Radionuclide Solid:liquid partitioning in an aged, reducing-grout wasteform recovered from a disposal facility.

The Saltstone Disposal Facility on the Savannah River Site in South Carolina disposes of Low-Level Waste in a reducing-grout waste form. Reducing grout is presently being evaluated as a subsurface disposal waste form at several other locations in the United States, as well as in Europe and Asia. The objective of this study was to collect core samples directly from the Saltstone Disposal Facility and measure desorption distribution coefficients (Kd; radionuclide concentration ratio of saltstone:liquid; (Bq/kg)/Bq/L)) and desorption apparent solubility values (ksp; radionuclide aqueous concentration (moles/L)). An important attribute of this study was that these tests were conducted with actual aged, grout waste form materials, not small-volume simulants prepared in a laboratory. The reducing grout is comprised of blast furnace slag, Class F fly ash, ordinary portland cement, and a radioactive salt waste solution generated during nuclear processing. The grout sample used in this study underwent hydrolyzation in the disposal facility for 30 months prior to measuring radionuclide leaching. Leaching experiments were conducted either in an inert (no oxygen) atmosphere to simulate conditions within the saltstone monolith prior to aging (becoming oxidized) or they were exposed to atmosphere conditions to simulate conditions of an aged saltstone. Importantly, these experiments were designed not to be diffusion limited, that is, the saltstone was ground finely and the suspensions were under constant agitation during the equilibration period. Under oxidized conditions, measured Tc Kd values were 10 mL/g, which was appreciably greater than the historical best-estimate value of 0.8 mL/g. This difference is likely the result of a fraction of the Tc remaining in the less soluble Tc(IV) form, even after extensive oxidation during the experiment. Under oxidized and reducing conditions, the measured Ba and Sr (both divalent alkaline earth metals) Kd value were more than an order of magnitude greater than historical best-estimate values of 100 mL/g. The unexpectedly high Ba and Sr Kd values were attributed to these radionuclides having sufficient time to age (form strong bonds) in the sulfur-rich saltstone sample. Apparent ksp values under reducing conditions were 10-9 mol/L Tc and 10-13 mol/L Pu, consistent with values measured with surrogate materials. Measured apparent Ba, Sr, and Th ksp values were significantly greater than historical best-estimates. The implications of the generally greater Kd values and lower ksp values in these measurements is that these cementitious waste forms have greater radionuclide retention than was previously estimated based on laboratory studies using surrogate materials. This work represents the first leaching study performed with an actual aged, reducing-grout sample and as such provides an important comparison to studies conducted with surrogate materials, and provides high pedigree data for other programs around the world evaluating reducing grouts as a wasteform for subsurface nuclear waste disposal.

Dramatic loss of microbial viability in bentonite exposed to heat and gamma radiation: implications for deep geological repository.

Bentonite is an integral part of the engineered barrier system (EBS) in deep geological repositories (DGR) for nuclear waste, but its indigenous microorganisms may jeopardize long-term EBS integrity. To predict microbial activity in DGRs, it is essential to understand microbial reactions to the early hot phase of DGR evolution. Two bentonites (BCV and MX-80) with varied bentonite/water ratios and saturation levels (compacted to 1600 kg.m- 3 dry density/powder/suspension), were subjected to heat (90-150 °C) and irradiation (0.4 Gy.h- 1) in the long-term experiments (up to 18 months). Molecular-genetic, microscopic, and cultivation-based techniques assessed microbial survivability. Exposure to 90 °C and 150 °C notably diminished microbial viability, irrespective of bentonite form, with negligible impacts from irradiation or sample type compared to temperature. Bentonite powder samples exhibited microbial recovery after 90 °C heating for up to 6 months but not 12 months in most cases; exposure to 150 °C had an even stronger effect. Further long-term experiments at additional temperatures combined with the mathematical prediction of temperature evolution in DGR are recommended to validate the possible evolution and spatial distribution of microbially depleted zones in bentonite buffer around the waste canisters and refine predictions of microbial effects over time in the DGR.

Decommissioning of a Medical Cyclotron Vault: The Case Study of the National Cancer Institute of Milano.

