• 방사성폐기물학회지
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• 제7권2호
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• pp.79-86
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• 2009

원자력시설의 콘크리트 폐기물은 서로 다른 메카니즘에 의해 다양한 핵종에 의해 방사화 되거나 오염된다. 우라늄 변환시설 및 연구로 해체 시 발생된 오염된 콘크리트의 부피감용을 위해 가열 분쇄 실험에 의해 자갈, 모래, 페이스트의 골재의 크기에 따른 핵종의 분배특성에 대해 고찰하였다. 실험결과 대부분의 방사성 핵종은 골재로부터 제거되어 페이스트에 존재하였으며 특히, 가열 온도는 방사성 핵종을 오염된 콘크리트 폐기물로부터 분리하는데 중요한 변수로 확인되었다. 즉, 콘크리트 표면에 오염된 물질은 밀도가 높은 자갈, 모래보다는 다공성 물질의 페이스트에 농축되었다. 방사화 콘크리트에서는 80%, 우라늄 변환시설의 콘크리트 폐기물에서는 약 75% 정도의 부피감용을 얻었다.

• Nuclear Engineering and Technology
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• 제25권3호
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• pp.413-423
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• 1993

본 논문에서는 경수로 노심 용융사고시 1차계통의 압력이 높은 경우에 발생하는 격납용기 직접가열 현상에 대한 실험연구를 하였다. 실험은 고리 1,2호기와 영광 3,4호기의 1/30 축소규모와 고리 1,2호기의 1/20 축소규모를 실험모형으로 하여 수행되었으며, 고리 1,2호기의 경우 축소 규모에 따른 검증도 시도하였다. 실험의 주요 변수는 초기 압력 용기의 압력, 파열면적 및 캐비티의 구조 등이다. 실험결과로부터 캐비티 외부로의 용융노심 분사비율은 높은 초기압력과 큰 파열면적을 가진 경우가 더 높으며 캐 비티의 구조가 분사비율에 큰 영향을 미침을 알 수 있었다. 본 연구의 실험결과를 이용하여 분사비율에 대한 실험관계식을 무차원 유효시간의 함수로 도출하여 제시하였으며, 이 실험관계식은 본 실험결과 뿐만 아니라 한국 과학기술원의 실험자료 및 미국 BNL 실험결과와도 잘 일치하였다.

• 공업화학
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• 제7권6호
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• pp.1164-1173
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• 1996

현재 국내에서 가동중인 원자력발전소 공급용 핵연료 분말제조 공정에서 발생되는 폐액의 물성과 처리방법에 대한 연구가 수행되었다. 중수로형과 경수로형 발생 폐액에 함유된 우라늄을 회수/처리하기 위하여, 공히 폐액 속의 탄산이온의 제거가 필수적이다. 중수로형은 ADU 형태로 경수로형의 경우 $UO_4$ 화합물 형태로 처리하는 것이, 최종 폐액의 우라늄 농도를 최소화할 수 있었다. 처리후 폐액의 우라늄 농도는 중수로형 폐액의 경우, 폐액을 가열하여 ADU를 제조한 후 여액에 lime을 처리하는 방법으로 1ppm까지, 경수로형 폐액의 경우 $UO_4{\cdot}2NH_4F$형태로 우라늄을 침전시킬 경우 0.8ppm까지 여액중의 우라늄 농도를 낮출 수 있었다. 최적 처리조건은 중수로형 폐액의 경우 $101^{\circ}C$까지 단순 가열방법이, 경수로형 폐액의 경우 가열한 후 $60^{\circ}C$에서 암모니아로 pH를 9.5로 조절한 후 과산화수소 용액을 첨가하여 1시간 반응시키는 경우로 나타났다. 폐액으로부터 회수된 우라늄 화합물은, 중수로형 폐액인 경우 pH가 낮을수록 회수된 ADU 입자의 크기가 증가하였으며, 경수로형 폐액인 경우 회수된 uranium peroxide 화합물을 공기분위기에서 열분해시킨 결과 기존의 AUC 분말이 열분해되어 나타내는 특성과 동일한 특성을 보임에 따라 핵연료분말 제조공정으로 recycle이 가능한 것으로 판단되었다.

