Search

A pre-compacted MX80 bentonite/Callovo-Oxfordian (COx) claystone mixture has been proposed as a candidate sealing material in a French deep geological repository for high-level radioactive waste. An in-depth understanding of the hydraulic property of this material is essential for accessing the over-pack corrosion and nuclide migration along the gallery. In this study, water flow in the compacted bentonite/claystone mixture was investigated by carrying out infiltration and water retention tests under constant-volume conditions alongside detailed microstructure observation. The evolution of total water hydraulic conductivity with suction was experimentally determined based on the instantaneous profile method. It was observed that the measured total hydraulic conductivity increased slowly at high suctions, then quickly at low suctions and finally approached the saturated value. A new method was developed based on the evolving pore structure, accounting for both vapour and liquid-water hydraulic conductivities. The total hydraulic conductivity predicted by the proposed method agreed well with the measured value, indicating the reliability of the predictive method. The vapour hydraulic conductivity was larger than the liquid-water one at a suction larger than 15·2 MPa, while liquid-water hydraulic conductivity became dominant at a suction lower than 15·2 MPa. In addition, based on the predicted vapour and liquid-water hydraulic conductivities, a numerical model was set up to simulate the water infiltration process. Comparison between the simulation and the measurement showed close agreement, validating the two-phase water flow mechanism in the bentonite/claystone mixture.

The paper presents results of studies on the possibility of using the ultrafiltration method supported by sorption on low-cost, easily accessible aluminosilicates to purify water contaminated with radionuclides. An aqueous solution contaminated with radionuclides in the form of cations at different oxidation states—Cs(I)-137, Co(II)-60 and Am(III)-241—as well as pertechnetate anions—TcO4−-99m—was treated by the proposed hybrid method. In the presented work, the influence of the important process parameters (i.e., pH, sorbent dosage, temperature and feed flow rate) on the removal efficiency of radionuclides was studied. The obtained results showed that hazardous impurities, both in the form of cations and anions, may be effectively removed from water by the application of sorption-assisted UF (SAUF) using the clay-salt slimes as a sorbent. As a final stage of the work, we treated the simulated liquid radioactive waste using the SAUF method, also showing satisfactory results in its purification efficiency.

Underground repositories for nuclear fuel waste will eventually become filled, and access shafts will have to be sealed. In many international studies, the preferred sealant is a compacted sand–bentonite mixture in which montmorillonite in the bentonite swells as it absorbs groundwater and develops swelling pressure. Lowered porosities in the bentonite lead to reduced hydraulic conductivity. Sealing is most effective when the seal is saturated and swelling pressure has been maximised. The time required to reach complete hydration is important. This paper describes the development of swelling pressures during hydration of a field-scale shaft-sealing experiment at Canada’s underground research laboratory. Numerical modelling used the finite-element program Code_Bright and hydro-mechanical parameters from laboratory tests to determine time-dependent porosities, total pressures and degrees of saturation in the full-scale field project. Data from laboratory-scale ‘wetting’ tests on specimens of seal material were then used in models by Liu and Dueck and Börgesson to evaluate changes in swelling pressures with time. Six years of data from field instruments indicate that the seal was functioning as designed. In the future, when the seal becomes fully hydrated and reaches stress equilibrium, the modelling suggests that swelling pressures in the seal will be in the range of 0.43–0.86 MPa. This is broadly similar to the value of 0.8 MPa suggested by a separate series of laboratory swelling tests. Full saturation and maximum swelling pressure can be expected in 17–19 years after the seal has been completed.

Geological repository of high-level nuclear waste (HLNW), the most feasible approach for the safe and permanent treatment of HLNW without human intervention, has been investigated for years by many researchers. In some countries, real facilities for the geological repository are already under construction. In Japan, however, due to complicated geologic conditions, especially underground water and fractured rock masses, and the high risk of natural disasters, ensuring the long-term stability of the method remains a struggle. The influential factors include underground water, heat generation from radioactive waste, and thermal and chemical weathering of the surrounding rock mass as a natural barrier. It is difficult to estimate and verify the long-term stability for up to one hundred thousand years, a complicated thermal–hydraulic-mechanical-chemical coupling behavior, via any field test. The objective of the study is to develop a numerical method for predicting the long-term stability of geological repositories. As the first step toward realizing this objective, heating and loading tests on cave model made of man-made rock specimens that are composed of diatoms, gypsum, and water, which are viscoplastic materials, were conducted, based on which a newly proposed numerical method with finite element method (FEM) was used to describe the thermal, mechanical, and time-dependent behavior of the model tests for a geological repository. To ensure the accuracy of the numerical calculations, all the material parameters in the thermoelasto-viscoplastic model with consideration of overconsolidation, the structure, and the influence of intermediate stress were determined via triaxial compression/creep tests under various temperatures, confining stresses, and loading rates. Finally, the validity of the numerical method was demonstrated by model tests over a limited time span.

