Low And Intermediate Level Radioactive Waste Research Articles

Post-closure safety assessment of near surface disposal facilities for disused sealed radioactive sources

Great attention has been recently paid to the post-closure safety assessment of low- and intermediate-level radioactive waste (LILW) disposal facility for disused sealed radioactive sources (DSRSs) around the world. Although the amount of volume of DSRSs generated from industry, medicine and research and education organization was relatively small compared with radioactive wastes from commercial nuclear power plants, some DSRSs can pose a significant hazard to human health due to their high activities and long half-lives, if not appropriately managed and disposed. In this study, post-closure safety assessment was carried out for DSRSs generated from 1991 to 2014 in Korea in order to ensure long-term safety of near surface disposal facilities. Two kinds of disposal options were considered, i.e., engineered vault type disposal facility and rock-cavern type disposal facility. Rock-cavern type disposal facility has been under operation in Gyeongju city, republic of Korea since August 2015 and engineered vault type disposal facility will be constructed until December 2020 in the vicinity of rock-cavern disposal facility. Assessment endpoint was individual dose to the member of critical group, which was modeled by GoldSim, which has been widely used as probabilistic risk analysis software based on Monte Carlo simulation in the area of safety assessment of radioactive waste facilities. In normal groundwater scenario, the maximum exposure dose was extremely low, approximately 1×10−7mSv/yr, for both disposal options and satisfied the regulatory limit of 0.1mSv/yr. However, in the drinking well scenario, the maximum exposure dose for engineered vault type disposal facility was assessed as 2.022mSv/yr where the value exceeded the regulatory limit of 1mSv/yr. The maximum exposure dose for rock-cavern type disposal facility was calculated to be 0.634mSv/yr, whose results was relatively very close to the regulatory limit considering high uncertainty of long-term environmental conditions. It was demonstrated that DSRSs including the radionuclides of 14C, 226Ra and 241Am significantly contributed to the large portion of exposure dose to the public based on the long-term safety assessment. Therefore, it was recommended that the near surface disposal of DSRSs containing 14C, 226Ra and 241Am should be restricted and managed by long-term interim storage option in order to minimize their potential radiological health effects.

An evaluation of nuclide release from a trench-type LILW repository

A simple and effective model and a GoldSim (GoldSim, 2006) template program, by which a probabilistic safety assessment of a conceptual trench-type repository for low- and intermediate level radioactive waste (LILW) disposal can be carried out under various nuclide release scenarios, have been developed. To quantify the exposure dose rates due to nuclide release from the trench and transport through the various pathways possible in the near- and far-fields of the repository system under a base case and some alternative scenarios, illustrative evaluations for a comparison among the scenarios as well as a sensitivity of shortcut pathways generated due to earthquakes on the nuclide transport are made and demonstrated. To this end, by changing the conservative base case nuclide release scenario under which all portions of the cap of the trench are failed unconditionally and immediately after a closure of the repository, a total of four other alternative scenarios were separately evaluated for the total exposure dose rates to the farming exposure group and then compared to the base case results. Among them, an earthquake scenario shows a dominant behavior almost throughout the whole time span. To see the influence of all shortcut travel pathways that are assumed to be newly generated by this earthquake scenario, their sensitivities to the exposure dose rates to the farming exposure group were also made and compared to each other.

Sensitivity analysis of a final repository model with quasi-discrete behaviour using quasi-random sampling and a metamodel approach in comparison to other variance-based techniques

This paper contributes to the investigation of recent computationally efficient variance-based methods for sensitivity analysis and sampling schemes on the basis of a Performance Assessment (PA) model for a repository for Low- and Intermediate-Level Radioactive Waste (LILW) in an abandonned salt mine. The PA model takes account of typical characteristics of repository systems including a quasi-discrete nature.The results indicate best convergence of the State-Dependent-Parameter (SDP) metamodel method and the simple first-order sensitivity indices (SI1) calculation scheme (EASI) if combined with the LpTau low discrepancy sampling method. The SI1 indices obtained from the SDP and EASI methods are very similar. The Fourier Amplitude Sensitivity Test (EFAST) seems to have some convergence problems, likely due to the quasi-discrete behaviour of the PA model. Due to parameter interactions, none of the investigated methods, however, could directly determine all significant parameters. This is illustrated by means of a simplified way of factor fixing.

