Near-surface Disposal Research Articles

Human intrusion dose assessment for the safe disposal of radioactive spent resin mixed waste at disposal facilities

This study explores the feasibility of safely disposing of spent resin mixed waste from which pure beta emitters(14C, 3H) have been removed 95 % at near-surface disposal facilities and evaluates the radiation dose to the public in the event of human intrusion after the management period of the disposal site. The decommissioning of Wolsong Unit 1, Korea's first commercial heavy water reactor, marks a critical juncture in the field of nuclear energy, highlighting the imperative for advancements in the methodologies of decommissioning, decontamination, and stewardship of mixed spent resin radioactive waste. To this end, six human intrusion scenarios are examined, and a cross-validation is performed using GENII and RESRAD-ONSITE simulation codes to confirm that the public dose criterion of 1 mSv/y is not exceeded under any of the scenarios. This result suggests that the safe disposal of waste from which beta radionuclides have been removed is possible, carrying significant implications for waste management and environmental protection. Such assessments provide scientific evidence for establishing waste disposal strategies and can serve as foundational data for the sustainable utilization of nuclear energy.

Investigating advection–dispersion behavior for simulation of HTO and 238Pu transport in argillaceous shale with different varying degrees of weathering

Abstract Exploring the migration behavior of radioactive nuclides in the surrounding rock media of the disposal site is the basis for the safety evaluation of radioactive waste disposal sites. In this study, the column experiment was employed to evaluate the nuclide migration behavior in the surrounding rock medium of a near-surface disposal site in China and to investigate the advection–dispersion behavior of tritium (HTO) and plutonium-238 (238Pu) in highly weathered argillaceous shale with different degree of weathering. A reasonable numerical model was selected to fit the experimental breakthrough curves (BTCs) and obtain relevant migration parameters. The results showed that: (1) the breakthrough curves of HTO and 238Pu in fractured media with varying degrees of weathering exhibited clear “peak forward” and “tail dragging” phenomena; (2) the stream tube model can better fit the BTCs of the nuclides in the highly weathered fractures and obtain the average dispersion coefficient <D>, average distribution coefficient <K d>, and other parameters; (3) the two-region non-equilibrium model can better fit the BTCs of the nuclides in the penetrating fractures and weak weathering fractures and obtain the volumetric water content of the mobile liquid region (θ m), distribution coefficient (K d), and other parameters.

К вопросу выбора способа захоронения низко- и среднеактивных РАО

The paper overviews different options considered for the near-surface disposal of low- and intermediate-level waste. It evaluates the engineered safety barriers proposed under the near-surface disposal concepts, their functions and designs. The paper demonstrates that the practice providing for RW disposal in shallow near-surface disposal facilities can be feasibly replaced by the one of surface disposal. It considers different engineering designs proposed for such surface repositories, in particular, those preventing water accumulation in the repository structures. It summarizes the findings o some broad safety assessments focused on such designs.

Study of γ-Ray Spectrum Measurement Analysis Algorithm of Radioactive Solid Waste Steel Boxes

Abstract In accordance with safety requirements for the near-surface disposal of low and medium-level radioactive solid waste, accurate identification and measurement of radionuclide types and activities within steel boxes containing such waste are necessary. Conventional sampling techniques are impractical due to nonuniform internal filling and radioactive distribution in the steel boxes used for storage. Nondestructive measurement analysis techniques offer significant advantages by providing comprehensive measurements without causing damage to the steel boxes or generating secondary waste. Therefore, nondestructive techniques are widely applicable and superior to chemical analysis methods. This study presents an algorithm for analyzing γ-ray spectrum measurements of radioactive solid waste steel boxes by combining the traditional computed tomography (CT) principle with a discrete treatment approach. The progressive split voxel measurement method is utilized to discretely handle the radioactive solid waste steel boxes during the measurement and calculation process while considering the impact of voxel interactions on detection efficiency. A detection efficiency calibration model is established using Monte Carlo (MC) simulation in conjunction with the algebraic reconstruction technique (ART). Additionally, an inversion correction of the efficiency for the physically coincident portion of the detected voxel volume is performed during the measurement process. The proposed algorithm is initially applied to the analysis platform of steel boxes that store FA-IV-type radioactive solid waste. The algorithm's performance is further investigated and validated using standard radioactive sources with known activity levels. The results indicate that the relative deviation between the reconstruction results and the nominal values of radioactive activities in the steel boxes is less than 30%. These findings align with the expected results and satisfy the measurement analysis and warehousing processing requirements for waste generated from the decommissioning of China's 101 nuclear reactor. The proposed algorithm holds great potential for widespread application in related detection work.