Cold-crucible induction melting (CCIM) technology has been intensively studied as an advanced vitrification process for the immobilization of highly radioactive waste. This technology uses high-frequency induction to melt a glass matrix and waste, while the outer surface of the crucible is water-cooled, resulting in the formation of a frozen glass layer (skull). In this study, for the fabrication of borosilicate glass waste form, CCIM operation test with 60 kg of glass per batch was conducted using surrogate wastes composed of Cs, Sr, and Nd as a representative of highly radioactive nuclides generated during spent nuclear fuel management. A 60 kg-scale glass waste form was successfully fabricated through melting and draining processes using a CCIM system, and its physicochemical properties were analyzed. In particular, to enhance the controllability and reliability of the draining process, an air-cooling drain control method that can control draining through air-cooling near drain holes was developed, and its validity for draining control was verified. The method can offer controllability on various draining processes, such as molten salt or molten metal draining processes, and can be applied to a process requiring high throughput draining.

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.

Since 2009, ESA has been conducting R&D activities leading towards the development of a European capability in radioisotope power systems for space. The programme is focused on the use of americium-241 as an innovative alternative to the plutonium-238 currently used by the USA and Russia. Two different technologies are being developed by the University of Leicester in the UK: radioisotope thermoelectric generators and radioisotope heater units. The development of the fuel containment architectures, known as heat sources, has recently evolved from the design stage to more representative prototypic systems. Multiple containment layers ensure that the heat sources can survive accident conditions from launch failures to Earth re-entry, minimising the risk of inadvertently releasing radioactive material into the environment. Validated accident models are necessary for the design iteration process of these systems, and for the creation of a safety case for their launch. The research project here presented was a collaboration, supported by ESA , between the University of Leicester in the UK and ArianeGroup in France. The goals of this activity were to perform the first-ever impact tests for the European Am-based heat sources, and to start developing computer models to simulate the behaviour of the containment systems under different impact conditions in accident scenarios. Tested samples were then characterized both at ESA/ESTEC and at the University of Leicester, in order to understand the behaviour of the fuel containment materials under the most relevant accident conditions.

Rare earth (RE)-containing crystals have been investigated as media for the immobilization of RE fission products from nuclear waste streams. During reprocessing of spent nuclear fuel, fission products including REs, alkalis, and alkaline earths released from the fuel. One viable option to immobilize the RE fission products is to incorporate them into chemically durable crystalline phases in specific waste forms. This study summarizes the crystal structures and synthesis methods of six RE-containing compounds that have applications in remediation of RE fission products. These compounds include oxyapatite (AE2RE8(SiO4)6O2), oxychloride (REOCl), borosilicate (RE3BSi2O10), pyrochlore (RE2M2O7), monazite (REPO4), and perovskite [REMO3] where AE denotes alkaline earth metals and M denotes transition metals. This review provides an overview of literature on the usage of these six compounds for immobilizing RE fission products and summarizes different synthesis methods for producing these compounds. Comparisons of structural parameters with different REs in each compound are also discussed.

The Waste Treatment and Immobilization Plant (WTP) will process and stabilize nuclear waste stored in tanks on the Hanford Site. At the WTP, the tank waste will be combined with glass-forming chemicals to make a melter feed slurry that can be vitrified in joule-heated melters. Technetium-99 (99Tc), a long-lived radionuclide present in tank waste, is semi-volatile from a glass melt at elevated temperatures. A small laboratory-scale melter system has been designed by PNNL to operate under radioactive conditions, giving it the ability to vitrify actual tank waste and gain information about melter feed processability and the partitioning of components of interest. This study describes two runs performed in a duplicate system with non-radioactive simulants of Hanford tanks 241-AP-107 and 241-AN-105 to gain insight into the relationship between the retentions of 99Tc and its non-radioactive surrogate Re while also investigating the effects of increasing the reducing agent such as sucrose during vitrification on processing and the retention of semi-volatile components in the glass product. The results from these runs align with the general trends of greater retention in the glass of Re compared to 99Tc and the improved retention in the glass of Re with increased reducing agent.

