• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.250-255
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• 1997

중수로형(PHWR) 원자력발전소는 감속재와 냉각재로 중수를 사용하고 있어 방사성 수소동위원소인 삼중수소 생성량이 경수로에 비해 크며 계통내 삼중수소 축적량은 운전년수에 따라 증가하게 된다. 중수로형 원전에서 삼중수소 저감화를 위한 장기 대책으로 Tritium Removal Facility를 적용하는 경우, 우선적으로 괴려하여야 할 사항은 적절한 TRF의 용량을 결정하는 것이다. 이는 초기 시설 투자비뿐만 아니라 설비 및 운전의 신뢰도와 이용율에도 영향을 미치므로 연속운전이 가능하도록 용량을 결정하는 것이 중요하다. 이를 위해 감속재를 대상으로 삼중수소 농도 목표치, 삼중수소 농도 목표치 도달기간, 탈 삼중수소율, TRF 적용시점이 TRF 처리량과 촉매탑 높이에 미치는 영향을 분석하였다. 삼중수소 농도 목표치는 5~15Ci/kg, 도달기간은 3~8년, 탈 삼중수소율은 0.05~0.4, TRF 적용시점은 가동 후 10~20년이 적절한 것으로 확인되었다.

• Proceedings of the Korean Vacuum Society Conference
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• 2010.02a
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• pp.314-314
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• 2010

The design of the blanket and shield play a key role in determining the size of a reactor since it has an impact on the various reactor components. The blanket should produce enough tritium for tritium self-sufficiency and the shield should provide sufficient protection for the superconducting TF coil. Neutronic optimization of the blanket and the shield is necessary, and we coupled the system analysis with a neutronic calculation to account for the interrelation of the blanket and shield with the plasma performance of a reactor system in a self-consistent manner. By using the coupled system analysis code, the operational space for a low aspect ratio (LAR) tokamak reactor with a superconducting toroidal field (TF) coil is investigated with an spect ratio in the range of 1.5 - 2.5. The minimum major radius which satisfies all the physics and engineering requirements increases with the magnetic field at the magnetic axis. A required inboard shield thickness is mainly determined by the requirement on the protection of the TF coil against radiation damage. It is shown that to have a fusion power bigger than 3,000 MW in the LAR tokamak with a superconducting TF coil, a major radius bigger than 4.0 m is required.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2019.05a
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• pp.307-308
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• 2019

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.315-316
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• 2018

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.10a
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• pp.335-336
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• 2017

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4274-4281
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• 2023

Blanket is of vital importance for engineering application of the fusion reactor. Nuclear heat deposition in materials is the main heat source in blanket structure. In this paper, the three-dimensional method for thermal-hydraulics/neutronics coupling analysis is developed and applied for the full-scale module of the helium-cooled ceramic breeder tritium breeding blanket (HCCB TBB) designed for China Fusion Engineering Test Reactor (CFETR). The explicit coupling scheme is used to support data transfer for coupling analysis based on cell-to-cell mapping method. The coupling algorithm is realized by the user-defined function compiled in Fluent. The three-dimensional model is established, and then the coupling analysis is performed using the paralleled Coupling Analysis of Thermal-hydraulics and Neutronics Interface Code (CATNIC). The results reveal the relatively small influence of the coupling analysis compared to the traditional method using the radial fitting function of internal heat source. However, the coupling analysis method is quite important considering the nonuniform distribution of the neutron wall loading (NWL) along the poloidal direction. Finally, the structure optimization of the blanket is carried out using the coupling method to satisfy the thermal requirement of all materials. The nonlinear effect between thermal-hydraulics and neutronics is found during the blanket structure optimization, and the tritium production performance is slightly reduced after optimization. Such an adverse effect should be thoroughly evaluated in the future work.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.2
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• pp.155-169
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• 2007

A transport container for a 500 kCi tritium storage vessel was developed, which could be used for the transport of metal tritide from Wolsong TRF facility to a disposal site. The structural, thermal, shielding, and confinement analyses were performed for the container in a view of Type B. As a result of structural analysis, the developed container sustained its integrity under normal and accidental conditions. The maximum temperature increase of the inner storage vessel by radiation was evaluated at $134.8^{\circ}C at room temperature. In $800^{\circ}C$ fire test, The thermal barrier of container sustained the inner vessel at $405^{\circ}C after 30 min, which temperature was allowable for the container integrity since maximum design temperature of inner vessel was $550^{\circ}C. In the evaluation of the shielding, the activity of radiation was nearly zero on the outer surface of inner vessel. Consequently the transport container for a 500 kCi tritium was evaluated to pass all the safety tests including accidental condition, so it was concluded that the designed transport container is proper to be used.

• Journal of Radiation Protection and Research
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• v.45 no.3
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• pp.142-146
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• 2020

Background: Tritium (³H) analysis in groundwater was difficult because of its low activity. Therefore, the electrolytic enrichment method was used. To improve the detection limit and for performing simple analysis, a high-volume counting vial with the available liquid scintillation counter (LSC) was investigated. Further, it was compared with a conventional 20-mL counting vial. Materials and Methods: The LSC with the electrolytic enrichment method was used ³H analysis in groundwater. A high-volume 145-mL counting vial was compared with a conventional 20-mL counting vial to determine the counting characteristics of different LSCs. Results and Discussion: When a Quantulus LSC was used, the counting window between channels 35 and 250 was used. The background count was approximately 1.86 cpm, and the counting efficiency increased from 8% to 40% depending on the mixing ratio of the volume of sample and cocktail solution. For LSC-LB7, the optimum counting window was between 1 and 4.9 keV, which was selected by the factory (Hitachi Aloka Medical Ltd., Japan) by considering quenching using a standard external gamma source. The background count of LSC-LB7 was approximately 3.60 ± 0.29 cpm when the 145-mL vial was used and 2.22 ± 0.17 cpm when the 20-mL vial was used. The minimum detectable activity (MDA) of the 20-mL vial was greater for LSC-LB7 than for Quantulus. The MDA with the 145-mL vial was improved to 0.3 Bq/L when compared with the value of 1.6 Bq/L for the 20-mL vial. Conclusion: The counting efficiency when using the 145-mL vial was 27%, whereas it was 18% when using the 20-mL vial. This difference can be attributed to the vial volume. The figure of merit (FOM) of the 145-mL vial was four times greater than that of the 20-mL vial because the volume of the former vial is approximately seven times greater than that of the latter. Further, the MDA for ³H decreased from 1.6 to 0.3 Bq/L. The counting efficiency and FOM of LSC-LB7 was slightly less than those of Quantulus when the 20-mL vial was used. The background counting rate of the Quantulus was lower than that of the LSC-LB7.

• 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.

• Journal of the Korean Society of Environmental Restoration Technology
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• v.16 no.3
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• pp.55-62
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• 2013

Generally, the radioactivity from NPP(Nuclear Power Plants) operation can be released below 3% of DRLs(Derived Release Limits) to environment. It was tried to understand which plant was efficient for absorbing radioactivity in this study. Pinus thunbergii Parl. and Viburnum awabuki K. KOCH were analyzed for radioisotope absorption. The samples were collected at three different locations depending on the distance from NPP at the vicinity 10km away, and 30km away. Gamma radionuclide was not detected from the samples, which means that the direct transition into the plant was not significant. Meanwhile, the very low level of radioactive tritium was detected in the samples. One remark was that every plant has different ability for tritium absorption. These results are expected to be applied to propagation and transplanting in radioactively contaminated area or reducing radioactivity in the soil and water near the plants.