• Corrosion Science and Technology
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• 제3권5호
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• pp.198-208
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• 2004

Although there has been no general agreement on the mechanism of primary water stress corrosion cracking (PWSCC) as one of major degradation modes of Ni-base alloys in pressurized water reactors (PWR's), common postulation derived from previous studies is that the damage to the alloy substrate can be related to mass transport characteristics and/or repair properties of overlaid oxide film. Recently, it was shown that the oxide film structure and PWSCC initiation time as well as crack growth rate were systematically varied as a function of dissolved hydrogen concentration in high temperature water, supporting the postulation. In order to understand how the oxide film composition can vary with water chemistry, this study was conducted to characterize oxide films on Alloy 600 by an in-situ Raman spectroscopy. Based on both experimental and thermodynamic prediction results, Ni/NiO thermodynamic equilibrium condition was defined as a function of electrochemical potential and temperature. The results agree well with Attanasio et al.'s data by contact electrical resistance measurements. The anomalously high PWSCC growth rate consistently observed in the vicinity of Ni/NiO equilibrium is then attributed to weak thermodynamic stability of NiO. Redox-induced phase transition between Ni metal and NiO may undermine the integrity of NiO and enhance presumably the percolation of oxidizing environment through the oxide film, especially along grain boundaries. The redox-induced grain boundary oxide degradation mechanism has been postulated and will be tested by using the in-situ Raman facility.

• 방사성폐기물학회지
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• 제2권2호
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• pp.125-133
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• 2004

최대 선출력 61 ㎾/m 및 평균 연소도 1,770 ㎿d/tU의 조건으로 하나로에서 조사한 DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) 핵 연료를 EPMA (Electron Probe Micro Analyzer)를 이용하여 핵분열 생성물을 분석하였다. EPMA의 정확한 분석 방법을 확립하고자, 핵분열생성물 대신 시약을 첨가하여 제조한 모의 DUPIC 핵연료로 EPMA 분석을 수행하였고, 그 결과를 습식 화학 분석의 결과와도 비교하여 평가하였다. 모의 DUPIC 핵연료 중심부의 금속 석출물은 약 1 $\mu\textrm{m}$ 정도의 크기로 관찰되었으며, 이들의 조성은 Mo-53.89 at.%, Ru-37.40 at.% 및 Pd+Rh-8.71 at%이었다. 모의 DUPIC 핵연료 시험에서 정립한 시험방법으로 조사한 DUPIC 핵연료 시편의 금속 석출물 특성을 분석하였다. 핵연료 중앙부에서 관찰된 금속 석출물들의 크기는 2∼2.5 $\mu\textrm{m}$ 정도이었으며, Mo-47.34 at.%, Ru-46 at.%, Pd+Rh-6.65 at.%의 조성임을 확인하였다. 이 실험을 위하여, 특별히 시료의 전도성을 향상시키기 위한 처리를 하였으며, 작은 금속 석출물에 EPMA의 전자빔을 정확히 조사할 수 있는 실험 조건을 제시하였다.

• Nuclear Engineering and Technology
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• 제25권3호
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• pp.361-370
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• 1993

국내 가압 경수로는 핵연료 재장전후 해당 주기 노심핵설계의 타당성 및 안선 제한치의 만족 여부를 확인하기 위하여 저출력에서 노물리 시험을 수행한다. 그러나 고리 3호기 7주기를 포함한 일부 저출력 노물리 시험 중 step 반응도를 삽입한 후에도 반응도가 서서히 증가하는 기이한 현상이 나타났다. 이러한 현상은 시험시 중성자속 준위가 낮고 노외 핵계측기로 비보상형 전리함을 사용하기 때문에 감마 background가 존재하여 생기는 것이다. 이로 인해 노물리 시험 결과는 많은 오차를 포함할 수도 있는 것이다. 본 연구에서는 반응도가 증가하는 현상을 정량적으로 분석하고 기준 제어봉 제어능 측정 시험을 모사함으로써 노물리 시험 결과의 오차를 줄일 수 있는 방법을 제시하고 이후의 노물리 시험에 적용하여 확인하였다. 또한 감마 background 준위를 산정한 후 중성자속 준위를 조정하여 기준 제어봉 제어능 측정 시험을 통해 감마 background의 영향을 받지 않는 중성자속 준위를 결정하였다. 결정된 중성자속 준위는 핵가열이 발생하는 중성자속의 3/10이다. 이것은 기존의 상한치보다 3배 증가된 것이다. 이 결과는 고리 4호기 7주기 및 영광 1호기 7주기 노물리 시험에 성공적으로 적용되었다.

