• Proceedings of the Korean Nuclear Society Conference
• /
• 1999.10a
• /
• pp.278.1-278
• /
• 1999

• Nuclear Engineering and Technology
• /
• v.35 no.5
• /
• pp.482-496
• /
• 2003

effectiveness of the PAM graphite-electrode technology for the treatment of many types of low-level radioactive waste including : combustible material, solidified resins in cement, inorganic materials, steel, glass, and solidified boric acid cement. The objectives of PAM-200 evaluation were to verify that 1) the facility meets air emission regulations, 2) the facility can be safely operated when processing hazardous and radioactive materials and 3) satisfactory final waste forms can be produced. Results, derived from KAERI's(Korea Atomic Energy Research Institute) analyses for samples of vitrified product, scrubbing solution and offgas collected during test period, show that PAM-200 can treat radioactive wastes as well as hazardous wastes with toxic constituents and radionuclides contained in the offgas exiting from the stack to the environment controlled to be far lower than the limit regulated by air conservation law and atomic law.

• Nuclear Engineering and Technology
• /
• v.9 no.4
• /
• pp.211-222
• /
• 1977

The decomposition of all the individual chemicals used in the Harwell inactive vitrification pilot plant has been studied by means of a thermal balance. Weight loss curves to 110$0^{\circ}C$ have been obtained. The four materials (sodium nitrate, cesium nitrate, lithium nitrate and ruthenium nitroso-nitrate solution) showed a greater weight loss than that based on an oxide yield, and hence these compounds of their products of decomposition are volatile below 110$0^{\circ}C$. The remaining materials suffered a weight loss no more than that corresponding to a full yield of the oxide, and hence they were not volatile below 110$0^{\circ}C$. Most of chemicals begin to decompose at less than 75$^{\circ}C$ but the nitrates of cesium, strontium, barium and sodium not until 295$^{\circ}C$ to 59$0^{\circ}C$. The results obtained can be used in the analysis of process conditions in the vitrification and calcination of highly radioactive wastes and also of the thermal decomposition behaviour of mixtures containing those materials.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.8 no.3
• /
• pp.221-227
• /
• 2010

The solidification methods for powder wastes dried at CWDS(Concentrate Waste Drying System) in PWR have been studied in a variety of ways both at home and abroad. The solidification for these wastes has been performed using stabilization agents such as cement, paraffin and polymer. The applicability studies to maximize the reduction ratio of wastes and operational effectiveness for wastes treatment have been carried out, recently. It is necessary to pretreat the powder wastes before feeding wastes to vitrification facility because the fines flying brings about clogging of feeding pipes and off-gas treatment system or workers' exposure to radiation during maintenance. This paper describes an effective method for treatment of powder wastes to improve safety and stability of vitrification facilities.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2005.11a
• /
• pp.69-78
• /
• 2005

Appropriate numerical models for the simulation of off-gas flow in hot area of the vitrification plant have been developed in this study. The models have been applied to analyze the effect of design parameters of real plant and numerical analyses have been performed for CCM(Cold Crucible Melter), pipe cooler and HTF(High Temperature Filter) At first, the effect of excess oxygen and the ratio of oxygen distribution on combustion characteristics in the CCM has been studied. Next, solidification behavior of radio nuclide In the pipe tooler has been numerically modeled and scrutinized. Finally, flow pattern In accordance with the location of off-gas entrance of the HTF has been compared.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.2 no.3
• /
• pp.201-209
• /
• 2004

Important operation parameters and performance of a high temperature ceramic candle filter system were evaluated through a series of demonstration tests at a pilot-scale vitrification plant. At the initial period of each test, due to the growth of dust cake on the surface of ceramic candles, the pressure drop across the filter media increased sharply. After that it became stable to a certain range and varied continuously proportion to the face velocity of off-gas. On the contrary, at the initial period of each test, the permeability of filter element decreased rapidly and then it became stable. Back flushing of the filter system was effective under the back flushing air pressure range of 3∼5 bar. Based on the dust concentrations measured by iso-kinetic dust sampling at the inlet and outlet point of HTF, the dust collection efficiency of HTF evaluated. The result met the designed performance value of 99.9%. During the demonstration tests including a hundred hour long test, no specific failure or problem affecting the performance of HTF system were observed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.2 no.4
• /
• pp.231-238
• /
• 2004

The conditions are obtained for the decomposition of solid radioactive wastes, including ion exchange resin, zeolite, charcoal, and sludge from nuclear power plant. In the process of decomposing the radioactive wastes was used the microwave acid digestion method with mixed acid. The solution after acid digestion by the following method was colorless and transparent. Each solution was analyzed with ICP-AES and AAS and the recovery yield for 5 different elements added into the simulated radioactive wastes were over $94{\%}$. The elemental analysis of destructive low and medium level radioactive wastes by the proposed microwave acid digestion conditions concerning the chemical characteristics of each radioactive waste are expected to be useful basic data for development of optimal glass formulation.

• Nuclear Engineering and Technology
• /
• v.29 no.4
• /
• pp.269-279
• /
• 1997

The errors associated with incorrect sampling and transport of radioactive aerosol from radwaste thermal process off-gas are analyzed and the conditions of representative sampling and correct transport of radioactive aerosol for off-gas system evaluation are discussed. An estimation method of sampling errors for individual radionuclides is proposed and applied to simulated vitrification melter aerosols. Prediction methods for particle deposition in sample transport tube under laminar as well as turbulent flow conditions are also described by example calculations with simulated incinerator off-gas From the results of example calculations and plots, instrumental and operational conditions of radioactive aerosol sampling system with minimized errors and correction methods for nonideal sampling and transport are recommended.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.10 no.4
• /
• pp.247-254
• /
• 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 the Korean Ceramic Society
• /
• v.39 no.6
• /
• pp.576-581
• /
• 2002

In order to simultaneously vitrify the ton Exchange Resin(IER) and Dry Active Waste(DAW) generated from the Nuclear Power Plants, a vitrification pilot test was conducted using an induction cold crucible melter. The PCT result evaluating the chemical durability of the vitrified from showed that the final glass was more durable than the benchmark glass. Liquidus temperature for the final vitrified form was 1048 K(775$\^{C}$) fur heat treatment experiments. The value of the compressive strength for the vitrified form was ninety times higher than the regulation limit, 34 kg/㎠. The glasses on bottom, middle and top of the CCM were homogeneous with no secondary phase. The precipitation of the magnetic metal phase was able to be avoided by simultaneously fEeding of DAW with IER containing strongly reducing organics. Volume reduction factor of 74 was achieved through the vitrification Pilot test for mixed waste.