ABSTRACT: In the widespread use of medical cyclotrons for isotope production, radiological and economic consequences related to the decommissioning of particle accelerators are often neglected. However, decommissioning regulation and its related procedures always demand efforts and costs that can unexpectedly impact on budgets. The magnitude of this impact depends strongly on the residual radioactivity of the accelerator and of the vault, and more specifically on the kind and activity concentration of residual radionuclides. This work reports and discusses a case study that analyzes in detail the characterization activities needed for optimized management of the decommissioning of a medical cyclotron vault. In particular, this paper presents the activities carried out for assessing the activity concentrations and for guiding the disposal of the cyclotron vault of the Italian National Cancer Institute of Milano (INT). An unshielded 17 MeV cyclotron vault was characterized by high resolution gamma-ray spectrometry both in-situ and in-laboratory on extracted samples. Monte Carlo simulations were also carried out to assess the overall distribution of activation in the vault. After a few months from the final shutdown of the accelerator, activity concentrations in the concrete walls due to neutron activation exceeded the clearance levels in many regions, especially close to the cyclotron target. Due to the relatively long half-lives of some radionuclides, a time interval of about 20 y after the end of bombardment is necessary for achieving clearance in some critical positions. Far from the target or in positions shielded by the cyclotron, activation levels were below the clearance level. The comparison between Monte Carlo simulations and experimental results shows a good agreement. The in-situ measurements, simpler and economically advantageous, cannot completely replace the destructive measurements, but they may limit the number of required samples and consequently the decommissioning costs. The methodology described and the results obtained demonstrated that it is possible to obtain accurate estimations of activity concentrations with cheap and quick in-situ measurements if the concentration profile in-depth inside the wall is well known. This profile can be obtained either experimentally or numerically through suitably validated Monte Carlo simulations.

Progress with the regulation of radiation safety during recovery and removal of spent nuclear fuel from the site for temporary storage at Andreeva Bay on the Kola Peninsula.

In the 1960s, a shore technical base (STB) was established at Andreeva Bay on the Kola Peninsula, in northwest Russia. The STB maintained nuclear submarines and the nuclear icebreaker fleet, receiving and storing fresh and spent nuclear fuel (SNF) as well as solid and liquid radioactive waste (RW). It was subsequently re-designated as a site for temporary storage (STS) for SNF and RW. Over time, the SNF storage facilities partly lost their containment functions, leading to radioactive contamination of workshops and the site above permitted values. The technological and engineering infrastructure at the site was also significantly degraded as well as the condition of the stored SNF. At present, the STS Andreeva Bay facility is under decommissioning. This paper describes progress with the creation of safe working measures for workers involved in site remediation and SNF recovery operations, including the determination of safe shift times in high radiation areas, as part of overall optimization of safety. Results are presented for the successful application of these measures in the period 2019-2021, during which time significant SNF recovery and removal operations were completed without incident. Significant important experience has been gained to support safe removal of remaining SNF, including the most hazardous degraded fuel, as well as recovery of other higher level RW and decommissioning of the old storage buildings and structures.

Health risks from radioactive particles on Cumbrian beaches near the Sellafield nuclear site.

A monitoring programme, in place since 2006, continues to recover radioactive particles (<2 mm diameter) and larger objects from the beaches of West Cumbria. The potential risks to members of the public using the beaches are mainly related to prolonged skin contact with or the inadvertent ingestion of small particles. Most particles are classified as either 'beta-rich' or 'alpha-rich' and are detected as a result of their caesium-137 or americium-241 content. Beta-rich particles generally also contain strontium-90, with90Sr:137Cs ratios of up to about 1:1, but typically <0.1:1. Alpha-rich particles contain plutonium isotopes, with Pu:241Amαratios usually around 0.5-0.6:1. 'Beta-rich' particles have the greatest potential to cause localised skin damage if held in stationary contact with the skin for prolonged periods. However, it is concluded that only particles of >106Bq of137Cs, with high90Sr:137Cs ratios, would pose a significant risk of causing acute skin ulceration. No particles of this level of activity have been found. Inadvertent ingestion of a particle will result in the absorption to blood of a small proportion of the radionuclide content of the particle. The subsequent retention of radionuclides in body organs and tissues presents a potential risk of the development of cancer. For 'beta-rich' particles with typical activities (mean 2 × 104Bq137Cs, Sr:Cs ratio of 0.1:1), the estimated committed effective doses are about 30µSv for adults and about 40µSv for 1 year old infants, with lower values for 'alpha-rich' particles of typical activities. The corresponding estimates of lifetime cancer incidence following ingestion for both particle types are of the order of 10-6for adults and up to 10-5for infants. These estimates are subject to substantial uncertainties but provide an indication of the low risks to members of the public.