• Journal of Mechanical Science and Technology
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• 제17권8호
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• pp.1171-1184
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• 2003

An experimental study on critical heat flux (CHF) has been performed in an internally heated vertical annulus with non-uniform heating. The CHF data for the chopped cosine heat flux have been compared with those for uniform heat flux obtained from the previous study of the authors, in order to investigate the effect of axial heat flux distribution on CHF. The local CHF with the parameters such as mass flux and critical quality shows an irregular behavior. However, the total critical power with mass flux and the average CHF with critical quality are represented by a unique curve without the irregularity. The effect of the heat flux distribution on CHF is large at low pressure conditions but becomes rapidly smaller as the pressure increases. The relationship between the critical quality and the boiling length is represented by a single curve, independent of the axial heat flux distribution. For non-uniform axial heat flux distribution, the prediction results from Doerffer et al.'s and Bowling's CHF correlations have considerably large errors, compared to the prediction for uniform heat flux distribution.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.621-625
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• 2020

Cast austenitic stainless steel (CASS) is used for fabricating different components of the primary reactor coolant system of pressurized water reactors. However, the thermal embrittlement of CASS resulting from long-term operation causes structural safety problems. Ultrasonic testing for flaw detection has been used to assess the thermal embrittlement of CASS; however, the high scattering and attenuation of the ultrasonic wave propagating through CASS make it difficult to accurately quantify the flaw size. In this paper, we present a different approach for evaluating the thermal embrittlement of CASS by assessing changes in the material properties of CASS using a nonlinear ultrasonic technique, which is a potential nondestructive method. For the evaluation, we prepared CF8M specimens that were thermally aged under four different heating conditions. Nonlinear ultrasonic measurements were performed using a contact piezoelectric method to obtain the relative ultrasonic nonlinearity parameter, and a mini-sized tensile test was performed to investigate the correlation of the parameter with material properties. Experimental results showed that the ultrasonic nonlinearity parameter had a correlation with tensile properties such as the tensile strength and elongation. Consequently, we could confirm the applicability of the nonlinear ultrasonic technique to the evaluation of the thermal embrittlement of CASS.

• 한국세라믹학회지
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• 제46권6호
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• pp.574-582
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• 2009

$UO_2$ microsphere particles, core material of HTGR(High Temperature Gas Reactor) nuclear fuel, were prepared using by the GSP(Gel Supported Precipitation) method which is a modified-method of the wet sol-gel process. The spherical shape of initial liquid ADU droplets from the vibration nozzle system was continuously kept till the conversion to the final $UO_2$ microsphere. But the size of a final $UO_2$ microsphere was shrunken to about 25% of an initial ADU droplet size. Also, we found that the composition of dried-ADU gel particles was constituted of the very complicated phases, coexisted the U=O, C-H, N-H, N-O, and O-H functional groups by FT-IR. The important factors for obtain the no-crack $UO_2$ microsphere during the thermal treatment processes must perfectly wash out the remained-$NH_4NO_3$ within the ADU gel particle in washing process and the selections of an appropriate heating rate at a suitable gas atmosphere, during the calcining of ADU gel particles, the reducing of $UO_3$ particles, and the sintering of $UO_2$ particles, respectively.

• Nuclear Engineering and Technology
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• 제47권5호
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• pp.608-616
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• 2015

