Radioactive waste generated in large quantities from NPP decommissioning has various physicochemical and radiological characteristics, and therefore treatment technologies suitable for those characteristics should be developed. Radioactively contaminated concrete waste is one of major decommissioning wastes. The disposal cost of radioactive concrete waste is considerable portion for the total budget of NPP decommissioning. In this study, we developed an integrated technology with thermomechanical and chemical methods for volume reduction of concrete waste and stabilization of secondary waste. The unit devices for the treatment process were also studied at bench-scale tests. The volume of radioactive concrete waste was effectively reduced by separating clean aggregate from the concrete. The separated aggregate satisfied the clearance criteria in the test using radionuclides. The treatment of secondary waste from the chemical separation step was optimally designed, and the stabilization method was found for the waste form to meet the final disposal criteria in the repository site. The final volume reduction rates of 56.4~75.4% were possible according to the application scenario of our processes under simulated conditions. The commercial-scale system designs for the thermomechanical and chemical processes were completed. Also, it was found that the disposal cost for the contaminated concrete waste at domestic NPP could be reduced by more than 20 billion won per each unit. Therefore, it is expected that the application of this technology will improve the utilization of the radioactive waste disposal space and significantly reduce the waste disposal cost.

Sulfate-rich waste powder containing a radioactive nuclide is generated from chemical decontamination process and radioactive liquid waste treatment using ion exchange resin. The radioactive sulfate-rich waste powder should be stabilized for final disposal. The techniques for immobilization of the radioactive sulfate-rich waste powder such as hydraulic cement, geopolymer, and iron phosphate glass have been applied, however, there are limitation in these techniques. Firstly, the hydraulic cement cannot applied to the wastes containing high concentration of sulfate because the expansion, cracks, and disintegration can be happened in the waste form. Geopolymer has a low density although they can be used as a good binder. The iron phosphate glass can be utilized, however, a considerable amount of SO2 gas is emitted due to the high sintering temperature. In this study, immobilization of radioactive sulfate-rich waste powder was carried out to resolve above problems by applying low temperature sintering method using a low-melting glass. As a result, it was confirmed that the waste form has a high bulk density. The compressive strength of the waste form was over 40 MPa, which is higher than the acceptance criteria (≥ 3.44 MPa). From ANS 16.1 test, it was verified that the waste form met the acceptance criteria of the leachability index (≥ 6). It was also confirmed that the waste form was chemically durable through product consistency test (PCT). In addition, the chemical stabilities of waste forms were compared following the sintering condition and the composition of the waste forms. The difference of the chemical stability was explained by difference in the abundance of chemical form obtained from the sequential extraction test.

The nuclear legacy that remains in the United Kingdom (UK) is complex and diverse. Consisting of legacy ponds and silos, redundant reprocessing plants, research facilities, and non-standard or one-off reactor designs, the clean-up of this legacy is under the stewardship of the Nuclear Decommissioning Authority (NDA). Through a mix of prompt and delayed decommissioning strategies, the NDA has made great strides in dealing with the UK’s nuclear legacy. Fuel debris and sludge removal from the legacy ponds and silos situated at Sellafield, as part of a prompt decommissioning strategy for the site, has enabled intolerable risks to be brought under control. Reactor defueling and waste retrievals across the Magnox fleet is enabling their transition to a period of care and maintenance; accelerated through the adopted ‘Lead and Learn’ approach. Bespoke decommissioning methods implemented by the NDA have also enabled the relevant site licence companies to tackle non-standard reactor designs and one-off wastes. Such approaches have potential to influence and shape nuclear decommissioning decision making activities globally, including in Korea.

The challenges facing companies and institutions surrounding civil nuclear decommissioning are diverse and many, none more so than those faced in the United Kingdom. The UK’s Generation I nuclear power plants and early research facilities have left a ‘Nuclear Legacy’ which is in urgent need of management and clean-up. Sellafield is quite possibly the most illfamed nuclear site in the UK. This complex and challenging site houses much of what is left from the early days of nuclear research in the UK, including early nuclear reactors (Windscale Piles, Calder Hall, and the Windscale Advanced Gas Cooled Reactor) and the UK’s early nuclear weapons programme. Such a legacy now requires careful management and planning to safely deal with it. This task falls on the shoulders of the Nuclear Decommissioning Authority (NDA). Through a mix of prompt and delayed decommissioning strategies, key developments in R&D, and the implementation of site licenced companies to enact decommissioning activities, the NDA aims to safety, and in a timely manner, deal with the UK’s nuclear legacy. Such approaches have the potential to influence and shape other such approaches to nuclear decommissioning activities globally, including in Korea.

