Intermediate Level Radioactive Waste Research Articles

Impact of carbonation on the durability of cementitious materials: water transport properties characterization

Within the context of long-lived intermediate level radioactive waste geological disposal, reinforced concrete would be used. In service life conditions, the concrete structures would be subjected to drying and carbonation. Carbonation relates to the reaction between carbon dioxide (CO2) and the main hydrates of the cement paste (portlandite and C-S-H). Beyond the fall of the pore solution pH, indicative of steel depassivation, carbonation induces mineralogical and microstructural changes (due to portlandite and C-S-H dissolution and calcium carbonate precipitation). This results in the modification of the transport properties, which can impact the structure durability. Because concrete durability depends on water transport, this study focuses on the influence of carbonation on water transport properties. In fact, the transport properties of sound materials are known but they still remain to be assessed for carbonated ones. An experimental program has been designed to investigate the transport properties in carbonated materials. Four hardened cement pastes, differing in mineralogy, are carbonated in an accelerated carbonation device (in controlled environmental conditions) at CO2 partial pressure of about 3%. Once fully carbonated, all the data needed to describe water transport, using a simplified approach, will be evaluated.

Corrosion susceptibility of steel drums to be used as containers for intermediate level nuclear waste

The present work is a study of the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins contaminated with different types and concentrations of aggressive species. A special type of specimen was manufactured to simulate the cemented ion-exchange resins in the drum. The evolution of the corrosion potential and the corrosion rate of the steel, as well as the electrical resistivity of the matrix were monitored over a time period of 900 days. The aggressive species studied were chloride ions (the main ionic species of concern) and sulphate ions (produced during radiolysis of the cationic exchange-resins after cementation). The work was complemented with an analysis of the corrosion products formed on the steel in each condition, as well as the morphology of the corrosion products. When applying the results obtained in the present work to estimate the corrosion depth of the steel drumscontaining the cemented radioactive waste after a period of 300 years (foreseen durability of the Intermediate Level Radioactive Waste facility in Argentina) , it is found that in the most unfavourable case (high chloride contamination), the corrosion penetration will be considerably lower than the thickness of the wall of the steel drums.

Corrosion behaviour of steel rebars embedded in a concrete designed for the construction of an intermediate-level radioactive waste disposal facility

The National Atomic Energy Commission of the Argentine Republic is developing a nuclear waste disposal management programme that contemplates the design and construction of a facility for the final disposal of intermediate-level radioactive wastes. The repository is based on the use of multiple, independent and redundant barriers. The major components are made in reinforced concrete so, the durability of these structures is an important aspect for the facility integrity. This work presents an investigation performed on an instrumented reinforced concrete prototype specifically designed for this purpose, to study the behaviour of an intermediate level radioactive waste disposal facility from the rebar corrosion point of view. The information obtained will be used for the final design of the facility in order to guarantee a service life more or equal than the foreseen durability for this type of facilities.

Modification to the Flow Properties of Repository Cement as a Result of Carbonation

The current UK repository concept for low- and intermediate-level radioactive wastes involves using large quantities of cementitious materials for both construction and buffer/backfill. CO2 generated from degradation of waste organic material results in cement carbonation and associated mineralogical changes. Hydraulic and gas permeability tests were performed at 4MPa and 40°C. Carbonation halved the permeability of Nirex Reference Vault Backfill (NRVB) cement under these repository conditions. Mineralogical changes occurred in a narrow zone along the leading edge of a migrating reaction front and decreased porosity. The small reduction in permeability (and slightly elevated pressures) that resulted is considered to have minor impact on the permeable characteristics of the NRVB.

