The transport of radionuclides at oceanic scales can be assessed using a Lagrangian model. In this review an application of such a model to the Atlantic, Indian and Pacific oceans is described. The transport model, which is fed with water currents provided by global ocean circulation models, includes advection by three–dimensional currents, turbulent mixing, radioactive decay and adsorption/release of radionuclides between water and bed sediments. Adsorption/release processes are described by means of a dynamic model based upon kinetic transfer coefficients. A stochastic method is used to solve turbulent mixing, decay and water/sediment interactions. The main results of these oceanic radionuclide transport studies are summarized in this paper. Particularly, the potential leakage of 137Cs from dumped nuclear wastes in the north Atlantic region was studied. Furthermore, hypothetical accidents, similar in magnitude to the Fukushima accident, were simulated for nuclear power plants located around the Indian Ocean coastlines. Finally, the transport of radionuclides resulting from the release of stored water, which was used to cool reactors after the Fukushima accident, was analyzed in the Pacific Ocean.

Several countries, including Korea, are considering the direct disposal of spent nuclear fuels. The radiological safety assessment results published after a geological repository closure indicate that the instant release is the main radiation source rather than the congruent release. Three Safety Case reports recently published were reviewed and the IRF values of seven long-lived radionuclides, including relevant experimental results, were compared. According to the literature review, the IRF values of both the CANDU and low burnup PWR spent fuel have been experimentally measured and used reasonably. In particular, the IRF values of volatile long-lived nuclides, such as 129I and 135Cs, were estimated from the FGR value. Because experimental leaching data regarding high burnup spent nuclear fuels are extremely scarce, a mathematical modelling approach proposed by Johnson and McGinnes was successfully applied to the domestic high burnup PWR spent nuclear fuel to derive the IRF values of iodine and cesium. The best estimate of the IRF was 5.5% at a discharge burnup of 55 GWd tHM−1.

Currently, in domestic nuclear power plants (NPP), the spent filters (SFs) used for the purpose of reducing and purifying the radiation of the primary cooling water system are temporarily stored in an untreated state. In order to dispose of SFs, radioactive nuclide analysis (RNA) of SFs is required to be conducted. As segmented gamma scanner (SGS) is already being used in Kori NPP, utilizing SGS for RNA of SFs would be practical and economical. In this paper, factors required to be considered to improve accuracy of SGSs for RNA of SFs are studied. The analysis of the nuclide inventory of the packaging drum for radioactive waste should be performed by the indirect drum nuclide analysis method. The material of the SFs is iron (SS304) on the outside, and paper on the inside. In addition, to meet disposal acceptance criteria, radioactive waste drums are packaged in thick grouting or shielding drums. Therefore, it is necessary to derive an appropriate correction method for high inhomogeneity and thick media. Considering these factors, evaluating radionuclides inventory plans to measure gamma rays in SGS mode. Correct the gamma ray measurement by examining the medium attenuation factor and error factors. In this way, the inventory of gamma nuclides is calculated, and the specific radioactivity of beta ray and alpha particle emitting nuclides other than gamma rays is planned to be calculated by applying scaling factors.

Several previous simulation studies using various geochemical models have been carried out in several major analogue sites. The cases are beneficial when these studies provided the possibility of testing the geochemical models to be used to describe the migration of radionuclides in a future radioactive waste repository system. It was possible to interpret the complex transport behaviour of radionuclides such as uranium and thorium in an environment. We organize major natural analogue study sites from the previous literatures that provided information on the general geochemistry of the sites, in terms of groundwater composition and mineralogy. Also, we calculated aqueous speciation and the solid phases most likely to control their solubilities. The results obtained from the previous studies and this study vary depending on the tools used and on the conceptual models followed. Also, the results differed from the actual measured concentrations of trace metals or radionuclide analogues. The results obtained from these tests identify the main mathematical limitations of available geochemical models. However, the modelling results using a geochemical code with the thermodynamic database simulated well the observed behaviour of radionuclides, especially to identify the dominant processes controlling actinide mobilization and fixation. It was a useful outcome in terms of building confidence on the current geochemical tools to predict the concentrations of radionuclide analogues once the major geochemical characteristics were known. This study allows improving specific aspects of geochemical modelling using major natural analogue sites.

An objective of a safety assessment for geological disposal is to evaluate the radiological impact by radionuclides release from radioactive wastes. Computational estimation of all radionuclides transport in the disposal system, however, is not neccessary because some radionuclides has negligible effect on radiological doses. For this reason, prioritization of radionuclides list is preceded before the safety assessment. The Korea Atomic Energy Research Institue (KAERI) has assessed the long-term safety of a disposal system for spent nculear fuels. Currently, thirty eight radionuclides and twenty three elements are considered in the safety assessment activity of the KAERI. Nevertheless, a screening process for radionulides selection has not been articulated yet. In this study, we reviewed radionuclides selection process in forign countries to re-establish screening criteria for the KAERI’s radionuclides list. Screeing models of the Swedish Nuclear Fuel and Waste Management Company (SKB), the Deparment of Eenrgy (US DOE), and the Japan Nuclear Cycle Development Istitute (JNC) were compared. We found that each country developed different screening model depending on scenarios of radionuclides release. Nonetheless, there were common properties that determines the importance of radionuclides. These properties for radionuclides include halflife, radiotoxicity (or specific activity), and mobility in underground medium. Based on the review results, we proposed radionuclides selection process to prioritize the importance of radionucldies in the KAERI safety assessment.

