• 상하수도학회지
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• 제19권4호
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• pp.480-486
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• 2005

The main goals of this work are to investigate the effect of molecular weight distribution(MWD) and activated carbon adsorption capacity after conventional coagulation and enhanced coagulation. The ozonation was very effective to decompose the NOM to smaller size and to remove molecular smaller than 1,000. The concentration of DOC was reduced 0.25mg/L and 0.56mg/L by the conventional coagulation and the enhanced coagulation, respectively The conventional coagulation was not effective to remove NOM. However, the enhanced coagulation was effective to remove MW bigger than 10,000. The higher MW was shifted to smaller weight by ozonation in the raw water and the after conventional coagulation. After enhanced coagulation the MW had not changed significantly by ozonation. Also, it was observed that the ozone dosage did not have significant impact on MW shifting to smaller size. The adsorption capacity simulated by IAST comparing K values showed that the adsorption capacity was not impacted by ozone doses. There was very strong correlation between MW smaller than 10,000 and the mid- and strongly adsorbable fractions.

• Nuclear Engineering and Technology
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• 제41권2호
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• pp.187-198
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• 2009

In this paper, three power ramp tests performed on high burn-up Re-crystallized Zircaloy2 - UO2 BWR fuel rods (56 to 63 MWd/kgU) within the SCIP project are simulated with METEOR and ALCYONE 3D. Two of the ramp tests are of staircase type up to Linear Heat Rates of 420 and 520 W/cm and with long holding periods. Failure of the 420 W/cm fuel rod was observed after 40 minutes. The third ramp test consisted of a more standard ramp test with a constant power rate of 80 W/cm/min up to 410 W/cm with a short holding time. The tests were first simulated with the METEOR 1D fuel rod code, which gave accurate results in terms of profilometry and fission gas releases. The behaviour of a fuel pellet fragment and of the cladding piece on top of it was then investigated with ALCYONE 3D. The size and the main characteristics of the ridges after base irradiation and power ramp testing were recovered. Finally, the failure criteria validated for PWR conditions and fuel rods with low-to-medium burn-ups were used to analyze the failure probability of the KKL rodlets during ramp testing.

• Nuclear Engineering and Technology
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• 제41권2호
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• pp.163-170
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• 2009

In order to evaluate high-temperature oxidation behavior of the advanced alloy cladding under LOCA conditions, isothermal oxidation tests in steam were performed with cladding specimens prepared from high burnup PWR fuel rods that were irradiated up to 79 MWd/kg. Cladding materials were $M5^{(R)}$ and $ZIRLO^{TM}$, which are Nb-containing alloys. Ring-shaped specimens were isothermally oxidized in flowing steam at temperatures from 1173 to 1473 K for the duration between 120 and 4000s. Oxidation rates were evaluated from measured oxide layer thickness and weight gain. A protective effect of the preformed corrosion layer is seen for the shorter time range at the lower temperatures. The influence of pre-hydriding is not significant for the examined range. Alloy composition change generally has small influence on oxidation in the examined temperature range, though $M5^{(R)}$ shows an obviously smaller oxidation constant at 1273 K. Consequently, the oxidation rates of the high burnup $M5^{(R)}$ and $ZIRLO^{TM}$ cladding are comparable or lower than that of unirradiated Zircaloy-4 cladding.

• 폴리머
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• 제39권1호
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• pp.165-168
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• 2015

하이드록시 그룹을 지닌 온도민감형 고분자들이 원자전이라디칼중합법(ATRP)과 클릭반응(click reaction)에 의해 합성되어졌다. 고분자들의 분자량과 분자량 분포도는 gel permeation chromatography(GPC)에 의하여 얻어졌고, 고분자들의 분자량은 잘 제어되었으며 분자량 분포도도 낮게 유지되었다. 클릭반응의 효율은 $^1H$ NMR spectroscopy에 의해 얻어졌으며, 높은 효율을 나타내었다. 고분자 사슬 곁가지의 아민 그룹의 종류와, 치환된 알코올 그룹의 종류에 따라 저임계 용액 온도(LCST)의 제어가 가능했다.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 1998년도 춘계 학술발표회 논문집
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• pp.221-228
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• 1998