유리섬유 및 알루미늄 금속 혼합물 유리조성 개발

원전에서 발생되는 방사성폐기물에 대한 고화처리 방법 중 하나인 유리화기술이 일부 가연성폐기물에 대해 적용되고 있다. 국내외적으로 중저준위 방사성폐기물의 효과적인 감용과 안정적인 처분을 위해 다양한 폐기물에 대한 유리화기술 적용방안이 확대 연구되고 있으며, 최근에는 가연성폐기물 뿐만 아니라 알루미늄 금속과 같은 비가연성폐기물에도 유리화 연구가 활발하게 진행되고 있다. 공기조화계통 (HVAC)에는 주로 필터가 이용되고 있으며, 사용 후 필터는 여과재 (유리섬유 및 알루미늄)를 이용하여 배기체를 흡착하기 때문에 방사성폐기물로 처리가 되어야 한다. 본 연구는 필터에 대한 처리기술 연구를 위해 유도가열식 저온용융로 (Cold Crucible Induction Melter: CCIM)를 이용한 유리화 타당성 연구를 수행하였다. 사용후 필터에 대한 유리화 (Vitrification)는 먼저 유리섬유 및 알루미늄 함량을 고려한 최적의 유리조성을 개발 하였으며, 개발된 유리 조성을 이용하여 최적의 폐기물 저감을 위한 용융변수와 최종 생성된 유리고화체의 특성을 분석하였다. 사용후 필터 유리화용 조성유리는 주로 <TEX>$SiO_2$</TEX>와 <TEX>$B_2O_3$</TEX>로 구성되어 있다. 전기로를 이용한 용융물 특성시험에서는 폐기물 투입률 및 최종 생성물인 유리고화체의 특성이 검토되었다. 본 연구에서는 알루미늄 금속과 유리섬유로 구성된 필터에 대한 유리조성 개발과 이를 통해 생성된 유리고화체의 물리화학적 특성을 검토하고 유리화 타당성을 확인하였다. Vitrification technology has been widely applied as one of effective processing methods for wastes generated in nuclear power plants. The advantage of vitrifying for low- and intermediate-level radioactive wastes has a large volume reduction and good durability for the final products. Recently, a filter using on HVAC(Heating Ventilating & Air Conditioning System) is composed with media (glass fiber) and separator (aluminum film) has been studied the proper treatment technology for meeting the waste disposal requirement. Present paper is a feasibility study for the filter vitrification that developing of the glass compositions for filter melting and melting test for physicochemical characteristic evaluation. The aluminum metal of film type is preparing with 0.5 cm size for proper mixing with glass frit, glass fiber is also preparing with 1 cm size within crucible. The glass compositions should be developed considering molten glass are related with wastes reduction. Glass compositions obtained from developing on glass formulation are mainly composed of <TEX>$SiO_2$</TEX> and <TEX>$B_2O_3$</TEX> for aluminum metal. A variety of factors obtained from the glass formulation and melting test are reviewed, which is feeding rate and glass characteristics of final products such as durability for implementing the wastes disposal requirement.

Dissolution of vitrified wastes in a high-pH calcium-rich solution

The current baseline for the conditioning of most UK intermediate-level radioactive waste (ILW) is immobilisation using cement. However, vitrification of some UK ILW is being considered as an alternative. One option for the disposal of the resulting vitrified ILW would be to place it in a geological disposal facility in a high-pH environment with cemented ILW and a cement-based backfill. Therefore, the potential effects of such a high pH (∼12.5), calcium-rich cement-based environment on the dissolution behaviour of simulant ILW glasses have been studied using the product consistency test (PCT). Three non-radioactive waste compositions were assessed: a laboratory simulant ILW vitrified in a borosilicate glass and two full-scale simulant vitrified products (a slag containing simulant plutonium-contaminated material and Magnox sludge; and a glass containing clinoptilolite). Powdered samples were leached in saturated Ca(OH)2 solutions for up to 42days at temperatures between 30 and 90°C.In general the rates of dissolution were lower than expected at such a high pH compared to studies in the literature under alkaline conditions. In contrast to the typical dissolution behaviour of high level waste (HLW) glasses, dissolution of the simulant borosilicate ILW glass was initially slow, followed by a period of faster boron and alkali metal release. The saturation/residual regime was not reached within experimental timescales. The rate of dissolution during the period of faster release increased with increasing temperature; the activation energy for this stage of dissolution was calculated to be 47±2kJmol−1 based on boron release. The two full-scale simulant glasses, which contained negligible boric oxide, exhibited conventional static dissolution profiles, and entered the residual rate regime after 7-14days at 50°C. The greater durability of the full-scale simulants is thought to be due to the greater content of network-forming oxides in these glasses compared to the borosilicate glass. It is also suggested that the formation of calcium borates may delay initial hydration of the borosilicate glass in a similar manner to the retardation of cement hydration by soluble borates. More generally, the formation of calcium- and magnesium-containing precipitates on the surface of the vitrified wastes, and agglomeration of the powder, appeared to act to reduce the dissolution rate. Overall these results suggest that calcium has an important role in the long-term durability of vitrified wastes at high pH.

