• Nuclear Engineering and Technology
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• 제53권1호
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• pp.304-313
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• 2021

Nickel base Alloy X-750, which is used as fastener parts in light-water reactor (LWR), has experienced many failures by environmentally assisted cracking (EAC). In order to improve the reliability of passive components for nuclear power plants (NPP's), it is necessary to study the failure mechanism and to predict crack growth behavior by developing a probabilistic failure model. In this study, The Bayesian inference was employed to reduce the uncertainties contained in EAC modeling parameters that have been established from experiments with Alloy X-750. Corrosion fatigue crack growth rate model (FCGR) was developed by fitting into Paris' Law of measured data from the several fatigue tests conducted either in constant load or constant ΔK mode. These parameters characterizing the corrosion fatigue crack growth behavior of X-750 were successfully updated to reduce the uncertainty in the model by using the Bayesian inference method. It is demonstrated that probabilistic failure models for passive components can be developed by updating a laboratory model with field-inspection data, when crack growth rates (CGRs) are low and multiple inspections can be made prior to the component failure.

• 방사성폐기물학회지
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• 제21권3호
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• pp.397-410
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• 2023

This study examined the efficacy of various chlorinating agents in partitioning light water reactor spent fuel, with the aim of optimizing the chlorination process. Through thermodynamic equilibrium calculations, we assessed the outcomes of employing MgCl₂, NH₄Cl, and Cl₂ as chlorinating agents. A comparison was drawn between using a single agent and a sequential approach involving all three agents (MgCl₂, NH₄Cl, and Cl₂). Following heat treatment, the utilization of MgCl₂ as the sole chlorinating agent resulted in a moderate separation. Specifically, this method yielded a solid separation with 96.9% mass retention, 31.7% radioactivity, and 44.2% decay heat, relative to the initial spent fuel. In contrast, the sequential application of the chlorinating agents following heat treatment led to a final solid separation characterized by 93.1% mass retention, 5.1% radioactivity, and 15.4% decay heat, relative to the original spent fuel. The findings underscore the potential effectiveness of a sequential chlorination strategy for partitioning spent fuel. This approach holds promise as a standalone technique or as a complementary process alongside other partitioning processes such as pyroprocessing. Overall, our findings contribute to the advancement of spent fuel management strategies.

• KSBB Journal
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• 제27권2호
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• pp.97-102
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• 2012

In the present study, the biomass productivity of two green microalgae (Chlorella sp. and Dunaliella salina DCCBC2) were assessed in a 12 L tubular photobioreactor under optimum culture conditions. In the batch culture optimization process, the Chlorella sp. biomass was obtained as 1.2 g/L under atmospheric air as a sole $CO_2$ source and other culture conditions as follows: light intensity, temperature, pH, $NH_4Cl$ and $K_2HPO_4$ were 100 ${\mu}E/m^2/s$, $27^{\circ}C$, 7.0, 20.0 mM and 2.0 mM, respectively. On the other hand, 2.9 g/L of D. salina DCCBC2 biomass production was observed under the following conditions: light intensity, temperature, pH, $KNO_3$ and $K_2HPO_4$were 80 ${\mu}E/m^2/s$, $27^{\circ}C$, 8.0, 3.0 mM and 0.025 mM, respectively. At 1% $CO_2$ supply to the reactor, the Chlorella sp. production was reached 1.53 g/L with 25% increment under the same operating conditions. In addition, the maximum D. salina DCCBC2 biomass was observed as 3.40 g/L at 3% $CO_2$ concentration. Based on the aforementioned optimized conditions, the dilution rate and maximal biomass productivity of Chlorella sp. and D. salina DCCBC2 in the continuous cultivation were 0.4/d and 0.6 g/L/d and 0.6/d and 1.5 g/L/d, respectively.

• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1120-1130
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• 2020

