• Nuclear Engineering and Technology
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• 제35권6호
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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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• 제22권2호
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• pp.119-125
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• 1997

가압중수로는 감속재와 냉각재로 중수를 채택함으로써 높은 중성자 경제성을 달성하는 대신 중수소의 중성자 포획반응 때문에, 경수로에 비해, 다량의 삼중수소가 발생한다. 한편 원자로심에서, 삼중수소의 ${\beta}$-붕괴결과 발생된 $^3He$는, 열중성자를 포획하여 다시 삼중수소로 변환된다. 중수로에서 삼중수소의 생성에 대한 기존의 계산모형은, $^3He$가 상대적으로 낮은 용해도를 가지므로, 그 기여도를 무시해왔다. 그러나 $^3He$의 중성자 포획단면적은 중수소의 그것에 비해 $1.6{\times}10^7$ 배가 된다. 즉 $^3He$가 중수내에 0.03 ppm만 녹아있다 하더라도 $^3He$에 의해 생성되는 삼중수소의 양은 전체 중수에 의한 삼중수소의 양에 필적하게 된다. 본 연구에서는 월성1호기를 대상으로, 중수로에서 $^3He$가 삼중수소의 생성에 미치는 영향을 평가하였으며 결과를 실측치와 비교하였다. 연구의 결과, 감속재에서는 $^3He$의 용해도가 낮고 $^4He$ Cover gas 때문에 $^3He$의 기여도는 무시할 수 있음이 밝혀졌다. 반면 냉각재의 경우 $^3He$ 삼중수소의 생성에 지대한 영향을 미치는 것으로 나타났다. 또한 본 연구의 계산방법은 원전 운전초기의 냉각재내 삼중수소 생성량은 과대평가 하는 것으로 나타났으나 운전기간이 증가함에 따라 실측치와 잘 일치하는 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제52권2호
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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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.211-212
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• 2005

• Nuclear Engineering and Technology
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• 제41권5호
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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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 추계학술발표회요약집
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• pp.283.1-283
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• 1999

• 한국세라믹학회지
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• 제54권1호
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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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• 제24권2호
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• pp.141-153
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• 1992

상온핵융합의 실험적 검증을 위하여 가공조건 및 기하학적조건이 다른 7종류의 팔라디움전극을 사용하여 24~28시간, 전류밀도 83~600 mA/$\textrm{cm}^2$의 조건하에 전기분해를 실시하였다. 상기조건 하에서 삼중수소의 농축에 기인한 분리팩타(separation factor)를 측정하였고 핵융합의 부산물일수도 있는 삼중수소 증가량을 측정하였다. 또한 초과열 계 산과 관련된 K(net Faradic efficiency)를 측정하여 산소/중수소 가스의 재결합정도를 조사하였다. 양전자소멸측정장치 및 일정체적 가스주입장치를 이용하여 팔라디움전극에서 격자결함과 수소의 반응 및 거동에 대하여 조사하였다. 전기분해하는 동안 삼중수소 농축현상이 관찰되었으나 핵융합의 증거가 될만한 삼중수소양은 검출되지 않았다. 한편 산소/수소 가스의 재결합 정도는 32%로 나타났다. 이는 재결합과정이 발열반응이므로 전기분해과정에서 핵반응과 관계없이 초과엔탈피가 발생할 수 있음을 의미한다. 양전자소멸측정장치를 이용하여 양전자수명, 양전자소멸밀도, P/W 및 R 파라메터의 측정을 통하여 전극의 격자결함(전위 및 공공)에 수소가 집적 (trap)되며 수소집적은 공공에서 보다 전위에 약간 더 선호하는 것으로 나타났다. 전극의 수소화물형성에 수반하여 대부분 전위가 발생한 것으로 나타났다. 또한 팔라디움수소화물의 등시소둔실험을 통하여 소량의 미소공동 형태의 결함이 존재하는 것으로 추정하였고 그 결함의 크기는 수 $\AA$정도인 것으로 생각된다.