Waste container durability: monitoring of a stainless steel 4 metre box over 12 years

AbstractIn the UK, most intermediate level radioactive waste is packaged in stainless steel containers. It is important that containers should maintain their integrity for extended periods of storage prior to final disposal. A prototype 4 metre box fabricated from 304S11 stainless steel was monitored at an inland rural building for over twelve years to gain experience in monitoring environmental conditions and corrosion during storage. A wide range of different types of aerosol particles were observed on the surfaces and in the air. They contained corrosive ions such as chloride, as well as other anions such as sulfate and nitrate. The chloride deposition rate was between 0.1 and 4 μg cm−2 yr−1 depending on the orientation and location of the surface, with a chloride/sulfate ratio ranging from ∼0.3–2. The temperature and relative humidity typically ranged from 0–30°C and 40–90%, respectively. The greatest degree of corrosion developed on the base of the box, which had elevated chloride concentrations (10–140 μg cm−2). However, there was little evidence of corrosion propagation on the surface of the box and no evidence of stress corrosion cracking. The average pit propagation rates on a 316L coupon mounted next to the 4 m box were &lt;1 μm yr−1.

Long-term cement corrosion in chloride-rich solutions relevant to radioactive waste disposal in rock salt – Leaching experiments and thermodynamic simulations

Low- and intermediate-level radioactive wastes are frequently solidified in a cement matrix. In a potential repository for nuclear wastes, the cementitious matrix is altered upon contact with solution and the resulting secondary phases may provide for significant retention of the radionuclides incorporated in the wastes. In order to assess the secondary phases formed upon corrosion in chloride-rich solutions, which are relevant for nuclear waste disposal in rock salt, leaching experiments were performed. Conventional laboratory batch experiments using powdered hardened cement paste in MgCl2-rich solutions were left to equilibrate for up to three years and full-scale cemented waste products were exposed to NaCl-rich and MgCl2-rich solutions for more than twenty years, respectively. Solid phase analyses revealed that corrosion of hardened cement in MgCl2-rich solutions advanced faster than in NaCl-rich solutions due to the extensive exchange of Mg from solution against Ca from the cementitious solid. Thermodynamic equilibrium simulations compared well to results at the final stages of the respective experiments indicating that close to equilibrium conditions were reached. At high cement product to brine ratios (>0.65gmL−1), the solution composition in the laboratory-scale experiments was close to that of the full-scale experiments (cement to brine ratio of 2.5gmL−1) in the MgCl2 systems. The present study demonstrates the applicability of thermodynamic methods used in this approach to adequately describe full-scale long-term experiments with cemented waste simulates.