Impact of oxidation and carbonation on the release rates of iodine, selenium, technetium, and nitrogen from a cementitious waste form

Evaluation of the long-term retention mechanisms and potential release rates for the primary constituents of potential concern (COPCs) (i.e., Tc, I, Se, and nitrate) is necessary to determine if Cast Stone, a radioactive waste form, can meet performance objectives under near-surface disposal scenarios. Herein, a mineral and parameter set accounting for the solubility of I and Se in Cast Stone was developed based on pH-dependent and monolithic diffusion leaching test results, to extend a geochemical speciation model previously developed. The impact of oxidation and carbonation as environmental aging processes on the retention properties of Cast Stone for primary COPCs was systematically estimated. Physically, the effective diffusion coefficients of 4 COPCs in Cast Stone were increased after carbonation and/or oxidation, reflecting an increase in permeability to diffusion. Chemically, i) pH & pe conditions in the original Cast Stone were favorable for the stabilization of Tc, but not for I, Se, and N; ii) oxidation (with/without carbonation) of Cast Stone changed the pe & pH conditions to be detrimental for Tc stabilization; and iii) carbonation (with/without oxidation) of Cast Stone modified the pH & pe conditions to be beneficial for the stabilization of I (in system with Ag added) and Se.

Advection–Dispersion Behavior for Simulation of H-3 and Pu-238 Transport in Undisturbed Argillaceous Shale of a Near-Surface Repository

In this study, a column experiment was employed to evaluate the nuclide migration behavior in the surrounding rock medium of a near-surface disposal site in China and to investigate the advection-dispersion behavior of tritium (H-3) and plutonium-238 (Pu-238) in highly weathered argillaceous shale. A reasonable numerical model was selected to fit the experimental breakthrough curves (BTCs) and to obtain the relevant migration parameters. The results show the following: (1) the internal structure of the highly weathered argillaceous shale exhibited heterogeneity, and the nuclide migration BTC showed characteristics of a "curve peak moving forward" and a "tail curve trailing"; (2) compared with other models, the stream tube mode could better fit the BTCs and obtain the average dispersion coefficient <D>, average distribution coefficient <Kd>, and other parameters; (3) compared to the results of the batch experiment, the distribution coefficient Kd obtained from the column experiment was smaller than that obtained from the batch experiment, which is speculated to be due to the influence of contact time and the contact area between the nuclide and the medium.

Micro- and Nanoscale Surface Analysis of Late Iron Age Glass from Broborg, a Vitrified Swedish Hillfort

AbstractArchaeological glasses with prolonged exposure to biogeochemical processes in the environment can be used to understand glass alteration, which is important for the safe disposal of vitrified nuclear waste. Samples of mafic and felsic glasses with different chemistries, formed from melting amphibolitic and granitoid rocks, were obtained from Broborg, a Swedish Iron Age hillfort. Glasses were excavated from the top of the hillfort wall and from the wall interior. A detailed microscopic, spectroscopic, and diffraction study of surficial textures and chemistries were conducted on these glasses. Felsic glass chemistry was uniform, with a smooth surface showing limited chemical alteration (&amp;lt;150 nm), irrespective of the position in the wall. Mafic glass was heterogeneous, with pyroxene, spinel, feldspar, and quartz crystals in the glassy matrix. Mafic glass surfaces in contact with topsoil were rougher than those within the wall and had carbon-rich material consistent with microbial colonization. Limited evidence for chemical or physical alteration of mafic glass was found; the thin melt film that coated all exposed surfaces remained intact, despite exposure to hydraulically unsaturated conditions, topsoil, and associated microbiome for over 1,500 years. This supports the assumption that aluminosilicate nuclear waste glasses will have a high chemical durability in near-surface disposal facilities.