• Environmental Engineering Research
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• 제23권1호
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• pp.28-35
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• 2018

$17{\alpha}-ethinylestradiol$ (EE2), a synthetic estrogen which interfere the endocrine and reproductive function in living organisms, has been found extensively to be deposited into municipal wastewater treatment plants and the environment via human excretion. EE2 has long been known to be efficiently cometabolized by ammonia-oxidizing bacteria (AOB) during ammonia ($NH_3$) oxidation. Current study aims to investigate the effect of culture history on the biotransformation of EE2 by nitrifying sludge which was enriched under different ammonia loading rates in continuous flow reactors. Result showed that past growth condition largely affected not only the metabolic rate of $NH_3$ oxidation but also EE2 cometabolism. Sludge previously acclimated with higher $NH_3$ loads as well as sludge dominated with AOB belong to high growth cluster (Nitrosomonas europaea-Nitrosococcus mobilis) showed higher rate of EE2 biotransformation than those one being acclimated with lower $NH_3$ loads because of its ability to provide more reducing power from $NH_3$ oxidation. Moreover, the correlation between the degradation rates of $NH_3$ and EE2 was higher in sludge being acclimated with higher load of $NH_3$ in comparison with other sludge. Implication of the findings emphasized the role of volumetric $NH_3$ loading rate in determining EE2 removal in wastewater treatment system.

• Corrosion Science and Technology
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• 제12권6호
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• pp.280-287
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• 2013

Flow accelerated corrosion(FAC) of the carbon steel piping in pressurized water reactors(PWRs) has been major issue in nuclear industry. Severe accident at Surry Unit 2 in 1986 initiated the worldwide interest in this area. Major parameters influencing FAC are material composition, microstructure, water chemistry, and hydrodynamics. Qualitative behaviors of FAC have been well understood but quantitative data about FAC have not been published for proprietary reason. In order to minimize the FAC in PWRs, the optimal method is to control water chemistry factors. Chemistry factors influencing FAC such as pH, corrosion potential, and hydrazine contents were reviewed in this paper. FAC rate decreased with pH up to 10 because magnetite solubility decreased with pH. Corrosion potential is generally controlled dissolved oxygen (DO) and hydrazine in secondary water. DO increased corrosion potential. FAC rate decreased with DO by stabilizing magnetite at low DO concentration or by formation of hematite at high DO concentration. Even though hydrazine is generally used to remove DO, hydrazine itself thermally decomposed to ammonia, nitrogen, and hydrogen raising pH. Hydrazine could react with iron and increased FAC rate. Effect of hydrazine on FAC is rather complex and should be careful in FAC analysis. FAC could be managed by adequate combination of pH, corrosion potential, and hydrazine.

• Nuclear Engineering and Technology
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• 제43권1호
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• pp.19-24
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• 2011

These experimental studies are carried out to build a database for analyzing fuel performance in nuclear power plants. In particular, this study focuses on the mechanical and irradiation properties of three kinds of zirconium alloy (Alloy A, Alloy B and Alloy C) irradiated in the HANARO (High-flux Advanced Neutron Application Reactor), one of the leading multipurpose research reactors in the world. Yield strength and ultimate tensile strength were measured to determine the mechanical properties before and after irradiation, while irradiation growth was measured for the irradiation properties. The samples for irradiation testing are classified by texture. For the irradiation condition, all samples were wrapped into the capsule (07M-13N) and irradiated in the HANARO for about 100 days (E > 1.0 MeV, $1.1{\times}10^{21}\;n/cm^2$). These tests and results indicate that the mechanical properties of zirconium alloys are similar whether unirradiated or irradiated. Alloy B has shown the highest yield strength and tensile strength properties compared to other alloys in irradiated condition. Even though each of the zirconium alloys has a different alloying content, this content does not seem to affect the mechanical properties under an unirradiated condition and low fluence. And all the alloys have shown the tendency to increase in yield strength and ultimate tensile strength. Transverse specimens of each of the zirconium alloys have a slightly lower irradiation growth tendency than longitudinal specimens. However, for clear analysis of texture effects, further testing under higher irradiation conditions is needed.

• Nuclear Engineering and Technology
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• 제45권7호
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• pp.847-858
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• 2013

Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea) in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4-5 $g-U/cm^3$ were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr), additional protective coatings (silicide or nitride), and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

• 조명전기설비학회논문지
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• 제15권3호
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• pp.57-66
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• 2001

본 논문에서는 전력계통의 안정도를 향상시키기 위하여 동기 발전기와 정지형 무효전력 보상기예 대한 퍼지-PI 제어기를 설계하기 위한 제어 기법을 설명하였다. 정지형 무효전력 보상기는 고정된 용량의 커패시터와 싸이리스터 제어에 의하여 용량이 가변되는 인덕터가 병렬로 연결된 구조를 가지고 있으며, 시스템 전압을 제어할 뿐만 아니라 동기 발전기의 제동을 개선하기 위해 설계되었다. 본 논문에서 제안한 SVC 계통의 퍼지-PI 제어기의 파라미터는 퍼지 추론 기법에 의해 자동 동조되어진다. 퍼지 추론 기법은 일반적인 기법과는 달리 인간의 경험과 전문가의 지식을 제어 규칙으로 제어 동작을 결정하였다. 그리하여 인간의 추론 과정과 매우 유사한 MMGM을 이용하여 PI 이득의 퍼지 추론 기법을 SVC 계통에 적용하여 설명하였다. 제안된 방법의 강인성을 입증하기 위해 중부하시, 정상부하시 및 경부하시에 초기 전력을 변동시킨 경우에 대하여 시스템의 회전자각, 각속도 편차 특성 및 단자전압의 동특성을 고찰하여 기존의 전력시스템안정화장치보다 응답특성이 우수함을 보였다.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.959-964
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• 2022

