a simple model that describes the process and flow of tritium, by which we figure out the locations of tritium inventory and their vulnerability. We construct event tree models that lead to various levels of tritium release from abnormal initiating events. Branch parameters on the event tree are assessed from the fault tree analysis. Based on the event tree models we construct influence diagram models which are more useful for the parameter updating and analysis. We briefly discuss the parameter updating scheme, and finally develop the methodology to obtain the predictive distribution of consequences resulting from the operating a fusion power plant. We also discuss the way to utilize the results of testing on sub-systems to reduce the uncertain ties on over all system.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.117-123
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• 2003

In succession to the previous paper, the tritium injection system and the regeneration system of the tritium handling system are presented. Both systems should be placed inside glove boxes since there can be potential leakage of tritium from these systems. The tritium injection system should be capable of measuring the exact amount of the injected tritium to keep track of the tritium inventory. The tritium injection system is designed to recover the remaining tritium from the system after injection for the minimization of tritum release to the environment as well as for the recovery of precious resource. TRS equipment such as MS, Ni catalyst bed, and metal getter are regenerated with a standalone regeneration system. Unlike other equipments which can be regenerated by heating and purging with appropriate gas, regeneration of the metal getter used to recover tritium is somewhat complicated.

• Transactions of the Korean hydrogen and new energy society
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• v.23 no.1
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• pp.56-64
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• 2012

The ITER fuel cycle is designed for DT operation in equimolar ratio. It involves not only a group of fuelling system and torus cryo-pumping system of the exhaust gases through the divertor from the torus in tokamak plant, but also from the exhaust gas processing of the fusion effluent gas mixture connected to the hydrogen isotope separation in cryogenic distillation to the final safe storage & delivery of the hydrogen isotopes in tritium plant. Tritium plant system supplies deuterium and tritium from external sources and treats all tritiated fluids in ITER operation. Every operation and affairs is focused on the tritium inventory accountancy and the confinement. This paper describes the major fuel cycle processes and interfaces in the tritium plant in aspects of upcoming technologies for future hydrogen and/or hydrogen isotope utilization.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.10a
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• pp.323-324
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• 2017

• Journal of Korean Society for Quality Management
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• v.26 no.1
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• pp.201-212
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• 1998

A methodology of the safety analysis on the fusion power plant is introduced. It starts with the understanding of the physics and engineering of the plant followed by the assessment of the tritium inventory and flow rate. We a, pp.y the probabilistic risk assessment. An event tree that explains the propagation of the accident is constructed and then it is translated in to an influence diagram, that is accident is constructed and then it is translated in to an influence diagram, that is statistically equivalent so far as the parameter updating is concerned. We follow the Bayesian a, pp.oach where model parameters are treated as random variables. We briefly discuss the parameter updating scheme, and finally develop the methodology to obtain the predictive distribution of time to next severe accident.