초록 ▼

1. 연구개발 목표 및 내용
대형 원자력시설 제염해체 주요 핵심기술 개발 및 환경복원 기술 기반 구축

2. 연구결과
• 원자력시설 고도 제염기술 개발
- 제염에 의한 이차폐기물 발생량 저감 목적의 독창적인 원전 일차계통 무착화성 화학제염 공정(HYBRID & HYBRID-D) 및 대면적 나노입자 함유 복합유체 제염기술의 개발
• 해체공정 통합평가 및 원격 제어기술 개발
- 최적의 해체공정을 도출할 수 있는 해체 시뮬레이터 개발 및 현장 접근이 불가능한 고방사능 시설 해체작업을 원격으로 정밀하게

1. 연구개발 목표 및 내용
대형 원자력시설 제염해체 주요 핵심기술 개발 및 환경복원 기술 기반 구축

2. 연구결과
• 원자력시설 고도 제염기술 개발
- 제염에 의한 이차폐기물 발생량 저감 목적의 독창적인 원전 일차계통 무착화성 화학제염 공정(HYBRID & HYBRID-D) 및 대면적 나노입자 함유 복합유체 제염기술의 개발
• 해체공정 통합평가 및 원격 제어기술 개발
- 최적의 해체공정을 도출할 수 있는 해체 시뮬레이터 개발 및 현장 접근이 불가능한 고방사능 시설 해체작업을 원격으로 정밀하게 제어하는 요소 기술 개발
• 해체특수폐기물 처리 기술개발
- 원전 중대사고 시 대용량/고방사성 발생 폐액에 대한 신개념 처리공정의 도출 및, 난처리성 알파유기혼성/우라늄폐기물의 친환경적/혁신적 부피감용 기술을 확보
• 해체 및 오염부지 환경복원 기술 개발
- 해체 부지의 방사선적 오염분포의 현장 측정기술 개발 및 중대사고 초기 오염의 확산을 신속하게 억제할 수 있는 확산 억제용 소재의 개발

(출처 : 보고서 요약서 4p)

Abstract ▼

Ⅳ. Results and Proposal for Application

1. Development of advanced decontamination technology for nuclear facilities
○ Development of chelating reagent-free chemical decontamination system for the primary coolant system of nuclear power plants
- A new chemical decontamination agent with

Ⅳ. Results and Proposal for Application

1. Development of advanced decontamination technology for nuclear facilities
○ Development of chelating reagent-free chemical decontamination system for the primary coolant system of nuclear power plants
- A new chemical decontamination agent with no use of organic acid or organic chelating agents has been developed in KAERI. The chemical decontamination agent is composed of hydrazine as a reducing agent, metal ion of Cu²⁺/Cu⁺ as a catalyst and inorganic acids such as HNO₃ or H₂SO₄ as a H⁺ ion donator. This agent was named HYBRID(HYdrazine Based Reductive metal Ion Decontamination) which is the attribute of reducing decontamination agent. With the series of dissolution experiments form magnetite powder, metal oxides (e.g. type 304 stainless steel and Inconel 600) produced under the condition simulating the pressure and temperature of operating reactors and radioactive specimens extracted from operating commercial or research reactor, the dissolution performance of metal oxides and decontamination capability of radionuclides from the specimens have been proven and the optimal composition of chemical agents was derived. Especially the decontamination factor obtained from the FTL(Fuel Test Loop of HANARO research reactor) was exceeded 200, which demonstrated the decontamination performance.
- HYBRID decontamination agent was devised to generate less amount of the secondary waste due to the total decomposition of hydrazine into nitrogen and water. To prove this concept, we treated decontamination liquid waste by adding hydrogen peroxide continually not all at once, and confirmed the total decomposition was reached to 100% when the addition of hydrogen peroxide was 30% to the volume of liquid waste.
- Besides, HYBRID decontamination agents was claimed highly resistive against the radioactivity which must be inevitably experienced during the decontamination around the reactor vessel. Comparison with other agents used in commercial decontamination technologies such as CAN-DEREM or CORD under 10,000 Gy radioactivity supported that HYBRID agents is four fold resistive than other methods with the lowest 6.8 % decomposition but 33 and 45 % for other agents of EDTA and oxalic acid respectively.

○ Development of foam decontamination technology containing nanoparticles for large components of nuclear facilities
- Decyl glucoside having a higher foam stability was selected as main surfactant components compared to conventional nonionic surfactant, and mesoporous silica nanoparticles was determined for stabilizer among various silica nanoparticle structures such as mesoporous, core-shell and non-porous structures to increase foam stability. The mesoporous silica nanoparticles prepared by HCl-EtOH solvent extraction were developed by KAERI, featuring to the excellent dispersion and foam stability. The foam decontamination agents containing KAERI silica nanoparticles, which has 20% higher foam stability and decontamination efficiency than that of conventional nanoparticles based foaming agents, show high DF over 10 from the demonstration test using contaminated components in KAERI Research Reactor. From above R&D project, the fundamental results concerning the foam decontamination has been established to apply the large components of nuclear facilities.

