Low-level Waste Disposal Facility Research Articles

Discerning β-emitting radioactivity from caesium-137 γ radiation via bremsstrahlung for in-situ groundwater monitoring

A qualitative approach to discern the relative contributions of 137Cs and 90Sr in-solution with a monitoring probe is described, based on a comparison of X- and γ-ray photon spectra. The bremsstrahlung yield from 90Sr has been measured as a function of distance (x) from a cerium bromide scintillation detector relative to the 137Cs γ-ray full-energy peak response. The utility of two, count-independent shape factor parameters has been compared: the ratio of the sum of the counts in the 60–800 keV region to the full-energy peak response associated with the 137Cs 662 keV γ-ray line, termed for the benefit of this paper as SF1, and the ratio of counts in the 60–350 keV region to those in the 350–450 keV region, defined in this work as SF2. These parameters probe the presence of isotopes associated with clear full-energy spectral lines and bremsstrahlung, respectively. They have been calibrated as a function of x with sealed sources in the laboratory and both are observed to be asymptotic as x→0. This approach has been tested with a variety of open, laboratory-based liquid samples comprising 90Sr and 137Cs, and in a water-based, contaminated medium in the field associated with a sump at a low-level waste disposal facility on the former fast reactor site at Dounreay, UK. The highest response in both SF1 and SF2 measured in this work is observed in these on-site measurements, consistent with x→0, implying both 90Sr and 137Cs presenting together in solution. This is consistent with prior, laboratory-based analysis of the sump media, and suggests that the β−-emitter contribution (anticipated to comprise 90Sr and 90Y) is very near the probe, consistent with the solution being in contact with it. This is of relevance where the β−/γ composition of contaminated groundwater, in-situ, is not well understood or where migratory changes in activity are suspected.

Cliff Retreat Rates Associated with a Low-Level Radioactive Waste Disposal Facility in Los Alamos, New Mexico, USA

We present an analysis and interpretation of potential cliff stability at a low-level waste disposal facility at Los Alamos National Laboratory, New Mexico, using cliff morphologic and fracture characteristics coupled with carbon-14 surface exposure dating. Our study is important as it directly bears on the licensing criteria for low-level radioactive waste sites. We find that future characteristic cliff failures will likely not breach disposal pits and shafts over the 1000-year minimum regulatory period. Further, we find, using a multivariate regression model, that slope angle and cliff face aspect are sub-equal in importance to predict regions of high risk of failure when combined with surface exposure ages and assuming that old exposure ages are most indicative of stability (instability) and therefore can aid decision making in final design implementation.

Sorption and mobility of radioactive Ni in Wolsong waste repository site, South Korea

Radioactive nickels are significant environmental concerns because of their long half-lives (59Ni = 76,000 years and 63Ni = 100.1 years) and high proportions in nuclear wastes in the repositories. Therefore, the investigation of sorption and mobility of Ni at the disposal facility sites is important for the prediction and evaluation of radiological risks to the public. Herein, the batch sorption and column experiments were performed under various geochemical conditions using site-specific rock and groundwater samples collected at the Wolsong low and intermediate level waste (LILW) disposal facility site in South Korea. The batch sorption results revealed that the sorption of Ni was strongly affected by the pH. Approximately 20% Ni adsorbed at pH 5, which enhanced over 90% at pH > 8. The batch sorption results were used to develop the surface complexation model (SCM) to predict the Ni sorption and its mobility behavior. The non-electrostatic generalized composite (GC) SCM approach simulated the Ni sorption data well. We believe, our study can interpret and extrapolate the sorption and transport behavior of Ni in the underground repository conditions.

Creation of Very Low-Level Waste Disposal Facility at Rostov NPP

The paper considers the construction of a very low-level waste disposal facility (VLLW) at the Rostov NPP with a volume of 11,000 m 3 . Information about formation and management of VLLW at Rostov NPP is given. Characteristics of 11,000 m 3 VLLW disposal facility is presented. The design documentation provides for construction of the VLLW disposal facility intended for disposal of very low-level waste of hazard classes IV and V formed as a result of Rostov NPP production activities and meeting the criteria of disposal acceptability.

