In South Korea, low- and intermediate-level radioactive waste (LILW) has been stored in on-site storage buildings after being super compacted, dried or solidified. Such on-site storage capacity is almost exhausted. If a permanent repository for the LILW is not available in the near future, current on-site storage capability should be expanded to accept the waste drums from further operations. The Table below shows the present status of the on-site storage for the Korean LILW, which indicates that a permanent repository must be constructed at least before 2008. Thus, it is important to develop a technology that will significantly reduce the volume of the LILW and enhance their disposal stability.
The Nuclear Environment Technology Institute (NETEC), a division of Korea Hydro & Nuclear Power (KHNP) has investigated and evaluated various thermal treatment technologies for LILW. NETEC has focused on a treatment technology that will result in a large reduction of volume, enhance the stability of the waste form, and can treat all waste streams generated from Korean nuclear plants.
It was decided in early 1994 that vitrification technology to treat LILW was the most promising technology. The vitrified radwaste is expected to remain stable in the repository environment for over one million years. In addition, vitrification technology contributes to a waste volume reduction factor of greater than 50.
Research and development
The multi-step R&D projects to vitrify the Korean LILW are shown in Figure 1.
During 1994 and 1995, a feasibility study was carried out to evaluate the application of vitrification technology to the LILW through laboratory tests to assess the melter technologies, to examine how innovative high temperature technologies could be implemented to achieve a large volume reduction, and to evaluate and compare these technologies from a technical and economic viewpoint. As a result, a combination of the cold crucible melter (CCM) and the plasma torch melter (PTM) with an off-gas treatment system (OGTS) was selected to vitrify the combustible radwaste and melt the non-combustible radwaste respectively.
An international joint R&D programme started in July 1996, with the support of SGN and Mobis, to develop the cold crucible vitrification of the Korean LILW using a pilot CCM through a multi-step programme. The first stage, completed in 1998, was to optimise the processing of the Korean LILW in the CCM and to design the off-gas treatment system. The second stage, completed in September 1999, was the design and construction of the pilot vitrification plant in Daejeon, Korea. This vitrification plant has been in operation since October 1999. The pilot test performed by NETEC with the support of SGN and Mobis simulated the behaviour of a commercial vitrification plant as closely as possible, and demonstrated its reliability and high productivity. Finally, the pilot vitrification tests were intended to assist in finalisation of the design parameters, design options, radiological environment assessment data, shielding and licence data, and so on. As a part of the project, the glass formulation study was conducted in parallel with the pilot vitrification tests and the pilot plant using simulated LILW.
Pilot vitrification plant and tests
The innovative combined vitrification process used for vitrification of the LILW is shown in Figure 2.
The pilot vitrification plant consists of a feeding system, a 300kW CCM, a 200kW PTM, and an OGTS. The feeding system consists of a glass frit feeder and two waste feeders. The CCM is a water-cooled melter in which electric currents are directly induced by an external high frequency generator. This allows high thermal power to be generated in the glass melt. Thus, the glass is melted and the combustible waste is thermally decomposed. It comprises 6 sections: main body; upper chamber; oxygen supply system; glass drain system; water cooling system; and high frequency generator. Combustible wastes are burned and pyrolysed by the heat of the glass melt and oxygen. The PTM system consists of a power generator, a water cooling system, a plasma gas supply system, a waste feeding system, upper and bottom chambers, and a control system. The upper chamber is connected to the plasma torch, the waste feeder, and the off-gas outlet. The plasma torch generates plasma with nitrogen gas. The torch can rotate and move up and down to melt wastes uniformly.
The OGTS handles the off-gas generated from both the CCM and the PTM. It will be tested to demonstrate the capability of capturing and recycling radioactive particles in the off-gas at the upper front of the OGTS, and compliance with guidelines for the safe dealing of toxic gases such as dioxin, nitrogen oxides, and hydrogen chloride. The OGTS includes a pipe cooler, a high temperature ceramic filter, a secondary combustion chamber, a venturi and a 4-stage packed tower scrubber, a HEPA filter, a selective catalytic reduction (SCR), and the associated equipment and an instrument and control system.
A total of 59 pilot vitrification tests using the CCM were carried out with simulated dry active waste (DAW), ion exchange materials including high and low activity ion exchange resin (HA/LA IER), and zeolite, and other waste materials for the all waste streams generated from nuclear plants. In addition, 21 pilot melting tests using the PTM were carried out with simulated concrete, soil, spent glass, metal scraps, and spent filter. The result of the pilot tests proved that environmentally stable vitrified forms for all the wastes could be produced. All regulated gases in the stack were well below environmental regulation limits. Excellent waste volume reduction was achieved.
