• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.693-698
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• 1995

원자로 조사 핵연료의 제원거동을 조사하기 위하여 고리원자력1호기 핵연료(평균연소도:17,000-38,000MWD/MTU, 농축도: 2.122-3.199 wt.%) 대한 제원을 측정하였다. 핵연료 연소도에 따른 핵연료봉의 길이신장률과 집합체 길이신장률이 각각 0.4-0.6sc, 0.1-0.2%였다. 조사 핵연료의 길이신장과 핵연료 집합체의 휨은 주로 핵연료 연소도에 의존하였으나 핵연료집합체의 비틀림은 핵연료 연소도와 거의 무관하였다.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.135-140
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• 1997

국내개발 액체금속로 KALIMER 노심설계 개발의 일환으로 전기출력 333 MWe(열출력 840 MWth)의 노심설계를 수행하고, 이에 대한 핵ㆍ열수력 특성을 분석하였다. 설계노심은 2농축 U-Zr(14.0/l8.9%) 이원 합금핵연료의 균질노심으로 구성하였다. 핵연료 재장전주기는 18개월, 평균증식비는 0.64로서 평형주기에서의 최대연소도는 125.2 MWD/kg이며, 특히 음의 소듐 void 반응도가를 가짐으로써 노심안전성 확보측면에서 매우 양호함을 보였다.

• Korean Journal of Adult Nursing
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• v.28 no.3
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• pp.266-278
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• 2016

Purpose: The purpose of this study was to identify the effects of reinforced walking exercise on dyspnea-fatigue symptoms, daily activities, walking ability and health related quality of life (HRQoL) in heart failure patients. Methods: This study used a randomized controlled trial design. The participants (experimental group=16, control group=25) were recruited from a university hospital in Kyeong-nam area. Data were collected from March to September, 2015. The reinforced walking exercise included goal setting and feedback (telephone and text message) provided for 12 weeks. Dyspnea-Fatigue Index, Korean Activity Scale/Index (KASI), six-minute walking distance (6MWD) and HRQoL were measured. Data were analyzed using descriptive statistics, t-test, Fisher's exact test, $x^2$ test, and Kolmogrove-Smirnov test. Results: Prior to the intervention there were no differences in the research variables between two groups. The exercise compliance in the experimental group was 100% (walking for 50 minutes per day, 5 times per week). The experimental group had improved dyspnea-fatigue symptoms (t=8.63, p<.001), daily activities (t=-4.92, p<.001), longer 6MWD (t=-5.66, p<.001), and increased HRQoL (t=-9.05, p<.001) compared to the control group. Conclusion: The reinforced walking exercise could be a cost-effective intervention in heart failure patient, which could enhance patients' outcomes, such as improving dyspnea-fatigue symptoms, daily activities, walking ability, and quality of life.

• Journal of Radiation Protection and Research
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• v.37 no.4
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• pp.197-201
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• 2012

Annual dose at the boundary of the interim storage facility at normal condition was calculated to estimate the site area of the facility of PWR spent nuclear fuel. In this work, source term was generated by ORIGEN-ARP for 4.5 wt% initial enrichment, 45,000 MWd/MTU burnup and 10 years cooling time. Modeling of the storage facilities and radiation shielding evaluations were conducted by MCNP code depending on the storage capacity. In the case of the centralized storage system, the required site area was found to have the radius of more than 700 m.

• Analytical Science and Technology
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• v.16 no.2
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• pp.110-116
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• 2003

A study has been done on the separation of iodide from spent pressurized water reactor (PWR) fuels and its quantitative determination using ion chromatography. Spent PWR fuels were dissolved with mixed acid of nitric and hydrochloric acids (80 : 20 molL%) which can oxidize iodide to iodate to prevent it from be vaporized. After reducing ${IO_3}^-$ ­to $I_2$ in 2.5 M $HNO_3$ with $NH_2OH{\cdot}HCl$, Iodine was selectively separated from actinides and all other fission products with carbontetrachloride and back-extracted with 0.1 M $NaHSO_3$. Recovered iodide was determined using the ion chromatograph of which the column was installed in a glove box for the analysis of radioactive materials. In practice, spent PWR fuel with 42,000~44,000 MWd/MtU was analyzed and its quantity was compared to that calculated by burnup code, ORIGEN2. The agreement was achieved with a deviation of -8.3~-0.5% from the ORIGEN 2 data, $324.5{\sim}343.6{\mu}g/g$.