In the present study, radiation shielding and protection ability of prepared Flyash-lime-Gypsum (FaLG) bricks has been studied in terms of energy exposure build up factors and dose parameters. The energy exposure build up factors of Flyash-lime-Gypsum (FaLG) bricks have been calculated for the energy range of 0.015 MeV–15 MeV and for penetration depth upto 40 mfp directly using a new and simplified Piecewise Linear Spline Interpolation Method (PLSIM). In this new method, the calculations of G.P fitting parameters are not required. The verification and accuracy of this new method has been checked by comparing the results of exposure build up factor for NBS concrete calculated using present method with the results obtained by using G.P fitting method. Further, the relative dose distribution and reduced exposure dose rate for various radioactive isotopes without any shielding material and with Flyash-lime-Gypsum (FaLG) bricks have been calculated in the energy range of 59.59–1332 keV. On the basis of the obtained results, it has been reported that the prepared Flyash-lime-Gypsum (FaLG) bricks possess satisfactory radiation shielding properties and can be used as environmentally safe storage facilities for low level nuclear waste.

Intermediate-level radioactive waste (ILW) is not subject to legal approval for cave disposal in Korea. To solve this problem, a spent resin treatment device that separates 14C-containing resin from zeolite/activated carbon and desorbs 14C through a microwave device has been developed. In this study, we evaluated the radiological safety of the operators performing repair work in the event of a failure in such a device treating 1 ton of spent resin mixture per day. Based on the safety evaluation results, it is possible to formulate a design plan that can ensure the safety of workers while developing a commercialized device. When each component of the resin treatment device can be repaired from the outside, the maximum and minimum allowable repair times are calculated as 263.2 h and 27.7 h for the 14C-detached resin storage tank and zeolite/activated carbon storage tank, respectively. For at least 6 h per quarter, the worker's annual dose limit remains within 50 mSv/year; further, over 5 years, it remained within 100 mSv. At least 6 h of repair time per quarter is considered, under conservative conditions, to verify the radiological safety of the worker during repair work within that time.

Accurate and precise determination of plutonium and uranium in Fast Breeder Reactor (FBR) fuels is one of the most important steps in the chemical quality control of these fuels. Along with high precision, minimization of radioanalytical waste generated during measurements is of equal importance. Till date, routine wet radioanalytical techniques used for this purpose produce a lot of radioactive waste which needed an extensive recovery procedure. This paper describes a novel and simple micro-analytical method for determination of the Pu percentage with respect to uranium in fast reactor nuclear fuels, both carbide and oxide, using micro X-Ray Fluorescence (μ-XRF) spectrometry. The developed methodology requires ∼200 ng of sample for analysis and the radioactivity in such sample specimens is below 500 becquerel. Therefore, these sample specimens could be isolated with a layer of Scotch tape and analyzed. The detection limit for U and Pu, obtained using the developed methodology was ∼60–64 ng mL−1. Using a single calibration plot, the Pu percentage in fuel samples could be determined in the entire concentration range from 10 to 90% for both the fuels. The average precision obtained was 0.23% (1σ; n = 3) for samples having a (U/Pu) concentration of

Detecting radioactive materials in mixtures is challenging due to low concentration, environmental factors, sensor noise, and others. This paper presents new results on nuclear material identification and mixing ratio estimation for mixtures of materials in which there are multiple isotopes present. Conventional and deep learning-based machine learning algorithms were compared. Both simulated and actual experimental data were used in the comparative studies. It was observed that some newly developed machine learning and deep learning methods have better performance than a conventional method based on region of interest (ROI). Extensive experiments also demonstrated that trained models using low signal-to-background ratio (SBR) data can better generalize to other SBR conditions.