Radiological assessment of the disposal of bulk oil NORM waste: Case study from Brazil.

The oil and gas industries are the largest producers of NORM wastes that are continuously generated during production and extraction activities. In addition, an increasing trend is observed in waste production worldwide due to the decommissioning of oil platforms. The problem is that most of these wastes are in activity concentration levels above the exemption and the clearance limits and are being accumulated in storage sites because no repositories exist in Brazil for NORM wastes generated by oil industries. There are regulations for radioactive wastes and for the licensing of repositories for managing wastes with low and intermediate levels of radiation but the current regulations apply only to packaged wastes. Therefore an initial radiological assessment was carried out with the RESRAD-OFFSITE code to show that bulk NORM wastes (not packaged wastes) could be disposed of in repositories near the surface without causing additional risk to the public above the criteria used. The results can also support decision-making by the Regulatory Authority to change the current regulations and allow for the disposal of wastes in bulk form.

SAFETY IMPROVEMENT OF THE RADIOACTIVE WASTE STORAGE FACILITY IN ALBANIA.

The Albanian National Radioactive Waste Storage Facility (NRWSF) constructed in 1999 consists of an interim waste storage facility for very low level waste, low level waste, intermediate level waste and disused sealed radioactive sources (DSRS) coming from research institutions, agriculture and industrial activities and from applications in nuclear medicine in Albania. The safety assessment of this facility is performed considering its impact to workers, public and environment. In 2018, Albania started a new National project Alb 9010 titled 'Upgrading the Radioactive Waste Storage Building According to International Standards'. The two fundamental objectives are to ensure the protection of the public, workers and the environment and to ensure retrievability of waste packages for final disposal. The project focuses on ensuring the safety of the treatment, conditioning and storage of radioactive waste and DSRS. Under this project has been done the improvement of the safety of the RWSF and upgrading of the existing building infrastructure. In the end of the project the objective is to improve the safety of the personnel in the RWSF, protection of public and environment, Upgrade the Equipment's used in the RW Management Activities, Upgrade the existing Infrastructure of the RWSF Building and also ensure retrievability of waste packages for final disposal.

Practical learning through radiation physics problem solving.

INTRODUCTION: The identification of unknown radionuclides is an authentic practical activity for students that provides the foundations for clinical problem solving, especially in the storage and management of radioactive waste. As different radionuclides have different half-lives, some of which are quite long, the storage of waste material has to accommodate the longest of these. Cross contamination requires a method of identifying the radionuclide samples in a mixed sample to safely and appropriately manage disposal. Similarly, identifying a single unknown sample of a radionuclide allows correct handling and disposal. METHODS: Performing a systematic investigation of the physical properties of unknown radioactive samples is a rich learning opportunity to instil understanding of important physics principles among students in nuclear medicine. RESULTS: This manuscript outlines an investigation developed that would allow students to identify single unknown radionuclides based on physical properties and identify the constituent radionuclides of a mixed sample using some additional mathematical curve stripping. CONCLUSION: The processes and solutions are provided with real data and this practical activity can be replicated by students generating their own data. IMPLICATIONS FOR PRACTICE: This paper provides a template and analysis/interpretation guideline for educators and clinicians to deepen understanding of foundation physics. Enhanced and deeper understanding are a vehicle for improved problem solving in clinical and research practice.

Discharges of Nuclear Medicine Radioisotopes: The Impact of an Abatement System.