2. Development of the integrated assessment system and remote control technology for decommissioning process
○ Development of tangible decommissioning simulator technology
- 3D models from 2D drawings of Kori-1 were manufactured using expert knowledge over nuclear facilities and 3D digital mock-up with functions of geometric and kinematic analysis was constructed to investigate geometric structures of decommissioning environment and kinematic motions of decommissioning processes. Database of physical and radiological characteristic evaluation over highly radioactive and heavy components of the nuclear power plant (Reactor Pressure Vessel, Steam Generator, Reactor Coolant Pump, Pressurizer) was constructed and spatial dose mapping around nuclear facilities using MCNP(Monte Carlo N-Particle) code was achieved. The core technical modules of decommissioning process simulation and the integrated decommissioning assessment process were organized through the decommissioning scenario analysis of nuclear facilities. The platform of software to develop the software frame of the integrated decommissioning assessment process was selected based on requirements of software frame.
- The dismantling process simulation module of remarkably enhanced reliability over multiple and repetitive cutting operations using the CAD kernel was developed. The user interface module minimizing mental transformation of the operator with the virtual fixture was developed. The visualization module for spatial dose calculated by MCNP code were developed. The software frame to satisfy requirements for the integrated decommissioning assessment process was developed by CAA C⁺⁺ extended development tool.
- The digital mock-up of the seamless remote dismantling system was developed to handle the whole decommissioning process over the heavy nuclear facilities without replacement of dismantling equipment by the CAD-based dismantling process simulation technology. The remote dismantling system technology capable of reducing considerable time and cost for decommissioning process was achieved. The tangible decommissioning simulator was developed to simulate decommissioning process of heavy nuclear facilities in real time and to train a remote operator by using the tangible remote control system capable of force reflection and immersive visualization system.

○ Development of the safety assessment technology for decommissioning of nuclear facilities
- The methodology of safety assessment for decommissioning accident from human errors was developed. Through analyzing the state-of-the-art technologies, requirements of the safety assessment were established and procedure of the safety assessment was established. The procedure consists of assessment step the safety of decommissioning scenarios, counter-measuring step for safety reduction and assuring step of the decommissioning safety. The measurement equipment for characteristics of decommissioning worker’s behavior was manufactured. The manufacturing criteria for capable of generating the input data and linking with the integrated assessment system was designed. The measurement equipment under selection of the best hardware and software tools was developed.
- The model of the human error assessment for decommissioning was developed. The mathematical model capable of assessing both working environments and human errors was developed. The safety assessment of working environments can be quantitatively estimated based on both dose distribution and working time. The safety assessment of human errors can be quantitatively estimated based on accidents, working time, and counter-measuring time.
- The preventing technology of reducing human errors was developed. The safety improvement technology was developed to reduce the hazards of human errors. The safety assessment system for decommissioning of nuclear facilities was developed. The safety assessment system was able to measure and assess the working scenarios under virtual environments like real decommissioning situations.

○ Development of heavy-duty manipulator and remote control technology
- The technology requirements of remote dismantling manipulator system were established. The conceptual design of the remote dismantling manipulator system was performed by using 3D digital mock-up in terms of workability and kinematics under the given geometric conditions.
- Original two electric actuator modules and three hydraulic actuator modules were developed for modularization design of remote dismantling manipulator system. To supply hydraulic power to hydraulic actuator modules, variable hydraulic supply device that can variably supply pressure and flow was developed. High-precision control algorithm that makes a precise control of hydraulic actuator module was also developed.
- To evaluate the developed hydraulic actuator modules, an assessment device was manufactured and a signal processing controller for handling various sensor signals and control inputs was developed. And a variable loading device that can supply variable load was manufactured.
- The kinematic parameters of remote dismantling manipulator were also defined through analyzing the working space and kinematically simulating decommissioning process. Various linear and rotary actuator modules were designed for making a dismantling manipulator to have modularity. Based on the defined kinematic parameters and designed actuator modules, the detailed design of a dismantling manipulator was carried out through dynamic analysis and stress analysis.
- To develop various dismantling manipulators for a nuclear power plant, two hollow-type electrical actuator modules that make a possible to insert power and signal cable into the actuator modules were developed for preventing the exposure of cables. And a small and high-torque electric-hydraulic actuator module was developed to make light-weight dismantling manipulator.
- A robust control algorithm for reducing a vibration and precise control of hydraulic system was developed using a frequency-shaped integral sliding surface and time delay control with switching action. The design of a remote control system for controlling a dismantling manipulator was carried out through establishment of the technology requirement of the remote control system.