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

Neptunium(V) sorption to vadose zone sediments: Reversible, not readily reducible, and predictable based on Fe-oxide content

Abstract Neptunium (Np) is among the key risk-drivers at the Savannah River Site's (SRS) low-level waste disposal facility located in South Carolina. A series of studies were undertaken to identify and to model the key geochemical processes controlling Np sorption to SRS vadose zone sediments. The approach was to conduct Np(V) laboratory batch and flow studies using two sediments representing end-member depositional facies recovered beneath the disposal facility. Baseline distribution coefficients (Kd values; Np concentration ratio of sediment:porewater) were 9.05 ± 0.61 L kg−1 and 4.26 ± 0.24 L kg−1 for the clayey and sandy end-member sediments, respectively. The addition of natural organic matter (NOM) to the sediment resulted in only a two fold increase in the Kd values, most likely due to the formation of ternary sediment-NOM-Np complexes. None of the reduction treatments (ascorbic acid, dithionite, zero valent iron, hydrogen peroxide, and anaerobic atmosphere), including some long-term (71-day) equilibration experiments, resulted in significant increases in Kd values. This indicated that little to no reduction of Np(V) to the more strongly sorbing Np(IV) occurred. Among the key novel findings in this research was that batch desorption tests and stop-flow stir-cell kinetic experiments indicated that the Np(V) sorption was completely reversible. These observations were used to develop a simple conceptual model describing Np(V) sorption. The conceptual model described NpO2+ reversibly complexation to iron oxide coatings on the sediments. The model was successfully applied without any adjustable parameters to an independent set of experimental data, requiring only the dithionite-extractable Fe concentration from the independent dataset. There is no standard approach for quantifying reactive sorption site concentrations on composite materials (e.g., soils and sediments). In this study we proposed such a method based on an operationally defined fraction of the extractable iron. This parameterization approach was calibrated with two sediments then used to blindly and successfully predict sorption on another end-member soil. The successful modeling approach taken in this work 1) identified key reactions that are or are not influencing the system to develop a simple but appropriate conceptual model and 2) calibrated the fraction of extractable iron required for surface site density determination and used the calibrated model for blind predictions. This modeling approach could be used for other composite materials to allow for comparisons of the fraction of surface reactive extractable metals.

Risk and Reliability Analysis of Multibarrier System for Near-Surface Disposal Facilities

AbstractThe performance of near-surface disposal facilities (NSDFs) for low-level radioactive waste depends on the type of materials used in the construction of barrier system, the properties of geological formations around the facility, and the type of waste containment system. By taking into account the effect of the factors mentioned here, a mathematical model can be developed to evaluate the migration of contaminants and radionuclides (via the contaminant transport model) from NSDFs to the nearest geosphere and biosphere, and assess the performance of the whole system. The model for migration of contaminants in a multibarrier system is considered in this paper for the safety assessment of NSDFs. The model considers scenarios during dumping period, after the dumping period, and after the closure of the NSDF. The annual radiation dose due to a radionuclide through drinking water pathway (for the single dump mode and multiple dump mode) is evaluated with the model. In the present work, a comprehensive an...

ICONE19-43161 CURRENT STATUS OF THE DEMONSTRATION TEST OF UNDERGROUND CAVERN-TYPE DISPOSAL FACILITIES

In Japan, the underground cavern-type disposal facilities for low-level waste (LLW) with relatively high radioactivity, mainly generated from power reactor decommissioning, and for certain transuranic (TRU) waste, mainly from spent fuel reprocessing, are designed to be constructed in a cavern 50-100 m underground and to employ an engineered barrier system (EBS) made of bentonite and cement materials. To advance a disposal feasibility study, the Japanese government commissioned the Demonstration Test of Underground Cavern-Type Disposal Facilities in fiscal year (FY) 2005. Construction of a full-scale mock-up test facility in an actual subsurface environment started in FY 2007. The main test objective is to establish the construction methodology and procedures that ensure the required quality of the EBS on-site. A portion of the facility was constructed by 2010, and the test has demonstrated both the practicability of the construction and the achievement of quality standards: low permeability of less than 5×10^<-13> m/s and low-diffusion of less than 1×10^<-12> m^2/s at the completion of construction. This paper covers the test results from the construction of certain parts using bentonite and cement materials.