Experimental study to evaluate the effect of polymeric encapsulants on the corrosion resistance of intermediate-level radioactive waste packages

AbstractIntermediate-level radioactive waste is normally encapsulated in cementitious grout. However, for some wastes, grout may not be suitable and polymeric encapsulants are being considered as an alternative. One concern with such encapsulants is their long-term chemical stability and the possibility that release of aggressive degradation products could cause corrosion.This paper evaluates the potential for three polymeric encapsulants: two epoxy resins (the APS polymer system, (APS) and Alchemix 4760 (ALC)) and a vinyl-ester styrene (VES); to cause internal corrosion of stainless steel waste containers. The corrosion behaviour of stainless steel 316L in contact with each encapsulant was studied in saturated Ca(OH)2 solutions and deionized (DI) water, at 80°C, under non-irradiated and γ-irradiated conditions.In aerated, alkaline conditions, 316L was resistant to corrosion in all the conditions tested. However, in DI water, the pH fell to values as low as three due to release of acidic species from the polymers. The two epoxy materials (particularly APS) also released significant levels of chloride; VES did not. Chloride release appeared to be increased by γ-irradiation. As a result of the low pH chloride-containing environment created by the APS encapsulant, 316L experienced localized corrosion, whereas coupons in Alchemix 4760 and VES did not. Weight loss measurements correlated with visual observations. γ-irradiation appeared to increase the degree of corrosion.

Cracking in a cementitious backfill and its implications for flow and chemistry

AbstractIn the UK, disposal of packaged intermediate-level radioactive waste may involve waste packages being placed in a geological disposal facility (GDF) and surrounded by a cementitious backfill. Cracking of the backfill could occur due to a number of mechanisms, and this could affect the post-closure performance of the GDF.This work has assessed potential cracking in the backfill during the backfilling and early post-closure period of GDF vaults with an open crown space in a higher strength rock. From the comprehensive range of processes considered, three were identified as potentially causing cracking: (1) during backfilling, plastic settlement under solid horizontal surfaces could result in horizontal gaps beneath waste packages; (2) within days of backfilling, early-age thermal contraction of the backfill could result in primarily vertical cracks; (3) over a number of years, expansion of waste packages could result in large horizontal cracks.A groundwater flow model incorporating a representation of the cracks was used to calculate flows through a backfilled GDF vault, and through the cracks themselves. Including cracks increased the flow rate significantly. A reactive transport model was used to estimate the evolution of the pore water chemistry as groundwater flows through the cracked backfill. Calcite and brucite were predicted to precipitate, with brucite subsequently dissolving. Calcite build-up could seal some cracks.

Rate and speciation of volatile carbon-14 and tritium releases from irradiated graphite

AbstractThe release and migration of gaseous carbon-14 has been identified as a key issue for geological disposal of intermediate-level radioactive wastes in the UK. A significant fraction of carbon-14 in the UK inventory is in irradiated graphite. This paper describes measurements of gaseous carbon-14 releases from irradiated graphite on immersion in alkaline solution. Apparatus has been developed to discriminate organic and inorganic (14CO/14CO2) species in the gas phase by means of selective oxidation and capture. In the initial experiment, small amounts of gaseous carbon-14 (∼4 Bq) were released from 9 g of crushed graphite within a two-week period. In a long-term experiment, cumulative releases were measured periodically from an intact specimen of graphite over a 14 month period. A small fraction of the graphite carbon-14 inventory was released to the gas phase (∼0.004% as CO/CO2 and ∼0.001% associated with organic compounds). A larger quantity of carbon-14, about 0.1%, was released to the solution phase and was thought to be mainly 14CO2, with some possible organic component. In general, the rate of gaseous carbon-14 release decreased with time. The results suggest a small initial release of relatively labile, accessible carbon-14, with longer term release occurring at a much slower rate. Tritium (T) releases were also measured.

Radionuclide diffusion into undisturbed and altered crystalline rocks

AbstractAn extensive set of porosity, ε, effective diffusion coefficient, De, and hydraulic conductivity, K, data were obtained from 45 granitic samples from the Bohemian Massif, Czech Republic. The measured dataset can be used to define parameter ranges for data to be used in safety assessment calculations for a deep (&gt;400 m) radioactive waste repository, even though the samples originated from shallower depths (&lt;108 m). The dataset can also be used for other purposes, such as evaluating the migration of contaminants in granitic rock (e.g. from shallow intermediate-level radioactive waste repositories and chemical waste repositories).Sample relaxation and ageing processes should be taken into account in research otherwise migration parameters might be overestimated in comparisons between lab results and those determined in situ.

