• 방사성폐기물학회지
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• 제3권1호
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• pp.67-75
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• 2005

원자력시설 핫셀 (Hot Cell)내에서 핵종실험 시 발생하는 고방사능 분진(Hot Particulate)의 크기는 0.5300 ${\mu}m$이고 주 핵종은 UO$_2$였다. 핫셀 내의 고방사능 분진을 제거하기 위해 사이클론과 Bag/HEPA필터로 구성된 장치를 고안하였고, 이 장치의 사이클론에 의해 고방사능 분진을 최대로 포집할 수 있는 실험조건을 제시했다. 모의입자의 크기가 클수록 입자의 포집효율은 높았다. 모의 입자의 크기가 5${\mu}m$ 이상일 때, 입자의 포집효율은 $80\%$보다 높았다. 모의 입자의 크기가 1.0 ${\mu}m$ 보다 작을 때, 포집효율은 $70\%$ 보다 작았다. 모의 입자의 유입속도가 12 m/sec보다 클 때, 포집효율은 $70\%$보다 높았다. 그러나 유입속도가 17 m/sec 보다 클 때 포집효율의 증가율은 크지 않았다. 모의입자의 포집효율은 Vortex Finder의 길이가 7.2 cm이하일 때, 길이의 증가와 함께 높아졌지만 7.2 cm 이상일 때는 낮아지기 시작했다. 그러므로 Vortex Finder의 길이가 7.2 cm 일 때, 최대포집효율을 나타냈다. 사이클론 밑에 보조콘 부착 시 모든 속도 범위에서 약 평균 $2\%$ 정도 포집효율이 증가하므로 보조콘 부착효과가 크지 않았다.

• Nuclear Engineering and Technology
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• 제39권5호
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• pp.603-616
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• 2007

Roy Huddle, having invented the coated particle in Harwell 1957, stated in the early 1970s that we know now everything about particles and coatings and should be going over to deal with other problems. This was on the occasion of the Dragon fuel performance information meeting London 1973: How wrong a genius be! It took until 1978 that really good particles were made in Germany, then during the Japanese HTTR production in the 1990s and finally the Chinese 2000-2001 campaign for HTR-10. Here, we present a review of history and present status. Today, good fuel is measured by different standards from the seventies: where $9*10^{-4}$ initial free heavy metal fraction was typical for early AVR carbide fuel and $3*10^{-4}$ initial free heavy metal fraction was acceptable for oxide fuel in THTR, we insist on values more than an order of magnitude below this value today. Half a percent of particle failure at the end-of-irradiation, another ancient standard, is not even acceptable today, even for the most severe accidents. While legislation and licensing has not changed, one of the reasons we insist on these improvements is the preference for passive systems rather than active controls of earlier times. After renewed HTGR interest, we are reporting about the start of new or reactivated coated particle work in several parts of the world, considering the aspects of designs/ traditional and new materials, manufacturing technologies/ quality control quality assurance, irradiation and accident performance, modeling and performance predictions, and fuel cycle aspects and spent fuel treatment. In very general terms, the coated particle should be strong, reliable, retentive, and affordable. These properties have to be quantified and will be eventually optimized for a specific application system. Results obtained so far indicate that the same particle can be used for steam cycle applications with $700-750^{\circ}C$ helium coolant gas exit, for gas turbine applications at $850-900^{\circ}C$ and for process heat/hydrogen generation applications with $950^{\circ}C$ outlet temperatures. There is a clear set of standards for modem high quality fuel in terms of low levels of heavy metal contamination, manufacture-induced particle defects during fuel body and fuel element making, irradiation/accident induced particle failures and limits on fission product release from intact particles. While gas-cooled reactor design is still open-ended with blocks for the prismatic and spherical fuel elements for the pebble-bed design, there is near worldwide agreement on high quality fuel: a $500{\mu}m$ diameter $UO_2$ kernel of 10% enrichment is surrounded by a $100{\mu}m$ thick sacrificial buffer layer to be followed by a dense inner pyrocarbon layer, a high quality silicon carbide layer of $35{\mu}m$ thickness and theoretical density and another outer pyrocarbon layer. Good performance has been demonstrated both under operational and under accident conditions, i.e. to 10% FIMA and maximum $1600^{\circ}C$ afterwards. And it is the wide-ranging demonstration experience that makes this particle superior. Recommendations are made for further work: 1. Generation of data for presently manufactured materials, e.g. SiC strength and strength distribution, PyC creep and shrinkage and many more material data sets. 2. Renewed start of irradiation and accident testing of modem coated particle fuel. 3. Analysis of existing and newly created data with a view to demonstrate satisfactory performance at burnups beyond 10% FIMA and complete fission product retention even in accidents that go beyond $1600^{\circ}C$ for a short period of time. This work should proceed at both national and international level.

• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1714-1723
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• 2020

This study aims at characterizing the rheological properties of uranium oxide powders for nuclear fuel pellets manufacturing. The flowability of these powders must be compatible with a reproducible filling of press molds. The particle size distribution is known to have an impact on the rheological properties and fine particles (<100 ㎛) are suspected to have a detrimental effect. In this study, the impact of the particle size distribution on the rheological properties of UO₂ powders was quantified, focusing on the influence of fine particles. Two complementary approaches were used. The first approach involved characterizing the powder in a static state: density, compressibility and shear test measurements were used to understand the behavior of the powder when it is transitioned from a static to a dynamic state (i.e., incipient flow conditions). The second approach involved characterizing the behavior of the powder in a dynamic state. Two zones, corresponding to two characteristic behaviors, were demonstrated for both types of measurements. The obtained results showed the amount of fines should be kept below 10 % wt to ensure a robust mold filling operation (i.e., constant mass and production rate).

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2007년도 학술논문요약집
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• pp.305-306
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• 2007

ACP(Advanced Spent Fuel Conditioning Process)의 금속전환로에 $U_3O_8$을 공급하기 위하여 20 kgHM/batch의 $UO_2$ 펠릿(pellets)을 처리할 수 있는 건식분말화 장치가 개발되고있다. 건식분말화 장치는 500 $^{\circ}C$온도에서 공기를 공급하여 일정한 입도범위의 균질한 $U_3O_8$을 만든다. 이런 건식 분말화 장치의 효율을 높이기 위해서는 반웅로에 불어 넣어주는 공기의 유량을 증가시킬 필요가 있다. 하지만 공기와 반응하여 생성되는 $U_3O_8$ 입자는 그 크기가 최소 3 ${\mu}$m 정도로 매우 미세하여，반응로 출구를 통해 외부로 빠져나갈 가능성 이있다. 이를 방지하기 위해 분말화 장치 출구 바깥에는 필터가 설치되어 있으나 공기와 함께 $U_3O_8$ 입자가 계속해서 빠져 나갈 경우 입자로 인해 필터가 막혀 제 기능을 할 수 없게 된다. 따라서 건식 분말화 장치는 미세한 $U_3O_8$ 입자가 반응로 밖으로 빠져나가지 않도록 입구에서의 공기 유량을 일정 수준 이하로 조절해주는 것이 필요하다. 이 연구의 목적은 초기 유량으로부터 유량을 점점 증가시키면서 시간변화에 따른 입자 거동 특성을 해석하며， 결과로부터 주어진 크기의 타원입자에 대해 최대 허용 공기 유량을 결정하고자한다. 이 해석을 위해 유동과 입자를 동시에 해석할 수 있는 ANSYS-CFX 5.7.1과 ANSYS-CFX 10.0 두 가지의 소프트웨어가 사용되었다. 해석 결과를 바탕으로 좀더 정확한 유량 한계치 계산을 위해 추가로 수행되어야 할 해석에 대해 제안하였다.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1442-1450
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• 2017