ABSTRACT: Clinical uses of radiopharmaceuticals imply the administration of radioactive substances that are mainly excreted through urine. The Nuclear Medicine Department at the Instituto Nacional de Cancerología (INC-COL) in Bogota, Colombia, administers radiopharmaceuticals for diagnostics and treatment to many patients, resulting in tens of cubic meters of radioactive waste water (WW) every day. As Colombian regulatory limits for liquid radioactive discharges to the sewer system are lower than in other countries, longer WW decay times are required, even when an in-house waste water treatment plant (WWTP) is used. To fulfill the requirements for controlled disposal of radioactive discharges, a complementary abatement system was implemented to retain WW for periods as long as 360 d, and was connected to the hospital´s WWTP. These holding times can cause major changes in the WW physicochemical parameters, reaching levels higher than acceptable. In this study, we evaluate the decontamination and decay efficiency of the retention system using water quality parameters and the amount of radioactivity in the effluents stored in the tanks and the WWTP. According to the results, to maintain the physicochemical parameters below acceptable levels, biological and chemical treatment of decayed WW is necessary before discharging it into urban waste water. Using the principles of dilution, retention, and decay, an integral radioactive WW management system was implemented favoring the quality of discharges and activity levels to the sewer system, with efficiencies close to 100% for WW from discharges in diagnostic procedures ranging from 98% (131I) to 100% (177Lu) for WW from discharges in therapeutic procedures. Activity concentration assessment in medically-derived radionuclides using an in-house waste water treatment plant (WWTP) and a complementary abatement system; an in-house WWTP could be used as an abatement system for short-lived radionuclides; and a tank-based abatement system attached to the in-house WWTP showed higher efficiencies for long-lived radionuclides and adequate physicochemical parameters for the discharge to the city sewage system.

Environmental radiation and health risk assessment in the neighborhood of a radioactive waste management facility.

The Radioactive Waste Management Center (RWMC) of the Ghana Atomic Energy Commission (GAEC) operates a licensed radioactive waste management facility known as the Centralized Radioactive Waste Management Facility (CRWMF). The Center undertakes environmental radiation monitoring in which indoor dose rates at various microenvironments, and nearby ambient environments of the facility are measured. A 2-year radiation dose data (i.e., 2017 and 2018) obtained from the monitoring exercise was used to determine whole-body exposure and cancer risk analysis for adult and child age groups. With the exception of the high dose area of the facility, observed doses in all microenvironments of the facility as well as the ambient environment were below the regulatory dose limits of 1 mSv/y and 20 mSv/y, set for radiation workers and the general public, respectively. Dose rate variation for the 2017 and 2018 datasets were not significant (p > 0.05) at 95% confidence interval (CI). Cancer risks due to exposure to alpha, neutron, and gamma radiation sources for both adult and child age groups were above the global average value of 2.90 × 10-4 reported by the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR). Gamma sources recorded the highest cancer risk followed by neutron and alpha sources with risk values of 3.95 × 10-1 and 3.92 × 10-2; 4.06 × 10-2 and 4.03 × 10-3; and 7.96 × 10-4 and 7.91 × 10-5 for the adult and child age groups, respectively. Radium (226Ra) recorded the highest activity concentration (9.62 × 1010 Bq) with 4 quantities in the inventory while plutonium-beryllium (as alloyed source) recorded the lowest activity concentration (9.82 × 1001) with 12 quantities in the inventory.

External dose reconstruction at the shore of the Metlinsky Pond in the former village of Metlino (Techa River, Russia) based on environmental surveys, luminescence measurements and radiation transport modelling.

The cohorts of people formerly living at the Techa River shoreline in the Southern Urals, Russia, are widely studied cohorts for the investigation of low-dose radiation effects to human health. The nuclear facilities of the Mayak Production Association (PA) discharged their radioactive effluents into the nearby Techa River, especially in the first years of operation. Health status of cohort member data is constantly being improved and updated. Consequently, there is a need to also improve and verify the underlying dosimetry, which gives information about the dose of cohort members. For the Techa River population, the dosimetry is handled in the Techa River Dosimetry System (TRDS). The present work shows results of a feasibility study to validate the TRDS at the location of the village of Metlino, a village just 7 km downstream from the Mayak PA. For this settlement there were two sources of external exposure, the contaminated banks of the Techa River and the contaminated shoreline of the nearby Metlinsky Pond. In the present study the north-western wall of a granary was used as a dose archive to validate dose estimates. Measurements of doses in brick accumulated over many decades and measurements of the current dose rate in bricks were combined with dose rate measurements in air above ground in front of the granary, historical contamination data and Monte-Carlo simulations. Air kerma estimates for 1949-1956 significantly different from zero could not be reconstructed for the Metlinsky Pond shoreline near the granary, but an upper dose limit could be estimated. Implications for TRDS-2016 are discussed.