Application of risk-based approach to post-closure safety assessment in radioactive waste disposal: An integration of complex radiation exposure situations

Based on the concept of risk, the post-closure safety criteria for High Level Radioactive Waste (HLW) disposal are under development in Korea. For compliance with the safety criteria of disposal, the risk-based safety assessment methodology should be implemented. In this paper, the authors suggest a risk-based approach to safety assessment for Korea radioactive waste disposal to evaluate an aggregative radiological risk for scenarios of complex radiation exposure situations; accordingly, the most plausible three cases are carefully selected. This risk-based approach deals with the scenarios from the viewpoint of the receptor to estimate the total risk. For respective exposure situations, it considers the occurrence probabilities of the relevant exposure scenarios as the probability of giving rise to doses to estimate the total risk to a representative person by aggregating the respective risks. In addition, the probability of an exposure scenario is estimated on the assumption that the initiating events influence release mechanisms and transport pathways. This integrated approach enables a systematic risk assessment and is informative when judging the probable overall risk for complex exposure situations of radioactive waste disposal. In this paper, the risk-based approach is applied for the case of the Low and Intermediate Level Radioactive Waste (LILW) Disposal, and also the results should be applicable to the HLW disposal facilities.

방사성폐기물드럼 핵종재고량 평가시설 구축에 따른 방사선차폐 영향평가

한국원자력연구원 내 저장중인 중 저준위방사성폐기물의 안전한 영구처분을 위하여 방사성폐기물처리시설에서는 이들 저장폐기물의 핵종재고량 평가를 위한 드럼핵종분석장치의 도입을 기획하고 있으며, 이 장치를 운영하고 유지보수하기 위한 전용의 평가시설 구축을 추진하고 있다. 이 평가시설 외부에는 제1방사성폐기물저장시설이 인접하여 있으며 내부에는 평가대상드럼들과 드럼핵종분석장치의 밀도보정용 선원 등이 존재함으로 이들이 방사선원항으로 작용하여 평가시설 내 외부의 선량률에 영향을 미치게 된다. 따라서 방사선원항 주변의 콘크리트 구조물에 대한 방사선차폐 영향을 평가하여야 한다. 본 연구에서는 MCNP 코드를 이용하여 해당 방사선원항으로부터 평가시설 내부 방사선관리구역을 둘러싸고 있는 콘크리트 구조물에 대한 방사선 안전성을 평가하였다. 평가결과 현재 고려되고 있는 콘크리트 구조물의 약 30 cm 두께는 해당 방사선원항으로부터의 방사선을 차폐하기에 충분한 것으로 확인되었다. In order to dispose of the LILW(low and intermediate level radioactive waste) stored at KAERI, the radwaste drum assay system will be introduced to evaluate the radioisotopes inventory of stored drums. At present, the construction project of the dedicated assay facility to operate it and carry out routine maintenance of that equipment has been conducting at the radwaste treatment facility. Since that facility will be constructed in front of a 1st radwaste storage facility as well as the radwaste drums to be assayed and the transmission source in the radwaste drum assay system are in that facility, they could act as the radioactive sources and then, would affect the dose rate at the inside and the outside of the facility. Therefore, the radiation shielding should be evaluated through the concrete wall near to the radioactive sources whether the wall thickness is sufficient against the regulations. In this study, the radiation safety for the concrete wall around the radiation controlled area in the radwaste drum assay facility was evaluated by the MCNP code. From the evaluation results, the thickness of those concrete walls which are under consideration of about 30 cm was enough to shield the radiation from the radioactive sources.

