• 방사성폐기물학회지
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• 제15권3호
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• pp.281-299
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• 2017

핵종재고량 관리는 처분시설의 안전한 관리를 위해서는 필수적이다. 본 논문에서는 원자력발전소의 잡고체폐기물에 대하여 기존 발생된 폐기물 자료를 반영한 예측방사능량과 실제 처분시설을 운영하면서 인수되어 처분검사까지 완료된 실제방사능량을 비교분석하였다. 극저준위방사성폐기물에서는 $^{137}CS$, $^{90}Sr$, $^{99}Tc$ 그리고 $^{129}I$ 핵종이 예측방사능량보다 실제방사능량이 높게 평가되었으며, 저준위방사성폐기물에서는 모든 핵종에서 예측방사능량이 실제 방사능량보다 높게 평가되었다. 또한 척도인자에 의한 예측방사능량의 민감도 분석을 통해 준위별 수량 및 총방사능량의 변화추이를 분석하였다. 향후 중저준위방사성폐기물 처분시설의 안전한 운영과 Safety Case 구축을 위한 기초자료로 활용될 것으로 판단된다.

• Nuclear Engineering and Technology
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• 제50권2호
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• pp.313-317
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• 2018

Licenses for the storage of spent nuclear fuel (SNF) and vitrified highly active waste in casks under dry conditions are limited to 40 years and have to be renewed for prolonged storage periods. If such a license renewal has to be expected since as in accordance with the new site selection procedure a final repository for spent fuel in Germany will not be available before the year 2050. For transport and possible unloading and loading in new casks for final storage, the integrity and the maintenance of the geometry of the cask's inventory is essential because the SNF rod cladding and the cladding of the vitrified highly active waste are stipulated as a barrier in the storage concept. For SNF, the cladding integrity is ensured currently by limiting the hoop stress and hoop strain as well as the maximum temperature to certain values for a 40-year storage period. For a prolonged storage period, other cladding degradation mechanisms such as inner and outer oxide layer formation, hydrogen pick up, irradiation damages in cladding material crystal structure, helium production from alpha decay, and long-term fission gas release may become leading effects driving degradation mechanisms that have to be discussed.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2418-2426
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• 2022

The grouping analysis is a method guided by the Korea Radioactive Waste Agency for efficient analysis of radioactive waste for disposal. In this study, experiments to verify the adequacy of grouping analysis were conducted with radioactive soil, concrete, and dry active waste in similar environments. First, analysis results of the major radionuclide concentrations in individual waste samples were reviewed to evaluate whether wastes from similar environments correspond to a single waste stream. As a result, the soil and concrete waste were identified as a single waste stream because the distribution range of radionuclide concentrations was "within a factor of 10", the range that meet the criterion of the U.S. Nuclear Regulatory Commission for a single waste stream. On the other hand, the dry active waste was judged to correspond to distinct waste streams. Second, after analyzing the composite samples prepared by grouping the individual samples, the population means of the values of "composite sample analysis results/individual sample analysis results" were estimated at a 95% confidence level. The results showed that all evaluation values for soil and concrete waste were within the set reference values (0.1-10) when five-package and ten-package grouping analyses were conducted, verifying the adequacy of the grouping analysis.

• Nuclear Engineering and Technology
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• 제33권4호
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• pp.355-364
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• 2001

In order to evaluate and finally optimize the vitrification condition for combustible dry active waste (DAW), dust and gas generation characteristics were investigated for PE, cellulose, and mixed waste Tests were conducted by varying the operation variables such as melter configuration, excess oxygen amount, and waste feeding rate. Results showed that dust generation characteristics were affected by the operation parameters and the melter＇s configuration is the dominant one. For all tested DAWs, dust generation was reduced by increasing the waste feeding rate and the excessive oxygen amount in the melter. Among waste types, dust amount was decreased by the order of mixed wastes, PE, and cellulose. Other parameters such as temperature variation and operation time have also affected the dust generation. The optimum condition for the DAW vitrification was determined as the melter＇s configuration equipped for minimizing the waste dispersion with 20 kg/h of waste feeding rate and 100% of excessive oxygen supply. CO gas concentration in the off-gas was immediately influenced by the combustion state in the melter, but showed similar trend as the dust generation. For the NOx production during the vitrification process, thermal NOx, which is generated from the Post Combustion Chamber (PCC), rather than fuel NOx was assumed to be dominant. The gas cleaning of efficiencies of the PCC, wet scrubber, and Selective Catalytic Reduction system (SCR) were found to be high enough to keep the concentration of pollutants (CO, NOx, SOx, HCI) in the stack below their relevant emission limits.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2017년도 추계학술논문요약집
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• pp.177-178
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• 2017

• 방사성폐기물학회지
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• 제2권3호
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• pp.175-180
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• 2004

In order to vitrify the combustible dry active waste (DAW) generated from Korean Nuclear Power Plants, a glass formulation development based on waste composition was performed. A borosilicate glass, DG-2, was formulated to vitrify the DAW in an induction cold crucible melter (CCM). The processability, product performance, and volume reduction effect of the candidate glass were evaluated using a computer code and were measured experimentally in the laboratory and CCM. The glass viscosity and electrical conductivity as the process parameters were in the desired ranges. Start-up and maintaining glass melt of the candidate glass were favorable in the CCM. The product of the glass product such as chemical durability, phase stability, and density was satisfactory. The vitrification process using the candidate glass was also evaluated assuming that it was operated as economically as possible.