Most research activities in the nuclear mining industries are directed to separation of heavy and radioactive minerals from rocks and sediments. This process is completed without utilizing tailing and mineral waste materials. Tailing materials containing radioactive wastes are considered as a major environmental problem; especially with increasing the exploration of metals and lower grade mineral deposits. In the present work, borosilicate glass could be prepared for immobilizing these nuclear wastes in the designing compositions by melting at 1200–1250 °C. NaI (TI) gamma-ray spectroscopy system was used to detect 238U, 232Th, 226Ra, and 40K concentrations in tailing materials and prepared glass to emphasize their safe dose. The produced glass was characterized by XRD, SEM, FTIR, and UV spectrophotometer. Vickers hardness and leaching properties were also studied. The results showed that compact - structured glass with high durability towards groundwater was obtained. All studied environmental parameters indicated that the glass radiation dose was lower compared to international averages, thus the glass can be safely used in nuclear waste immobilization.

Bentonite is a favourable candidate as a sealing material in the deep geological repository to safe storage of highly radioactive nuclear waste in several countries. The high swelling characteristic made bentonite famously type of clay that extracted from smectite group with the main mineral is montmorillonite that has a double diffusion layer. The benefits of sodium bentonite like Wyoming bentonite has proven in the industry. Research on modifying the properties of calcium bentonite to high swelling clay like sodium bentonite for it be able to benefit the industry has been done. Most of the research attempted to improve the workability of calcium bentonite as an alternative material for replacing sodium bentonite in construction industry; however, very little is known about using raw calcium bentonite to improve weak soil subgrades. This paper highlighted the usage of calcium bentonite adapted in construction industry especially in soil stabilization activity and some experimental analysis has been made on the type of calcium bentonite found in British Columbia mine.

Pre-compacted MX80 bentonite/Callovo-Oxfordian (COx) claystone mixture has been considered as a promising sealing/backfilling material in deep geological repository for high-level radioactive waste in France. When the pre-compacted blocks are emplaced in the gallery, technological voids can remain. After the infiltration of groundwater from the host rock, they will swell freely, filling the technological voids, and then undergo further hydration under constant-volume conditions. From the perspective of the storage safety, it is essential to understand the effects of technological voids and hydration time on the hydro-mechanical behaviour of such pre-compacted blocks. In this work, a series of infiltration tests at various hydration times was carried out on compacted MX80 bentonite/COx claystone mixture with different technological voids. The evolutions of the swelling pressures in axial and radial directions as well as the hydraulic conductivity were monitored while wetting. After the predetermined hydration times, the dry density, water content and suction at different positions were determined, together with the microstructure investigation using mercury intrusion porosimetry. It was observed that the soils close to initial voids swelled and filled the initial voids upon contact with water, with significant increases in large-pore and medium-pore void ratios. From the variation of dry density profile with time, compression and swelling zones could be identified: in the compression zone, the soils corresponding to the initial soils were subjected to compression, with decrease in large-pore and medium-pore void ratios over time, whereas in the swelling zone, the soils with a higher dry density than the expected final one underwent further swelling, with large-pore void ratio increasing until saturation and then a slight decrease due to water redistribution in the soil. Owing to the soil density heterogeneity, the axial swelling pressure and hydraulic conductivity of samples with voids were slightly larger than those of samples without voids; contrarily, the radial swelling pressure was lower than that of the samples without voids.

In this chapter, different methods of disposal of radioactive waste adopted in nuclear medicine are elaborately discussed, namely, decay-in-storage, disposal into sewage, incineration, and transfer to authorized recipients. Sewer disposal and incineration are restricted by maximum permissible concentration limit allowed by the NRC in the sewer and the air. Transfer of radioactive waste to authorized recipients requires a manifest, which gives the details of shipment such as type and quantity of material, date, etc. Disposal of sealed sources and cadavers with radioactivity is mentioned.

The Department of Transportation regulates the transportation of radioactive materials. Types of packaging, sturdiness of packages, and categories of materials for shipment are discussed.

This chapter is an abridged version of NRC 10CFR35 that describes the conditions for the uses of byproduct materials in humans for medical purposes. Issues of license requirement, dosage measurement, calibration of equipment, radiation safety officer, radiation safety committee, and written directives are discussed. Survey meter and measurement of exposure rate are described. Conditions of medical mobile service are spelled out. Medical events are defined and requirements for their reporting are elucidated. Also conditions for reporting the fetal dose are given.