Analysis of Uncertainties Affecting Numerical Transport Models for a Potential Radioactive Waste Disposal Site

Numerical models have been commonly used to study solute transport behaviours both in laboratory-and field-scale cases. However, numerical interpretation and predictions of solute transport are complicated by various uncertainties and therefore can be misleading and lead to incorrect understanding of solute transport behaviours in hydrogeological systems. Such uncertainties have been discussed with a case study of radionuclides transport at a potential site for disposal of Low-and Intermediate-Level radioactive Wastes (LILW) in southern China. Firstly, pre-modelling assumptions cause uncertainties for model performance and have a significant influence on the model output. Therefore, such assumptions and related factors should be thoroughly considered and minimized in laboratory experiments or investigated in the field. Secondly, difference in the calculated plumes caused by uncertainties in the development of conceptual models suggests that the hydrogeological conceptual model should be developed based on intrinsic understanding of the transport problems and relevant data collection. Thirdly, relative sensitivity (RS) has been calculated to analyse the parameter uncertainties. Results indicate that the uncertainties are tracer-, parameter-and time-dependent. Therefore, model uncertainties arising from parameters should be considered for each specific parameter with the specific period. Finally, prediction scenarios are associated with uncertainties in the model prediction stage. Results suggest that the uncertainties in scenarios of future transport conditions should be recognized prior to model prediction. This study illustrates some of the uncertainties that might affect numerical transport models and therefore will be a useful reference for application of solute transport models to assessment of contamination risks from a LILW disposal site.

A field-scale long-term study on radionuclide transport through weathered granites at a site in southern China

A regional strategy for safety disposal of low- and intermediate-level radioactive wastes (LILW) has been implemented in China to protect humans and the environment. A joint onsite and laboratory investigation was conducted for a field site in southern China to assess the probability for safe disposal of LILW, which requires the understanding of long-term radionuclide transport behaviors under field conditions. This study presents the field-scale modeling of radionuclide transport through weathered granites for a conservative, a weakly sorbing and a strongly sorbing tracer by incorporating laboratory and field data. The field-scale radionuclide transport model was developed on the basis of a validated long-term groundwater flow model and field-measured dispersion coefficient, as well as laboratory-characterized strontium and cesium distribution coefficients in the weathered granites. The model was then used to perform the long-term transport prediction and risk assessment of radionuclide pollution for both the natural site setting and the graded site setting. Model simulation reveals that the numerical sensitivities of calculated concentrations are tracer dependent and changing with time. The conservative radionuclide is most sensitive to changes in hydraulic conductivity (K) while slightly sensitive to changes in effective porosity (φ), specific yield (μ) and longitudinal dispersion coefficient (D L), indicating advection is the main transport process of conservative radionuclide. The weakly and strongly sorbing tracers, on the other hand, are most sensitive to changes in the distribution coefficient (K d) and less sensitive to changes in the rest of model parameters, revealing that sorption is the main process for controlling the transport of sorbing tracers. A conservative radionuclide plume moves at an average velocity of about 54 m/a, which is too fast to be considered as safe disposal under the natural site setting. However, the plume of the conservative radionuclide could be slowed down to a velocity around 5.3 m/a due to the reduction of the hydraulic gradient under the graded site setting. Therefore, the conservative radioactive wastes could be disposed at the mid-eastern part of the site under the graded site setting because the transport path has been prolonged and thus no conservative radionuclides could migrate out of the site in a reasonable timeframe. For the sorbing tracers, however, results of the computed transport distance are 40 and 2 m at 500 years, respectively. Therefore, they can be disposed safely at the site under both natural and graded site settings. This study provides an insight to the field-scale long-term behaviors of radionuclide transport. The integrated modeling method presented in this study is most useful for the environmental impact assessment of the site conditions relevant to the safe disposal of hazardous wastes.

