Ammonia is considered a promising hydrogen carrier due to its high hydrogen density and liquefaction temperature. Considering that the energy efficiency generally decreases as chemical conversion is repeated, it is more efficient to directly use ammonia as a fuel for fuel cells. However, catalysts in direct ammonia fuel cells have the critical issues of sluggish ammonia oxidation reaction (AOR) rate and poisoning of reaction intermediates. In particular, the use of precious metal as cathodic catalysts has been limited due to ammonia crossover and poisoning. In this study, we introduce Fe-based single-atom catalysts with selective activity for the oxygen reduction reaction (ORR) even in the presence of ammonia. As the Fe content increased, the single-atom structure of the catalysts changed into Fe nanoparticles or carbides. Among our Fe–N–C catalysts, FeNC-50 with a Fe loading amount of 0.34 wt% showed the highest ORR performance regardless of the ammonia concentration. In particular, the difference in activity between the catalysts increased as the concentration increased. The FeNC-50 catalyst showed remarkable stability after 1000 cycles. Therefore, we believe that single-atom dispersion is an important factor in the development of stable non-precious catalysts with high activity and inactivity for the ORR and AOR, respectively.

A Cu-15Ag-5P filler metal (BCuP-5) is fabricated on a Ag substrate using a high-velocity oxygen fuel (HVOF) thermal spray process, followed by post-heat treatment (300oC for 1 h and 400oC for 1 h) of the HVOF coating layers to control its microstructure and mechanical properties. Additionally, the microstructure and mechanical properties are evaluated according to the post-heat treatment conditions. The porosity of the heat-treated coating layers are significantly reduced to less than half those of the as-sprayed coating layer, and the pore shape changes to a spherical shape. The constituent phases of the coating layers are Cu, Ag, and Cu-Ag-Cu3P eutectic, which is identical to the initial powder feedstock. A more uniform microstructure is obtained as the heat-treatment temperature increases. The hardness of the coating layer is 154.6 Hv (as-sprayed), 161.2 Hv (300oC for 1 h), and 167.0 Hv (400oC for 1 h), which increases with increasing heat-treatment temperature, and is 2.35 times higher than that of the conventional cast alloy. As a result of the pull-out test, loss or separation of the coating layer rarely occurs in the heat-treated coating layer.

In this study, a new manufacturing process for a multilayer-clad electrical contact material is suggested. A thin and dense BCuP-5 (Cu-15Ag-5P filler metal) coating layer is fabricated on a Ag plate using a high-velocity oxygen-fuel (HVOF) process. Subsequently, the microstructure and bonding properties of the HVOF BCuP-5 coating layer are evaluated. The thickness of the HVOF BCuP-5 coating layer is determined as 34.8 μm, and the surface fluctuation is measured as approximately 3.2 μm. The microstructure of the coating layer is composed of Cu, Ag, and Cu-Ag-Cu3P ternary eutectic phases, similar to the initial BCuP-5 powder feedstock. The average hardness of the coating layer is 154.6 HV, which is confirmed to be higher than that of the conventional BCuP-5 alloy. The pull-off strength of the Ag/BCup-5 layer is determined as 21.6 MPa. Thus, the possibility of manufacturing a multilayer-clad electrical contact material using the HVOF process is also discussed.

It is necessary to fabricate uniformly dispersed nanoscale catalyst materials with high activity and long-term stability for polymer electrolyte membrane fuel cells with excellent electrochemical characteristics of the oxygen reduction reaction and hydrogen oxidation reaction. Platinum is known as the best noble metal catalyst for polymer electrolyte membrane fuel cells because of its excellent catalytic activity. However, given that Pt is expensive, considerable efforts have been made to reduce the amount of Pt loading for both anode and cathode catalysts. Meanwhile, the atomic layer deposition (ALD) method shows excellent uniformity and precise particle size controllability over the three-dimensional structure. The research progress on noble metal ALD, such as Pt, Ru, Pd, and various metal alloys, is presented in this review. ALD technology enables the development of polymer electrolyte membrane fuel cells with excellent reactivity and durability.