Ageing management program for the Spanish low and intermediate level waste disposal and spent fuel and high-level waste centralised storage facilities

The generic design of the centralised spent fuel storage facility was approved by the Spanish Safety Authority in 2006. The planned operational life is 60 years, while the design service life is 100 years. Durability studies and surveillance of the behaviour have been considered from the initial design steps, taking into account the accessibility limitations and temperatures involved. The paper presents an overview of the ageing management program set in support of the Performance Assessment and Safety Review of El Cabril low and intermediate level waste (LILW) disposal facility. Based on the experience gained for LILW, ENRESA has developed a preliminary definition of the Ageing Management Plan for the Centralised Interim Storage Facility of spent Fuel and High Level Waste (HLW), which addresses the behaviour of spent fuel, its retrievability, the confinement system and the reinforced concrete structure. It includes tests plans and surveillance design considerations, based on the El Cabril LILW disposal facility.

Predicting hydraulic conductivity of sand–bentonite mixture backfill before and after swelling deformation for underground disposal of radioactive wastes

Mixtures of sand and bentonite will be used as bentonite-based backfill materials for high-level radioactive waste repositories. They are also used in Japan as artificial barriers in disposal facilities for low-level radioactive waste. Very low hydraulic conductivity of the sand–bentonite mixture is expected to keep wastes separated from the surrounding environment. Some researchers have investigated hydraulic conductivities of sand–bentonite materials at different dry densities and at different mass ratios of sand and bentonite to design and develop specifications of the materials described above. However, the function of low permeability of those materials is degraded by the altered hydraulic conductivity of sand–bentonite mixtures after swelling deformation. This study investigates experimentally hydraulic conductivity changes before and after swelling for sand–bentonite materials. Furthermore, results confirm the applicability of theoretical equations for evaluating the hydraulic conductivity of compacted bentonites and sand–bentonite mixtures, which have already been proposed in Komine (2008), by comparing calculated results and experimental results.

Long term water flow scenario in low-level waste disposal vaults, with particular regard to concrete structures in El Cabril, Cordoba, Spain

This paper deals with the main durability design objectives adopted for the Spanish Low-level waste disposal facility of El Cabril. The presentation summarizes the studies and models developed to represent the performance of the reinforced concrete vaults. Particular attention is paid to recent developments in modelling the water flow through the disposal system and its humidity saturation and their relation to the long term behaviour of the concrete barriers. It also describes the work being carried out to improve the existing models as a part of the required effort to maintain up to date the performance assessment of the facility.

Some principles of service life calculation of reinforcements and in situ corrosion monitoring by sensors in the radioactive waste containers of El Cabril disposal (Spain)

Reinforced concrete is the most usual material used in engineered barriers in low-level nuclear waste disposal facilities. The record of modern concrete is no longer than about 100 years. During this time, it has been noticed that the material gives a good performance in many environments, however several chemical aggressive species in water, soil or the atmosphere may react with the cement mineralogical phases and perturb its integrity. El Cabril repository has a design life objective of longer than 300 years and therefore, these structures should maintain their main characteristics during this target service life. The potential aggressive conditions that the cement-based materials can suffer have been identified to be: carbonation, water permeation (leaching) and reinforcement corrosion. More unlikely may be the biological attack. Chlorides are not in the environment but they are inside the drums as part of analytical wastes. Vaults and containers are made of a very similar concrete composition while the mortar is specifically designed to be pumpable, with low hydration heat, low shrinkage and of low permeability. In this paper results of concrete characteristics are given as well as the monitoring of the behaviour of reinforcement corrosion parameters from 1995 on the same environmental conditions of the actual waste. This monitoring has been made in a buried structure with embedded sensors. The effect of temperature is commented.