In liquid metal fast breeder reactors, postulated failures of the plant protection system may lead to serious unprotected accidental consequences. Unprotected transients are generically categorized as transient overpower accidents and transient under cooling accidents. In both cases, core meltdown may occur and this can lead to a molten fuel coolant interaction (MFCI). The understanding of MFCI phenomena is essential for study of debris coolability and characteristics during post-accident heat removal. Sodium is used as coolant in liquid metal fast breeder reactors. Viewing inside sodium at elevated temperature is impossible because of its opaqueness. In the present study, a methodology to depict MFCI phenomena using a flat panel detector based imaging system (i.e., real time radiography) is brought out using a woods metal-water experimental facility which simulates the $UO_2-Na$ interaction. The developed imaging system can capture attributes of the MFCI process like jet breakup length, jet front velocity, fragmented particle size, and a profile of the debris bed using digital image processing methods like image filtering, segmentation, and edge detection. This paper describes the MFCI process and developed imaging methodology to capture MFCI attributes which are directly related to the safe aspects of a sodium fast reactor.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.603-610
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• 2021

The major drawback of zirconia-based materials, in view of their applications as targets for minor actinide transmutation, is their poor thermal conductivity. The addition of MgO, which has high thermal conductivity, to zirconia-based materials is expected to improve their thermal conductivity. On these grounds, the present study aims at phase characterization and thermophysical property evaluation of neodymium-substituted zirconia (Zr_0.8Nd_0.2O_1.9; using Nd₂O₃ as a surrogate for Am₂O₃) and its composites with MgO. The composite was prepared by a solid-state reaction of Zr_0.8Nd_0.2O_1.9 (synthesized by gel combustion) and commercial MgO powders at 1773 K. Phase characterization was carried out by X-ray diffraction and the microstructural investigation was performed using a scanning electron microscope equipped with energy dispersive spectroscopy. The linear thermal expansion coefficient of Zr_0.8Nd_0.2O_1.9 increases upon composite formation with MgO, which is attributed to a higher thermal expansivity of MgO. Similarly, specific heat also increases with the addition of MgO to Zr_0.8Nd_0.2O_1.9. Thermal conductivity was calculated from measured thermal diffusivity, temperature-dependent density and specific heat values. Thermal conductivity of Zr_0.8Nd_0.2O_1.9-MgO (50 wt%) composite is more than that of typical UO₂ fuel, supporting the potential of Zr_0.8Nd_0.2O_1.9-MgO composites as target materials for minor actinides transmutation.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1644-1650
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• 2023

In spent fuel reprocessing, UO₂(DBP)₂ (U-DBP) can be deposited in stainless steel equipment. U-DBP must be removed by dissolution and the process must not cause corrosion to stainless steel. This study was conducted to find the best scheme for dissolution. U-DBP was manufactured by the titrimetric sedimentation method. The effects of different factors on the dissolution of U-DBP were investigated. For example, solid-liquid ratio, hydrazine carbonate solutions with different mass components, mixed solutions containing different concentrations of H₂O₂, and different carbonates. The results indicated that U-DBP does not have a regular crystal morphology. With the increase of the solid-liquid ratio and the mass fraction of hydrazine carbonate, the concentration of U(VI) at the dissolution equilibrium increases gradually. The addition of H₂O₂ has a great promotion effect on the dissolution. However, when the concentration of H₂O₂ is greater than 0.5 M, the dissolution solution may have an erosive effect on the stainless steel. (NH₄)₂CO₃ can increase the dissolution capacity of dissolved U-DBP, but it may also accelerate the corrosion of stainless steel.

• Nuclear Engineering and Technology
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• 제52권2호
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• pp.248-257
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• 2020