최근 선박 연료유는 고점도화 되고 슬러지분이 증가되고 있는 추세이며, 선박에서 발생한 슬러지의 처리 및 보일러 연료유로의 재활용 방안 등에 대해서 많은 연구가 수행되고 있다. 이러한 연구 중 특히 슬러지를 미립화하여 분쇄하기 위한 초음파 유화기는 가장 현실성 있는 재활용 장치로 알려져 있다. 이러한 관점에서, 이 연구는 초음파 유화기 개발에 대한 기초연구로서 슬러지의 유온과 유압이 따른 여과효율을 조사하였다. 실험결과는 보일러 인젝터에 슬러지를 분사할 경우 적절한 온도와 압력을 결정하거나, 또한 초음파 유화기에 의한 실험결과와 비교할 수 있는 자료로 활용될 수 있다. 아울러 유온과 유압의 영향에 따라 분쇄된 슬러지 입자의 여과효율 등을 연구하는데 있어서 기초자료로 활용될 수 있을 것이며, 궁극적으로 선박에서 발생한 슬러지를 자체 처리하여 보일러의 연료유로 사용함으로써 해양유류오염을 방지하는데 기여할 수 있을 것이다.

본 논문에서는 MDO기법에 의한 핵연료교환장치의 구조해석 단계 중 핵연료교환장치의 휨 변형을 구하는 재료역학해석을 수행하였다. 이는 액체 금속로(LMR) 핵연료교환장치의 기본설계를 위하여 매우 중요하다. 해석대상 핵연료교환장치의 정적구조는 기 수행한 핵연료교환장치의 기구 동역 학 해석 결과를 활용하였다. 네 가지 핵연료교환동작에 대하여 핵연료 봉의 무게를 100㎏에서 500㎏까지 100㎏씩 증가시켜 휨 변형의 크기를 구하였다. 그 결과 회전 중심 축에서 가장 멀리 있는 핵연료 봉을 교환하는 핵연료교환동작에서 최대 휨 변형이 발생함이 밝혀졌다. 또한 이 최대 휨 변형이 발생하는 핵연료교환장치구조에 대하여 부재의 단면두께를 축소하면서, 또 단면형상을 여러 가지로 바꾸면서 휨 변형크기를 구하여 비교하였다. 비교결과 비교대상 단면형상 중에서 중공직사각형 단면이 최소 휨 변형이 발생하는 최적단면형상임이 밝혀졌다.

액체 금속로(LMIR) 핵연료교환장치의 기본설계를 위해서는 여러 분야(예를 들면, 기구학, 동역 학, 재료역학 등)의 해석을 동시에 수행해야 한다. 그러나 이와 같은 해석들은 각각 별개로 연속적으로 수행되는 것이 아니라, 상호 유기적인 연관을 갖고 수행되어야 한다. 이와 같은 해석에 적합한 기법이 MDO 기법이다. 본 논문에서는 MDO기법에 의한 핵연료교환장치 구조해석의 한 단계로 핵연료교환장치의 기구 동역 학 해석을 수행하여 핵연료 교환장치 작동에 대한 기구운동학적 특성 및 동역학적 특성을 분석하였다. 분석결과 해석대상 핵연료교환장치는 예상한대로 원활하게 작동됨이 확인되었다. 아울러 이 분석 결과를 토대로 핵연료교환장치의 정적 휨 변형을 구하기 위한 재료역학해석에서 요구되는 정적구조를 결정하였다.

The transportation of spent nuclear fuel between management stages is expected, and the transportation workers may be exposed to radiation. When transporting spent nuclear fuel, the ALARA principle must be observed for the workers. The objective of this study is to assess a radiation dose for workers transporting spent nuclear fuel using metal overpacks. For this objective, the cask to be handled was selected and the radiation source term was set. Then, the radiation exposure scenario for the transportation workers was defined. Finally, the dose rates for each location of operation were assessed using Monte Carlo simulations, and collective doses were derived for each operation considering the radiation exposure scenario. Each worker performed 11 operations to transport spent nuclear fuel to other facilities and was exposed to a total of 1.138 man-mSv. The operation of removing the bottom shield ring resulted in the highest radiation exposure at 0.503 man-mSv. In contrast, the operation of installing the impact limiter resulted in the lowest radiation exposure at 0.0009 man-mSv. The results of this study can be used to strengthen radiation protection measures for workers transporting spent nuclear fuel in dry storage facilities using metal overpacks.