Carbon-14 release and speciation during corrosion of irradiated steel under radioactive waste disposal conditions.

Carbon-14 is a key radionuclide in the safety assessment of deep geological repositories (DGR) for low- and intermediate-level radioactive waste (L/ILW). Irradiated metallic wastes generated during the decommissioning of nuclear power plants are an important source of 14C after their disposal in a DGR. The chemical form of 14C released from the irradiated metallic wastes determines the pathway of migration from the DGR into the environment. In a long-term corrosion experiment with irradiated steel simulating the hyper-alkaline, anoxic conditions of a cement-based DGR, total inorganic (TI14C2) and organic 14C contents (TO14C) in the liquid and gas phases (TG14C), as well as individual 14C-bearing carbon compounds by compound-specific radiocarbon analysis (CSRA), were quantified using accelerator mass spectrometry (AMS). The AMS-based quantification allows the determination of 14C in the pico- to femtomolar concentration range. An initial increase in TO14C was observed, which could be attributed partially to the release of 14C-bearing oxygenated carbon compounds. In the long term, TO14C and the TI14C remain constant, while TG14C increases over time according to a corrosion rate of steel of 1 nm/yr. In solution, 14C-bearing carboxylic acids (CAs) contribute ~40% to TO14C, and they are the main 14C carriers along with 14C-bearing carbonate (14CO32-). The remaining fraction of TO14C (~ 60%) is likely due to the presence of as yet non-identified polymeric or colloidal organic material. In the gas phase, 14CH4 accounts for more than 80% of the TG14C, while only trace amounts of 14CO, and other small 14C-bearing hydrocarbons have been detected. In a DGR, the release of 14C will be mainly in gaseous form and migrate via the gas pathway from the repository near field to the surrounding host rock and eventually to the environment.

Sorption of beryllium in cementitious systems relevant for nuclear waste disposal: Quantitative description and mechanistic understanding.

Beryllium has applications in fission and fusion reactors, and it is present in specific streams of radioactive waste. Accordingly, the environmental mobility of beryllium needs to be assessed in the context of repositories for nuclear waste. Although cement is widely used in these facilities, Be(II) uptake by cementitious materials was not previously investigated and was hence assumed negligible. Sorption experiments were performed under Ar-atmosphere. Ordinary Portland cement, low pH cement, calcium silicate hydrated (C-S-H) phases and the model system TiO2 were investigated. Sorption kinetics, sorption isotherms and distribution ratios (Rd, in kg⋅L-1) were determined for these systems. Molecular dynamics were used to characterize the surface processes driving Be(II) uptake. A strong uptake (5 ≤ log Rd ≤ 7) is quantified for all investigated cementitious systems. Linear sorption isotherms are observed over three orders of magnitude in [Be(II)]aq, confirming that the uptake is controlled by sorption processes and that solubility phenomena is not relevant within the investigated conditions. The analogous behaviour observed for cement and C-S-H support that the latter are the main sink of beryllium. The two step sorption kinetics is explained by a fast surface complexation process, followed by the slow incorporation of Be(II) in C-S-H. Molecular dynamics indicate that Be(OH)3- and Be(OH)42- are sorbed to the C-S-H surface through Ca-bridges. This work provides a comprehensive quantitative and mechanistic description of Be(II) uptake by cementitious materials, whose retention properties can be now reliably assessed for a wide range of boundary conditions of relevance in nuclear waste disposal.

Evaluation of Migration Radiological Equivalence for Dual Component Nuclear Waste in a Deep Geological Repository.