Performance Assessment of Low-Level Waste Disposal Facilities Using Coupled Unsaturated Flow and Reactive Transport Simulators

ABSTRACTRecent advances in the development of reactive chemical transport simulators have made it possible to use these tools in performance assessments (PAs) for nuclear waste disposal. Reactive transport codes were used to evaluate the impacts of design modifications on the performance of two shallow subsurface disposal systems for low-level radioactive waste. The first disposal system, located at the Hanford site in Richland, Washington, is for disposal of lowlevel waste glass. Glass waste blocks will be disposed in subsurface trenches, surrounded by backfill material. Using different waste package sizes and layering had a small impact on technetium release rates to the vadose zone. The second disposal system involves a hypothetical repository for low-level waste in Italy. A model of uranium release from a grout waste form was developed using the STORM reactive transport code. Uranium is predicted to be relatively insoluble for several hundred years under the high-pH environment of the cement pore water. The effect of using different filler materials between the waste packages on uranium flux to the vadose zone proved to have a negligible impact on release rates.

Evaluation of Hydrogeo-chemical Conditions for Development of Nuclear Criticality in Low-Level Waste Disposal Facilities

The under ground disposal of fissile isotope-bearing wastes poses some unique issues. Specifically, radionuclides such as 235U disposed in low-level waste facilities, if present in the correct concentration and geometry, can create a nuclear criticality event that releases radioactivity to ground water. This paper reports the results of a study of the potential for 235U to be transported by ground water within low-level waste facilities and to concentrate to form a critical mass. Coupled hydrologic and geochemical modeling was used to investigate two possible mechanisms for concentrating mobile uranium: sorption on high capacity sites and precipitation in a reducing zone. The hydrogeochemical modeling showed that (1) it is difficult to mobilize uranium from sorption sites then re-deposit it; (2) if uranium is already in solution it can accumulate in zones of high sorption, and (3) reducing zones can accumulate sufficient uranium in the presence of oxygenated ground water. Site-specific disposal practices, such as the commingling of large quantities of depleted (nonfissile) uranium and the difficulty of bringing sufficient mass together in the correct geometry, limit the potential for criticality safety concerns. In order to determine appropriate disposal practices, hydrogeochemical modeling can be used to study the future mobility and accumulation of the waste.

Depleted Uranium Disposal Options

Depleted uranium hexafluoride (UF6) has been produced in the United States since the 1940s as part of both the military program and the civilian nuclear energy program. The U.S. Department of Energy (DOE) is the agency responsible for managing most of the depleted UF6 that has been produced in the United States. The total quantity of depleted UF6 that DOE has to or will have to manage is approximately 700,000 Mg. Studies have been conducted to evaluate the various alternatives for managing this material. This paper evaluates and summarizes the alternative of disposal as low-level waste (LLW). Results of the analysis indicate that UF6 needs to be converted to a more stable form, such as U3O8, before disposal as LLW. Estimates of the environmental impacts of disposal in a dry environment are within the currently applicable standards and regulations. Of the currently operating LLW disposal facilities, available information indicates that either of two DOE facilities—the Hanford Site or the Nevada Test Site—or a commercial facility—Envirocare of Utah—would be able to dispose of up to the entire DOE inventory of depleted UF6.

Probabilistic safety assessment model for near surface radioactive waste disposal facilities

A probabilistic safety assessment model has been developed for assessing the performance of near surface disposal facilities for low-level radioactive waste. Two modes of disposal such as single dump and multiple dump are considered in the model. The model is composed of four components: source term, repository failure, geosphere transport and radiological assessment. The source term contains low-level radioactive waste equivalent to 50 GWe.y energy production (1 GWe.y=3.2×10 16 J of electrical energy production) either disposed at an instant or distributed over 50 years. The endpoints of assessment are expressed as radioactivity release rate, radionuclide concentration in ground water, radiation dose to a member of the critical group through drinking water pathway and total risk to critical group due to disposal practice. Sensitivity analysis is carried out to identify the critical parameters, which have maximum effect on the assessment endpoints. Uncertainty analysis, based on random selection of all main parameters, is also carried out to identify the effect of overall variation of parameters used in the model in relation to a reference level. The reference level delivers a maximum annual effective dose of 1.2×10 −2 mSv to a member of the critical group after about 1.25×10 2 years of disposal mostly from 129I, the long-lived and less sorbing radionuclide. The corresponding risk to the public due to the disposal practice (9.0×10 −7 y −1) is found to be lower than that due to natural background radiation. The most critical parameters as indicated by the sensitivity analysis are the distribution coefficient of radionuclides, seepage velocity in the unsaturated zone between the facility and the water table, dispersivity in ground water and thickness of the unsaturated zone. The uncertainty analysis shows that 129I is the critical radionuclide delivering maximum dose in most cases though it constitutes a low percentage in the low-level radioactive waste inventory. The annual effective doses derived from the uncertainty analysis fall in a log normal distribution with a geometric mean of 3.6×10 −2±3.9 mSv and the most probable annual effective dose to a member of the critical group works out to be 2.8×10 −4 mSv.