Noble metal phase, present in used fuel, are fission products that can be found as metallic precipitates in used nuclear fuel. They exist as small particles (nm~um) in grain boundaries of the used fuels. Since they are particles deposited between the grain structures, they can be considered as defects in the pellet structure. Thermal expansion of fuels with noble metal is slightly higher than that of bare fuels. The fuels at high temperature, such as immediately after being discharged from nuclear reactors, may be subject to fuel failure if sufficient cooling is not provided. Recent research has shown that the noble metals can migrate into the rim space between the pellet and the cladding, and be deposited in the inner layer of the claddings. therefore, the mechanical integrity of the cladding can be degraded by noble metals, as well as the pellets. The concentration of the noble metal phase should be considered to evaluate the effect of the noble metals on the fuel integrity, after discharge from the reactors. SCALE/ORIGEN code was used to evaluate the noble metals in fuel assembly-scale, and the radial distribution in the fuel assembly. The radial distribution of the reactor power was derived from the SCALE/TRITON, considering Westinghouse 17×17. Square cell model was chosen for the geometry and 1/4 model was applied to reduce the computation time.

Noble metal precipitates are fission products that can be found as metallic alloys in used nuclear fuel. They do not exist homogenously inside the fuel pellets, but exists in grain boundaries in the form of immiscible particles. The first drawback that comes because they exist in grain boundaries is the degradation of mechanical integrity. The particles in the grain boundaries can be considered as defect n solid solution of uranium oxide pellets, and they can change the lattice volume. Therefore, it is known that it can cause stress corrosion cracking of fuel pellets. Furthermore, there is a negative effect from the perspective of used fuel management. However, they also have a positive effect on used fuel management. Since the noble metal has galvanic reduction effect, the particles serve as an oxidation inhibitor for uranium. There are many other effects regarding to the noble metal precipitates. However, in any case, quantifying the particles is important in order to quantitatively analyze these effects from the perspective of used fuel management. SCALE/TRITON code was applied to calculate the noble metal isotopes including Mo, Tc, Ru, Rh and Pd. In order to calculate the distribution inside the pin, the multiregion cell model was selected. In particular, a cylindrical geometry was used, and the pellet was divided into several layers. In addition, coolant and cladding surrounded the pellet. Finally, the radial distribution was evaluated using the computational code, along with neutron flux map.

When the recycling technology of spent nuclear fuels (SNF) for future nuclear reactor systems and the treatment technology of SNF for disposing of in a disposal site use a molten salt such as LiCl-KCl eutectic as a processing medium one of the essential unit processes is a distillation process that remove the salt component mixed with fission products recovered. Especially, in case of Pyro-SFR recycling system the recovered nuclear fuel materials such as U, TRU and some of rare earths come from main three processes (electro-refining, electro-winning, and drawdown processes) for recycling of SNF. These recovered fuel materials contain large portion of molten salt or liquid cadmium which requires removal of them by distillation. In spent nuclear fuels discharged from PWR the portion of composing element is as follows. Uranium is about 95%, other actinides such as transuranic elements (TRU; Np, Pu, Am, Cm) is about 1%, the rare earths (lanthanides) is about 1%, and the other elements is about 3%. For example, americium (Am) in the recovered fuel materials has a problem that the reported loss of Am inevitably occurs during the vacuum salt distillation operation. A new segregation method of AMM (actinide metal mixture)–salt system is based on the difference in melting point of the actinide elements. It is possible to apply this segregation method to recovering other actinides from AMM with accompanied salt because of relatively large amount and lower melting point of a specific element in other actinides avoiding vacuum salt distillation. This new segregation method successfully tested using a surrogate element such as aluminum due to its similar melting point with a specific element. The segregation principle is solid-liquid separation, thus the solidified actinides mixture ingot can take out of a molten salt medium.

