• 한국해양환경ㆍ에너지학회지
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• 제8권4호
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• pp.179-185
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• 2005

다환방향족 탄화수소는 주요한 독성환경오염물질풍의 하나이며 해양퇴적물에 축적이 된다. 다환방향족 탄화수소가 가지고 있는 위해성으로 인해 해양퇴적물내의 다환방향족 탄화수소 제거는 많은 관심사가 되어오고 있다. 본 연구에서는 다환방향족 탄화수소로 오염된 해양퇴적물의 생물정화를 위해 bioslurry reactor를 사용하기 위한 전단계로 ex situ bioremediation을 위한 최적 조건을 찾기 위해 멸균 유무의 퇴적물에서 균주접종과 혼합 cyclodextrin (M-CD)공급의 효과를 알아보았다. 비멸균상태의 퇴적물에서 분해균주의 접종여부에 상관없이 M-CD가 첨가된 실험구의 전자전달계활성이 증가되었고 멸균상태의 퇴적물에서는 분해균주가 접종되고 M-CD가 첨가된 실험구의 전자전달계활성이 증가되었다. 멸균상태의 퇴적물에서 M-CD와 함께 단일 균주가 접종되었을 때 퇴적물내의 다환방향족 탄화수소의 분해율이 9-20%로 나타났으며 동일한 조건에서 혼합균주가 접종되었을 때는 24-37%로 나타났다. 비멸균상태의 퇴적물에서 동일한 조건으로 실험하였을 때는 유의한 분해가 일어나지 않았으며 군집분석 결과 접종 균주도 감소하였다. 또한 M-CD 첨가 없이는 2일후 접종균주를 찾아볼 수 없었다. 결론적으로 퇴적물내의 다환방향족 탄화수소 제거를 위해 퇴적물내의 토착미생물과 포식자가 제거된 상태에서 적절한 탄소원과 함께 균주를 접종하는 것이 효과적인 것으로 나타났다.

• 한국산학기술학회논문지
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• 제11권11호
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• pp.4656-4663
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• 2010

본 연구에서는 복합 후처리장치의 최적 모델 개발을 위해 반응기 내부 유동 특성 및 유량 분배 수준, 공기 연령, 체류시간 등과 관련지어 3차원 전산유체역학(CFD)을 수행하였다. 백필터의 각 격실별로 유량 분배가 크게 차이가 발생해 반건식 반응기(SDR)에 편류현상이 발생할 것으로 예측되어 백필터에 균등한 유량 분배를 위한 구조 개선이 시급한 것으로 나타났다. 또한 SDR의 유입구 구조 변경을 통해 반응기 내부의 속도장과 체류시간 분포 특성을 개선할 수 있는 것으로 나타났다. 본 해석을 이용하여 수정 보완한 배가스 복합 후처리장치는, 반건식 반응기와 백필터 처리부가 일체화되어 있어 장치가 콤팩트하고 설치면적이 적으며 운영 및 관리가 매우 편리할 것으로 판단된다.

• 청정기술
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• 제16권2호
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• pp.124-131
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• 2010

슬러리 기포탑을 이용하여 수산화칼슘 농도(0.16~0.64 wt%), 계면활성제 농도(2~16 wt%), 총 부피유량(3~6 L/min) 및 $CO_2$ 유량의 부피분율(0.3~0.6)이 탄산칼슘의 morphology, 결정구조, 입자의 크기, 입자 간의 응집화, 비표면적에 미치는 영향을 알아보았다. 실험에 사용한 반응기는 높이가 1.0 m이고 직경이 0.11m, 그리고 중앙에는 직경 4 cm인 튜브가 들어있는 슬러리 기포탑이다. 실험에 사용한 음이온 계면활성제 Dispex N40은 탄산칼슘 합성에 있어 반응 속도에 영향을 주어 반응 종결 시간을 감소시켰다. Dispex N40의 농도가 2 wt%일 때 수산화칼슘의 포화농도인 0.16 wt%에서 이산화탄소의 유량에 따른 침강성 탄산칼슘의 morphology를 살펴보면 반응 속도가 증가할수록 결정의 형태는 단일 결정으로 존재하는 입자들이 많아졌다. Dispex N40은 탄산칼슘의 crystal의 성장과 입자와 입자간의 응집 현상에 영향을 주어 탄산칼슘의 평균 입도를 변화시켰다. 또한 $0.9 L/min\;CO_2$ 유량에서 수산화칼슘의 농도가 0.64 wt%일 때 2 wt%의 계면활성제 첨가로 인해 비표면적을 $35m^2/g$에서 $44m^2/g$로 크게 증가시켰다.