Assessment of radioactive wastes from a DCLL fusion reactor: Disposal in El Cabril facility

Under the Spanish Breeding Blanket Technology Programme TECNO_FUS a conceptual design of a DCLL (Dual-Coolant Lithium–Lead) blanket-based reactor is being revised. The dually cooled breeding zone is composed of He/LiPb and SiC as material of the liquid metal flow channel inserts. Structural materials are ferritic-martensitic steel (Eurofer) for the blanket and austenitic steel (SS316LN) for the vacuum vessel (VV) and the cryostat.In this work, radioactive wastes are assessed in order to determine if they can be disposed as low and intermediate level radioactive waste (LILW) in the Spanish near surface disposal facility of El Cabril. Also, unconditional clearance and recycling waste management options are studied.The neutron transport calculations have been performed with MCNPX code, while the ACAB code is used for calculations of the inventory of activation products and for activation analysis, in terms of waste management ratings for the options considered.Results show that the total amount of the cryostat can be disposed in El Cabril joined to the outer layer of both VV and channel inserts, whereas only concrete-made biological shield can be managed through clearance and none of the steels can be recycled. Those results are compared with those corresponding to French regulation, showing similar conclusions.

Deployment of Radioactive Waste Disposal Facility with the Introduction of Nuclear Power Plants (NPP) in Kenya

This paper describes basic plans for the development of a radioactive waste disposal facility with the introduction of Nuclear Power Plants (NPPs) for Kenya. The specific objective of this study was to estimate the total projected waste volumes of low- and intermediate-level radioactive waste (LILW) expected to be generated from the Kenyan nuclear power programme. The facility is expected to accommodate LILW to be generated from operation and decommissioning of nuclear power plants for a period of 50 years. An on-site storage capacity of 700 <TEX>$m^3$</TEX> at nuclear power plant sites and a final disposal repository facility of more than 7,000 <TEX>$m^3$</TEX> capacity were derived by considering Korean nuclear power programme radioactive waste generation data, including Kori, Hanbit, and APR 1400 nuclear reactor data. The repository program is best suited to be introduced roughly 10 years after reactor operation. This study is important as an initial implementation of a national LILW disposal program for Kenya and other newcomer countries interested in nuclear power technology.

Glasses for immobilization of low- and intermediate-level radioactive waste

Reprocessing of spent nuclear fuel (SNF) for recovery of fissionable elements is a precondition of long-term development of nuclear energetics. Solution of this problem is hindered by the production of a great amount of liquid waste; 99% of its volume is low- and intermediate-level radioactive waste (LILW). The volume of high-level radioactive waste (HLW), which is characterized by high heat release, does not exceed a fraction of a percent. Solubility of glasses at an elevated temperature makes them unfit for immobilization of HLW, the insulation of which is ensured only by mineral-like matrices. At the same time, glasses are a perfect matrix for LILW, which are distinguished by low heat release. The solubility of borosilicate glass at a low temperature is so low that even a glass with relatively low resistance enables them to retain safety of under-ground LILW depositories without additional engineering barriers. The optimal technology of liquid confinement is their concentration and immobilization in borosilicate glasses, which are disposed in shallow-seated geological repositories. The vitrification of 1 m3 liquid LILW with a salt concentration of ∼300 kg/m3 leaves behind only 0.2 m3 waste, that is, 4–6 times less than by bitumen impregnation and 10 times less than by cementation. Environmental and economic advantages of LILW vitrification result from (1) low solubility of the vitrified LILW in natural water; (2) significant reduction of LILW volume; (3) possibility to dispose the vitrified waste without additional engineering barriers under shallow conditions and in diverse geological media; (4) the strength of glass makes its transportation and storage possible; and finally (5) reliable longterm safety of repositories. When the composition of the glass matrix for LILW is being chosen, attention should be paid to the factors that ensure high technological and economic efficiency of vitrification. The study of vitrified LILW from the Kursk nuclear power plant with high-power channel reactors (HPCR; equivalent Russian acronym, RBMK) and the Kalinin nuclear power plant with pressurized water reactors (PWR; equivalent Russian acronym VVER) after their 14-yr storage in the shallow-seated repository at the MosNPO Radon testing ground has confirmed the safety of repositories ensured by confinement properties of borosilicate matrix. The most efficient vitrification technology is based on cold crucible induction melting. If the content of a chemical element in waste exceeds its solubility in glass, a crystalline phase is formed in the course of vitrification, so that the glass ceramics become a matrix for such waste. Vitrified waste with high Fe; Na and Al; Na, Fe, and Al; Na and B is characterized. The composition of frit and its proportion to waste depends on waste composition. This procedure requires careful laboratory testing.