ABSTRACT: The paper is concerned with the issue of achieving the radiological equivalence (the equivalence of radiation risks) of radioactive waste of nuclear reactors and corresponding mass of natural uranium, taking into account the different migration ability of radionuclides in geological formations and soil. This migration radiological equivalence is being investigated for the deep burial of radioactive waste in the case of the development of a two-component nuclear power system with the concurrent use of thermal neutron reactors and fast neutron reactors. Calculations were performed of radiation doses and radiation risks of cancer death arising from consumption of drinking water from a well above a disposal site. The radiation risk relating to a two-component nuclear power system is lower than that from natural uranium; i.e., after reaching the radiological equivalence (100 y of storage) over the timescale of 109 y, the principle of migration radiological equivalence is satisfied. It would take 106 y after radioactive waste disposal to reach the migration radiological equivalence if only thermal reactors were operated. As regards consumption of well drinking water, the radiation risk does not exceed 10-5 y-1 for a two-component nuclear power system, while being 10-3 y-1 (socially unacceptable level) for a power system using only thermal reactors. Radionuclides 241Am, 239Pu, and 240Pu in drinking water make the main contribution to the doses and radiation risks of people for 104 y after the disposal of radioactive waste.

Specific features of medical care provision to the population of the Techa riverside settlements.

This paper is devoted to the issue of medical care provision to the residents of the Techa riverside settlements affected by long-term radiation exposure. The river was contaminated due to operational and accidental releases of liquid radioactive waste (LRW) by the 'Mayak' Production Association from 1949 to 1956. Contamination of the river and its floodplain with radionuclides, including long-lived90Sr and137Cs, caused long-term external and internal exposure of the population, predominantly of the bone marrow. Protective countermeasures (resettlement of residents, introduction of restrictions on the use of the river and floodplain, construction of wells, etc) did not manage to prevent relatively high exposure doses to the population. The mean dose value of bone marrow exposure in residents of the riverside settlements was 0.35 Gy, whereas the maximum values were up to 7.92 Gy. The first medical examinations by mobile teams of the Moscow Institute of Biophysics were started approximately two years after the onset of LRW releases. Since 1955, exposed residents have been followed up and are undergoing medical treatment at the Clinic of the Urals Research Center for Radiation Medicine of the Federal Medical and Biological Agency (URCRM). This center was established in response to the necessity to study the biological effects of the combined external γ-exposure and exposure due to90Sr in order to arrange medical care for the exposed population. The URCRM Clinic focuses on the provision of hematological care since cases of chronic radiation syndrome were registered among the exposed population in the early period, and increased leukemia incidence has been observed in the long-term period.

Radioecological consequences of radioactive releases due to weapons-grade plutonium production at the 'Mayak' facility in the Russian Federation.

The process of nuclear weapons production from 1949 to 1987 was accompanied by the generation of a great amount of radioactive waste. Waste processing operations and controls on discharges at this time were not to the same standard as today. Because of this, vast areas of the Urals region of Russia surrounding the Mayak Production Association (MPA) were exposed to routine and accidental radioactive contamination. The greatest contribution to the contamination was gas aerosol emissions from the MPA in the 1950s (total activity 38 PBq, mainly131I), releases of liquid radioactive waste into the Techa River from 1949 to 1956 (total activity 115 PBq, including long-lived90Sr and137Cs) and accidental atmospheric releases as a result of the thermochemical explosion of the storage tank for liquid radioactive waste in 1957 (74 PBq, relatively short-lived radionuclides being the main contributors). Protective measures helped to relieve the pressing problem of population safety in the 1950s and 1960s, but they led to the appearance of new sources of contamination in the territory surrounding the MPA-Lake Karachay (total activity of beta-emitters 4400 PBq) and the Techa Cascade of Reservoirs (TCR; total activity 8 PBq). Owing to natural radioactive decay and rehabilitation measures, the radiation situation in the East Urals Radioactive Trace (EURT) has improved considerably over the years. Economic activity has been partially restored in these territories. Only the most contaminated territory of the East Urals Radioactive Reserve cannot be used for any economic activities up to the present day. Marked non-uniformity of radioactive contamination of the EURT and the Techa River floodplain, as well as radionuclide washout from Lake Karachay and the TCR into the underground waters and the Techa River require on-going radioecological monitoring, management and regulatory supervision.

Nickel Retention on Callovo-Oxfordian Clay: Applicability of Existing Adsorption Models for Dilute Systems to Real Compact Rock.