Safety analysis of near surface nuclear waste repository constituting an aquifer in proximity

Probabilistic safety assessment involves the calculation of risk associated with possible radioisotope contamination of Groundwater. The overall scenario and the consequences that may occur on the failure of the external barriers are identified using an Event Tree analysis to be the multi-barrier failure due to rainwater intrusion into the near-surface disposal site and leaching out of radionuclides into groundwater which is extracted as drinking water by critical groups present around the repository. A program is developed in Python 3.0 for such a scenario to assess necessary parameters such as radioactive release rate, groundwater radionuclide concentration, dose ingested by members and total risk associated with each radioisotope. The initiation is with the assessment of the failure density function of the multi-barrier system operating in redundancy, using the failure density functions. The radionuclide concentrations of the near-surface disposal facility are necessary to be calculated to know the probable release rate of each radionuclide into groundwater. The concentration depletion with Spatio-temporal variance is also considered a factor for radioisotope concentration variation in groundwater. The concentration of radionuclides in groundwater when ingested at a rate of 2.2 L/day helps us to calibrate the Radiation dose to humans through this pathway by using the Ingestion dose coefficient for each isotope fixed by International Atomic Energy Agency [IAEA]. The final risk assessment is accomplished by multiplying the radiation dose ingested by members of the critical group with a risk factor for fatal or non-cancers and critical hereditary effects by the International Commission on Radiological Protection. In the present work, a subjective analysis is done to calibrate the dosage of radiation, of five radioisotopes (3H, 60Co, 59Ni, 90Sr, 129I) for disposal. Radioactive Iodine delivered maximum risk despite a low concentration in the inventory of radioactive waste. However, much lower than the permissible dosage suggested by ICRP i.e.,1 mSv in a scenario where an exclusion zone is considered to be existing, and a drinking water well is established at a distance of 1.6 km from the repository and the risk is computed over a period of 150 years since the multi-barrier failure.

Возможности изоляции модульных сооружений ППЗРО термопластичными полимерами

High-molecular compounds have been studied, as well as their relationship with other materials and modifications made based on various additives introduced. Their findings helped to initiate the development of technologies providing control over the adsorption process at the phase boundary and polymerization under thermal conditions. The paper discusses a method developed involving thermoplastic polymer of the polyolefin (polyethylene) class applied to concrete and metal surfaces. This material has high adhesive capacity, is resistant to abrasion, corrosion and other destructive factors. Thanks to its improved characteristics this polymer can be considered as an additional waterproofing material in the designs of engineered safety barriers for radioactive waste Class 3 and 4 disposal in near-surface disposal facilities (NSDF).

Neutron activation analysis of novel space shielding for terrestrial applications

Appropriate disposition of nuclear waste is a high priority in the production of nuclear energy and its minimization is of great interest to the public. Employing the use of metal oxide infused conformal coatings as a shielding material in a terrestrial reactor should ideally produce nuclear wastes no higher than class B while also reducing the neutron activations (and thus waste classification) of any materials behind the shield. This allows for the increased likelihood of the ability to use on-site sequestration or minimal near-surface disposal, in lieu of geological repositories, and thus the potential for considerable cost savings.

Impact of Concrete Degradation on the Long-Term Safety of a Near-Surface Radioactive Waste Disposal Facility in Korea