• 비파괴검사학회지
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• 제34권3호
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• pp.254-259
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• 2014

중수로 원자로는 한 개의 원자로용기로 구성된 경수로와는 달리 약 380여개의 연료채널(fuel channel)로 구성되어 있다. 연료채널을 구성하는 압력유지 기기인 압력관(pressure tube)은 지르코니움 합금(Zr-2.5wt% Nb) 재질로서 치수는 내경이 103.4 mm, 두께가 약 4.19 mm, 길이가 6.36 m인 튜브 형태의 관이다. 압력관은 내부에 핵연료 다발과 냉각재가 내장되며 압력관의 기능은 연료를 지지하고 열수송 유체인 중수($D_2O$)를 이송한다. 압력관의 단순한 기하학적인 형상으로 인하여 자동화 비파괴검사가 가능하고 접근성이 우수하다. 연료채널은 경수로형 원전과 동일하게 설치전과 운전중에 원자력안전위원회 법령 요건에 따라 주기적으로 엄격한 비파괴검사를 수행하여 건전성을 확인한다. 연료채널의 주기적 비파괴검사에는 초음파탐상 및 와전류탐상검사 기법을 적용한 체적 비파괴검사 기술이 적용된다. 이중에서 와전류탐상검사 기법은 초음파탐상검사에서 검출된 결함의 확인을 위한 보충검사기술로 적용되고 있지만 표면결함에 대한 검출능이 초음파탐상검사 기법보다 우수한 장점을 가지고 있다. 본 논문에서는 압력관 내부 표면 비파괴검사에 적용되고 있는 와전류탐상검사 기술의 압력관 내면에 발생할 수 있는 결함의 검출 및 깊이 측정 특성에 대한 연구결과를 기술하였다. 즉, 와전류검사 기술은 압력관 내면에 발생할 수 있는 아주 미세한 결함을 매우 우수한 분해능으로 검출할 수 있으므로 초음파탐상검사 결과 확인을 위한 보충기술로서 매우 유용하지만, 결함의 깊이 측정은 오차가 매우 크게 발생하므로 결함 깊이 측정에는 적합하지 않고 오직 표면결함 검출에만 적용하는 것이 바람직하다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.513-518
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• 1996

This report presents a demonstration of application of realistic evaluation methodology to a posturated cold leg large break LOCA in a Westinghouse three-loop pressurized water reactor with 17$\times$17 fuel. The new method of this analysis can be divided into three distinct step: 1) Best Estimate Code Validation and Uncertainty Quantification 2) Realistic LOCA Calculation 3) Limiting Value LOCA Calculation and Uncertainty Combination RELAP5/MOD3/K ［1］, which was improved from RELAP5/MOD3.1, and CONTEMPT4/MOD5 code were used as a best estimate thermal-hydraulic model for realistic LOCA calculation. The code uncertainties which will be determined in step 1) were quantified already in previous study ［2］, and thus the step 2) and 3) for plant application were presented in this paper. The application uncertainty parameters are divided into two categories, i.e. plant system parameters and fuel statistical parameters. Single parameter sensitivity calculations were performed to select system parameters which would be set at their limiting value in Limiting Value Approach (LVA) calculation. Single run of LVA calculation generated 27 PCT data according to the various combinations of fuel parameters and these data provided input to response surface generation. The probability distribution function was generated from Monte Carlo sampling of a response surface and the upper 95$^{th}$ percentile PCT was determined. Break spectrum analysis was also made to determine the critical break size. The results show that sufficient LOCA margin can be obtained for the demonstration NPP.

• 소음진동
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• 제5권4호
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• pp.515-524
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• 1995

Tubes in heat exchanger of fuel rods in reactor core are supported at intemediate point by support p0lates or springs. Current practice is, in case of heat exchanger, to allow clearance between tube and support plate for design and manufacturing consideration. And in case of fuel rod the clearance in support point can be generated due to the support spring force relaxation. Flow-induced vibration of a tube can cause it to impact or rub against support plate or against adjacent tubes and can result in fretting-wear. The tube-to- support dynamic interaction is used to relate experimental wear data from single-span test rigs to real multi-span heat exchanger configurations. The dynamic interaction cna be measured during experimental wear tests. However, the dynamic interaction is difficult to measure in real heat exchangers and, therefore, analytical techniques are required to estimate this interaction. This paper describels the nonlinear impact model of DAGS(Dynamic Analysis of Gapped Structure) code which simulates the tube response to external sinusodial or step excitation and predicts tube motion and tube-to-support dynamic interaction. Three experimental measurements-two single span rods excited by sinusodial force and a two span rod impacted by a steel ball are compared from the simulation nonlinear model of DAGS code. The simulation results from DAGS code are in good agreement with measurements. Therefore, the developed model of DAGS code is good analytical tool for estimating tube-to-support dynamic interaction in real heat exchangers.

