• Nuclear Engineering and Technology
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• 제26권1호
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• pp.78-89
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• 1994

콘크리트 구조물과 점토성 되메움재로 구성된 중저준위 방사성폐기물처분장 인공방벽으로 부터의 핵종유출률이 평가되었다. 네 종류의 유출경로가 고려되었으며, 각 유출경로가 방사성핵종의 총유출률에 미치는 영향이 분석되었다. 입력변수 간의 불확실도가 핵종유출률 분석에 미치는 영향을 정량화하기 위해 Latin Hypercube 표본추출 방법이 이용되었으며, 그 결과 얻어진 유출률 분포는 적합도검증을 통하여 결정되었다. 마지막으로 최대유출률의 범위가 통계적방법에 의해 95% 신뢰도수준으로 추정되었다.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1385-1397
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• 2024

Bentonite is a recommended material for the multiple barriers in the final disposal of low-level radioactive waste (LLW) to prevent groundwater intrusion and nuclear species migration. However, after drying-wetting cycling during the repository construction stage and ion exchange with the concrete barrier in the long-term repository, the bentonite mechanical behaviors, including swelling capacity and hydraulic conductivity, would be further influenced by the groundwater intrusion, resulting in radioactive leakage. To comprehensively examine the factors on the mechanical characteristics of bentonite, this study presented scenarios involving MX-80 and KV-1 bentonites subjected to drying-wetting cycling and accelerated ion migration. The experiments subsequently measured free swelling, swelling pressure, and hydraulic conductivity of bentonites with intrusions of seawater, high pH, and low pH solutions. The results indicated that the solutions caused a reduction in swelling volume and pressure, and an increase in hydraulic conductivity. Specifically, the swelling capability of bentonite with drying-wetting cycling in the seawater decreased significantly by 60%, while hydraulic conductivity increased by more than three times. Therefore, the study suggested minimizing drying-wetting cycling and preventing seawater intrusion, ensuring a long service life of the multiple barriers in the LLW repository.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2003년도 가을 학술논문집
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• pp.109-116
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• 2003

국내 원자력 산업의 급속한 성장과 더불어 하나로 시설의 본격적인 가동 및 핵연료주기시험과 관련한 연구의 증가로 인하여 방사성폐기물의 발생량 및 누적량이 지속적으로 증가될 전망이며, 이에 따라 방사성폐기물의 안전성 확보 및 감용 처리를 위한 노력이 더욱 강조되고 있다. 조사후시험시설에서는 원자력발전소에서 발생하는 사용후핵연료봉의 결함원인 규명과 건전성 평가를 위한 조사후시험을 수행하고 있으며, 본 연구에서는 조사후시험시설에 설치되어 있는 방사성고체폐기물 처리설비를 활용하여 조사후시험에서 발생하는 폐기물의 압축, 파쇄, 절단기술 및 경험사례에 대하여 기술하였다. 고준위 방사성고체폐기물 처리는 특수 제작하여 핫셀에 설치되어 있는 100톤 압축기로 방사성고체폐기물을 압축하여 폐기물의 양을 1/12정도로 감용 처리하였으며, 중ㆍ저 위 방사성고체폐기물은 인터벤션에 설치된 60톤 압축기를 사용하여 가연성폐기물을 1/8정도로 압축 감용 처리하였다. 폐플라스틱 통은 파쇄기를 사용하여 절단처리 함으로써 1/5, 폐 필터는 1/6의 감용 비를 얻었으며, 비 가연성물질인 금속류 물질 또한 절단 처리하여 드럼의 양을 줄일 수 있었다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2003년도 가을 학술논문집
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• pp.381-387
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• 2003

가압 경수로 사용후핵연료봉을 재처리하는 과정에서 발생되는 hull은 고준위 방사성폐기물로 분류되고 있다. 본 논문에서는 연소도 32,000MWd/tU, 냉각기간 15년(고리 1호기 cycle 4-7)인 PWR 사용후핵연료의 건식처리 공정에서 발생한 hull에 대하여 방사능적 특성 실험을 수행하였고, 문헌 조사 및 관련 코드의 계산을 통하여 가압 경수로 사용후핵연료 hull에 대한 방사화학적 특성을 조사하였다. 이를 토대로 hull에 부착되어 있는 핵물질을 레이저 또는 플라즈마 등의 건식 방법으로 제거함으로써 hull의 방사능을 저감시켜 중저준위 폐기물화하는 방안을 제시하였다.

• 방사성폐기물학회지
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• 제22권1호
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• pp.37-44
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• 2024

In 2017, a decision was made to permanently shut down Kori Unit 1, and preparations began to be made for its decontamination and decommissioning. The dismantling of the biological shields concrete, reactor vessel (RV), and reactor vessel internals (RVI) is crucial to the nuclear decommissioning process. These components were radiologically activated by the neutron activation reaction occurring in the reactor during its operational period. Because of the radioactivity of the RV and RVI of Kori Unit 1, remotely controlled systems were developed for cutting within the cavity to reduce radiation exposure. Specialized equipment was developed for underwater cutting operations. This paper focuses on modeling related to RVI operations using the MAVRIC code and the dose calculation for a diver entering the cavity. The upper and lower parts of the RVI are classified as low-level radioactive waste, while the sides that came into contact with the fuel are classified as intermediate-level radioactive waste. Therefore, the modeling presented in this paper only considers the RVI sides because the upper and lower parts have a minimal impact on the radiation exposure. These research findings are anticipated to contribute to enhancing the efficiency and safety of nuclear reactor decommissioning operations.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.52-62
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• 2023