Radiological and geochemical characteristics of an ultramafic massif (NE Portugal) regarding the site aptness to host a near surface repository for low and intermediate level radwaste

The search for suitable sites to install near-surface repositories for low to intermediate level radioactive wastes has increased in the recent years, reflecting both the continuous rise in radioactive materials applications by the society and the increasing awareness of public perception regarding radwaste management. International recommendations towards the application of higher safety standards to low and intermediate level waste disposal sites being developed by the International Atomic Energy Agency (IAEA), and the European Commission decision that each Member State must safely dispose its own radwaste in its territory, according to national radwaste management plans, has leading to new soundly scientific studies. In Portugal, possible locations for that purpose are under study since 2006, following the pre-requisites indicated by IAEA. One of these sites overlaps a significant portion of an ultramafic massif, corresponding to an allochthonous and metamorphosed ophiolite complex. In the surveyed area of this massif, the prevailing rock type consists of (serpentinised) peridotites, occasionally containing irregular bodies of (amphibolitised) gabbros, over which an irregular/ill-developed regolith exists. The activities of 40K, 137Cs, 210Pb, 226Ra and 232Th measured in samples representing three consecutive levels of the regolith upper portion in 13 stations confirm the heterogeneity put in evidence through in situ γ-ray surveys previously performed. The available dataset also demonstrates that radionuclide contents decrease prominently with depth, becoming closer to the bedrock values (up to ca. 41 Bq kg−1 for 40K in flaser-gabbros and near zero for all the remaining cases) but never reaching them. Therefore, the low/moderate radionuclide activity in the regolith upper portion reflect the contribution of sources that are independent of the lithological background; i.e. they indicate external and miscellaneous contributions to the “soil”-forming components, most of them coming from (dry or wet) atmospheric deposition. The chemical composition of the (serpentinised) peridotites and (amphibolitised) gabbros in the surveyed area, as well as their alteration products included in regolith, is homogeneous enough to be used as a faithful reference in monitoring studies of any kind. In addition, the significant 137Cs activities measured in some top “soil” samples (up to 92 Bq kg−1) strongly suggest that there are mineralogical specificities of the peridotite-derived regolith able to fix that radionuclide. This particular feature reinforces the existing criteria that favour the peridotites of Morais as a suitable host for a near surface repository of low to intermediate level radwaste.

A hybrid modeling approach to evaluate the groundwater flow system at the low- and intermediate-level radioactive waste disposal site in Gyeong-Ju, Korea

The development and implementation of a hybrid discrete fracture network/equivalent porous medium (DFN/EPM) approach to groundwater flow at the Gyeong-Ju low- and intermediate-level radioactive waste (LILW) disposal site in the Republic of Korea is reported. The geometrical and hydrogeological properties of fractured zones, background fractures and rock matrix were derived from site characterization data and implemented as a DFN. Several DFN realizations, including the deterministic fractured zones and the stochastic background fractures, whose statistical properties were verified by comparison with in-situ fracture and hydraulic test data, were suggested, and they were then upscaled to continuums using a fracture tensor approach for site-scale flow simulations. The upscaled models were evaluated by comparison to in-situ pressure monitoring data, and then used to simulate post-closure hydrogeology for the LILW facility. Simulation results demonstrate the importance of careful characterization and implementation of fractured zones. The study highlighted the importance of reducing uncertainty regarding the properties and variability of natural background fractures, particularly in the immediate vicinity of repository emplacement.

The effects of γ-radiation on model vitreous wasteforms intended for the disposal of intermediate and high level radioactive wastes in the United Kingdom

The effect of γ-radiation on a variety of model vitreous wasteforms applied to, or conceived for, immobilisation of UK intermediate and high level radioactive wastes was studied up to a dose of 8MGy. It was determined that γ-irradiation up to this dose had no significant effect upon the mechanical properties of the wasteforms and there was no evidence of residual structural defects. FTIR and Raman spectroscopy showed no evidence of radiation directly affecting the silicate network of the glasses. The negligible impact of this γ-irradiation dose on the physical properties of the glass was attributed to the presence of multivalent ions, particularly Fe, and a mechanism by which the electron–hole pairs generated by γ-irradiation were annihilated by the Fe2+–Fe3+ redox mechanism. However, reduction of sulphur species in response to γ-radiation was demonstrated by S K-edge XANES and XES data.