원전 일차계통 HyBRID 제염공정에서 발생되는 제염폐액에는 황산이온과 방사성 핵종을 포함한 금속이온 및 발암성 물질의 하이드라진을 포함하고 있어 이를 안전한 수준으로 처리할 수 있는 기술개발이 필요하다. 본 연구에서는 모의 제염폐액 내 황산 및 금속이온의 제거와 하이드라진의 분해시험을 실시하여 황산이온, 금속이온 및 하이드라진을 효과적으로 제거할 수 있는 HyBRID 제염폐액 처리공정을 도출하였으며, 1 L 규모에서의 반복실험과 Pilot 규모(300 L/batch)에서의 평가시험을 통해 도출한 HyBRID 제염폐액 처리공정의 성능 재현성과 적용성을 검증하였다.

향후 원자력시설 해체 시 막대한 양의 해체 콘크리트 폐기물이 발생할 수 있음을 감안하였을 때, 방사성 콘크리트 폐기물의 최적 처리기술에 대한 면밀한 검토와 향후 기술개발 방향에 대한 논의는 반드시 필요하다. 본 논문에서는 방사성 콘크리트 폐기물의 국내외 발생 사례를 종합해 보고, 처리 대상이 되는 방사성 콘크리트 폐기물의 특성을 검토하였다. 또한, 종래의 방사성 콘크리트 처리기술로써 기계적 제염기술, 화학적 제염기술, 부피감용기술, 재활용 및 고화기술에 대한 국내외 적용 사례를 정리하고 기술 개발 동향을 살펴봄으로써 기존 기술의 한계점을 파악하고 기술 고도화 방향을 고찰해 보고자 한다.

본 연구는 고온 열분해를 통한 Cs, Sr 등 고방사성핵종의 고정화를 위하여 각각 Cs이 흡착된 CHA (K형 Chabazite zeolite)-Cs, CHA-PCFC (potassium cobalt ferrocyanide)-Cs 및 Sr이 흡착된 4A-Sr, BaA-Sr 등의 제올라이트 계에서 TGA 및 XRD에 의한 배소 온도 변화에 따른 상변환을 고찰하였다. CHA-Cs 제올라이트 계의 경우 900℃ 까지는 CHA-Cs의 형태를 유지하고 있으며, 1,000℃에서 무정형 단계를 거친 후 1,100℃에서 pollucite (CsAlSi2O6)로 재결정 되었다. 반면에 CHA-CFC-Cs 제올라이 트 계는 700℃ 까지는 CHA-PCFC-Cs 형태를 유지하고 있으나, 900∼1,000℃ 사이에서 구조가 파괴되어 무정형으로 상변환 된 후 1,100℃에서 pollucite로 재결정 되었다. 한편 4A-Sr 제올라이트 계의 경우 700℃ 까지는 4A-Sr의 구조를 유지하고 있 으며, 800℃에서 무정형으로 상변환 된 다음 900℃에서는 Sr-feldspar (SrAl2Si2O8, hexagonal)으로, 1,100℃에서 SrAl2Si2O8 (triclinic)로 재결정 되었다. 그러나 BaA-Sr 제올라이트 계의 경우는 500℃ 이하부터 구조가 파괴되기 시작하여 500∼900℃ 에서 무정형 단계를 거친 후, 1,100℃에서 Ba/Sr-feldspar (Ba0.9Sr0.1Al2Si2O8 및 Ba0.5Sr0.5Al2Si2O8 공존)로 재결정 되었다. 상기 제올라이트 계 모두 온도 증가에 따라 탈수/(분해)→ 무정형→ 재결정의 단계를 거쳐 광물상으로 재결정 되었으며, 고온 열 분해 과정에서의 Cs 및 Sr의 휘발성, 침출성 등의 추가 연구가 요구되지만 각 제올라이트 계에 흡착된 Cs 및 Sr은 pollucite나 Sr-feldspar, Ba/Sr-feldspar 등으로 광물화 하여 Cs과 Sr을 배소체/(고화체) 내에 완전히 고정화 시킬 수 있을 것으로 보인다.