• Journal of Microbiology and Biotechnology
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• 제14권6호
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• pp.1120-1128
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• 2004

The SCBFC was composed of bilayered cathode, the outside of which was modified with $Fe^{3+}$ (graphite-Fe(III) cathode) and the inside of which was porcelain membrane, and of an anode which was modified with $Mn^{4+}$ (graphite­Mn(lV) anode). The graphite-Fe(III), graphite-Mn(IV), and porcelain membrane were designed to have micropores. The outside of the cathode was exposed to the atmosphere and the inside was contacted with porcelain membrane. In all SCBFCS the graphite-Fe(III) was used as a cathode, and graphite-Mn(IV) and normal graphite were used as anodes, for comparison of the function between normal graphite and graphite-Mn(IV) anode. The potential difference between graphite-Mn(IV) anode and graphite-Fe(III) cathode was about 0.3 volt, which is the source for the electron driving force from anode to cathode. In chemical fuel cells composed of the graphite-Mn(IV) anode and graphite-Fe(III) cathode, a current of maximal 13 mA was produced coupled to oxidation of NADH to $NAD^{+}$ the current was not produced in SCBFC with normal graphite anode. When growing and resting cells of E. coli were applied to the SCBFC with graphite-Mn(IV) anode, the electricity production and substrate consumption were 6 to 7 times higher than in the SCBFC with normal graphite anode, and when we applied anaerobic sewage sludge to SCBFC with graphite-Mn(IV) anode, the electricity production and substrate consumption were 3 to 5 times higher than in the SCBFC with normal graphite anode. These results suggest that useful electric energy might possibly be produced from SCBFC without electron mediators, electrode-active bacteria, and extra energy consumption for the aeration of catholyte, but with wastewater as a fuel.

• 한국압력기기공학회 논문집
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• 제11권1호
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• pp.20-24
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• 2015

Steam generator(S/G) tubes in pressurized water reactor (PWR's) are subject to several types of degradation. This degradation includes denting, pitting, intergranular attack(IGA), intergranular stress corrosion cracking(IGSCC), fatigue, fretting and wear. Degradation can be derived from either the primary side(inside) or the secondary side(outside) of the tube. Recent issue for tube degradation in domestic steam generator is the tube end cracking on seal weld region. The seal weld region at the tube end and tube itself is regarded as a pressure boundary between the primary side and the secondary side. One of the Westinghouse Model-F S/G has experienced tube end cracking and its number of plugging approximately becomes to the operating limit up to 5% due to tube end cracking which was reported as SAI/MAI(single/multiple axial indication) or SCI/MCI(Single/multiple circumferential indication) from the results of eddy current testing. Eddy current mock-up test was carried out to determine the origin of cracking whether it is from weld zone area or parent tube. This result was helpful to analyze crack location on ECT data. Correct action on this problem was the installation of tube-end sleeve. Last year, after removing 340 installed plugs from tubes, selected 269 tubes took tube-end sleeve installation. Tube-end sleeve brought pressure boundary from parent tube to installed sleeve tube. Tube-end sleeve has the benefit of reducing outage period and increasing more revenue than replacing S/G. This paper is provided to assist interest parties in effectively understanding this issue.

• Journal of Radiation Protection and Research
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• 제41권4호
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• pp.384-388
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• 2016

Background: For a verification of newly-developed neutron absorbers, one of guidelines on the qualification and acceptance of neutron absorbers is the neutron attenuation test. However, this approach can cause a problem for the qualifications that it cannot distinguish how the neutron attenuates from materials. Materials and Methods: In this study, an estimation method of neutron absorption performances for materials is proposed to detect both direct penetration and back-scattering neutrons. For the verification of the proposed method, MCNP simulations with the experimental system designed in this study were pursued using the polyethylene, iron, normal glass and the vitrified form. Results and Discussion: The results show that it can easily test neutron absorption ability using single absorber model. Also, from simulation results of single absorber and double absorbers model, it is verified that the proposed method can evaluate not only the direct thermal neutrons passing through materials, but also the scattered neutrons reflected to the materials. Therefore, the neutron absorption performances can be accurately estimated using the proposed method comparing with the conventional neutron attenuation test. Conclusion: It is expected that the proposed method can contribute to increase the reliability of the performance of neutron absorbers.

• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3604-3613
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• 2022

The Physical Protection System (PPS) plays an important role and must effectively deal with various adversary attacks in nuclear security. In specific single adversary path scenarios, we can calculate the PPS effectiveness by EASI (Estimated Adversary Sequence Interruption) through Probability of Interruption (P_I) calculation. EASI uses a single value of the probability of detection (P_D) and the probability of alarm communications (P_C) in the PPS. In this study, we develop a multi-path analysis code based on EASI to evaluate the effectiveness of PPS. Our quantification method for P_I considers the variability and uncertainty of P_D and P_C value by Monte Carlo simulation. We converted the 2-D scheme of the nuclear facility into an Adversary Sequence Diagram (ASD). We used ASD to find the adversary path with the lowest probability of interruption as the most vulnerable paths (MVP). We examined a hypothetical facility (Hypothetical National Nuclear Research Facility - HNNRF) to confirm our code compared with EASI. The results show that implementing the variability extension can estimate the P_I value and its associated uncertainty. The multi-path analysis code allows the analyst to make it easier to assess PPS with more extensive facilities with more complex adversary paths. However, the variability of the P_D value in each protection element allows a significant decrease in the PI value. The possibility of this decrease needs to be an important concern for PPS designers to determine the P_D value correctly or set a higher standard for PPS performance that remains reliable.