Pilot vitrification tests of dry active waste
Eight pilot vitrification tests were carried out to determine the combustion, vitrification and off-gas generation characteristics of the DAW containing cellulose, polyethylene,
rubber, polyester, and polyvinyl chloride (PVC). One of these was a test for removal of dioxin during combustion of the DAW containing 20% PVC. The tests determined the optimum feeding rate of the DAW and the proper amount of excess oxygen. Control of the glass melt was found to be easy and the vitrified form was homogeneous without any secondary phase formation. Hazardous off-gases generated during the DAW feeding onto the glass melt were completely treated in the OGTS, and their concentrations at stack were well below the environmental regulation limits. The concentration of dioxin at stack was about 1/300 of the environmental permissible emission limit. The volume reduction factor of the DAW from the initial bulk volume was higher than 90.
Pilot vitrification tests of ion exchange materials
A total of 29 pilot vitrification tests for short-, mid- and long-term were carried out. The simulated IER containing high and low activity, and zeolite used for the test were similar to the spent resin and zeolite generated from nuclear plants. Several test parameters, such as the waste feed rate, excess oxygen amount, configuration of oxygen supply, glass melt temperature, and flow rate of bubblers were optimised to produce a homogeneous and durable vitrified form. The phase stability and leachability of the vitrified form were tested using scanning electron microscope/energy dispersive spectrometer (SEM/EDS) and several leaching methods.
A Mössbauer spectroscope was used to analyse the redox states of the glass. The IER and zeolite were successfully vitrified in the CCM and a homogeneous vitrified form was produced. Hazardous off-gases generated during the IER feeding onto the glass melt were completely treated in the OGTS and their concentrations at stack were well below the environmental regulation limits. The volume reduction factor was 21-35.
Long duration vitrification tests of mixed wastes
A total of 15 pilot vitrification tests were successfully carried out for three types of the mixed wastes containing mixtures of the DAW and HA IER, “W1” waste, and “W2” waste. Three long-duration tests were performed for each mixed waste. Two 120-hour duration tests were for a vitrified mixture of the DAW and HA IER, and W2 waste made by mixing the HA IER and W1 waste generated from Ulchin 5 & 6. One 180-hour duration test was carried out on vitrifying W1 waste made by mixing LA IER, zeolite and DAW from Ulchin 5 & 6. The combustion and melting status of each mixed waste under the operational parameters was observed while the mixed waste and glass frit were simultaneously fed onto the glass melt, to evaluate the system integrity of the pilot scale vitrification plant, and to simulate commercial operation.
Waste accumulation on the glass melt surface did not take place while the waste and glass frit were simultaneously fed into the CCM. The glass melt temperature and glass composition were constantly maintained during the whole waste feeding period, and the pilot vitrification plant showed its integrity. The concentration of hazardous gases emitted from the stack satisfied the permissible emission limits. The volume reduction factors were 33, 74, and 84 for W2 waste, mixture of the DAW and HA IER, and W1 waste, respectively.
Pilot vitrifiction tests of borated concentrate
The purpose of the pilot test of the borated concentrate was to optimise its vitrification condition by analysing the melting and dust generation characteristics of the dried borate concentrate in the melter when it was vitrified. To do this, several factors affecting the vitrification conditions were selected. Simulated borate concentrates were made by the different mole ratios for boron to sodium present in the real borated concentrate. The simulated borate concentrates were dried and pelletised to be easily fed into the CCM. Three different samples of borated concentrate were tested. Optimum melting temperature and the CCM internal pressure were determined. The volume reduction factor was about 8.
Pilot vitrification tests of sludge slurry
The sludge generated at nuclear plants contains different components, and so two different simulated sludges were made. One was similar to the sludge generated from the primary coolant loop of a reactor, and the other similar to the sludge generated from the waste treatment building. To easily feed the sludge into the CCM, the sludge was mixed with water and converted into slurry. The feed rate of the slurry affected the quality of the glass. Thus, an optimised feed rate was determined. The volume reduction factor was about 8.
