• Nuclear Engineering and Technology
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• 제50권7호
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• pp.1120-1130
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• 2018

An analysis of a pyroprocessing safeguards methodology employing the Pu-to-$^{244}Cm$ ratio is presented. The analysis includes characterization of representative used nuclear fuel assemblies with respect to computed nuclide composition. The nuclide composition data computationally generated is appropriately reformatted to correspond with the material conditions after each step in the head-end stage of pyroprocessing. Uncertainty in the Pu-to-$^{244}Cm$ ratio is evaluated using the Geary-Hinkley transformation method. This is because the Pu-to-$^{244}Cm$ ratio is a Cauchy distribution since it is the ratio of two normally distributed random variables. The calculated uncertainty of the Pu-to-$^{244}Cm$ ratio is propagated through the mass flow stream in the pyroprocessing steps. Finally, the probability of Type-I error for the plutonium Material Unaccounted For (MUF) is evaluated by the hypothesis testing method as a function of the sizes of powder particles and granules, which are dominant parameters to determine the sample size. The results show the probability of Type-I error is occasionally greater than 5%. However, increasing granule sample sizes could surmount the weakness of material accounting because of the non-uniformity of nuclide composition.

• 자원환경지질
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• 제33권3호
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• pp.195-204
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• 2000

order to investigate physical characteristics and element concentrations of sediments, coastal bottom sediments were collected at 20 stations in the vicinity of Youngkwang Nuclear Power Plant. After air drying of samples in the laboratory. article size distribution was examined by Master sizer (X-350F), radio-activity by HPGe ${\gamma}$-spectrphotometer, and element concentrations by ICP-AES and AAS. According to particle size analysis , sediments are mainly composed of silt fraction weith 23% of sand, 65% of silt and 12% of clay on average. Most sediments are derived from muddy environment that silt dominates with the characteristics of 5.3${\varsigma}$ mean particle size, poorly sorted, very fine skewed and lepto-kurtic. Only two sediments are well sorted with sandy silt owing to wind, winnowing action, tide and current andits complex reactions. Element concentrations in the coastal bottom sediments are relatively high at finer sediment and show significant relationship with grain size. Index of geoaccumulation by heavy metals at every sampling station is classified as practically unpolluted. The radioactivities of the sediments were measured for 15 isotope elements, and 2 elements of K-40 and Cs-137 were detected in most sediments. The K-40 is the natural nuclide and the artificial nuclide of Cs-137 was thought to be derived from the fallout of past nuclear weapon test. The results of correlation coefficient between grain size and radioactivity shows that the activity of Cs-137 significantly increases in finer grain.

• Nuclear Engineering and Technology
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• 제47권7호
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• pp.859-866
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• 2015

The aim of this work is to investigate the transport mechanism of radioactive nuclides through the reverse osmosis (RO) membrane and to estimate its effectiveness for nuclide separation from radioactive liquid waste. An analytical model is developed to simulate the RO separation, and a series of experiments are set up to confirm its estimated separation behavior. The model is based on the extended Nernst-Plank equation, which handles the convective flux, diffusive flux, and electromigration flux under electroneutrality and zero electric current conditions. The distribution coefficient which arises due to ion interactions with the membrane material and the electric potential jump at the membrane interface are included as boundary conditions in solving the equation. A high Peclet approximation is adopted to simplify the calculation, but the effect of concentration polarization is included for a more accurate prediction of separation. Cobalt and cesium are specifically selected for the experiments in order to check the separation mechanism from liquid waste composed of various radioactive nuclides and nonradioactive substances, and the results are compared with the estimated cobalt and cesium rejections of the RO membrane using the model. Experimental and calculated results are shown to be in excellent agreement. The proposed model will be very useful for the prediction of separation behavior of various radioactive nuclides by the RO membrane.

• Nuclear Engineering and Technology
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• 제42권2호
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• pp.175-182
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• 2010

As part of a fundamental study on the volume reduction of contaminated concrete wastes, the separation characteristics of the aggregates and the distribution of the radioactivity in the aggregates were investigated. Radioisotope $^{60}Co$ was artificially used as a model contaminant for non-radioactive crushed concrete waste. Volume reduction for radioactively contaminated dismantled concrete wastes was carried out using activated heavy weight concrete taken from the Korea Research Reactor 2 (KRR-2) and light weight concrete from the Uranium Conversion Plant (UCP). The results showed that most of the $^{60}Co$ nuclide was easily separated from the contaminated dismantled concrete waste and was concentrated mainly in the porous fine cement paste. The heating temperature was found to be one of the effective parameters in the removal of the radionuclide from concrete waste. The volume reduction rate achieved was above 80% for the KRR-2 concrete wastes and above 75% for the UCP concrete wastes by thermal and mechanical treatment.