The migration of radionuclides from radioactive waste into the environment poses a public safety concern. Thus, the long-term safety assessment for near-surface disposal sites for radioactive waste in South Korea entails providing reasonable assurance that the annual radiation dose exposure from radionuclide release from the waste repository into the biosphere will not exceed the regulatory limit of 0.1 mSv/yr. At the first near-surface disposal site in Gyeongju, concrete was a crucial component of the engineered barriers designed to contain radionuclides within the disposal site. The ability of concrete to retain radioactive waste within the disposal site is attributed to its high sorption capacity for radionuclides. However, research has shown that the degradation of concrete can affect its radionuclide retention capabilities, which are defined by sorption properties of distribution (Kd) and diffusion (Ds) coefficient parameters. As a result, changes in sorption properties may lead to radionuclides migrating out of the disposal vault. In light of the geochemical deterioration of engineered concrete barriers, this study assesses the long-term safety of near-surface disposal sites. To simulate the impact of concrete degradation on radionuclide migration, we employed RESRAD-OFFSITE’s extended source-term features, which can model the release of radionuclides from radioactive waste shielded by concrete barriers. Using carefully screened published sorption data of four radionuclides (14C, 137Cs, 90Sr and 99Tc) in different stages of concrete degradation, the results indicated that released radioactivity during the most degraded state of concrete will result in a maximum radiation exposure dose of 1.4 × 10−8 mSv/yr from 99Tc which is below the permissible limit of 0.1 mSv per year, thus demonstrating that concrete is a reliable component of the engineered designed barriers for near-surface disposal facilities.

Groundwater Flow Modeling of a Near-Surface Disposal Repository for Low- and Intermediate-Level Radioactive Waste in Southwest China

The influence and function of engineering facilities were increasingly concerned about the safety analysis of low- and intermediate-level radioactive waste (LILW). In terms of near-surface disposal, many artificial facilities were set down, such as drainage facilities, covering layers, and disposal units. To analyze the long-term impact of these artificial facilities on groundwater in the disposal site area, we built four time-continuing models by setting the time nodes of parameter or boundary changes using FEFLOW code, considering the possible aging and degradation state of these facilities. According to the models, the site area’s groundwater level situations for long-term safety assessment were predicted. The results showed the different regulating abilities of drainage facilities affected the groundwater level of the disposal site in different degradation states, which also reflected the necessity of artificial facilities simulation in groundwater modeling. In addition, the Monte Carlo method and surrogate model were adopted to analyze the influence of the uncertainty of model parameters on the output of groundwater flow models. This study could help in further understanding the groundwater flow modeling for long-term safety assessment of near-surface disposal engineering.

Physical characterization of fault rocks within the Opalinus Clay formation

Near-surface disposal of radioactive waste in shales is a promising option to safeguard the population and environment. However, natural faults intersecting these geological formations can potentially affect the long-term isolation of the repositories. This paper characterizes the physical properties and mineralogy of the internal fault core structure intersecting the Opalinus Clay formation, a host rock under investigation for nuclear waste storage at the Mont Terri Laboratory (Switzerland). We have performed porosity, density, microstructural and mineralogical measurements in different sections of the fault, including intact clays, scaly clays and fault gouge. Mercury intrusion porosimetry analysis reveal a gouge that has a pore network dominated by nanopores of less than 10 nm, yet a high-porosity (21%) and low grain density (2.62 g/cm3) when compared to the intact rock (14.2%, and 2.69 g/cm3). Thus, a more permeable internal fault core structure with respect to the surrounding rock is deduced. Further, we describe the OPA fault gouge as a discrete fault structure having the potential to act as a preferential, yet narrow, and localized channel for fluid-flow if compared to the surrounding rock. Since the fault gouge is limited to a millimetres-thick structure, we expect the barrier property of the geological formation is almost not affected.

The products transformation and leaching behavior in solidified cement matrices under the coupling actions of “water-heat-chemistry”

Solidified cement matrices for immobilization of intermedium- and low-level radioactive waste stream will be subjected to the coupling actions of water, heat and chemistry at the near-surface disposal. This study mainly researches the products transformation and leaching behavior of Ca2+ and Sr2+ in solidified cement matrices under the coupling actions. The products are characterized through 29Si Nuclear Magnetic Resonance and X-ray Photoelectron Spectroscopy while the ion concentrations are examined by Inductively Coupled Plasma. The results show that ettringite and AFm change from crystal to amorphous phase while the coordination number of Al decreases from 6 to 4 under the coupling actions of high temperatures and NH4Cl concentrations. Meanwhile, the decalcification and polymerization of solidified cement matrices occurred in deionized water with high temperature (60 °C and 80 °C). However, the maximum degree of decalcification and polymerization was reached at 60 °C instead of 80 °C because of the precipitation of large NH4Cl crystals in the pores on the surface of solidified cement matrices. Furthermore, the diffusion of Sr2+ in solidified cement matrices was accelerated by the decalcification, and the apparent diffusion coefficients of Ca2+ and Sr2+ under coupling actions showed linear relationship. The corresponding results obtained in this study can provide the basic data to the further building of reaction-transport model in the immobilization of intermedium- and low-level radioactive waste stream by cement-based materials.