In a waste management context, predicting the mobility of contaminants is essential. A key issue entails assessing the applicability of current knowledge on adsorption processes to natural systems. Such is the focus herein for nickel in interaction with Callovo-Oxfordian (COx) clay rock, a formation selected in France for possible radioactive waste disposal. The challenge is to link predictive modeling results with the experimental data characterizing the behavior of the labile and naturally occurring Ni fraction by implementing a new simple method. Retention studies on compact systems serve to complete this work. Combined electron microprobe and laser ablation high-resolution inductively coupled plasma mass spectrometry data show that natural Ni (∼39 mg kg-1) is homogeneously distributed within the clay matrix, which corresponds to the main reservoir (∼70%). Data interpretation of desorption tests yields an in situ Kd value of ∼80 L kg-1 and a labile Ni amount of ∼5 mg kg-1, that is, ∼13% of the Ni inventory. Predictive modeling explains the sorption data in considering that only weak clay fraction sites take part in the adsorption. The role of the clay matrix in Ni retention is confirmed by analyzing the Ni-spiked compact COx samples, whereby an increase of the Ni content in the clay fraction is observed following the retention experiment.

Investigation on gas permeability of compacted GMZ bentonite with consideration of variations in liquid saturation, dry density and confining pressure.

During the long term operation of a disposal repository, gas will be inevitably generated. Determination of gas permeability of compacted bentonite is of great importance for the safety assessment of the engineered barrier system. In the present work, the steady-state and residual pressure methods were employed to determine the gas permeability of GaoMiaoZi (GMZ) bentonite with consideration of variations in liquid saturation, dry density and confining pressure. Results show that, gas migration in saturated GMZ bentonite was mainly controlled by diffusion with an effective gas permeability of 1E-23 m2 - 1E-25 m2. While in unsaturated GMZ bentonite, significant gas flow rates could be observed, which increased stably with the increase of gas injection pressure. Klinkenberg effect was significant when gas flow through GMZ bentonite. The Klinkenberg corrected gas permeability decreased by 3.5-5.5 orders of magnitude as the liquid saturation increased from 10% to 70%. A decreasing magnitude of 1-2 orders in Klinkenberg corrected gas permeability was presented with the dry density increased from 1.5 Mg/m3 and 1.7 Mg/m3. The Klinkenberg corrected gas permeability decreased by 0-1 orders of magnitude as the confining pressure increased from 3 MPa to 7 MPa. By using the accessible porosity, gas measured intrinsic permeability could be determined with values ranged between 1E-15 m2 to 4E-15 m2, which was higher than the water measured one by 5 orders of magnitude. Additionally, a generalized power law was successfully adopted in this study to describe the evolution of gas relative permeability with the liquid saturation. Overall, the effective gas permeability, Klinkenberg corrected gas permeability, intrinsic and relative permeability determined in this study provided a comprehensive perspective to assess the buffering property of GMZ bentonite in multi-physical field coupling environment. The parameters obtained can be adopted in further simulation works for long-term safety analysis of the disposal repository from the viewpoint of gas migration.

Natural radioactivity and radiation hazard assessment of industrial wastes from the coastal phosphate treatment plants of Gabes (Tunisia, Southern Mediterranean Sea).

This work is a first contribution to the knowledge of natural radionuclides (226Ra, 238U, 40K, and 232Th) activities in phosphate rock (NORM), phosphogypsum, and phosphogypsum foam (TENORM) from the coastal fertilizer plants of Gabes (Southeastern Tunisia) and the assessment of their radiation hazards on human health and the surrounding environment. In the three studied materials, activities were found to be in the range of 35.4 (40K)-375.1 (226Ra), 10.0 (40K)-220.2 (226Ra), and 79.2 (232Th)-1168.6 Bq kg-1 (226Ra), respectively. Considering the studied radionuclides and materials, the corresponding decreasing activity orders were found to be 226Ra > 238U > 40K > 232Th and PGF > PR > PG, respectively. All human health hazard indices exceeded the worldwide recommended safety limits, which show that the workers in Gabes phosphate fertilizer plants as well as the neighboring human community may potentially be exposed to significant radiation, which may cause several diseases and malformations. It is therefore recommended to avoid and/or reduce the potential fertilizer industry radioactive impact in the area.