• 방사성폐기물학회지
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• 제19권2호
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• pp.267-270
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• 2021

Long-term experiments have been conducted on two important safety issues: long-term durability of a concrete barrier with the steel reinforcements and gas generation from low-and intermediate-level wastes in an underground research tunnel of a radioactive waste disposal facility. The gas generation and microbial communities were monitored from waste packages (200 L and 320 L) containing simulated dry active wastes. In the concrete experiment, corrosion sensors were installed on the steel reinforcements which were embedded 10 cm below the surface of concrete in a concrete mock-up, and groundwater was fed into the mock-up at a pressure of 2.1 bars to accelerate groundwater infiltration. No clear evidence was observed with respect to corrosion initiation of the steel reinforcement for 4 years of operation. This is attributed to the high integrity and low hydraulic conductivity of the concrete. In the gas generation experiment, significant levels of gas generation were not measured for 4 years. These experiments are expected to be conducted for a period of more than 10 years.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 춘계 학술대회
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• pp.523-530
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• 2005

본 연구에서는 IP형 운반용기의 개념설계를 위하여 소내 임시저장중인 방사성폐기물중 $200\;{\ell}$ 잡고체 드럼을 대상으로 운반용기에 적재 가능한 드럼의 최대표면선량률을 제시하고자 하였다. 이를 위해 잡고체 폐기물을 가연성과 비가연성으로 구분하였으며, 각각 혼합핵종이 균일 분포되어있는 경우와 단일핵종(Co-60)이 균일 분포되어있는 경우를 나누어 계산하였다. 연구결과, 가연성 잡고체 드럼에 혼합핵종이 분포되어 있는 경우, IP-1, IP-2-a, IP-2-b형 운반용기에 적재 가능한 최대표면선량률은 각각 3.60E-01, 8.85E-01, 1.27E+01 mSv/hr 이었으며, Co-60이 분포되어 있는 가연성 잡고체 드럼의 최대표면선량률은 각각 3.59E-01, 8.83E-01, 1.25E+01 mSv/hr 이었다. 비가연성 잡고체 드럼에 혼합핵종이 분포되어 있는 경우, IP-1, IP-2-a, IP-2-b형 운반용기에 적재 가능한 최대표면선량률은 각각 7.14E-01, 1.83E+00, 2.69E+01 mSv/hr이었으며, Co-60이 분포되어 있는 비가연성 잡고체 드럼의 최대표면선량률은 각각 7.13E-01, 1.81E-01, 2.62E+01 mSv/hr 으로 계산되었다. 이를 통해 운반가능 방사성내용물의 최대수량을 실측이 용이한 표면선량률 만으로 제시할 수 있었으며, 향후 다른 종류의 폐기물에 대해서도 차폐해석을 수행할 예정이다.

• Nuclear Engineering and Technology
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• 제29권3호
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• pp.218-228
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• 1997

Correlations between the amount of DAW (Dry Active Waste) generated from present Korean PWRs and their operating parameters were analyzed. As the result of multi-variable linear regressions, a model predicting the volume of DAW using the number of shutdowns ( $f_{FS}$ ) and total personnel exposure ( $P_{\varepsilon}$) was derived. Considering one standard error bound, the model could successfully simulate about 8575 of the real data. In order to predict the amount of DAW to be generated from a KNGR another model was derived by taking into account the additional volume reduction by supercompaction system. In addition, the volume of WAW (Wet Active Waste) to be generated from KNGR (Korean Next Generation Reactor) was calculated by considering conceptual design data and replacement effect of radwaste evaporator with selective ion exchangers. Finally, total volume of SRW (Solid Radioactive Waste) to be generated from KNGR was predicted by inserting design goal values of $f_{FS}$ and $P_{\varepsilon}$ into the model. The result showed that the expected amount of SRW to be generated from KNGR would be in the range of 33~44㎥. $y^{-1}$ . It was proved that the value would meet the operational target of KNGR proposed by KEPCO, that is, 50㎥. $y^{-1}$ .

• 방사성폐기물학회지
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• 제21권3호
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• pp.411-417
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• 2023

During the operation of a nuclear power plant (NPP), the generation of radioactive waste, including dry active waste (DAW), concentrates, spent resin, and filters, mandates the implementation of appropriate disposal methods to adhere to Korea's waste acceptance criteria (WAC). In this context, this study investigates the potential use of polymer concrete (PC) as a high-integrity container (HIC) material for solidifying and packaging these waste materials. PC is a versatile composite material comprising binding polymers, aggregates, and additives, known for its exceptional strength and chemical stability. A comprehensive analysis of PC's long-term integrity was conducted in this study. First, its compressive strength, which is crucial for ensuring the structural stability of HICs over extended periods, was evaluated. Subsequently, the resilience of PC was tested under various stress conditions, including biological, radiological, thermal, and chemical stressors. The findings of this study indicate that PC exhibits remarkable long-term properties, demonstrating exceptional stability even when subjected to diverse stressors. The results therefore underscore the potential viability of PC as a reliable material for constructing high-integrity containers, thus contributing to the safe and sustainable management of radioactive waste in NPPs.