Management and disposal of waste from sites contaminated by radioactivity

Various methods of managing and disposing of wastes generated by decontamination and decommissioning (D & D) activities are described. This review of current waste management practices includes a description of waste minimization and volume reduction techniques and their applicability to various categories of radwaste. The importance of the physical properties of the radiation and radioactivity in determining the methodology of choice throughout the D & D process is stressed. The subject is introduced by a survey of the common types of radioactive contamination that must be managed and the more important hazards associated with each type. Comparisons are made among high level, transuranic, low level, and radioactive mixed waste, and technologically-enhanced, naturally-occurring radioactive material (TENORM). The development of appropriate clean-up criteria for each category of contaminated waste is described with the aid of examples drawn from actual practice. This includes a discussion of the application of pathway analysis to the derivation of residual radioactive material guidelines. The choice between interim storage and permanent disposal of radioactive wastes is addressed. Approaches to permanent disposal of each category of radioactive waste are described and illustrated with examples of facilities that have been constructed or are planned for implementation in the near future. Actual experience at older, existing, low-level waste disposal facilities is discussed briefly.

Prediction of Solid-Aqueous Equilibria in Cementitious Systems Using Gibbs Energy Minimization: I. Multiphase Aqueous - Ideal Solid Solution Models

ABSTRACTConcrete and other cement-based materials are increasingly utilized as major structural components of the disposal facilities for low-level and intermediate-level radioactive waste (LLW/ILW). At the same time. cementitious materials function as engineered barriers against migration of radionuclides and other hazardous compounds [1]. Taking into account the expected operation times of such constructions, the long-term prediction of environmental interactions and stability of concretes is important for the development of reliable facilities for LLW/ILW disposal. Thermodynamic approach has been widely used to promote understanding of chemical phenomena in cementitious systems. Recently, several authors attempted to demonstrate usefulness of computer codes for description of the available experimental data and prediction of cement/water equilibria [1-5]. These calculations were based on the Law-of-Mass Action - Reaction Stoichiometry Algorithm (LMA-RSA), widely applied so far in speciation modeling [6]

Investigation of14C Release in an Engineered Low-Level Waste Disposal Facility

Atmospheric releases of I4C from a generic engineered low-level waste (LLW) disposal facility and its radiological impacts are investigated. A computer model that describes microbial gas generation...

95/05420 Total caesium-fixing potentials of acid organic soils

Various methods of managing and disposing of wastes generated by decontamination and decommissioning (D & D) activities are described. This review of current waste management practices includes a description of waste minimization and volume reduction techniques and their applicability to various categories of radwaste. The importance of the physical properties of the radiation and radioactivity in determining the methodology of choice throughout the D & D process is stressed.The subject is introduced by a survey of the common types of radioactive contamination that must be managed and the more important hazards associated with each type. Comparisons are made among high level, transuranic, low level, and radioactive mixed waste, and technologically-enhanced, naturally-occurring radioactive material (TENORM). The development of appropriate clean-up criteria for each category of contaminated waste is described with the aid of examples drawn from actual practice. This includes a discussion of the application of pathway analysis to the derivation of residual radioactive material guidelines.The choice between interim storage and permanent disposal of radioactive wastes is addressed. Approaches to permanent disposal of each category of radioactive waste are described and illustrated with examples of facilities that have been constructed or are planned for implementation in the near future. Actual experience at older, existing, low-level waste disposal facilities is discussed briefly.