Experience in radiological risk assessment of a surface waste disposal facility in Chişinău, Moldova

The long-term condition and potential radiological consequences of legacy radioactive waste stored in a RADON-type of near-surface disposal facility outside the city of Chişinău is of concern to the central government and health protection authorities of the Republic of Moldova. A ‘zero alternative scenario’ risk assessment has been undertaken in order to evaluate the potential radiological impact on humans and the environment of the facility, were it to be left in its current state with no remediation. The results have been used as a basis for regulatory decision making regarding remediation and decommissioning of the legacy radioactive waste facility. The aim of this study was two-fold: first to demonstrate a complete radiological risk assessment of a real site using a combination of methodologies developed by the IAEA (ISAM and BIOMASS), the second to illustrate the current state-of-the-art in respect of extracting site-specific information from site-descriptive material. We illustrate the practicality of employing geographic information systems techniques on site-specific topographic data to identify relevant biosphere dose objects, thereby allowing customisation of the generic ISAM model framework to site-specific conditions. As a result, a simple method is suggested to bound activity concentrations in well water based on an understanding of water balance in the local catchment area in which the biosphere dose object is embedded. With conservative assumptions, estimated doses from the calculation cases of the design scenario remain lower than the IAEA’s dose criteria and environmental screening values. However, the results also indicate that human intrusion activities after the institutional control period could lead to radiological exposures above the IAEA’s criteria for a period up to 100 000 years. The long-lived radionuclide 239Pu dominates doses for the on-site residence scenario. Remediation measures should be implemented were the waste to remain at its present place of disposal.

Определение прогнозных сроков долговременной безопасности пунктов приповерхностного захоронения радиоактивных отходов при различных сценариях их эксплуатации

The paper considers technical solutions providing the long-term safety of near-surface disposal facilities for radioactive waste (NSDF) based on a engineered barrier system s (EBS). The system of isolating engineered barriers provides the long-term safety and reduces the spread of ionizing radiation and radioactive material from NSDF into the environment.

Исследование осушки и оптимизация кондиционирования отработавших ионообменных смол без включения в матрицу для захоронения

The publication [1] discusses technical issues faced during the conditioning of intermediate- and low-level spent ion exchange resins by thermal vacuum drying to obtain packages that would meet waste acceptability criteria for nearsurface disposal facilities. This article is a continuation of another publication [1] which elaborated on a study of ionexchange resin drying process and provided some proposals on the conditioning optimization using NZK-150-1.5 P containers. It specifies optimal parameters for the pre-disposal treatment and conditioning of dried resins providing their safe storage and disposal. The study presents calculated permissible internal pressure providing adequate strength of the containers. The proposed pre-disposal ion exchange resin treatment and conditioning method can be recommended for industrial use.