• The Journal of Korean Medicine
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• v.41 no.3
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• pp.162-172
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• 2020

Objectives: The purpose of study was to report the clinical improvement of Chronic Obstructive Pulmonary Disease (COPD) patients treated with Korean medicine pulmonary rehabilitation. Methods: The patients were treated with Lung-conduction exercise, Chuna manual therapy, Exercise therapy. To assess the treatment outcomes, we used the pulmonary function test (PFT), modified medical research council scale (mMRC), 6-minute walk distance (6MWD), peak expiratory flow rate (PEFR), COPD assessment test (CAT), St. George respiratory questionnaire (SGRQ). Results: After treatments, the patient's clinical symptoms were improved with CAT, SGRQ's significant decrease and PFT, mMRC, 6MWD and PEFR were maintained or improved slightly. Conclusions: The Korean medicine pulmonary rehabilitation was effective in the treatment of COPD patients. This study suggested the possibility of Korean Medicine pulmonary rehabilitation program in the clinic.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.3
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• pp.233-244
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• 2008

A conceptual design study for a pyroprocesing facility, has been carried out in this study, which is available for the recovery of uranium and transuranic elemental group(TRU), that is, reusable as a nuclear fuel especially in a next generation-type reactor. The scale of this facility has been chosen as 20 kg HM/batch, comparatively small engineering size in order to collect scale-up data for the design of a commercial facility as well as to get operational experience. The spent fuel to be handled in this process is as follows : 3.5 % enriched uranium fuel, 35,000MWd/tU and 5-year cooled. The major items considered in the conceptual study are a building lay-out including various hot cells, safety management of the process operation in conjunction with material balance control, radiation safety, inert atmosphere control in shielded hot cells, and criticality control of uranium and TRU products.

• Macromolecular Research
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• v.8 no.5
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• pp.238-242
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• 2000

The new trisiloxane-bridged heterometallic dinuclear metallocenes, hexamethyltrisiloxanediyl(cyclopentadienyltitanium trichloride) (cyclopentadienylindenyl zirconium dichloride) , $C_3ITi-Cp(CH_3)_2Si-O-Si(CH_3)_2-O-Si(CH_3)_2-Cp-ZrIndCI_2$ (1) and hexamethyltrisiloxanediyl (cyclopentadienylindenylhafnium dichloride) (cyclopentadienylindenyl zirconium dichloride), $C_2IndHf-Cp(CH_3)_2Si-O-Si(CH_3)_2-Cp-ZrIndCl_2$ 2) connecting two dissimilar metallocenes were synthesized and used for ethylene polymerization in the presence of modified methylaluminoxane (MMAO) cocatalyst. The catalytic activity of heterometallic dinuclear metallocenes, 1 and 2 was lower than that of corresponding mononuclear metal-locene as well as two physically mixed catalysts, $CpTiCl_2/Cp_2ZrCl_2 and Cp_2HfCl_2/Cp_2ZrCl_2$. On the tither hand, MWD of PE obtained with 1 and 2 was remarkably broader ($M_w/M_n$) became up to 9.4) than those of PEs prepared with the corresponding mononuclear metallocenes and mixed catalysts. With analysis by GPC and CFC, it was found that PE produced by the heterometallic dinuclear metallocenes exhibited the definite bimodal GPC curves that should cause the broadening of MWD.

• Journal of Radiation Protection and Research
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• v.38 no.4
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• pp.189-193
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• 2013

Annual dose on the containment building wall of the interim storage facility at normal condition was calculated to estimate the dose rate transition of the facility of PWR spent nuclear fuel. In this study, source term was generated by ORIGEN-ARP with 4.5 wt% initial enrichment, 45,000 MWd/MTU burnup and 10 years cooling time. Modeling of the storage facility and the containment building and radiation shielding evaluations were conducted by MCNP code depending on the distance between the wall and the facility in the building. In the case of the centralized storage system, the distance required for the annual dose rate limit from 10CFR72 was estimated to be 50 m.

• Journal of Radiation Protection and Research
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• v.18 no.2
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• pp.27-35
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• 1993

A shielding analysis of the shipping cask for transporting the CANDU spent fuel bundles has been studied. Radiation source term has been calculated on spent fuel with burn-up of 7,800 MWD/MTU and 5 years cooling time by ORIGEN2 code. The shielding calculation for the cask capable of transporting 378 bundles of CANDU spent fuel has been made by use of 1-D ANISN and 2-D DOT 4.2 codes. As a result of analysis, the optimum shield thickness of cask was obtained. And it is proved that the safety in radiation shielding under normal transport and hypothetical accident conditions is confirmed to satisfy the allowable values specified in IAEA Safety Series No. 6 and the Korean Atomic Law.