Electrochemical migration technique to accelerate ageing of cementitious materials

Durability assessment of concrete structures for constructions in nuclear waste repositories requires long term service life predictions. As deposition of low and intermediate level radioactive waste (LILW) takes up to 100 000 years, it is necessary to analyze the service life of cementitious materials in this time perspective. Using acceleration methods producing aged specimens would decrease the need of extrapolating short term data sets. Laboratory methods are therefore, needed for accelerating the ageing process without making any influencing distortion in the properties of the materials. This paper presents an electro-chemical migration method to increase the rate of calcium leaching from cementitious specimens. This method is developed based on the fact that major long term deterioration process of hardened cement paste in concrete structures for deposition of LILW is due to slow diffusion of calcium ions. In this method the cementitious specimen is placed in an electrochemical cell as a porous path way through which ions can migrate at a rate far higher than diffusion process. The electrical field is applied to the cell in a way to accelerate the ion migration without making destructions in the specimen's micro and macroscopic properties. The anolyte and catholyte solutions are designed favoring dissolution of calcium hydroxide and compensating for the leached calcium ions with another ion like lithium.

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.

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.

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.

Assessment of site conditions for disposal of low- and intermediate-level radioactive wastes: A case study in southern China

Near surface disposal of low- and intermediate-level radioactive wastes (LILW) requires evaluating the field conditions of the candidate site. However, assessment of the site conditions may be challenging due to the limited prior knowledge of some remote sites, and various multi-disciplinary data requirements at any given site. These situations arise in China as in the rest of the industrialized world, particularly since a regional strategy for LILW disposal has been implemented to protect humans and the environment. This paper presents a demonstration of the site assessment process through a case study focusing mainly on the geologic, hydrogeologic and geochemical characteristics of the candidate site.A joint on-site and laboratory investigation, supplemented by numerical modeling, was implemented in this assessment. Results indicate that no fault is present in the site area, although there are some minor joints and fractures, primarily showing a north–south trend. Most of the joints are filled with quartz deposits and would thus function hydraulically as impervious barriers. Investigation of local hydrologic boundaries has shown that the candidate site represents an essentially isolated hydrogeologic unit, and that little or no groundwater flow occurs across its boundaries on the north or east, or across the hilly areas to the south. Groundwater in the site area is recharged by precipitation and discharges primarily by evapo-transpiration and surface flow through a narrow outlet to the west. Groundwater flows slowly from the hilly area to the foot of the hills and discharges mainly into the inner brooks and marshes. Some groundwater circulates in deeper granite in a slower manner. The vadose zone in the site was investigated specially for their significant capability for restraining the transport of radionuclides. Results indicate that the vadose zone is up to 38m in thickness and is made up of alluvial clay soils and very highly weathered granite. The vadose zone has low saturated hydraulic conductivities on the order of 10-5cm/s and in this respect is well-suited for the disposal of LILW. The saturated formations are primarily made up of silt and moderately-to-slightly weathered granite, which exhibit even lower hydraulic conductivities, on the order of 10−6cm/s, also favorable for restraining the transport of radionuclides. Chemical analyses indicate that the groundwaters at the site are of the HCO3–Na·Ca and HCO3·SO4–Na·Ca types and are weakly corrosive to concrete and steel. Geochemical analyses indicate that the rock and soil materials (particularly weathered granite) at the site contain very small fractions of colloidal particles and exhibit low Cation Exchange Capacities (CEC), and would therefore have limited capacity for sorption of radionuclides.Groundwater flow and solute transport models of the candidate site have been developed using MODFLOW and MT3DMS, incorporating the data obtained during the assessment program. Calibration was based on the available measured groundwater level fluctuations and tracer concentrations from in situ dispersion tests. The longitudinal dispersion coefficient as determined in calibration is equal to 5.0×10−3m2/d. Numerical sensitivity analyses indicate that the hydraulic conductivity and the longitudinal dispersion coefficient are the key parameters controlling the transport of radionuclides, while the numerical model is not sensitive to changes in the effective porosity and the specific yield. Preliminary predictions have been performed with the calibrated model both for the natural setting of the site and the graded site in which the valleys of the site are backfilled with low permeable materials. Results indicate that the proposed site grading increases the safety of the site for disposal of LILW by reducing both the groundwater level and the hydraulic gradient and that radionuclide transport would not likely be a problem or cause groundwater contamination. Although there are some problems remaining to be addressed in future work, the conclusion of the assessment is that the conditions at this site are appropriate for LILW disposal. This study provides an example of the procedures necessary in an assessment of site conditions relevant to the safe disposal of LILW. Such an assessment is crucial to the site selection process and to subsequent environmental impact assessment.