3. Development of the decommissioning waste treatment technologies
3-1. Development of treatment technology for the high liquid radwastes
○ Elemental and composition analysis for massive high-level liquid radwaste
- It was confirmed that the major high-level radionuclides dissolved in high salt-laden waste seawater were Cs, Sr, and I etc. It was derived from the analyses of nuclides’solubilities in irradiated nuclear fuel and their forms and chemical species in loading nuclear fuel. Also, the reported data of elemental composition of radioactive wastewater generated from the Fukushima accident was analyzed. Studies on adsorption and coagulation/flocculation treatment using inorganic materials (zeolite etc.) were performed for the research scope to develop the treatment technology for the massive high-level radioactive wastewater.

○ Evaluation of adsorbent selection for removal of high radioactive nuclides
- The selection standard of adsorbents or precipitants for high radioactive liquid waste was established. Overall 14 candidates of adsorbents or precipitants for removal of Cs and Sr from the high radioactive/high salt-laden wastewater were tested. For the Cs removal, IE911 (crystalline silicotitanate, CST), AW500/IE96 (Chabazite type zeolite) as adsorbents and K₄Fe(CN)₆, NaTPB as precipitants were efficient. For the Sr removal, 4A zeolite and BaSO₄-Sr isomorphous precipitation were efficient.

○ Removal of high radioactive nuclides (¹³⁷Cs, ⁹⁰Sr, ¹³¹I) from a massive liquid radwaste
- Based on the results from part 2, the specific evaluation for removal of high-level radionuclides, such as Cs (by IE911 and AW500), Sr (by 4A and Ba impregnated A zeolite (BaA)), and I (by Ag impregnated acidic active carbon (AgAAC) and basic active alumina (AgBAA)), was performed to understand the effects of V/m, pH of solution, shaking speed, adsorbent size, and solution salinity etc. with respects to the removal efficiencies of the target radionuclides. In this study, IE911, BaA, and AgBAA were evaluated to be effective adsorbents for the removal of Cs, Sr, and I, respectively. The radionuclides (Cs, Sr, and I) were removed more than 99 %.

○ Evaluation of the adsorption/precipitation efficiency using radioisotope
- In the radioactive wastewater generated from the severe nuclear accident similar to Fukushima accident, the concentration of Cs is less than 1 ppm, Sr does not exist in the presence of Ba and sulfate ions in solution, but gradually can increase up to about 0.5 ppm in solution by desalination of the solution. Therefore, the adsorption/precipitation efficiencies of the selected adsorbents, such as IE911, IE96, AW500 for Cs removal and 4A, BaA for Sr removal were evaluated by using radioisotopes, ¹³⁷Cs and ⁹⁰Sr, in the seawater conditions. The results confirmed that there was slight difference between the results from non-radioisotopic and radioisotopic experiments, in within an error range. More importantly, radioisotopic experiments are efficient and easy in terms of characterization.

○ Enhancement of treatment technology of massive radioactive wastewater
- For enhancement of treatment technology of massive high-level radioactive wastewater, Potassium Cobalt Ferrocyanide (PCFC) was introduced, and its adsorption properties in views of variables such as V/m ratio, pH of solution, adsorbent aging time, and solution salinity was evaluated, and its established optimal conditions was suggested.
- As an optimal material for Cs removal, a CHA-PCFC hybrid material having advantages of both CHA and PCFC was selected. Then, the optimal synthesis conditions for CHA-PCFC were derived through comparison with conventional Cs adsorbents. CHA-PCFC had decontamination factor (DF) higher than 10³, and it showed very fast Cs adsorption kinetics. Also, BaA was finally selected as the optimal remover for Sr in the suggested process, which was demonstrated by comparative experiments on Sr removal in the simulated condition. The synthesis theory and Sr removal mechanism were derived from the combined analyses and adsorption/desorption experiments using BaA.

○ Complex flocculation treatment of residual radionuclides
- To remove the residual radionuclides remaining after removing Cs, Sr, and I from radioactive wastewater generated, a coprecipitation-flocculation system using ferric hydroxide-anionic PAA was selected. It was demonstrated that ferric hydroxide-anionic PAA system is efficient, because it had DF higher than 100 and precipitated 99% of formed flocs within 5 minutes. The mechanism of coprecipitation-flocculation using Fe-PAA was studied for the first time, which elucidated that the extremely-small amount of target nuclide ions was removed through co-precipitation action by ferric hydroxide at pH 8., that the PAA organic flocculent made fine ferric co-precipitate particles grow in large flocs, and that the free PAA molecules were bound with the metal ions of opposite charge in solution, but they did not form flocs. The optimal amount of anionic PAA was ranging from 0.01 to 0.05 g/L, when 100 ppm of ferric ion was added in solution. Even if there was less than 5 ppm of ferric ion, it was effective enough to remove 99% of target nuclides.