The present article describes the decommissioning of a self-shielded 18 MeV medical cyclotron IBA Cyclone 18/9 after 14 years of operation. A Monte Carlo simulation of the possible nuclear reactions was performed in order to plan the decommissioning activities. During the cyclotron dismantling, the activities of the cyclotron components, concrete wall and floor samples were measured. Residual activities were analysed by means of an HPGe detector and liquid scintillation counting, and compared with simulation data. Dosimetry of the staff involved in the decommissioning procedure was monitored by individual TL dosimeters and/or digital dosimeter. The cyclotron component analysis confirmed the presence of gamma and pure beta emitters, 22Na, 54Mn, 60Co, 65Zn, 207Bi, 55Fe, 63Ni at different values of specific activity, depending on the positioning of the sample point and on the alloy of the sampled part. In these components the presence of gamma and pure beta emitters was measured 5 years after the shutdown at levels far above clearance limits as defined by the ‘Recommended radiological protection criteria for the recycling of metals from the dismantling of nuclear installations’ (RP89) guidelines. The simulation, carried out by FLUKA Code (version 2020.0.5) on the cyclotron components, provided good agreement with measurements, with a maximum discrepancy of the same order as the uncertainties. Four engineers of the cyclotron maintenance staff were involved in the dismounting of the hottest components and rigging of the cyclotron in the deposit 6 months after shutdown and two engineers were involved during the drilling phase 3.5 years after shutdown. The measured dose from external exposure of the involved staff was lower than 100 μSv person−1 during the first phase and lower than 20 μSv person−1 during the final drilling phase. Measured doses from intake were negligible. In conclusion, the decommissioning of the 18 MeV cyclotron does not represent a risk for the involved staff, but, due to the presence of long-lived radioisotopes, the cyclotron components are to be treated as low level radioactive waste, and stored in an authorised storage area for at least 25 years after shutdown.

Cesium radioisotopes are among the majors that are presented in liquid radioactive waste causing long-term environmental threats. The mandatory treatment of these streams is regenerative goal in wh...

Purification of fifteen NaCl samples from natural and different area in the middle and south of Iraq are prepared and studied the structural characteristics of samples by powder X-ray diffraction analysis at 𝛌=1.542Å. The present work considered the specific activities of naturally occurring radioactive materials in salt samples, which are measured and analyzed using high resolution HPGe system. Also, the radiological parameters have been considered in this work. The average specific concentration (in Bq.kg-1) of the salt samples is found 16.864.92 for 226Ra, 5.972.05 for 232Th and 9.852.8 for 40K, respectively, which are below the national levels, 40Bq.kg-1, 30Bq.kg-1 and 400Bq.kg-1, respectively (UNSCEAR, 2000). Similarly the absorb dose rates are in the range3.71 nGy.h-1 to 12.591nGy.h-1.The measured radiation hazard indices show that the concentrations of these radionuclides in all samples are within the allowable limits.

Radon exhalation from soil and ores is among the most dangerous risks for the public health care. The impact becomes even more powerful when technological enhanced naturally occurring radioactive materials (TENORM) are used for public and private building. Here, we built a down-scaled model (a 1.0 m × 1.0 m × 0.5 m parallelepiped) of a dwelling, whose construction materials contain TENORM harvested in a site in Crotone (Italy). We observed an increase of the radon activity in the model when TENORM residues are employed, reaching a value around 120 Bq/m3, i.e. up to three times higher than the typical values of Crotone indoor environment, which ranges around 40 Bq/m3. These results have then been compared to a real use case. The correspondence found between the values of radon activity concentration in the model and in the use case suggests that estimating the radon concentration is a useful method to target TENORM presence inside buildings.

The sorption of 241Am3+, 85Sr2+ and 137Cs+ by calcium, strontium and barium alginates has been studied under different operation conditions. The most prominent adsorption was found in the pH range of 5–6 for all systems, even if the observed dependence on the acidity of the solution was small. The most favourable time for the adsorption process was found to be about 2 h for calcium alginate and 4 h for the other two sorbents.