Lead is an important material, both for fusion or fission reactors. The cross sections of natural lead should be validated because lead is a main component of lithium-lead modules suggested for fusion power plants and it directly affects the crucial variable, tritium breeding ratio. The presented study discusses a validation of the lead transport libraries by dint of the activation of carefully selected activation samples. The high emission standard ²⁵²Cf neutron source was used as a neutron source for the presented validation experiment. In the irradiation setup, the samples were placed behind 5 and 10 cm of the lead material. Samples were measured using a gamma spectrometry to infer the reaction rate and compared with MCNP6 calculations using ENDF/B-VIII.0 lead cross sections. The experiment used validated IRDFF-II dosimetric reactions to validate lead cross sections, namely ¹⁹⁷Au(n, 2n)¹⁹⁶Au, ⁵⁸Ni(n,p)⁵⁸Co, ⁹³Nb(n, 2n)^92mNb, ¹¹⁵In(n,n')^115mIn, ¹¹⁵In(n,γ)^116mIn, ¹⁹⁷Au(n,γ)¹⁹⁸Au and ⁶³Cu(n,γ)⁶⁴Cu reactions. The threshold reactions agree reasonably with calculations; however, the experimental data suggests a higher thermal neutron flux behind lead bricks. The paper also suggests ²⁵²Cf isotropic source as a valuable tool for validation of some cross-sections important for fusion applications, i.e. reactions on structural materials, e.g. Cu, Pb, etc.

• 한국진공학회:학술대회논문집
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• 한국진공학회 1998년도 제14회 학술발표회 논문개요집
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• pp.77-77
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• 1998

Present silicon dioxide (SiOz) 떠m as intennetal dielectridIMD) layers will result in high parasitic c capacitance and crosstalk interference in 비gh density devices. Low dielectric materials such as f f1uorina뼈 silicon oxide(SiOF) and f1uoropolymer IMD layers have been tried to s이ve this problem. I In the SiOF ftlm, as fluorine concentration increases the dielectric constant of t뼈 film decreases but i it becomes unstable and wa않r absorptivity increases. The dielectric constant above 3.0 is obtain어 i in these ftlms. Fluoropolymers such as polyte$\sigma$따luoroethylene(PTFE) are known as low dielectric c constant (>2.0) materials. However, their $\alpha$)Or thermal stability and low adhesive fa$\pi$e have h hindered 야1리ru뚱 as IMD ma따"ials. 1 The concept of a plasma processing a찌Jaratus with 비gh density plasma at low pressure has r received much attention for deposition because films made in these plasma reactors have many a advantages such as go여 film quality and gap filling profile. High ion flux with low ion energy in m the high density plasma make the low contamination and go어 $\sigma$'Oss피lked ftlm. Especially the h helicon plasma reactor have attractive features for ftlm deposition 야~au똥 of i앙 high density plasma p production compared with other conventional type plasma soun:es. I In this pa야Jr, we present the results on the low dielectric constant fluorocarbonated-SiOF film d밑JOsited on p-Si(loo) 5 inch silicon substrates with 00% of 0dFTES gas mixture and 20% of Ar g gas in a helicon plasma reactor. High density 띠asma is generated in the conventional helicon p plasma soun:e with Nagoya type ill antenna, 5-15 MHz and 1 kW RF power, 700 Gauss of m magnetic field, and 1.5 mTorr of pressure. The electron density and temperature of the 0dFTES d discharge are measUI벼 by Langmuir probe. The relative density of radicals are measured by optic허 e emission spe따'Oscopy(OES). Chemical bonding structure 3I피 atomic concentration 따'C characterized u using fourier transform infrared(FTIR) s야3띠"Oscopy and X -ray photonelectron spl:’따'Oscopy (XPS). D Dielectric constant is measured using a metal insulator semiconductor (MIS;AVO.4 $\mu$ m thick f fIlmlp-SD s$\sigma$ucture. A chemical stoichiome$\sigma$y of 야Ie fluorocarbina$textsc{k}$영-SiOF film 따~si야영 at room temperature, which t the flow rate of Oz and FTES gas is Isccm and 6sccm, res야~tvely, is form려 야Ie SiouFo.36Co.14. A d dielec$\sigma$ic constant of this fIlm is 2.8, but the s$\alpha$'!Cimen at annealed 5OOt: is obtain려 3.24, and the s stepcoverage in the 0.4 $\mu$ m and 0.5 $\mu$ m pattern 킹'C above 92% and 91% without void, res야~tively. res야~tively.