○ Suggestion of a process combined by several unit operations for treatment of massive high-level radioactive wastewater and its evaluation
- The emergency operation system concept using sequential precipitation process suggested in this study was confirmed to effectively remove the nuclides in the radioactive wastewater generated from severe nuclear accident. Based on the several evaluations on that, a conclusion was that if such a system, to be rapidly installed after accident or prepared prior to the accident, is effectively operated, and then recycles the treated water into the reactors, it could remove many secondary problems such as installation and management of many storage tanks with high possibility of wastewater leakage from storage facilities into environment, and exposure to radiation workers to disentangle the accident, which consequently lead to huge capacity of normal wastewater treatment facility to be installed. Such a concept and method could be helpful to get better public acceptance in operating nuclear power plants as a safety measurement against a severe nuclear accident like Fukushima nuclear disaster.

3-2. Development of treatment technology for the organic mixed wastes
○ Establishment of an engineering model of integrated pyrolysis/steam reforming system for the treatment of α-bearing organic mixed waste
- A low-temperature pyrolysis process equipped with a two-stage non-flame oxidation system was proposed as an appropriate thermal treatment process for the target spent TBP. A proposed process is mainly consisted of a low-temperature pyrolyser, a steam reformer, and a catalytic oxidizer in series. The main function of each key process unit is as follows:
∙ Low-temperature pyrolyser has a function of pyrolysis of spent uranium-bearing TBP to convert organic constituents into gas while uranium and phosphate remains together with other inorganic constituents.
∙ The steam reformer has two different functions. One is partial oxidation of unburned hydrocarbons (UHCs) from the pyrolyser and the other is the reduction of nitrous oxide. A relatively small amount of N₂O can be converted into nitrogen and oxygen in the presence of a larger amount of H₂ and CO, both of which normally exist significantly in a hydrocarbon steam reformer.
∙ The catalytic oxidizer burns out UHCs and carbon monoxide from the steam reformer to meet emission standards.
○ Study of the design optimization of a bench-scale pyrolysis/steam reforming system
- The maximum pyrolysis temperature that guarantees without sublimation of phosphorus oxides was determined as 257 ℃ and the rate of organics destruction was relatively very slow at this lowered pyrolysis temperature. The required residence time for a substantial gasification of organic in the spent solvents at 257 ℃ was determined as 5 hours. At this low-temperature pyrolysis condition, butanol, butanal and butene generated from the thermal decomposition of TBP and volatilized gaseous dodecane are major pyrolysis gas species. Based on that, the design basis for the low-temperature pyrolysis process was established as follows:
∙ Pyrolysis temperature should be maintained at 257 ℃.
∙ Residence time of spent solvent in the pyrolysis chamber should be longer than 5 hours.
∙ Off-gas system should substantially oxidize unburned UHCs and decompose.

○ Installation of a bench-scale pyrolysis/steam reforming process units
- A bench-scale low-temperature pyrolysis/steam reforming process with a capacity of 1 kg spent TBP per hour was designed, manufactured and installed according to the determined optimum conditions of key process units. Required subsystems for the hot demonstration tests using actual uranium-bearing spent TBP, such as ventilation system, radiation monitoring system, and off-gas emission monitoring system were additionally installed to achieve a licence to treat real α-bearing spent solvent.
○ Pyrolysis and steam reforming tests using simulated spent solvents and optimization study
- After performance tests using inactive simulated spent solvents with an organic composition same as actual uranium bearing spent TBP, an advanced catalyst can be used in a high-temperature steam reformer for the treatment of low-temperature pyrolysis. The thermal stability of a rhodium-based steam reforming catalyst was greatly enhanced by adding cerium. The addition of cerium leaded to a greatly enhanced dispersion of rhodium nanoparticles and no agglomeration of rhodium was found for the samples heated even at 800℃. The XPS spectra, XRD patterns, TEM image, and TPR curves of ceria-promoted rhodium catalyst samples showed that an addition of ceria retarded the reduction of rhodium oxides at high temperature up to 800℃ and this enhanced the dispersion of rhodium nanoparticles with 15-30 nm in size. The results suggested that if the rhodium catalysts are thermally stabilized by the addition of cerium oxide, they can be properly used as a steam reforming catalyst at high temperatures up to 900℃.