This study presents an initial design for a novel system consisting in a coupled nuclear reactor and a phase change material-based thermal energy storage (TES) component, which acts as a buffer and regulator of heat transfer between the primary and secondary loops. The goal of this concept is to enhance the capacity factor of nuclear power plants (NPPs) in the case of high integration of renewable energy sources into the electric grid. Hence, this system could support in elevating the economics of NPPs in current competitive markets, especially with subsidized solar and wind energy sources, and relatively low oil and gas prices. Furthermore, utilizing a prismatic-core advanced high temperature reactor (PAHTR) cooled by a molten salt with a high melting point, have the potential in increasing the system efficiency due to its high operating temperature, and providing the baseline requirements for coupling other process heat applications. The present research studies the neutronics and thermal hydraulics (TH) of the PAHTR as well as TH calculations for the TES which consists of 300 blocks with a total heat storage capacity of 150 MWd. SERPENT Monte Carlo and MCNP5 codes carried out the neutronics analysis of the PAHTR which is sized to have a 5-year refueling cycle and rated power of 300 MW_th. The PAHTR has 10 metric tons of heavy metal with 19.75 wt% enriched UO₂ TRISO fuel, a hot clean excess reactivity and shutdown margin of $33.70 and -$115.68; respectively, negative temperature feedback coefficients, and an axial flux peaking factor of 1.68. Star-CCM + code predicted the correct convective heat transfer coefficient variations for both the reactor and the storage. TH analysis results show that the flow in the primary loop (in the reactor and TES) remains in the developing mixed convection regime while it reaches a fully developed flow in the secondary loop.

• 방사성폐기물학회지
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• 제1권1호
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• pp.11-23
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• 2003

국내의 가동 중지된 우라늄 변환시설의 해체 시 우라늄 화합물로 오염되어 대량으로 발생될 금속폐기물의 재활용 또는 자체처분을 위한 제염기술로 전해제염 공정의 적용성을 평가하였다. 이를 위하여 우라늄 변환시설 내부설비의 주 구성 재료인 SUS-304 와 Inconel-600 금속시편을 대상으로 전해용해 실험을 수행하였다. SUS-304 와 Inconel-600 금속시편에 대한 전해용해 성능에 있어서 중성염 전해용액으로 $Na_2$SO$_4$가 가장 효과적이었으나, 우라늄변환시설의 가동 시 질산 매질과 주로 접촉했던 설비 표면의 이력과 시설 가동 중 발생한 우라늄 폐액의 성상을 고려하여 $Na_2$SO$_4$ 전해용액 내에서의 SUS-304 시편에 대한 전해용해와 비교해서 약 30%, 그리고 Inconel-600 시편에 대해서는 거의 동등한 성능을 보인 NaNO$_3$ 중성염 용액을 금속성폐기물의 전해제염 용액으로 선정하였다. 본 연구에서는 NaNO$_3$ 중성염 전해용액에서 전류밀도, 전해시간 및 전해 용액의 농도가 SUS-304 및 Inconel-600 금속시편의 전해용해 성능에 미치는 영향을 조사하였다. 이 실험결과를 바탕으로 실제 우라늄 변환시설로부터 인출하여 $UO_2$, AUC 및 ADU 등의 우라늄 화합물로 오염된 시편에 대해 전류밀도 100mA/$\textrm{cm}^2$, IM NaNO$_3$ 전해용액 내에서 전해 제염 실증시험을 수행하였으며, 오염물의 종류 및 오염준위의 대소와는 관계없이 모든 시편에 대하여 10분 이내의 짧은 시간 내에 자체처분 기준치 이하로 $\alpha$ 및 $\beta$ 방사능 준위를 감소시킴으로써 본 중성염 전해제염이 매우 성공적임을 확인하였다.nely regimented hierarchical language. I try, in this paper, to develop the idea that hierarchical regimentation of Korean language uses is not humane. 1 of for the main argument for the thesis as what follows: How could one justify the hierarchical regimentation of a language like Korean\ulcorner Only if there is an essential structure in which the fine grades of differences of social positions of all the people are distinct; The essentialism here involved is not plausible. And I may add that language is to be used fur the purposes of communication, rationalization and expression. If true, language use is a genuine art of liberation or humanization. Any overt hierarchical language tends to damage those purposes and more to enforce those oppressive elements already existing in the community. Then, a hierarchical language is to defeat its own purpose.중 행정부가 북한에 대해 실시한 포용정책이 어떠한 성과를 거두고 어떠한 문제점을 간과하고 있는가에 대해 논의하고, 대북 정책의 새로운 지평을 논의하는 것을 목적으로 하고 있다. 1) 포용 정책은