The CREOLE experiment performed In the EOLE critical facility located In the Nuclear Center of CADARACHE - CEA have allowed us to get interesting and complete experimental information on the temperature effects in the light water reactor lattices. To analyze these experiments with accuracy an elaborate calculation scheme using the Monte Carlo method implemented in the MCNP6.1 code and the ENDF/B-VII.1 cross section library has been developed. We have used the ENDF/B-VII.1 data provided with the MCNP6.1.1 version in ACE format and the Makxsf utility to handle the data in the specific temperatures not available in the MCNP6.1.1 original library. The main purpose of this analysis is the qualification of the ENDF/B-VII.1 nuclear data for the prediction of the Reactivity Temperature Coefficient while ensuring the ability of the MCNP6.1 system to model such a complex experiment as CREOLE. We have analyzed the case of UO₂ lattice with 1166 ppm of boron in ordinary water moderator in specified temperatures. A detailed comparison of the calculated effective multiplication factors with the reference ones [1] in room temperature presented in this work shows a good agreement demonstrating the validation of our 3D calculation model. The discrepancies between calculations and the differential measurements of the Reactivity Temperature Coefficient for the analyzed configuration are relatively small: the maximum discrepancy doesn't exceed 1,1 pcm/℃. In addition to the analysis of direct differential measurements of the reactivity temperature coefficient performed in the poisoned UO₂ lattice configuration, we have also analyzed integral measurements in UO₂ clean lattice configuration using equivalency of the integral temperature reactivity worth with the driver core fuel reactivity worth and soluble boron reactivity worth. In this case both of the ENDF/B-VII.1 and JENDL.4 libraries were used in our analysis and the obtained results are very similar.

• 방사성폐기물학회지
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• 제18권spc호
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• pp.21-36
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• 2020

With respect to spent nuclear fuels, disposal containers and bentonite buffer blocks in deep geological disposal systems are the primary engineered barrier elements that are required to isolate radioactive toxicity for a long period of time and delay the leakage of radio nuclides such that they do not affect human and natural environments. Therefore, the thermal stability of the bentonite buffer and structural integrity of the disposal container are essential factors for maintaining the safety of a deep geological disposal system. The most important requirement in the design of such a system involves ensuring that the temperature of the buffer does not exceed 100℃ because of the decay heat emitted from high-level wastes loaded in the disposal container. In addition, the disposal containers should maintain structural integrity under loads, such as hydraulic pressure, at an underground depth of 500 m and swelling pressure of the bentonite buffer. In this study, we analyzed the thermal stability and structural integrity in a deep geological disposal environment of the improved deep geological disposal systems for domestic light-water and heavy-water reactor types of spent nuclear fuels, which were considered to be subject to direct disposal. The results of the thermal stability and structural integrity assessments indicated that the improved disposal systems for each type of spent nuclear fuel satisfied the temperature limit requirement (< 100℃) of the disposal system, and the disposal containers were observed to maintain their integrity with a safety ratio of 2.0 or higher in the environment of deep disposal.

• 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.

• 한국응용과학기술학회지
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• 제34권2호
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• pp.349-356
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• 2017

본 연구에서는 $TiO_2$ nanotube 광촉매의 고도산화처리능을 비교하기 위해서 OH 라디칼 생성력을 평가하고자 하였다. 자외선 조사에 따른 Probe compound인 4-Chlorobenzoic acid (pCBA)의 농도 감소에 따라 OH radical 생성량을 산정하는 방법으로 광촉매 효율을 평가하였는데, $TiO_2$ nanotube 표면에서의 전자의 흐름을 원활하게 하기 위하여 전기적 에너지를 주었을 시 광촉매 효율의 증가 가능성 또한 확인하고자 자외선 조사 시 전류밀도를 인가하는 방법으로 실험을 진행하였다. 실험에 사용된 $TiO_2$ nanotube는 전극효과를 부여하기 위해 양극산화법으로 티타늄판을 이용하여 제조하였으며, pCBA 용액에는 전도도를 부여하기 위하여 NaCl을 첨가하여 전해질로 사용하였다. 정전류 정전압 조건하에서 자외선조사 실험을 진행하였으며, 전류가 흐르는 광촉매에 자외선 조사 시 OH 라디칼 생성량은 광촉매 없이 자외선만 조사하였을 때에 비해 약 5.6배, $TiO_2$ 광촉매와 함께 자외선을 조사하였을 보다 약 2.2배 증가하였다. 결과적으로 광촉매반응에 전기적 에너지를 부여하였을 시 시너지효과를 가져올 수 있는 가능성을 확인할 수 있었다.

