• 대한화학회지
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• 제54권4호
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• pp.419-428
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• 2010

마이크로웨이브 반응 장치 (Synthos 3000 Aton Paar, GmbH, 1400 W maximum magnetron)를 이용하여, diisopropylcarbodiimide (DIC)/와 1-hydroxybenzotriazoles (HOXt) (X = A or B)를 반응시켜서 amino acid hydrazide를 효율적으로 합성할 수 있는 반응 조건을 개발하였다. 일반적인 가열반응과 마이크로웨이브 반응을 반응 시간, 반응 조건 등을 비교하였을 때에, 마이크로웨이브 반응이 보다 효율적으로 진행되었으며, diisopropylcarbodiimide (DIC)와 1-hydroxybenzotriazole (HOBt) 반응에서보다는 diisopropylcarbodiimide (DIC)와 7-aza-1-hydroxybenzotriazole (HOAt)를 반응시켰을 때에 좋은 수율 (95 - 98%)로 얻어졌다.

• Nuclear Engineering and Technology
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• 제30권4호
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• pp.342-352
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• 1998

A nuclear design feasibility of soluble boron free(SBF core for the medium-sized(600MWe) PWR was investigated. The result conformed that soluble boron free operation could be performed by using current PWR proven technologies. Westinghouse advanced reactor, AP-600 was chosen as a design prototype. Design modification was applied for the assembly design with burnable poison and control rod absorber material. In order to control excess reactivity, large amount of gadolinia integral burnable poison rods were used and B4C was used as a control rod absorber material. For control of bottom shift axial power shape due to high temperature feedback in SBF core, axial zoning of burnable poison was applied to the fuel assemblies design. The combination of enrichment and rod number zoning for burnable poison could make an excess reactivity swing flat within around 1% and these also led effective control on axial power offset and peak pin power, The safety assessment of the designed core was peformed by the calculation of MTC, FTC and shutdown margin. MTC in designed SBF core was greater around 6 times than one of Ulchin unit 3&4. Utilization of enriched BIO(up to 50w1o) in B4C shutdown control rods provided enough shutdown margin as well as subcriticality at cold refueling condition.

• 한국해양공학회지
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• 제15권2호
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• pp.88-93
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• 2001

In this paper, joints of Cu-1Cr-0.1Zr alloy to STS316L were performed by friction welding method. Particularly, Cu-1Cr-0.1Zr alloy is attractive candidate as nuclear power plant material and exibit the best combination of high strength and good electrical and thermal conductivity of any copper alloy examined. The stainless steel is a structural material while copper alloy acts as a heat sink material for the surface heat flux in the first wall. So, in this paper, not only the development of optimizing of friction welding with more reliability and more applicability but also the development of in-process real-time weld quality (such as strength and toughness) evaluation technique by acoustic emission for friction welding of such nuclear reactor component of Cu-1Cr-0.1Zr alloy to STS316L steel sere performed.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.419-433
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• 2016

One of the important severe accident management measures in the Light Water Reactors is water injection to the reactor core. The related phenomena are investigated by performing experiments and computer simulations. One of the most widely known is the QUENCH test-program. A number of analyses on QUENCH tests have also been performed by different computer codes for code validation and improvements. Unfortunately, any deterministic computer simulation is not free from the uncertainties. To receive the realistic calculation results, the best estimate computer codes should be used for the calculation with combination of uncertainty and sensitivity analysis of calculation results. In this article, the QUENCH-03 and QUENCH-06 experiments are modelled using ASTEC and RELAP/SCDAPSIM codes. For the uncertainty and sensitivity analysis, SUSA3.5 and SUNSET tools were used. The article demonstrates that applying the best estimate approach, it is possible to develop basic QUENCH input deck and to develop the two sets of input parameters, covering maximal and minimal ranges of uncertainties. These allow simulating different (but with the same nature) tests, receiving calculation results with the evaluated range of uncertainties.

• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3650-3659
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• 2022

We proposed a modification of a double cascade scheme to enrich reprocessed uranium. Such a cascade scheme represents a combination of one cascade with "broadening" of the flow and an ordinary three-flow cascade. A calculation and optimization method has been developed for the proposed scheme according to various efficiency criteria. It is shown that the proposed scheme makes it possible to obtain low-enriched uranium of commercial quality using reprocessed uranium of different initial compositions. For example, the enrichment of reprocessed uranium, which had gone through five consequent recycles, was considered. The proposed scheme allowed to enrich it with simultaneous fulfillment of restrictions on isotopes ²³²U, ²³⁴U, and ²³⁶U. Such results indicate the scheme's applicability under conditions of multiple recycling of uranium in reactor fuel. Computational experiments have shown that in the proposed modification, a noticeable saving of natural uranium in the cycle (~18%) can be achieved, provided that the additional consumption of separative work does not exceed 10%, compared with the case of enrichment of natural uranium to obtain LEU of equivalent quality.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.484-497
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• 2021

This paper proposes an extended time uncertainty analysis approach in Level 2 human reliability analysis (HRA) considering severe accident management (SAM) strategies. The method is a time-based model that classifies two time distribution functions-time required and time available-to calculate human failure probabilities from delayed action when implementing SAM strategies. The time required function can be obtained by the combination of four time factors: 1) time for diagnosis and decision by the technical support center (TSC) for a given strategy, 2) time for strategy implementation mainly by the local emergency response organization (ERO), 3) time to verify the effectiveness of the strategy and 4) time for portable equipment transport and installation. This function can vary depending on the given scenario and includes a summation of lognormal distributions and a choice regarding shifting the distribution. The time available function can be obtained via thermal-hydraulic code simulation (MAAP 5.03). The proposed approach was applied to assess SAM strategies that use portable equipment and safety depressurization system valves in a total loss of component cooling water event that could cause reactor vessel failure. The results from the proposed method are more realistic (i.e., not conservative) than other existing methods in evaluating SAM strategies involving the use of portable equipment.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1890-1901
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• 2022