○ Optimization of operating conditions of bench-scale system and demonstration of the process performance for the treatment of uranium-bearing spent TBP
- Optimum condition of steam reformer with 10 cm ID and 50 cm length was determined when Φ_H2O is 0.8 (25 % excess steam feed rate) and the temperature is 827℃. At this condition, high concentrations of dodecane and butene from the low-temperature pyrolyser will be substantially decomposed into smaller hydrocarbons without too much increase of gas volume. To meet the emission standard of the THCs (100 ppm), gas residence time of at least about 0.45 seconds at an air equivalence ratio of 0.5 (100 % of excess air) and temperature of 477℃. The optimum condition of catalytic oxidizer was determined as temperature of 477℃, an air equivalence ratio of 0.5 and a set of four square-column catalytic oxidizers with a 15 cm in width and 5 cm in height. Based on the optimized operating conditions, we successfully demonstrated the improved performance of phosphate retention in the pyrolysis residue and that of destruction and oxidation of organic constituents in the actual uranium-bearing spent TBP under the optimized conditions.
○ Study on the treatment method of pyrolysis residue containing uranium
- Residue from the pyrolysis of uranium-bearing spent TBP contains UP₄O₁₂, UP₂O₇ as well as a small amount of pyro-carbon and phosphorus oxide (P₄O₁₀). Based on the experimental investigation, a simple treatment process for pyrolysis residue to separate uranium pyrophosphate from pyro-carbon and phosphorus oxide was developed and demonstrated as follows; First, the residue was dissolved in water to remove water soluble P₄O₁₀ and then it was heated at high temperature in the presence of air to remove the pyrocarbon in the residue. After this, it was confirmed that we can obtain maximum volume reduction and mass reduction by remaining secondary radwaste just in the form of uranium pyrophsophate.

3-3. Development of treatment technology for the uranium complex
○ Evaluation of applicability of unit technologies used in the processes for treatment of uranium complex wastes, and suggestion of target-specific processes
- To secure the technology for treatment of uranium complex waste, two representative uranium complex wastes of heat-treated uranium sludge generated from KAERI and spent uranium catalyst generated from a domestic private company were selected. Then the physicochemical characteristics of the uranium complex wastes were evaluated, and then a series of unit technologies required for their treatment were devised and tested. Thereafter, processes customized according to the uranium complex wastes were finally suggested.
- The dissolution efficiencies of the thermally decomposed uranium sludge wastes were evaluated by several solutions of acids and bases. For efficient separation of uranium from dissolved solution, a sequential application of alkalization by using sodium carbonate with hydrogen peroxide and acidification by sodium nitrate was performed. More than 99 % of uranium was recovered as a form of UO₄∙4H₂O precipitate in the process. However, the UO₄ was precipitated together with Fe at pH >3, which requires the additional purification step of UO₄. From the various experimental results, an optimal process to treat the uranium thermal decomposed sludge was suggested.
- To develop a process for treatment of spent uranium catalyst, a new concept of acid-base swing dissolution for dissolution of spent uranium catalyst, where the supporter, Si, of the catalyst and uranium were sequentially dissolved in alkaline and acidic solutions, was introduced. A few unit steps to treat the wastewater and Si precipitate generated from the dissolution were devised to minimize the final secondary waste, which were coprecipitation of Fe and U, precipitation and purification of Si, removal of residual U, recovery of used acid and base, and volume reduction of precipitated solids by calcination. An optimal process to treat the spent uranium catalyst was suggested.

○ Evaluation of stability of uranium peroxo compound
- Studies on the behavior characteristics and stability of UO₂(O₂)(CO₂)₂^4-and UO₄ in solution, which are generated as intermediate or final products in the processes to treat uranium waste, were carried out in several ways including Raman spectroscopy. The uranyl peroxo carbonato complex ion was confirmed to be self-decomposed faster to uranyl tris-carbonato complex ion in the carbonate solution with temperature. and its decomposition rate equation was obtained In Na-U(VI)-CO₃-OH-H₂O₂ solution system, UO₂(O₂)(CO₂)₂^4- species existed, along with three other uranium species thought to be UO₂(O₂)₂^2-, UO₂(CO₃)₃^4-, and UO₂(CO₃)_x(OH)_y^z-. The UO₂(O₂)(CO₃)₂^4- species was not formed in the bicarbonate condition, and it could exist in only a narrow pH range from approximately 9 to 12. When the pH of the carbonate solution increased further, the UO₂(O₂)(CO₃)₂^4- species changed to UO₂(CO₃)₃^4- and UO₂(CO₃)_x(OH)_y^z- species. These results will be used to advance the process used for treatment of uranium waste. Uranyl peroxide was stable in distilled water at elevated temperatures, but dissolved other solutions at temperatures higher than 40 ℃, and a greater amount of uranyl peroxide dissolved in more acidic conditions at elevated temperatures. Uranyl ions that dissolved from uranyl peroxide were able to be recovered as uranyl peroxide in the solution where the dissolution occurs by adding hydrogen peroxide to the solution.