The overestimation and underestimation of the radioactivity concentration of difficult-to-measure radionuclides can occur during the implementation of the scaling factor (SF) method because of the uncertainties associated with sampling, radiochemical analysis, and application of SFs. Strict regulations ensure that the SF method as an indirect method does not underestimate the radioactivity of nuclear wastes; however, there are no clear regulatory guidelines regarding the overestimation. This has been leading to the misuse of the SF methodology by stakeholders such as waste disposal licensees and regulatory bodies. Previous studies have reported instances of overestimation in statistical implementation of the SF methodology. The analysis of the two most popular linear models of the SF methodology showed that severe overestimation may occur and radioactivity concentration data must be dealt with care. Since one major source of overestimation is the use of minimum detectable activity (MDA) values as true activity values, a comparative study of instrumental techniques that could reduce the MDAs was also conducted. Thermal ionization mass spectrometry was recommended as a suitable candidate for the trace level analysis of long-lived beta-emitters such as iodine-129. Additionally, the current status of the United States and Korea was reviewed from the perspective of overestimation.

2011년 동일본대지진에 의해 발생한 후쿠시마 원자력 발전소 사고와 최근 국내 지진 발생 빈도의 증가는 원자력 발전소의 지진 안전성에 대한 불안감을 야기하였다. 더불어 최근(2021년) 일본 동경전력은 후쿠시마 원전 오염수의 태평양 방류를 결정하였으며, 이로 인해 국내외 수산물을 통한 방사능 오염 가능성이 높아지면서 국민들의 우려가 급증하고 있다. 후쿠시마 원전사고 이후 해양으로 의 인공방사능 유입에 관한 연구는 국제적으로 많이 이루어졌으나, 한국인의 주요 식재료인 동아시아 연근해의 수산물에서 인공방사능의 분포 현황 및 축적에 대한 연구는 상대적으로 부족한 실정이다. 따라서 이 논문에서는 후쿠시마 원전사고 이후, 국내산 수산물에서의 원전 기원 인공방사능(예, 137Cs, 239,240Pu, 90Sr 등)의 분포 특성과 관련한 최근 연구 사례들을 소개하고자 한다. 또한, 후쿠시마 원전 오염수의 방류와 더불어 2030년까지 계획된 중국의 신규 원전 시설로 인한 향후 한반도 주변해역의 방사능 유출 영향에 대한 대비 및 사전 연구가 필요한 시점이기에 향후 연구 방향들을 제안하고자 한다.

A safety assessment of radioactive waste repositories is a mandatory requirement process because there are possible radiological hazards owing to radionuclide migration from radioactive waste to the biosphere. For a reliable safety assessment, it is important to establish a parameter database that reflects the site-specific characteristics of the disposal facility and repository site. From this perspective, solubility, a major geochemical parameter, has been chosen as an important parameter for modeling the migration behavior of radionuclides. The solubilities were derived for Am, Ni, Tc, and U, which were major radionuclides in this study, and on-site groundwater data reflecting the operational conditions of the Gyeongju low and intermediate level radioactive waste (LILW) repository were applied to reflect the site-specific characteristics. The radiation dose was derived by applying the solubility and radionuclide inventory data to the RESRAD-OFFSITE code, and sensitivity analysis of the dose according to the solubility variation was performed. As a result, owing to the low amount of radionuclide inventory, the dose variation was insignificant. The derived solubility can be used as the main input data for the safety assessment of the Gyeongju LILW repository in the future.

Numerical model was developed that simulates radionuclide (3 H 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. 3 H 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, 3 H 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.

Anionic radionuclides pose one of the highest risks to the long-term safety assessments of disposal repositories. Therefore, techniques to immobilize and separate such anionic radionuclides are of crucial importance from the viewpoints of safety and waste volume reduction. The main objective of this study is to design a separator with minimum pressure disturbance, based on the concept of a conventional cyclone separator. We hypothesize that the anionic radionuclides can be immobilized onto a nanomaterial-based substrate and that the particles generated in the process can flow via water. These particles are denser than water; hence, they can be trapped within the cyclone-type separator because of its design. We conducted particle tracking analysis using computational fluid dynamics (CFD) for the conventional cyclone separator and studied the effects due to the morphology of the separator. The proposed sandglass-like design of the separator shows promising results (i.e., only one out of 10,000 particles escaped to the outlet from the separation zone). To validate the design, we manufactured a laboratory-scale prototype separator and tested it for iron particles; the efficiency was ca. 99%. Furthermore, using an additional magnetic effect with the separator, we could effectively separate particles with ~100% efficiency. The proposed sandglass-like separator can thus be used for effective separation and recovery of immobilized anionic radionuclides.