Pilot vitrification tests of dust
Dust is generated as a result of the vitrification process of wastes. Such a dust contains sulfur, carbon, glass components and caesium. To minimise contamination of radioactive materials on the off-gas treatment system, radioactive dust is filtered on the HTF and then the dust is fed into the CCM. However, the sulfur in the dust may form a high conductive sulfate layer within the glass melt in the CCM, so the sulfur should be removed from the dust. To establish and assess a method of sulfur removal, one test was carried out and an assessment of vaporisation of caesium present in the dust was also carried out. To prevent transforming sulfur in the dust into sulfate, activated carbon was mixed with the dust as a reduction agent. The test was performed using different mole ratios of carbon to sulfur without a supply of oxygen. The granulated dust was fed onto the glass melt and vitrified without any secondary phase formation. The retention of caesium was also assessed and it was observed that more than 90% of caesium present in the dust was incorporated into the glass matrix.
Pilot melting tests of non-combustible wastes
Twenty-one melting tests were performed using the pilot batch-type plasma torch melter. Test points were changing a melting period, melting test under continuous feeding of wastes, a test of the heat transfer model for plasma torch melter, an assessment test of operational mode characteristics of plasma torch and caesium volatilisation due to its power change. The average volume reduction factor was about 3.
Glass formulation study
This study determined the glass formulation that could be applicable to both treatments of a single stream of radwaste and mixed radwaste generated from Korean reactors. To determine such glass formulations, the inorganic materials present in the wastes were analysed and the extent of their loading into the glass matrix during vitrification of the inorganic materials analysed. The glass formulation with the viscosity and the electrical conductivity appropriate to operation of the CCM was determined using GlassForm 1.1, a computer code jointly developed by KHNP and the Idaho National Engineering and Environmental Laboratory. The candidate glass was selected after an experimental verification of the glass formulation determined by the computer code. The waste characteristics of the glass formulations were verified by the pilot vitrification tests.
The leachability characteristics of the candidate glasses and vitrified forms produced through the pilot vitrification tests were analysed. The results showed that the vitrified forms had a lower leachability than the standard glass used in assessment of radioactive waste vitrification in the USA (Figure 3). The compressive strength of the vitrified forms produced in the pilot vitrification tests was up to 90 times higher than the 500psi required by the US Nuclear Regulatory Commission.
Safety and economic assessment
To assess the safety of the vitrification plant, data obtained from operational experience of the pilot vitrification plant was analysed and the safety assessment strategy prepared using the data of the relevant facilities as references. A present value analysis method was used to assess the economics of the vitrification plant. The economics of the vitrification plant were analysed assuming it was constructed within an existing nuclear site operating four 1000MWe PWR units.
A sensitivity analysis on construction cost of the vitrification plant, type of radwaste, volume reduction factor, and disposal cost was carried out.
To simulate the behaviour inside the CCM, the penetration depth, distribution of current, heat transfer, and bubbling were modeled, and particle image velocimetry (PIV) was adopted to verify the effect of bubbling and analyse the unsteady flow field. The effect of the bubbling and temperature distribution could be analysed from the results of the experiment on the bubbling effect. The experimental data was applied to analyse the heat transfer. A simulation code was developed, which included an effect of bubbling, characteristics of frequency affecting the heat transfer of the CCM, the fixed shapes of the CCM, heat characteristics of melting glass, and electromagnetic field.
The first LILW commercial vitrification plant
Every country operating or interested in operating nuclear plants has to consider how to resolve the problem of LILW.
A major issue is environmental safety and amount of LILW produced. Such an issue could be solved by the LILW vitrification technology we have developed. The IAEA has decided to support the planned construction of a commercial LILW vitrification plant at Ulchin, which will be built within the radwaste building at the Ulchin site. The capacities of the CCM and PTM in the plant are 300kW and 500kW, respectively.
The design of the vitrification plant will be started in 2003, and the procurement and construction will be completed by mid 2005, with commercial operation due in 2007. Several R&D activities to develop a universal glass canister for the LILW vitrified form, to develop an in situ glass composition measurement system (laser induced breakdown spectrometer), and to develop the advanced CCM simulation code will be performed in parallel with the design, construction and commissioning. Figure 4 shows the commercialisation milestones, including the design and construction, commissioning, and tests of the commercial LILW vitrification plant and related R&D.
It is expected that the commercial vitrification plant will have sufficient capacity to vitrify all LILW generated from six PWR units, each of 1000MWe. Accordingly, it is expected that the vitrification technology will not only enhance the safety of the waste disposal repository, but will also greatly contribute to the further promotion of Korea’s nuclear power generation programme.
TablesStatus of radioactive waste storage as of August 2002 (unit: 200 litre drum)