2차원 노심핵설계 코드 HELIOS를 이용하여 $^{7}$ LiF-BeF$_2$-ThF$_4$-$^{233}$ UF$_4$ 용융염 핵연료와 흑연(Graphite) 감속재로 구성된 AMBIDEXTER(Advanced Molten-salt Break-even Inherently-safe Dual-mission EXperimental and TEst Reactor) 원자로의 육각주형 로심격자에 대해 핵적 자활성 요건의 설계해석을 수행하였다. AMBIDEXTER 원자로는 액체 핵연료의 유동성을 이용한 온라인 핵연료 정화ㆍ처리ㆍ재생의 연속공정을 도입하여 노내의 잔류 핵분열 생성물질의 포화양을 최소로 유지시키고 중성자 경제성을 극대화하므로 높은 전환율을 얻는 설계이다. 핵연료 내에 잔류하는 핵분열생성물질의 포화농도에 대응하는 연소도를 등가연소도로 정의할 때, 열출력 250MW$_{th}$ AMBIDEXTER 원자로의 등가연소도 374MWD/TeH.E.의 평형 로심 모델에 대해 핵적 자활성을 지배하는 주요 핵설계 인자로서 용융염 핵연료의 $^{233}$ U Mole 분율, 흑연-대-용융염의 체적비, 노심격자 간격 및 출력 밀도의 변화에 따른 임계도 및 전환율을 평가하였다. 그 결과, $^{233}$ U Mole 분율과 혹연-대-용융염 체적비를 좌표축으로 하는 2차원상공간에서 핵적 자활성 요건 상태함수는 각 노심격자간격에 대해 완만한 선형 함수로 표현할 수 있음을 확인하였다.

• Nuclear Engineering and Technology
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• 제28권4호
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• pp.366-372
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• 1996

The simulated spent PWR fuel pellet which is corresponding to the turnup of 33,000 MWD/MTU is prepared by adding 11 fission-product elements to UO$_2$. The simulated spent fuel pellet is treated at 40$0^{\circ}C$ in air (oxidation), at 110$0^{\circ}C$ in air (high-temperature treatment), and at $600^{\circ}C$ in hydrogen (reduction). The product is treated through additional addition and reduction up to 3 cycles. Pellets are completely pulverized by the first oxidation, and the high-temperature treatment causes particle and crystallite to grow and surface to be smooth, and thus particle size significantly increases and surface area decreases. The reduction following the high-temperature treatment decreases much the particle size by means of the formation of intercrystalline cracks. The particle size decreases a little during the second oxidation and reduction cycle and then remains nearly constant during the third and fourth cycles. Surface area of pounder increases progressively with the repetition of oxidation and reduction cycles, mainly due to the formation of Surface cracks. The degradation of surface area resulting from high-temperature treatment is restored by too subsequent resulting oxidation and reduction cycles.

• 농촌계획
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• 제11권4호
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• pp.93-98
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• 2005

A aggregate size distribution of soils is an important in successful crop production in reclaimed tidelands. The aggregate size distribution for this study were determined of 0.1mm, 0.25mm, 0.5mm, 1.0mm, and 2.0mm by wet sieving method. Agricultural activity, period of reclamation showed significant effects on aggregate size distribution in reclaimed tidelands. Aggregate MWD was greater in SS and KH(above 1.0m) than in YSG, GHD, SMG, and DH(below 0.5mm) reclaimed tidelands and tidelands. The percentage of aggregates less than < 2mm for SMG, GHD, and SM reclaimed tidelands and tidelands were ranged 8.9%, 36.7%, and 38.0%, respectively. The percentage of > 0.1mm aggregates for SMG, GHD, and SM reclaimed tidelands were ranged 9.0%, 26.0%, and 48.9%, respectively. Results indicated that aggregate size distribution of reclaimed tidelands and tidelands under various agricultural systems vary with reclamation period and soil type.