• 한국전산유체공학회지
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• 제10권4호통권31호
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• pp.66-71
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• 2005

HANARO, a multi-purpose research reactor, 30 MWth open-tank-in-pool type, is planning to produce a fission moly-99 of radio isotopes, a mother nuclide of Tc-99m, a medical isotope and a target handling tool is under development for loading and unloading it in a circular flow tube (OR-5) of HANARO. A guide tube is extended from the reactor core to the top of the reactor chimney for easily loading the target under a normal operation of the reactor. But active coolant through the core can be quickly raised up to the top of the chimney through the guide tube. The jet flow was suppressed in the guide tube after reducing the inner diameter of a flow restriction orifice installed in the OR-5 flow tube for adding the pressure difference in the flow tube. This paper describes an analytical analysis to calculate the flow distribution in the core of HANARO after suppressing the jet flow of the guide tube. As results, it was confirmed through the analysis results that the flow distribution in the core of HANARO were not adversely affected.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2005년도 춘계 학술대회논문집
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• pp.70-73
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• 2005

The HANARO, multi-purpose research reactor, 30 MWth open-tank-in-pool type, is planning to produce a fission moly-99 of radio isotopes, a mother nuclide of Tc-99m, a medical isotope and is under developing a target handling tool for loading and unloading it in a circular flow tube (OR-5). A guide tube is extended from the reactor core to the top of the reactor chimney for easily loading the target under the reactor normal operation. But active coolant through the core can be quickly raised up to the top of the chimney through the guide tube. The jet flow was suppressed in the guide tube after reducing the inner diameter of a flow restriction orifice installed in the OR-5 flow tube for adding the pressure difference in the flow tube after unloading the target. This paper describes an analytical analysis to calculate the flow distribution in the core of the HANARO after suppressing the jet flow of the guide tube. As results, it was confirmed through the analysis results that the flow distribution in the core of the HANARO were not adversely affected.

• 방사성폐기물학회지
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• 제5권3호
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• pp.213-220
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• 2007

유리화공정 고온영역에서의 방사성 배기체 유동해석을 해석하기 위하여 상용 수치해석 범용 툴인 FLUENT를 이용하여 적용성을 검토하여 보았다. 수치해석을 통하여 유리화공정 원형설비에 영향을 미치는 인자를 파악하였는데, 저온용응로, 배관냉각기 및 고온필터 등의 세 단계로 나누어 해석을 수행하였다. 저온용융로의 경우 폐기물 처리용량에 따른 해석과 저온용융로 내부 과잉산소 공급 비에 따른 연소지연 가능성에 대한 수치해석을 수행하였다. 배관냉각기의 경우에는 각종 수치 모델 및 외벽 열전달계수를 확보하였으며 또한 방사성 핵종의 거동을 모사할 수 있는 수치적 기업을 검토하였다. 이러한 방법론을 적용하여 핵종의 열교환기 내부에서의 응고 특성에 대하여 고찰하였다. 수평 유입형식의 인입관이 있는 일반적인 형상과 유입구가 필터 내부에 수직으로 있는 고온필터의 수치해석을 통하여 인입관의 위치에 따른 고온필터의 작동 특성을 비교하였다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 추계 학술대회 논문집
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• pp.69-78
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• 2005

유리화공정 고온영역에서의 방사성 배기체 유동해석을 통하여 해석에 적합한 모델을 개발하였다. 개발된 모델을 이용한 수치해석을 통하며 유리화공정 원형설비에 영향을 미치는 인자를 파악하였는데, 저온용융로. 배관냉각기 및 고온필터 등의 세 단계로 나누어 해석을 수행하였다. 저온용융로의 경우 폐기물 처리용량에 따른 해석과 저온용융로 내부 과잉산소 공급 비에 따른 연소지연 가능성에 대한 수치해석을 수행하였다. 배관냉각기의 경우에는 각종 수치 모델 및 외벽 열전달계수를 확보하였으며 또한 방사성 핵종의 거동을 모사할 수 있는 수치적 모델을 개발하였다. 이러한 방법론을 적용하여 핵종의 열교환기 내부에서의 응고 특성에 대하여 고찰하였다. 수평 유입형식의 인입관이 있는 일반적인 형상과 유입구가 필터 내부에 수직으로 있는 고온필터의 수치해석을 통하여 인입관의 위치에 따른 고온필터의 작동 특성을 비교하였다.

• 한국정보통신학회논문지
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• 제17권4호
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• pp.1028-1033
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• 2013

전 세계적으로 해운물류 안전 보안체계가 강화됨에 따라 물류보안 체계구축을 위한 유비쿼터스 기술 기반의 해운물류 안전 보안 핵심기술 개발이 이루어지고 있다. 우리나라의 물류보안제도를 살펴보면 물류보안확보에 필수적인 기술 및 장비 개발이 이루어지지 않아 향후 선진국에 기술 종속 우려가 있다. 앞으로 물류 전 구간을 완벽하게 통합하는 물류보안제도를 도입하는 것이 시급하다. 이에 따라서 안전 보안 체계강화를 위한 감마선 핵종을 검출할 수 있는 휴대용 방사선 검출 장치의 개발 필요성이 높아지고 있다. 본 논문에서는 Cortex-A8을 이용한 휴대용 방사선 검출 장치 플랫폼 구현에 대한 연구를 제안하고자 한다.

• 방사성폐기물학회지
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• 제10권4호
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• pp.263-272
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• 2012

파이로처리 방사성 폐기물 처분 시스템에 대하여 골드심을 이용하여 개발된 확률론적 평가 프로그램을 이용하여 폐쇄후 방사선적 안전성 평가를 수행하였다. 처분장으로부터 핵종이 유출되어 다양한 처분 시스템 내 매질을 이동하는 것에 관련된 정상 시나리오에 대한 평가를 위하여, 평가 결과에 대한 민감도나 일반적으로 불확실성의 범위가 큰 입력자료 중 주요하다고 판단되는 파라미터를 9개로 선정하여 평가에 고려된 핵종 중 Tc, Sn, Pa, Cs 4개의 원소에 대하여 평가 결과를 논의해 보았다. 확률론적 안전성 평가와 함께 이들 각 입력 자료에 대한 최종 방사선 피폭선량에 대한 민감도도 분석하여 결과에 대한 각 입력 파라미터의 중요도도 비교하였다.