A Preliminary Study on the Transport Behavior for a Potential Disposal Site of LILW in Southern China

Near surface disposal of low and intermediate level radioactive waste (LILW) requires a detailed site investigation of the potential sites, in which the transport behavior of solutes in the groundwater system is one of the key processes that needed to be addressed. An investigation is undertaking for the disposal of LILW at a potential site in Southern China. In-situ test have been conducted to study the hydrogeologic characteristics of the site. Firstly, tests including pit permeability tests, injecting tests, water pressure tests, pumping tests and laboratory permeability tests have been performed according to the specific field conditions. Hydraulic conductivities for different layers of rock and soils have been calculated and their recommended values have been presented. And then in-situ dispersion tests have been performed at an area adjacent to the disposal site with non-sorbed tracers. A numerical model has been developed for the site based on data obtained during the site investigation, and has been calibrated with available measured groundwater level and measured tracer concentrations in the dispersion tests. Calibrated results indicate that the longitudinal dispersion coefficient is equal to 5.0×10-3 m2/d. Preliminary predictions have been performed for the groundwater flow and solute advection-dispersion behavior according to the design of the site, in which the center valley will be backfilled with low permeable materials. Predicted results indicate that the groundwater exhibits a lower water table and a smaller hydraulic gradient under designed site condition than under natural condition. All the tracers remain underground in the site and transport slowly mainly through advection and dispersion. Finally, conclusions for the study have been presented and the key tasks for future work have been discussed. This study provides an insight understanding of the hydrogeology characteristics of the disposal site and is useful for the assessment of environment impacts of the site under disposal of LILW.

Analysis of Leaching Behavior of Simulated LILW Glasses by Using the MCC-1 Test Method

Glasses developed for the treatment of low- and intermediate-level radioactive waste (LILW) from nuclear power plants were evaluated by using the Material Characterization Center-1 (MCC-1) leaching method. Tests were conducted at temperatures of 40, 70, and 90°C for three weeks in pH buffer solutions spanning the range from pH 4 to pH 11. Normalized mass losses and forward dissolution rates of major glass elements (B, Na, Al, Si, Co, Cs) were analyzed under each leaching condition. From these data, the forward rate equations depending on pH and temperature were defined using a nonlinear regression method. This equation provided an overall diagram of the leach rate with these parameters (i.e., pH and temperature). The forward dissolution rates of the glasses were found to have a V-shaped pH dependence. The glasses in the pH ranges were found to have a forward dissolution rate below 10 g/m2·d, when the temperatures were between 40 and 90°C and the leachant condition was pH 4–11. Except for the DG2 glass, the minimum forward dissolution rate (0.01–1 g/m2·d) was obtained at approximately pH 7–8. Compared with previously reported results, the developed glasses showed relatively high forward dissolution rates at the neutral region, while showing similar or lower rates compared with other glasses and ceramic waste forms at both extremes of pH.