• 대한토목학회논문집
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• 제29권5B호
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• pp.497-502
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• 2009

광촉매에 의한 수처리 방법은 수중에서 오염물질을 직접분해 처리하며 난분해성 유기물 또는 무기물의 분해가 용이하다. 특히 2차 오염물질의 생성이 거의 없는 것이 가장 큰 장점이라 할 수 있다. 하지만 광촉매 형태에 따라 여러 문제가 발생한다. 회전 광촉매 형태는 기존의 문제를 최소화시키고 회전원판법을 적용하여 고도산화처리가 가능하다. 회전광촉매 반응기의 적용을 위해서는 여러 가지 설계와 운전인자 및 특성에 대한 고찰이 필요하다. 본 연구에서는 회전 광촉매를 $TiO_2$ 고정화 작업으로 회전원판법에 적합하게 제작하였다. 이를 이용하여 회전 광촉매 반응에 의한 폐수처리를 수행하기 위한 운전인자들을 도출하였다. 회전 광촉매 $TiO_2$ 함량은 최대 70%가 한계로 나타났다. $TiO_2$ 함유량이 증가할수록 처리효율도 지속적으로 증가되고 있다. 적절한 회전 광촉매는 R4로 $TiO_2$ 함유량 36.8% 이다. 자외선 세기가 증가 할수록 TCODcr의 분해효과는 지속적으로 증가 된다. 다만 적절한 광원의 세기는 경제성을 고려해서 판단하여야 한다. 회전 광촉매의 회전속도가 증가할수록 처리효율은 향상된다. UV lamp를 반응조에 침지시키지 않을 때 회전 광촉매 수심변화는 수심이 50%, 30%, 10%, 70%, 100% 순으로 처리효율이 높게 나왔다. 본 실험을 바탕으로 태양광에서도 유기물을 처리할 수 있는 시스템을 개발에 바탕이 될 것이라 판단한다.

• 대한기계학회논문집A
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• 제39권3호
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• pp.249-257
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• 2015

원자력 발전소의 반응로에는 핵분열 에너지를 생성하고 방사성 물질의 유출을 막는 핵연료 집합체가 있으며, 이러한 집합체는 핵연료와 피복관으로 구성되어 있는 핵 연료봉으로 구성되어 있다. 원자로에서 핵연료봉 거동의 안전성을 평가하기 위해 해석적인 방법을 적용하며 이러한 평가 코드를 핵 연료 성능 코드라 한다. 경수로 핵연료 해석에서는 간극의 두께에 따라 열전도도가 크게 영향을 받는 간극 열전도도가 주요 거동해석에 영향을 미친다. 본 연구에서는 간극 두께에 따라 열전도도가 변화하는 3 차원 간극 요소(Gap element)를 제안하였으며, 이를 적용하기 위해 3 차원 열탄성 모듈을 FORTRAN90을 이용하여 개발하였다. 제안된 3 차원 간극 요소를 이용하여 핵 연료봉에서 발생할 수 있는 비대칭적인 형상인 핵 연료 표면에 결함이 생긴 경우 MPS(Missing Pellet Surface)와 핵연료봉의 편심(Eccentricity of the nuclear fuel rod) 형상에 대하여 3 차원 해석을 진행하였다.

• Nuclear Engineering and Technology
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• 제38권2호
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• pp.163-174
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• 2006

The $Ph{\acute{e}}bus$ FP project is an international reactor safety project. Its main objective is to study the release, transport and retention of fission products in a severe accident of a light water reactor (LWR). The FPT4 test was performed with a fuel debris bed geometry, to look at late phase core degradation and the releases of low volatile fission products and actinides. Post Test Analyses results indicate that releases of noble gases (Xe, Kr) and high-volatile fission products (Cs, I) were nearly complete and comparable to those obtained during $Ph{\acute{e}}bus$ tests performed with a fuel bundle geometry (FPT1, FPT2). Volatile fission products such as Mo, Te, Rb, Sb were released significantly as in previous tests. Ba integral release was greater than that observed during FPT1. Release of Ru was comparable to that observed during FPT1 and FPT2. As in other $Ph{\acute{e}}bus$ tests, the Ru distribution suggests Ru volatilization followed by fast redeposition in the fuelled section. The similar release fraction for all lanthanides and fuel elements suggests the released fuel particles deposited onto the plenum surfaces. A blockage by molten material induced a steam by-pass which may explain some of the low releases. The revaporisation testing under different atmospheres (pure steam, $H_2/N_2$ and steam /$H_2$) and up to $1000^{\circ}C$ was performed on samples from the first upper plenum. These showed high releases of Cs for all the atmospheres tested. However, different kinetics of revaporisation were observed depending on the gas composition and temperature. Besides Cs, significant revaporisations of other elements were observed: e.g. Ag under reducing conditions, Cd and Sn in steam-containing atmospheres. Revaporisation of small amounts of fuel was also observed in pure steam atmosphere.