As regulations on carbon emissions increase, the interest in renewable energy is also increasing. However, the efficiency of renewable energy generation is highly low and has limitations in replacing existing energy consumption. In terms of this view, nuclear power generation is highlighted because it has the advantage of not emitting carbon. And accordingly, the amount of spent nuclear fuel is going to increase naturally in the future. Therefore, it will be important to obtain the reliability of containers for transporting safely and storing spent nuclear fuel. In this study, a method for verifying the integrity and airtightness of a metal cask for the safe transportation and storage of spent nuclear fuel was studied. Non-destructive testing, thermal stability, leakage stability, and neutron shielding were demonstrated, and as a result, suitable quality for loading spent nuclear fuel could be obtained. Furthermore, it is meaningful in that it has secured manufacturing technology that can be directly applied to industrial field by verifying actual products.

The quality standards of solid refuse fuel (SRF) define the values for 12 physico-chemical properties, including moisture, lower heating value, and metal compounds, according to Article 20 of the Enforcement Rules of the Act on Resource Saving and Recycling Promotion. These parameters are evaluated via various SRF Quality Test Methods, but problems related to the heavy metal content have been observed in the microwave acid digestion method. Therefore, these methods and their applicability need improvement. In this study, the appropriate testing conditions were derived by varying the parameters of microwave acid digestion, such as microwave power and pre-treatment time. The pre-treatment of SRF as a function of the microwave power revealed an incomplete decomposition of the sample at 600 W, and the heavy metal content analysis was difficult to perform under 9 mL of nitric acid and 3 mL of hydrochloric acid. The experiments with the reference materials under nitric acid at 600 W lasted 30 minutes, and 1,000 W for 20 or 30 minutes were considered optimal conditions. The results confirmed that a mixture of SRF and an acid would take about 20 minutes to reach 180 oC, requiring at least 30 minutes of pre-treatment. The accuracy was within 30% of the standard deviation, with a precision of 70 ~ 130% of the heavy metal recovery rate. By applying these conditions to SRF, the results for each condition were not significantly different and the heavy metal standards for As, Pb, Cd, and Cr were satisfied.

국내 폐기물 발생량은 급격한 산업화와 인구 증가 등의 요인으로 인해 꾸준히 증가하고 있으며 이에 따라 다양한 폐기물 처리방법이 수행되고 있다. 폐기물 처리방법 중 하나인 고형연료제품 제작은 폐기물 발생을 최소화할 수 있고 폐기물 중 가용 자원의 재활용을 극대화 할 수 있기 때문에 신재생에너지로 간주되고 있다. 고형연료는 고체폐기물 중 폐합성수지류, 폐지류, 폐목재류 등 가연성 물질을 선별하여 파쇄, 건조 등의 처리과정을 거쳐 연료화시킨 것을 통칭하며 소각시설이나 발전시설에서 연료로 사용되고 있다. 하지만 최근 미세먼지 문제가 심각해지면서 고형연료에 대한 부정적 인식이 늘고 있으며 이를 극복하기 위해서는 고형연료가 안전한 제품으로 인식될 수 있도록 다수의 품질기준 적합성 검사가 필요하다. 고형연료 품질기준 중 중금속 함량 분석은 이러한 인식 제고에 반드시 필요한 시험 항목이기 때문에 정확성이 확보되어야 하며 현재 고형연료 품질시험방법에 따른 중금속 분석방법은 전처리 과정에서 고형연료 시료가 완전히 분해되지 않는 문제점이 발견되었다. 본 연구에서는 기존 고형연료의 중금속 함량 분석 방법을 개선하기 위해 마이크로파 전처리 조건의 산 종류, 마이크로파 전력(W), 반응시간에 변화를 주어 이에 따른 17종의 중금속(As, Cd, Pd, Ca, Co, Cr, Cu, Fe, Li, Mg, Mn, Ni, Sb, Sr, Ti, V, Zn)함량 변화를 확인하였다. 대상 시료는 인증표준물질 ERM-EC680k를 사용하였고 마이크로파 전처리를 통해 제조된 액상시료는 유도결합플라즈마 분광분석기(ICP-OES)를 통해 분석하였다.