Evaluation of nuclide release scenarios for a hypothetical LILW repository

A program for the safety assessment and performance evaluation of a low- and intermediate-level radioactive waste (LILW) repository system has been developed. Utilizing GoldSim (2006), the program evaluates nuclide release and transport into the geosphere and biosphere under various disruptive natural and manmade events and scenarios that can occur after a waste package failure. We envisaged and illustrated these events and scenarios as occurring after the closure of a hypothetical LILW repository, and they included the degradation of various manmade barriers, pumping well drilling, and natural disruptions such as the sudden formation of a preferential flow pathway in the far-field area of the repository. Possible enhancement of nuclide transport facilitated by colloids or chelating agents is also dealt with. We used the newly-developed GoldSim template program, which is capable of various nuclide release scenarios and is greatly suited for simulating a potential repository given the geological circumstances in Korea, to create the detailed source-term and near-field release scheme, various nuclide transport modes in the far-field geosphere area, and the biosphere transfer. Even though all parameter values applied to the hypothetical repository were assumed, the illustrative results, particularly the probabilistic calculations and sensitivity studies, may be informative under various scenarios. ( GoldSim, LILW, Nuclide transport, Safety assessment, Scenario).

Glassy Wasteforms for Nuclear Waste Immobilization

Glassy wasteforms currently being used for high-level radioactive waste (HLW) as well as for low- and intermediate-level radioactive waste (LILW) immobilization are discussed and their most important parameters are examined, along with a brief description of waste vitrification technology currently used worldwide. Recent developments in advanced nuclear wasteforms are described such as polyphase glass composite materials (GCMs) with higher versatility and waste loading. Aqueous performance of glassy materials is analyzed with a detailed analysis of the role of ion exchange and hydrolysis, and performance of irradiated glasses.

Usefulness of systematic in situ gamma-ray surveys in the radiometric characterization of natural systems with poorly contrasting geological features (examples from NE of Portugal)

This paper focuses on the starting point of various studies that are being carried out in two possible locations being considered to host a hypothetical site for a repository for low and intermediate level radioactive waste (LILW) produced in Portugal in compliance with international requirements on the long-term safety of this kind of repository. Previous studies concerning the geology of the much larger geographical areas where these locations are included were fundamental in the choice of these locations and for the design of the survey strategy. One of the fundamental assessment studies during the site-selection is the overall radiological characterization of the locations and its relation to the geology. This paper pretends to show the adequability of using a fast and reasonably inexpensive survey technique such as in situ gamma-ray portable detectors, to access the radiometric response of the systems in study by providing the radiometric mapping of the areas. The existence of adequate radiometric maps represents a critical pre-requisite to constrain both the number and spatial distribution of samples to be collected for further analysis, sustaining as well the subsequent extrapolation of results needed to fully characterise the surveyed system. Both areas were surveyed using portable gamma-ray spectrometers with NaI(Tl) detectors. In situ gamma-ray measurements have clearly shown not only the poorly contrasting geological features, but also their differences representing: (i) a deformed/metamorphosed ophiolite complex and (ii) a monotonous meta-sedimentary sequence. The radiometric maps obtained have show heterogeneities that reflect mostly changes in rock-forming mineral assemblages, even in the presence of small variations of gamma radiation. These maps support objective criteria about the number/distribution of samples to be collected for subsequent comprehensive studies and reinforce the valuable contribution of in situ gamma spectrometry to assess, in radiological terms, the prevalent geological features.

Use of strategic environmental assessment in the site selection process for a radioactive waste disposal facility in Slovenia

The benefits of strategic environmental considerations in the process of siting a repository for low- and intermediate-level radioactive waste (LILW) are presented. The benefits have been explored by analyzing differences between the two site selection processes. One is a so-called official site selection process, which is implemented by the Agency for radwaste management (ARAO); the other is an optimization process suggested by experts working in the area of environmental impact assessment (EIA) and land-use (spatial) planning. The criteria on which the comparison of the results of the two site selection processes has been based are spatial organization, environmental impact, safety in terms of potential exposure of the population to radioactivity released from the repository, and feasibility of the repository from the technical, financial/economic and social point of view (the latter relates to consent by the local community for siting the repository). The site selection processes have been compared with the support of the decision expert system named DEX. The results of the comparison indicate that the sites selected by ARAO meet fewer suitability criteria than those identified by applying strategic environmental considerations in the framework of the optimization process. This result stands when taking into account spatial, environmental, safety and technical feasibility points of view. Acceptability of a site by a local community could not have been tested, since the formal site selection process has not yet been concluded; this remains as an uncertain and open point of the comparison.