○ Sophistication of process for treatment of uranium complex waste and evaluation of optimal processes by the uranium complex waste characteristics
- Studied on minimization of secondary waste generated from the processes for treatment of uranium complex wastes were performed to enhance the environment-friendly characteristics of the processes and their unit operations. For them, removal of residual uranium in the wastewater generated from various forms of uranium waste by biosorbent materials was evaluated, and electro-dialysis efficiency to recover acid and base from effluent of the process was evaluated according to its configuration way of ED(electrodialysis) and EDBM (electrodialysis with bipolar membranes) and the cell resistance components affecting ED operation was analyzed.
- The uranium adsorption efficiency of non-living biomass of brown algae was evaluated in various adsorption experimental conditions in order to develop a technology for treatment of uranium liquid waste generated from the previous processes. Several different sizes of biomass were prepared using pretreatment and surface-modification steps. For the optimal particle size, contact time, and injection amount for the stable operation of the liquid waste treatment process were determined. From the results of comparative experiments using the biosorbents and other chemical adsorbents/precipitants, it was demonstrated that the brown algae biosorbent could replace the conventional chemicals for uranium removal. As a post-treatment for the final solid waste reduction, the ignition treatment could significantly reduce the weight of waste biosorbents. In conclusion, the brown algae biosorbent is shown to be a favorable adsorbent for uranium removal from liquid waste.
- The electrodialysis extent for HNO₃ and NaOH recovery from NaNO₃ solution were almost proportional to the total amount of electricity supplied to the system regardless of the operation mode and the electrodialysis systems. For the treatment of the same volume of feed solution, the energy consumption and current efficiency were depending on the operation mode and the electro-dialysis system. In both the ED and EDBM systems, the conductivity of the feed solution was evaluated to strongly affected the overall cell resistance after approximately 50 % of the ions in the feed solution had migrated.

○ Evaluation of optimal processes by the uranium complex waste characteristics
- Based on the suggested target-specific treatment process for the uranium complex wastes, such as thermally-treated uranium sludge and spent uranium catalyst, waste reduction efficiencies (%) from each step and final stage were evaluated. In the case of thermally-treated uranium sludge waste, 80.5 % of final weight reduction was achieved by the application of acid-base swing dissolution, and UO₄ purification. In the case of spent uranium catalyst, finally 76.7 % of weight reduction was achieved by the combination of unit technologies including acid-base swing dissolution, Si precipitant purification, and thermal treatment of secondary solid waste etc. It was equivalent to 89.5 % of final volume reduction of radioactive waste based on the tap density of generated solids.


4. Development of Site Remediation Technology for Decommissioning and Contaminated Site
4-1. Development of in site measurement technology for final status survey of decommissioning site
○ Evaluation technology development of representative samples for site survey
- Evaluation methods of a systemic survey and representative samples are developed to evaluate a residual radioactivity using a spatial analysis method of geostatistics on decommissioning site. The improved method from MARSSIM is established to evaluate a representative sample of final status survey using Kriging and simulation of spatial analysis in contaminated site. The developed method can reduce required numbers of samples up to 35% during the final status survey in Uranium Conversion Plant

○ Development of low level residual radioactivity measurement technology
- The low level measurement system is designed and constructed to improve the reliability by reducing the background using a compton suppression anti-coincidence system. The new concept detector with four block type guard scintillator is developed in order to improve an applicability of movement and assembly in fields. The results of optimized guard detector position and size are placing the guard detector in the front and rear of 4 cm and extension of guard detector size 20 ㎝ (L) × 20 ㎝ (W) × 11 ㎝ (H) (suppression ratio 2.89). The developed system can reduce the 50% of minimum detectable concentration and 1/2 of detector weight compared to the commercial detector.

○ Development of measurement technology for radioactivity depth distribution
- The diffusion model of radioactivity depth distribution is established using evaluation results of Chernobyl and Fukushima site. In situ measurement technology is developed by analyzing the corelation of Q(peak to valley ratios) values from detected spectrum with a depth distribution in a contaminated site. The experimental depth distribution models for in situ measurement in laboratory are constructed based on the Fukushima site survey results (exponential depth distribution(β) from 0.25 to 2.0. The evaluation program using a peak to valley method is developed to reduce the uncertainty of depth distributions by reducing background contribution(Bi-214), density and composition correction factors of soil. The in situ measurement technology is applied for undisturbed Jeju site, the radioactivity distributions with a exponential depth distribution(β = 0.67) are compared with in situ measurement results. The comparison results between sampling and in situ analysis show a depth distribution error was lower than 1% and initial activity error was 8% respectively.

