A study was conducted on the vitrification of the rare earth oxide waste generated from the PyroGreen process. The target rare earth waste consisted of eight elements: Nd, Ce, La, Pr, Sm, Y, Gd, and Eu. The waste loading of the rare earth waste in the developed borosilicate glass system was 20wt%. The fabricated glass, processed at 1,200℃, exhibited uniform and homogeneous surface without any crystallization and precipitation. The viscosity and electrical conductivity of the melted glass at 1,200℃ were 7.2 poise and 1.1 S·cm−1, respectively, that were suitable for the operation of the vitrification facility. The calculated leaching index of Cs, Co, and Sr were 10.4, 10.6, and 9.8, respectively. The evaluated Product Consistency Test (PCT) normalized release of the glass indicated that the glass satisfied the requirements for the disposal acceptance criteria. Furthermore, the pristine, 90 days water immersed, 30 thermal cycled, and 10 MGy gamma ray irradiated glasses exhibited good compressive strength. The results indicated that the fabricated glass containing rare earth waste from the PyroGreen process was acceptable for the disposal in the repository, in terms of chemical durability and mechanical strength.

Vitrification is one of the best ways to immobilize high-level radioactive waste (HLW) worldwide over the past 50 years. Since the glass matrix has a medium (3.0-5.5 A) and short (1.5-3.0 A) periodicity, it can accommodate most elements from the periodic table. Borosilicate glass is the most suitable glass matrix for vitrification due to its high chemical durability, high waste-loading capacity, and good radiation resistance. Mo is a fission product contained in liquid waste generated in the process of reprocessing spent nuclear fuel and exists in the form of MoO4 2- in the glass. MoO4 2- forms a depolymerization region without directly connecting with the glass network former. When the concentration of Mo increases in the depolymerization region, it combines with nearby alkali or alkaline earth cations to form a crystalline molybdate phase. Phase separation and crystallization in the glass can degrade the performance of the material because it changes the physical and chemical properties of the glass. In particular, since alkali molybdate has high water solubility when it forms crystals containing radioactive elements such as Cs, there is a risk of leakage of radionuclides by groundwater during deep underground disposal. Therefore, in this study, the most stable glass-ceramic composition was developed using various alkali elements, and the difference in phase separation and crystallization behavior in glass and the stability of the solidified body were analyzed by structural analysis of the glass network and alkali molybdate. The cause of the difference in crystallization of alkali molybdate according to the type of alkali cation is structurally analyzed, and using this, research is conducted to increase the Mo content in the glass without crystallization.

In this study, we have investigated a selective emitter using a UV laser on BBr3 diffusion doping layer. The selective emitter has two regions of high and low doping concentration alternatively and this structure can remove the disadvantages of homogeneous emitter doping. The selective emitters were fabricated by using UV laser of 355 nm on the homogeneous emitters which were formed on n-type Si by BBr3 diffusion in the furnace and the heavy boron doping regions were formed on the laser regions. In the optimized laser doping process, we are able to achieve a highly concentrated emitter with a surface resistance of up to 43 Ω/□ from 105 ± 6 Ω/□ borosilicate glass (BSG) layer on Si. In order to compare the characteristics and confirm the passivation effect, the annealing is performed after Al2O3 deposition using an ALD. After the annealing, the selective emitter shows a better effect than the high concentration doped emitter and a level equivalent to that of the low concentration doped emitter.

Effects of chemical compositions on the sintering behavior of the lead borosilicate glass developed for barrier ribs of plasma display panels were investigated in this study. Formation of pores during sintering of the glass was noted and their formation mechanism was investigated using XPS, TG/DTA, and XRD. The results indicated that pores are formed by the oxygen released from Pb-oxides during sintering.

This research is to investigate the effect of borosilicate glass powder on neutron shielding capability of cement mortar. The average particle size of the borosilicate glass powder was 13 μm. It was found that the addition of borosilicate glass powder increased 28 day compressive strength. In addition, neutron shielding capability of cement mortar also increased by the addition of borosilicate glass powder. Considering our earlier findings on enhanced thermal neutron shielding of cement mortar by borosilicate glass powder, the use of borosilicate glass powder was found to be effective to shield neutron when the cement mortar was exposed to the neutron radiation. It can be concluded that borosilicate glass powder is a good alternative material for neutron shielding purposes.

핵폐기물을 고화시키는 재료로 사용하는 붕규산염(borosilicate) 유리의 용해는 지층 처분장에 처리된 고준위 방사성 폐기물의 생태계 유출을 결정할 수 있는 중요한 화학반응이다. 습식 실험에서 유리의 용해속도(dissolution rate)는 유리 화학조성에 의해 크게 좌우되는 것이 관찰된다. 유리의 bulk 구조를 규명한 분광분석 실험에 의하면 유리의 화학조성과 분자수준(molecular-level) 구조(예: SiO4 사면체의 연결구조와 B 원소의 배위구조) 사이의 상관관계가 존재한다. 따라서 화학조성에 따른 유리 용해도의 차이는 조성에 따른 bulk 내부구조의 변화로 이해되어 왔다. 그런데 유리 표면은 수용액과 계면을 이루면서 용해 과정에서 가장 직접적으로 반응하는 부분이기 때문에, 화학조성에 따른 표면구조 변화에 대한 지식 또한 필요하다. 본 논문에서는 분자 동역학(molecular dynamics, MD) 시뮬레이션을 사용하여 4가지의 다른 화학조성을 가지는 소듐붕규산염 유리(xNa2O·B2O3·ySiO2 화학조성)에 대하여 bulk 구조와 실험으로 얻기 어려운 표면(surface) 구조를 연구하였다. MD 시뮬레이션은 유리 표면의 화학조성과 분자수준 구조가 bulk의 것과 매우 상이한 결과를 보여준다. 본 연구의 MD 시뮬레이션 결과는 화학조성에 따른 유리 용해도(특히 초기 용해과정)는 bulk 구조의 변화보다 유리 표면구조의 변화에 의해 크게 좌우될 수 있다는 표면구조에 대한 이해의 중요성을 역설한다.