Various decontamination technologies have been developed for removing contaminated areas in industries. Although it is important to consider parameters such as safety, cost, and time when selecting the decontamination technology, till date their comparative study is missing. Furthermore, different decontamination technologies influence the decontamination effects in different ways. Therefore, this study compares different decontamination techniques for the steam generator using a multicriteria decision-making method. A steam generator is a large device comprising both low- and very low-level waste (LLW, VLLW) and reflects the difference in weights of the standards according to the classification of the waste. For LLW and VLLW decontaminations, chemical oxidizing reduction decontamination (CORD) and decontamination grit blasting were used as the preferred techniques, respectively, considering the purpose of decontamination differs based on the initial state of waste. An expert survey revealed that safety in LLW and waste minimization in VLLW exhibited high preference. This evaluation method can be applied not only to the comparison between each process, but also to the creation of process scenarios. Therefore, determining the decontamination approach using logical decision-making methods may improve the safety and economic feasibility of each step in the decommissioning process and ensure a public acceptance.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 Proceedings of The 6th korea-china joint workshop on nuclear waste management
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• pp.236-244
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• 2005

Radioactive wastes arising from a wide range of human activities are in many different physical and chemical forms, contaminated with varying radioactivity. Their common feature is the potential hazard associated with their radioactivity and the need to manage them in such a way as to protect the human environment. The geological disposal is regarded as the most reasonable and effective way to safely disposal high-level radioactive wastes in the world. The conceptual model of geological disposal in China is based on a multi-barrier system that combines an isolating geological environment with an engineered barrier system. The buffer is one of the main engineered barriers for HLW repository. The buffer material is expected to maintain its low water permeability, self-sealing property, radio nuclides adsorption and retardation property, thermal conductivity, chemical buffering property, overpack supporting property, stress buffering property over a long period of time. Benotite is selected as the main content of buffer material that can satisfy above. GMZ deposit is selected as the candidate supplier for Chinese buffer material of High Level Radioactive waste repository. This paper presents geological features of GMZ deposit and basic property of GMZ Na bentonite. GMZ bentonite deposit is a super large scale deposits with high content of Montmorillonite (about $75\%$) and GMZ-l, which is Na-bentonite produced from GMZ deposit is selected as reference material for Chinese buffer material study.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 춘계학술발표회논문집(2)
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• pp.833-838
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• 1995

This study presents a method to predict plant-specific and operational history dependent scaling factors. Realistic and detailed approaches are taken to find scaling factors at reactor coolant. This approach begins with fission product release mechanisms and fundamental release properties of fuel-source nuclide such as fission product and transuranic nuclide. Scaling factors at various waste streams are derived from the predicted reactor coolant scaling factors with the use of radionuclide retention and build up model. This model makes use of radioactive material balance within the radioactive waste processing systems. According to input parameters of plant operation history, scaling factors predicted at reactor coolant and waste streams are well brought out the effects of plant operation history.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2003년도 가을 학술논문집
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• pp.244-250
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• 2003

본 논문에서는 황사빗물이 중ㆍ저준위 방사성 폐기물 시멘트 고화체에 미치는 영향을 알아보았다. 실험은 ANS 16.1 실험법을 채택하였다. Co 핵종을 포함한 시멘트 고화체를 제작한 후, 대기 중 황사성분의 질량농도를 이용해 침출수의 부피, 이온 및 금속의 농도 등을 결정한다. 실험을 위해 대기 중 황사 부하량이나 강수에 포함되는 황사성분의 양, 처분장의 면적 등은 적합한 가정을 통해 결정하였다. 본 논문에서는 황사의 특성에 대해 간략히 소개하고 침출 실험의 준비과정으로 실험 조건을 결정한 후에, 90일간의 침출실험을 통해 나온 결과로 황사빗물에 의한 시멘트 고화체의 영향을 평가ㆍ분석하였다.

• 방사성폐기물학회지
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• 제18권4호
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• pp.439-455
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• 2020

Numerical model was developed that simulates radionuclide (3H and 14C) transport modeling at the 2nd phase facility at the Wolsong LILW Disposal Center. Four scenarios were simulated with different assumptions about the integrity of the components of the barrier system. For the design case, the multi-barrier system was shown to be effective in diverting infiltration water around the vaults containing radioactive waste. Nevertheless, the volatile radionuclide 14C migrates outside the containment system and through the unsaturated zone, driven by gas diffusion. 3H is largely contained within the vaults where it decays, with small amounts being flushed out in the liquid state. Various scenarios were examined in which the integrity of the cover barrier system or that of the concrete were compromised. In the absence of any engineered barriers, 3H is washed out to the water table within the first 20 years. The release of 14C by gas diffusion is suppressed if percolation fluxes through the facility are high after a cover failure. However, the high fluxes lead to advective transport of 14C dissolved in the liquid state. The concrete container is an effective barrier, with approximately the same effectiveness as the cover.