○ Development of mobile radiation survey technology in contaminated site
- The mobile residual radioactivity measurement system is developed to reduce the cost and time for surveying large contaminated area. The system consists of NaI detector (3 × 3 inch), GPS sensors, digital signal processor, power supply, and data transmission module for long distance. The survey conditions, which are optimized to detect hot spots, are evaluated during the final state survey. The optimized scan survey condition, which meets the release criteria and site investigation goals (10 ha/day), is a scanning height lower than 50 cm from ground and a slower scan speed than 5 km/h. We could confirmed the excellent detection efficiency from the cross validation results in decommissioning KRR site and Jeju site.

4-2. Development of site remediation technology in contaminated sites
○ Development of soil fixation technology for suppression of contaminant spreading
- Properties of contaminated areas from nuclear accidents and the advantages and disadvantages of applied remediation technologies were analyzed to develop suitable and effective remediation technologies on contaminated sites. The base data were also produced to develop optimal fixation and remediation technologies through the analysis of weather environments around the national nuclear power plants and properties of contaminated soil. Polymer materials were widely used as fixation materials to form a polymer barrier or to bind soil particles through formation of polymer complexes having strong binding force between the polymers. Polyelectrolytes were selected as a raw materials because they are suitable for mass production and easy disposal of secondary waste materials.
- The preparation condition of fixation materials were determined through analysis of the preparation factor such as, pH, water contents, and salt concentration. After optimization of preparation condition, the fabrication steps were simplified, and the fabrication time were also shortened. The structural stability and long-term integrity (larger than 1 month) of fixed soil were secured from the evaluation of fixation ability by changing the salt concentration and concentration and molecular weight of polyelectrolyte polymer. The separation method between fixation materials and soil were additionally determined through investigation of adsorption property of fixation materials with the soil surface, micro-structure of fixed soil, and interaction between soil and polyelectrolyte complex. Additionally, we developed the new natural polymer based fixation materials which complement the disadvantages of polyelecryolyte complex based fixation materials such as water erosion found in the study of long-term integrity and we also confirmed an applicability of the developed fixation materials through the performance test using the developed spray equipment.

○ Development of fabrication technology of contaminated surface remediation materials in residential areas
- Hydrogel/absorbent composite materials for the remediation of contaminated surface were basically designed and established the concept for effective surface remediation. The hydrogel/absorbent composite materials are consist of natural polymer (alginate) based hydrophlic hydrogel and nano-absorbent embedded in the hydrogel. The metal-ferrocyanide which can selectively remove the radioactive cesium were coated on the surface of mangetic nanoparticles to prepare the magnetic absorbent. As an first method, a silane coupling agent was used as a mediating materials. The wt% content of Cu in the absorbent was calculated to be 0.98 wt% and the magnetic absorbent has a good removal efficiency that exceeds 98% of Cs-137 (17.75 Bq/g). As a second method, polyethyleneimine (PEI) was used as a mediating materials. The wt% content of Cu in the absorbent was calculated to be 1.54 wt%. The removal efficiency of magnetic absorbent for Cs-137 was 99.59% at 50.0 Bq/g. As an third method, polyvinylpyrrolidone (PVP) was used as a mediating materials. The wt% content of Cu in the absorbent was calculated to be 2.40 wt%. The removal efficiency of magnetic absorbent for Cs-137 was 99.41% at 161.0 Bq/g. As an forth method, iron-ferrocyanide was directly synthesized on the surface of magnetic nanocluster under acid condition. The wt% content of Fe-ferrocyanide were incerased upto 50.35 wt%. the magnetic absorbent displayed excellent removal efficiency (99.76% at 63.0 Bq/g).
- The preparation method of alginate based hydrogel/absorbent composite materials was determined via an analysis of formation condition and correlation of ionic complex network structure by changing the formation factor, such as polymer contents and concentration of Ca ion. The maximum value of decontamination factor (DF) for removal of radioactive cesium on the surface of paint coated cement was calculated to be 10.5. As a second hydrogel/absorbent composite materials, we developed the stimuli sensitive hydrogel/absorbent composite materials. The stimuli sensitive hydrogel/absorbent composite materials having semi-IPN structure are viscous and easily removed from the contaminated surface via morphology change to film through the formation of Ca-alginate after addition of Ca ion (external stimulus). The maximum value of DF was measured to be 28.84. We also developed a method to reduce waste products through magnetic separation of absorbent from the hydrogel/absorbent composite materials after breaking cross-linkers in the hydrogel using a Ca chelating agent.

