• Proceedings of the Korean Nuclear Society Conference
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• 1995.05b
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• pp.781-786
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• 1995

석탄화력발전소 산업부산물인 Fly ash를 이용한 고준위방사성폐기물의 붕규산 유리고화 가능성을 분석하였다. Fly ash SiO$_2$, NaNO$_3$, B$_2$O$_3$에 DUPIC 핵연료 제조공정으로부터 발생되는 모의 scrap waste를 20 wt% 혼합하여, l15$0^{\circ}C$ 에서 3시간 용융시켜 붕규산유리화시켰다. 또한 붕규산유리고화체의 침출성을 평가하기 위하여 2일동안의 soxhlet 침출실험결과 양호한 내침출성을 보였다. 또한 고체폐기물의 안정화물질로 fly ash를 사용할 경우 fly ash 함량을 57%까지 첨가하여도 붕규산유리고화체의 제조가 가능함을 확인하였으며, fly ash의 첨가로 인한 유리화원료 재료비를 30% 까지는 절감시킬 수 있을 것으로 예상된다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.4
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• pp.335-340
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• 2005

The forward dissolution rate of a borosilicate waste glass was determined as an interlaboratory study(ILS) testing program for the evaluation of precision in the measurement of the dissolution rate or a waste glass using a single-pass flow-through(SPFT) test, whose conducting practice has been written for standardization through American Society for Testing and Materials (ASTM). A simulated low-activity waste glass powder with a size of 100/200 mesh was dissolved by lithium buffer solution (pH=10) at 70? under Ar atmosphere. By plotting the dissolution rates as a function of silicon and boron concentration in eluate, the forward dissolution rate of the glass was obtained as about $2.7\times10^{-5}g{\cdot}m{\cdot}s^{-1}$ in our laboratory.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.2
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• pp.143-150
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• 2010

Vitrification technology has been gradually recognized as one of effective solidification methods for concentrated boric acid wastes generated in PWR. Vitrification for low- and intermediate-level radioactive wastes has a large volume reduction and good durability for the final products. A feasibility study for the vitrification of concentrated boric acid wastes has been performed with developing the pre-treatment methods of powdered wastes, glass compositions using glass formulation and demonstration test. The pre-treatment method is pelletizing the powder type for stable feeding within cold crucible melter. The glass compositions should be developed considering molten glass are related with wastes reduction. High contents of sodium and boron within borate wastes give influence to waste loading. A variety of factors obtained from the demonstration test are reviewed, which is wastes feeding rate, off-gas characteristics on stack and glass characteristics of final products such as durability for implementing the wastes disposal requirement. The aim of this paper is to present the feasibility of vitrification and review the solidification method for concentrated boric acid wastes and obtain the physicochemical characteristics of solidified glass.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.4
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• pp.247-254
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• 2012

Vitrification technology has been widely applied as one of effective processing methods for wastes generated in nuclear power plants. The advantage of vitrifying for low- and intermediate-level radioactive wastes has a large volume reduction and good durability for the final products. Recently, a filter using on HVAC(Heating Ventilating & Air Conditioning System) is composed with media (glass fiber) and separator (aluminum film) has been studied the proper treatment technology for meeting the waste disposal requirement. Present paper is a feasibility study for the filter vitrification that developing of the glass compositions for filter melting and melting test for physicochemical characteristic evaluation. The aluminum metal of film type is preparing with 0.5 cm size for proper mixing with glass frit, glass fiber is also preparing with 1 cm size within crucible. The glass compositions should be developed considering molten glass are related with wastes reduction. Glass compositions obtained from developing on glass formulation are mainly composed of $SiO_2$ and $B_2O_3$ for aluminum metal. A variety of factors obtained from the glass formulation and melting test are reviewed, which is feeding rate and glass characteristics of final products such as durability for implementing the wastes disposal requirement.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.2
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• pp.99-105
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• 2010

This paper investigated the characteristics of wasteform containing a spent zeolite used as a separating agent of FPs for recycling LiCl waste which would be generated from pyrochemical process of spent PWR fuel. In this study, a conventional borosilicate and Ca-rich glass were used as a consolidating agent for spent zeolite and it's mixing ratio was changed in the range, $25{\sim}35wt%$. The leach rates of Cs and Sr had about $0.1{\sim}0.01g/m^2day$ and $0.001{\sim}0.0001g/m^2day$, respectively. The leach resistance of Cs increased with amount of SAP and it showed about 10 times higher in the Ca-rich glass wasteform than in the conventional borosilciate glass wasteform. The compressive strength of wasteform was affected with the amount of glass. Thermal expansion rate of containing 30wt% glass has relatively lower than others. Also, the melting temperature was little changed in given mixing ratio of glass.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.3
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• pp.177-181
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• 2005

Yellowish uranium compounds were enriched at the interface between a Ca-bentonite block and a waste glass, containing about $20\%$ uranium oxide, in contact with the block due to the dissolution of uranium by a synthetic granitic groundwater in Ar atmosphere. The uranium compound formed for 6 years leach time was identified as a beta-uranophane $[Ca(UO_2)_2(SiO_{3}OH)_{2}5H_{2}O]$ using XRD, IR and mass spectrometer. The solubility of the beta-uranophane was measured to be about $10^{-6}\;mole/L$ in de-mineralized water at $80^{\circ}C$.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2007.11a
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• pp.121-122
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• 2007

• Journal of the Korean Ceramic Society
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• v.38 no.9
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• pp.839-845
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• 2001

In order to vitrify the Ion-Exchange Resin(IER), Dry Active Waste(DAW), and borate concentrate generated from the commercial nuclear facilities, the glass formulation study based on the their compositions was performed. Two glasses named as RG-1 and DG-1 were formulated as the candidate glasses for the vitrification of hte IER and DAW, respectively. A glass named as MG-1 was also formulated as a candidate glass for the vitrification of the mixed wastes containing the IER, DAW, and borate concentrate. The process parameters, product qualities, and economics were evaluated for the candidate glasses and confirmed experimentally for the some properties. The glass viscosity and electrical conductivity as the process parameters were in the desired ranges. the product qualities such as glass density, chemical durability, phase stability, etc. were satisfactory. In case of vitrifying the wastes using our developed glass formulation study, the volume reduction factors for the IER, DAW and mixed wastes were evaluated as 21, 89 and 75, respectively.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.294.1-294.1
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• 2003

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2006.06a
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• pp.99-100
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• 2006