Confinement matrices for low- and intermediate-level radioactive waste

Mining of uranium for nuclear fuel production inevitably leads to the exhaustion of natural uranium resources and an increase in market price of uranium. As an alternative, it is possible to provide nuclear power plants with reprocessed spent nuclear fuel (SNF), which retains 90% of its energy resource. The main obstacle to this solution is related to the formation in the course of the reprocessing of SNF of a large volume of liquid waste, and the necessity to concentrate, solidify, and dispose of this waste. Radioactive waste is classified into three categories: low-, intermediate-, and high-level (LLW, ILW, and HLW); 95, 4.4, and 0.6% of the total waste are LLW, ILW, and HLW, respectively. Despite its small relative volume, the radioactivity of HLW is approximately equal to the combined radioactivity of LLW + ILW (LILW). The main hazard of HLW is related to its extremely high radioactivity, the occurrence of long-living radionuclides, heat release, and the necessity to confine HLW for an effectively unlimited time period. The problems of handling LILW are caused by the enormous volume of such waste. The available technology for LILW confinement is considered, and conclusion is drawn that its concentration, vitrification, and disposal in shallow-seated repositories is a necessary condition of large-scale reprocessing of SNF derived from VVER-1000 reactors. The significantly reduced volume of the vitrified LILW and its very low dissolution rate at low temperatures makes borosilicate glass an ideal confinement matrix for immobilization of LILW. At the same time, the high corrosion rate of the glass matrix at elevated temperatures casts doubt on its efficient use for immobilization of heat-releasing HLW. The higher cost of LILW vitrification compared to cementation and bitumen impregnation is compensated for by reduced expenditure for construction of additional engineering barriers, as well as by substantial decrease in LLW and ILW volume, localization of shallow-seated repositories in various geological media, and the use of inexpensive borosilicate glass.

Determination of a representative composition of dry active waste from Korean nuclear power plants

As part of the following-up action pertaining to the construction and operation permit for the first stage of the LILW (Low- and Intermediate-Level Radioactive Waste) repository, the preparation for a large-scale in situ experiment is underway by the Korea Radioactive Waste Management Corporation (KRMC) for a realistic assessment of the characteristics of gas generation after the post-closure phase of a repository. In this paper, we discuss a method of determining the representative composition of simulated dry active waste and a related fabrication plan for this material. After a comparison with experimental gas generation results from Finland, dry active waste content was chosen for a large-scale gas generation experiment. Six different types of materials and details on their simulated dry active waste contents are derived with the total mass and density for both a 200L and a 320L drum.

Modelling of a Generic Near-Surface Disposal System

ABSTRACTDevelopment of numerical tools for performance assessment studies of radioactive waste disposal facilities, must address the management of the wide-ranging uncertainties associated with the long-term behaviour of these complex systems. Different approaches and assumptions are made in order to identify and describe relationships between the disposal system and its environment. They take into account, among other factors, the uncertainties associated with temporal evolution of the system within a proposed scenario; the landscape changes arising from future human actions, climate and geological events and processes; the relationships between components of the disposal system and its immediate environment; the behaviour and characteristics of radionuclides within the system and their role in contributing to radiation exposure. In all cases, the different scenario-based models are typically used to determine the radiological significance of potential future discharges from waste disposal facilities. However, it is important to keep always in mind that in any specific case, the purpose of developing and/or applying a model may vary from a simple calculation (e.g. to support concept development) to detailed site-specific performance assessment in support of a disposal license application. The assumptions and modelling simplifications that are appropriate to one type of calculation may not be so easily justified in different circumstances. In order to develop the capability of modelling different long-term scenarios for a generic disposal site for low and intermediate level radioactive wastes, implementation of models of both the near-field/geosphere and biosphere were performed using general approaches for geosphere-biosphere interface, with sub-models for the whole system.