CLIMATE CHANGES AND SITE DISPOSAL OF OBJECTS ON THE CHRONOBYL STATION SITE

Currently, in the Exclusion Zone at the site of the Chornobyl NPP, work is being carried out on the final closure and conservation of power units 1–3, the feature of which is the presence of a high-capacity channel uranium-graphite reactor (RVPK). During operation, the reactor graphite becomes radioactive due to the accumulation of the long-lived radionuclide 14C in the graphite. An additional contribution is made by the indicated radioactivity of technological impurities (36Cl, 60Co) and spills of fission products and fragments of nuclear fuel. The main problem of the decommissioning of nuclear installations from the RVPK is related to the need to choose optimal methods of handling large volumes of spent graphite. The possibility of applying the technology of near-surface disposal of decommissioned uranium-graphite reactors of the RVPK, known as the “green mound”, to power units 1–3 of the Chernobyl NPP was considered. Natural factors have been identified, the possible action of which during the time required for the decay of radionuclides 14C and 36Cl to an acceptable level, may lead to the destruction of near-surface storage facilities for disposal at the Chernobyl NPP site. These factors determine the rehabilitation of the river valley of the Pripyat River and the vulnerability of surface disposal repositories to the effects of cyclical climate changes. Climatic changes should be taken into account when making decisions on on-site disposal, given that the processes of decay of long-lived radionuclides, degradation of matrices of incorporated radioactive waste, and the duration of protective barriers in terms of time are relevant to the dynamics of climate change. The need to take into account the impact of global climate changes on the intensity of changes in the geomorphological characteristics of the locations of “green mounds” was noted. A particularly dangerous and powerful natural factor that will inevitably lead to the destruction of the “green mounds” is the influence of probable glaciation. The relevance of solving the scientific task of a complex combination of knowledge regarding the formation of the physical and geographical features of Polissia in the last 200 thousand years with the modeling of climate changes in the future 100 thousand years is emphasized, which will allow to reasonably accept or reject the very idea of near-surface burial of long-lived radioactive waste at the Chernobyl NPP site.

Study on Rainfall Infiltration Into Vault of Near-surface Disposal Facility Based on Various Disposal Scenarios

Study on Rainfall Infiltration Into Vault of Near-surface Disposal Facility Based on Various Disposal Scenarios

ЩОДО ЗАСТОСУВАННЯ БЕНТОНІТОВИХ ГЛИН ЧЕРКАСЬКОГО РОДОВИЩА ДЛЯ УБЕЗПЕЧЕННЯ ЗАХОРОНЕННЯ РАВ У ПОВЕРХНЕВИХ / ПРИПОВЕРХНЕВИХ СХОВИЩАХ НИЗЬКОАКТИВНИХ ВІДХОДІВ: ДОСЛІДЖЕННЯ СОРБЦІЙНИХ ВЛАСТИВОСТЕЙ БЕНТОНІТІВ

The article is devoted to the study of the barrier properties of bentonite clays from the Cherkasy deposit (Ukraine) as the most promising material for engineered barriers in near-surface disposal facilities for low-level radioactive waste. The work presents a description of the Cherkasy bentonite deposit, in particular, of the most promising II layer of the Dashukivka site, and the composition and properties of the natural and soda-modified (PBA20) bentonites which can be used for safe long-term storage and disposal of radioactive waste. The scanning electron microscope (SEM) images of the samples were made and the chemical and mineral composition of the samples was determined. Based on the obtained data, the Dashukivka bentonite was classified as Al-Fe-montmorillonit. The main rock-forming minerals of the bentonite are montmorillonite (75 ± 3 wt. %) and quartz (20 - 25 wt. %). Infrared spectra of the bentonites in the range of 4000 - 400 cm-1 are typical for dioctahedral Al-smectites and are almost identical both for the natural and modified samples. The derivatographic analysis (DTA) of the bentonite samples allows us to conclude that the modified bentonite PBA-20 has greater thermal stability than the natural bentonite. The total weight loss of the samples is about 10.3% and 17.8%, respectively. From the experimental study of 137Cs adsorption from a simulated ChEZ water solution (mineralization – 0.2 g/dm3), it was found that the equilibrium in the sorbent-solution system is established within 12 to 14 h, and the degree of adsorption on the modified bentonite PBA-20 reaches 93%. The degree of adsorption on the natural bentonite is 7 - 10% lower. Analysis of the partition coefficients (Kd) also indicates a higher efficiency of the Na-modified bentonite compared to the natural, and that ion exchange is the main mechanism of caesium adsorption. Increasing of the pH of the initial solution results only in a slight increase of the degree of adsorption. The degree of 137Cs adsorption on Na-modified bentonite is practically independent of the initial concentration of Ca2+ (from 16 to 160 mg/dm3) and Na+ (from 6 to 60 mg/dm3) in the solution. However, 137Cs adsorption on natural bentonite decreases with the increase in Na+ concentration. Thus, the bentonite clays from the Cherkasy deposit (Dashukivka site, II layer) having good barrier properties can be recommend as anti-migration engineered barriers for radioactive waste repositories, in particular, near-surface storage facilities I and II line of Vector complex