(출처 : SUMMARY 33p)

목차 Contents

• 표지 ... 1
• 제 출 문 ... 3
• 보고서 요약서 ... 4
• 요 약 문 ... 5
• SUMMARY ... 27
• CONTENTS ... 54
• 목차 ... 56
• 표목차 ... 60
• 그림목차 ... 64
• 제 1 장 연구개발과제의 개요 ... 77
• 제 1 절 연구배경 ... 77
• 1. 원자력시설 고도 제염기술 개발 ... 77
• 2. 해체공정 통합평가 및 원격 제어 기술 개발 ... 79
• 3. 해체 특수 폐기물 처리 기술 개발 ... 79
• 4. 해체 및 오염부지 환경복원 기술 개발 ... 80
• 제 2 절 추진현황 및 연구내용 ... 82
• 1. 원자력시설 고도 제염기술 개발 ... 82
• 2. 해체공정 통합평가 및 원격 제어 기술 개발 ... 84
• 3. 해체 특수 폐기물 처리 기술 개발 ... 87
• 4. 해체 및 오염부지 환경복원 기술 개발 ... 89
• 제 2 장 국내외 기술개발 현황 ... 93
• 제 1 절 원자력시설 고도 제염기술 개발 ... 93
• 1. 원전 일차계통 화학제염 기술 개발 ... 93
• 2. 대형기기 나노복합유체 제염 기술 개발 ... 94
• 제 2 절 해체공정 통합평가 및 원격 제어 기술 개발 ... 96
• 1. 실감형 해체 시뮬레이터 기술 개발 ... 96
• 2. 원자력시설 해체 안전성 평가 기술 개발 ... 96
• 3. 고하중 취급 원격 제어기술 개발 ... 96
• 제 3 절 해체 특수 폐기물 처리 기술 개발 ... 97
• 1. 대용량 고방사성 폐액 처리기술 개발 ... 97
• 2. 유기혼성 폐기물 처리기술 개발 ... 99
• 3. 우라늄 복합폐기물 처리 처리기술 개발 ... 100
• 제 4 절 해체 및 오염부지 환경복원 기술 개발 ... 102
• 1. 해체 부지 규제해제 측정 및 평가기술 개발 ... 102
• 2. 오염 부지 환경 복원 기술 개발 ... 102
• 제 3 장 연구개발수행 내용 및 결과 ... 105
• 제 1 절 원자력시설 고도 제염기술 개발 ... 105
• 1. 원자력시설 일차계통 화학제염 기술 개발 ... 105
• 2. 대형기기 나노복합유체 기술 개발 ... 134
• 제 2 절 해체공정 통합평가 및 원격 제어기술 개발 ... 161
• 1. 실감형 해체 시뮬레이터 기술 개발 ... 161
• 2. 원자력시설 해체 안전성 평가 기술 개발 ... 179
• 3. 고하중 취급 원격 제어기술 개발 ... 195
• 제 3 절 해체 특수 폐기물 처리 기술 개발 ... 223
• 1. 대용량 고방사성 폐액 처리기술 개발 ... 223
• 2. 유기혼성폐기물 처리기술 개발 ... 270
• 3. 우라늄복합폐기물 처리기술 개발 ... 285
• 제 4 절 해체 및 오염부지 환경복원 기술 개발 ... 315
• 1. 해체 부지 규제해제 측정 및 평가기술 개발 ... 315
• 2. 오염 부지 환경 복원 기술 개발 ... 335
• 제 4 장 목표달성도 및 관련분야에의 기여도 ... 359
• 제 1 절 원자력시설 고도 제염기술 개발 ... 359
• 제 2 절 해체공정 통합평가 및 원격 제어기술 개발 ... 361
• 제 3 절 해체 특수 폐기물 처리 기술 개발 ... 363
• 제 4 절 해체 및 오염부지 환경복원 기술 개발 ... 366
• 제 5 장 연구개발결과의 활용계획 ... 369
• 제 1 절 원자력시설 고도 제염기술 개발 ... 369
• 제 2 절 해체공정 통합평가 및 원격 제어기술 개발 ... 370
• 제 3 절 해체 특수 폐기물 처리 기술 개발 ... 371
• 제 4 절 해체 및 오염부지 환경복원 기술 개발 ... 372
• 제 6 장 연구개발과정에서 수집한 해외과학기술정보 ... 375
• 제 1 절 원자력시설 고도 제염기술 개발 ... 375
• 제 2 절 해체공정 통합평가 및 원격 제어기술 개발 ... 376
• 제 3 절 해체 특수 폐기물 처리 기술 개발 ... 377
• 제 4 절 해체 및 오염부지 환경복원 기술 개발 ... 379
• 제 7 장 참고문헌 ... 381
• 끝페이지 ... 385