The migration of silver (Ag) in silicon carbide (SiC) and $^{110m}Ag$ through SiC of irradiated tristructural isotropic (TRISO) fuel has been studied for the past three to four decades. However, there is no satisfactory explanation for the transport mechanism of Ag in SiC. In this work, the diffusion coefficients of Ag measured and/or estimated in previous studies were reviewed, and then pre-exponential factors and activation energies from the previous experiments were evaluated using Arrhenius equation. The activation energy is $247.4kJ{\cdot}mol^{-1}$ from Ag paste experiments between two SiC layers produced using fluidized-bed chemical vapor deposition (FBCVD), $125.3kJ{\cdot}mol^{-1}$ from integral release experiments (annealing of irradiated TRISO fuel), $121.8kJ{\cdot}mol^{-1}$ from fractional Ag release during irradiation of TRISO fuel in high flux reactor (HFR), and $274.8kJ{\cdot}mol^{-1}$ from Ag ion implantation experiments, respectively. The activation energy from ion implantation experiments is greater than that from Ag paste, fractional release and integral release, and the activation energy from Ag paste experiments is approximately two times greater than that from integral release experiments and fractional Ag release during the irradiation of TRISO fuel in HFR. The pre-exponential factors are also very different depending on the experimental methods and estimation. From a comparison of the pre-exponential factors and activation energies, it can be analogized that the diffusion mechanism of Ag using ion implantation experiment is different from other experiments, such as a Ag paste experiment, integral release experiments, and heating experiments after irradiating TRISO fuel in HFR. However, the results of this work do not support the long held assumption that Ag release from FBCVD-SiC, used for the coating layer in TRISO fuel, is dominated by grain boundary diffusion. In order to understand in detail the transport mechanism of Ag through the coating layer, FBCVD-SiC in TRISO fuel, a microstructural change caused by neutron irradiation during operation has to be fully considered.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2011년도 정기 학술대회
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• pp.415-418
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• 2011

Thermal fatigue is one of the life-limiting damage mechanisms in the nuclear power plant conditions. The turbulent mixing of fluids of different temperatures induces rapid temperature changes to the pipe wall. The successive thermal transients cause varying cyclic thermal stresses. These cyclic thermal stresses cause fatigue crack nucleation and growth similar to the cyclic mechanical stresses. The aim of this study was to fulfil the need by developing an real crack manufacturing method, which would produce realistic cracks. The test material was austenitic STS 304, which is used as pipelines in the reactor coolant system of a nuclear power plants. In order to fabricate thermal fatigue crack similar to realistic crack, successive thermal transients were applied to the specimen. Thermal transient cycles were combined with heating (60sec) and cooling cycle (30sec). And, In order to identify ultrasonic characteristic, it was performed the ultrasonic reflection measuring method for the fabricated specimen. From the results of ultrasonic reflection measuring testing, it was conformed that A-scan results(average 83% of real crack depth) for the TFC reference specimen was more enhanced NDT reliability than results(average 38% of real crack depth) for the EDM notch reference specimen.

• Nuclear Engineering and Technology
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• 제36권6호
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• pp.559-570
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• 2004

A computational fluid dynamics model for predicting moderator circulation inside the Canada deuterium uranium (CANDU) reactor vessel has been developed to estimate the local subcooling of the moderator in the vicinity of the calandria tubes. The buoyancy effect induced by the internal heating is accounted for by the Boussinesq approximation. The standard $k-{\varepsilon}$ turbulence model with logarithmic wall treatment is applied to predict the turbulent jet flows from the inlet nozzles. The matrix of the calandria tubes in the core region is simplified to a porous media in which the anisotropic hydraulic impedance is modeled using an empirical correlation of pressure loss. The governing equations are solved by DFX-4.4, a commercial CFD code developed by AEA technology. The resultant flow patterns of the constant-z slices containing the inlet nozzles and the outlet port are "mined-type", as observed in the former 2-dimensional experimental investigations. With 103% full power for conservatism, the maximum temperature of the moderator is $82.9^{\circ}C$ at the top of the core region. Considering the hydrostatic pressure change, the minimum subcooling is $24.8^{\circ}C$.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3126-3132
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• 2023

To meet the increasing heating needs of the China Fusion Experimental Tokamak Reactor (CFETR), the output power in each Lower Hybrid Current Drive (LHCD) transmission line should be increased from 250 kW to 500 kW. Therefore, a new high-power water load must be developed for the 4.6 GHz 500 kW LHCD system. This paper aims to report the most recent research progress of the water load: aluminum nitride (AlN) ceramic is used as the media material to isolate the water and vacuum, and the radio frequency (RF) simulation results show that the return loss of the water load is less than -25dB at 4.6 GHz over a wide temperature range. Under 500 kW continuous wave (CW) operation, the maximum temperatures of the ceramic and water are separately 67 ℃ and 62 ℃, resulting in thermal deformation of the ceramic of approximately 0.003 mm. Moreover, the AlN water load was tested on the 4.6 GHz 250 kW high-power test bench and found to work well with low reflected power.