• 한국터널지하공간학회 논문집
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• 제17권5호
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• pp.563-573
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• 2015

For hard rock subsea tunnels the most challenging rock mass conditions are in most cases represented by major faults/weakness zones. Poor stability weakness zones with large water inflow can be particularly problematic. At the pre-construction investigation stage, geological and engineering geological mapping, refraction seismic investigation and core drilling are the most important methods for identifying potentially adverse rock mass conditions. During excavation, continuous engineering geological mapping and probe drilling ahead of the face are carried out, and for the most recent Norwegian subsea tunnel projects, MWD (Measurement While Drilling) has also been used. During excavation, grouting ahead of the tunnel face is carried out whenever required according to the results from probe drilling. Sealing of water inflow by pre-grouting is particularly important before tunnelling into a section of poor rock mass quality. When excavating through weakness zones, a special methodology is normally applied, including spiling bolts, short blast round lengths and installation of reinforced sprayed concrete arches close to the face. The basic aspects of investigation, support and tunnelling for major weakness zones are discussed in this paper and illustrated by cases representing two very challenging projects which were recently completed (Atlantic Ocean tunnel and T-connection), one which is under construction (Ryfast) and one which is planned to be built in the near future (Rogfast).

• 대한간호학회지
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• 제37권3호
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• pp.343-352
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• 2007

Purpose: The purpose of this study was to examine the effects of an inpatient pulmonary rehabilitation program on dyspnea, exercise capacity, and health related quality of life in inpatients with chronic lung disease. Method: This quasi experimental study was designed with a nonequivalent control group pre-post test time series. Twenty three patients were assigned to the experimental group and nineteen to the control group. The inpatient pulmonary rehabilitation program was composed of upper and lower extremity exercise, breathing retraining, inspiratory muscle training, education, relaxation and telephone contacts. This program consisted of 4 sessions with inpatients and 4 weeks at home after discharge. The control group was given a home based pulmonary rehabilitation program at the time of discharge. The outcomes were measured by the Borg score, 6MWD and the Chronic Respiratory Disease Questionnaire(CRQ). Results: There was a statistically significant difference in dyspnea between the experimental group and control group, but not among time sequence, or interaction between groups and time sequence. Also significant improvements in exercise capacity and health related quality of life were found only in the experimental group. Conclusions: An Inpatient pulmonary rehabilitation program may be a useful intervention to reduce dyspnea, and increase exercise capacity and health related quality of life for chronic lung disease patients.

• Journal of Radiation Protection and Research
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• 제36권4호
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• pp.174-182
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• 2011

With the reactor operation conditions - 4.3 wt% $^{235}U$ initial enrichment, burn-up 55,000 MWd/MTU, average power 34 MW/MTU for three periods burned time for 539.2 days per period and cooling time for 100 hours after shut down, to set up the condition to determine the minimum height (depth) of spent fuel storage pool to shut off the radiation out of the spent fuel storage pool and to store spent fuels safely, the dose rate on the specific position directed to the surface of spent fuel storage pool induced by the neutron and gamma-ray from spent fuels are evaluated. The length of spent fuel is 381 cm, and as the result of evaluation on each position from the top of spent fuel to the surface of spent fuel storage pool, it is difficult for neutrons from spent fuels to pass through the water layer of maximum 219 cm (600 cm from the floor of spent fuel storage pool) and 419 cm (800 cm from the floor of spent fuel storage pool) for gamma-ray. Therefore, neutron and gamma-ray from spent fuels can pass through below 419 cm (800 cm from the floor) water layer directed to the surface of spent fuel storage pool.