We have assessed the applicability of the thermal-hydraulic system analysis code, SPACE, to a small modular reactor called SMART. For the assessment, the experimental data from a scale-down integral-test facility, SMART-ITL, were used. It was conformed that the SPACE code unrealistically calculates the safety injection flow rate through the CMT and SIT during a small-break loss-of-coolant experiment. This unrealistic behavior was due to the overprediction of interfacial heat transfer at the steam-water interface in a vertically stratified flow in the tanks. In this study, a special thermal-hydraulic component model has been developed to realistically calculate the interfacial heat transfer when a strong non-equilibrium two-phase flow is formed in the CMT or SIT. Additionally, we developed a special heat structure model, which analytically calculates the heat transfer from the hot steam to the cold tank wall. The combination of two models for the tank are called the special component model. We assessed it using the SMART-ITL passive safety injection system (PSIS) test data. The results showed that the special component model well predicts the transient behaviors of the CMT and SIT.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.536-545
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• 2024

Radioactive iodine is a representative fission product to be quantified for the safety assessment of nuclear facilities. In integral severe accident analysis codes, the iodine behavior is usually described by a multi-physical model of iodine chemistry in aqueous phase under radiation field and mass transfer through gas-liquid interface. The focus of studies on iodine source term evaluations using the combination approach is usually put on the chemical aspect, but each contribution to the iodine amount released to the environment has not been decomposed so far. In this study, we attempted the decomposition by revising the two-film theory of molecular-iodine mass transfer. The model involves an effective overall mass transfer coefficient to consider the iodine chemistry. The decomposition was performed by regarding the coefficient as a product of two functions of pH and the overall mass transfer coefficient for molecular iodine. The procedure was applied to the EPICUR experiment and suppression chamber in BWR.

• 대한환경공학회지
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• 제33권1호
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• pp.39-46
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• 2011

$250{\sim}400^{\circ}C$ 범위에서 $NO_x$ 제거를 위해 운영되는 선택적 촉매 환원법의 반응 온도를 $200^{\circ}C$ 이하로 낮추기 위해서는 NO를 $NO_2$로 산화시키는 전처리 공정을 필요로 한다. 이번 연구에서는 분말 $NaClO_2(s)$를 이용하여 NO를 $NO_2$로 산화시킨 후, 탄소 분산형 촉매를 이용한 저온에서의 $NO_x$, $SO_2$ 동시 제거에 관한 실험실 규모 실험과 제철소 소결 공장에서 실제 배기가스를 이용하는 bench 규모 실험을 진행하였다. 실험실 규모 실험에서는 반응기에 $NaClO_2(s)$ (2.4~3.6 g)를 충진 하여 $NO_x$ 200 ppm, $SO_2$ 75 ppm, $H_2O$ 10%, $O_2$ 15%의 모사가스(2.6 L/min)를 통과시켰으며, $NaClO_2(s)$와 반응 후의 모사가스를 탄소 분산형 촉매가 충진 된 반응기(공간 속도 = $2,000hr^{-1}$)로 주입하였다. bench 규모 실험에서는 $50Nm^3/hr$의 배기가스 유량에 screw feeder로 $NaClO_2(s)$ 분말을 주입하여 NO를 $NO_2$로 산화 시킨 후, $1,000hr^{-1}$ 탄소 분산형 촉매를 통과하여 $NO_x$ 제거 가능성을 확인하였다. 실험실 규모와 bench 규모 실험 모두 $SO_2$를 측정하며 $NO_x$, $SO_2$ 동시 제거 가능성을 확인하였다. 그 결과 실험실 규모와 bench 규모 실험 모두 $NaClO_2(s)$에 의하여 NO가 $NO_2$로 산화되었고, 이를 결합한 탄소 분산형 촉매에서 90% 이상의 $NO_x$, $SO_2$ 제거 효율을 나타내는 것을 확인하였다. 이상의 실험 결과로부터 $NaClO_2(s)$와 탄소 분산형 촉매의 결합은 저온에서 $NO_x$와 $SO_2$를 동시에 제거할 수 있음을 알 수 있었다.

• 대한기계학회논문집B
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• 제35권4호
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• pp.391-397
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• 2011

수소가 매력적인 연료로 각광받기 시작하면서 수요가 급증하였으며 이에 대응하여 수소 생산 기술에 대한 연구가 필요하다. 본 연구에서는 산소-메탄올 비율에 따른 연료전지용 메탄올 개질기의 반응 효율을 알아보았다. 각각의 촉매 배열에 따른 산소-메탄올의 비율($O_2/CH_3OH$)의 영향을 알아보기 위해 $O_2/CH_3OH$를 0.1에서 0.4까지 0.05씩 증가시켜 반응기의 온도, 변환율, 효율에 관한 실험을 수행하였다. $O_2/CH_3OH$가 0.15에서 0.2로 증가할 때 촉매층(catalyst bed)의 온도도 증가하며, 흡열 반응이 발열반응으로 변하여 반응기의 온도를 상승시켜 촉매 점화에 따라 온도는 $235^{\circ}C$정도 급상승한 $500^{\circ}C$가 된다. 반응기의 성능은 $O_2/CH_3OH$에 크게 의존하며 이론적 연구에서 $O_2/CH_3OH$는 0.23이었으나 실험 결과는 30 % 높은 0.30일 때 최적의 성능을 나타내었다. 이것은 혼합기체의 농도차이, 반응속도, 촉매, 반응기의 열손실, 반응 시 생성된 생성물 등의 변화 때문인 것으로 여겨진다.