Biosorption has been examined for the treatment of aqueous solutions containing uranium, a radiotoxic pollutant. Nevertheless, the evaluation of the role of process variables by experimental design on the use of hydroxyapatite and bone meal as biosorbents for uranium has not yet been previously addressed. In this study, the effects of adsorbent dosage (M), initial uranium concentrations ([U]0), and solution pH were investigated, using a two-level factorial design and response surface analysis. The experiments were performed in batch, with [U]0 of 100 and 500 mg L−1, pH 3 and 5, and adsorbent/uranium solution ratios of 5 and 15 g L−1. Contact time was fixed at 24 h. Removal rates were higher than 88%, with a maximum of 99% in optimized conditions. [U]0 and M were found to be the most influential variables in U removal in terms of adsorption capacity (q). The experiments revealed that bone meal holds higher adsorption capacity (49.87 mg g−1) and achieved the highest uranium removal (~ 100%) when compared to hydroxyapatite (q = 49.20 mg g−1, removal = 98.5%). The highest value of q for both biomaterials was obtained for [U]0 = 500 mg L−1, pH 3, and M = 5 g L−1. Concerning the removal percentage, bone meal achieved the best performance for [U]0 = 500 mg L−1, pH 3, and M = 15 g L−1. Further experiments were made with real radioactive waste, resulting in a high uranium adsorption capacity for both materials, with 22.11 mg g−1 for hydroxyapatite and 22.08 mg g−1 for bone meal, achieving uranium removal efficiencies higher than 99%.

Over the last 50 years, nuclear energy has reduced US energy-related CO2 emissions by over 30 gigatons compared to if the same electricity were produced by fossil fuels such as coal and natural gas. However, many kilotons of spent nuclear fuel have accumulated at different sites across the country, and sociopolitical factors have frustrated efforts to address the challenge of nuclear waste disposal. Presently, a consolidated interim storage facility in Andrews, Texas, provides a promising temporary solution. In this paper, we compare the technical and policy risks of the project to continued storage at independent spent fuel storage installations. Our results indicate that the cost of the radiological risk is low (

We investigated interfacial reactions of U(VI) in the presence of Suwannee River natural organic matter (NOM) at acidic and neutral pH. Laboratory batch experiments show that the adsorption and precipitation of U(VI) in the presence of NOM occur at pH 2 and pH 4, while the aqueous complexation of U by dissolved organic matter is favored at pH 7, preventing its precipitation. Spectroscopic analyses indicate that U(VI) is mainly adsorbed to the particulate organic matter at pH 4. However, U(VI)-bearing ultrafine to nanocrystalline solids were identified at pH 4 by electron microscopy. This study shows the promotion of U(VI) precipitation by NOM at low pH which may be relevant to the formation of mineralized deposits, radioactive waste repositories, wetlands, and other U- and organic-rich environmental systems.

The accident at Fukushima Daiichi Nuclear Power Plant caused radioactive materials to spread outside the plant. To limit exposure in the area, air dose rates have been measured. When the dose rate exceeded the acceptable upper bound, the area was decontaminated by stripping the soil and washing roads. Immediately after the accident, it was not clear how much of an effect outdoor contamination had on dose rates inside houses. This paper discusses the effect of outdoor contamination on indoor air dose rates and factors influencing those dose rates based upon actual measurements taken of a wooden house standing on flat land in the area around the power plant. In addition, the effects of indoor horizontal distribution, height distribution, and decontamination are discussed using simple model simulations. As a result, the following was found, although within the range of air dose rates and location conditions measured in this study. Horizontal variation in indoor air dose rates decreases monotonically toward the centre of the house. Once decontamination has been performed, the overall dose rate decreases and variation becomes smaller. This trend is thought to be mainly due to expansion of a low-dose area. Vertical changes in indoor air dose rates are not straightforward. It is estimated that the dose rate decreases near the ground surface, then tends to increase going higher and finally decreases even higher up. This change in dose rates is thought to be due to the effect of radioactive materials permeating the soil. Recognition of the effects of these factors is important when estimating indoor air dose rates because indoor air dose rates are affected by the degree of decontamination, house size, and degree of penetration of radionuclides into the soil.

Fractured basements are not only potential reservoirs for hydrocarbon resources, but they also provide significant storage space for carbon dioxide ([Formula: see text]) sequestration and radioactive waste disposal. However, fractured basements are challenging for seismic imaging methods due to the complexities in their fault and fracture networks, strong heterogeneity, highly variable structural dip, and strong impedance contrasts between the basement rocks and the surrounding sediments. We present a field case in which a walkaway vertical seismic profiling (VSP) survey was conducted at a fractured basement play located in the Bohai Bay Basin, China, to improve the resolution compared with a preexisting surface seismic profile. Using the advanced random-space-shift (RSS) reverse time migration (RTM), we obtain a high-resolution image with a clear delineation of the highly faulted dipping basement. From numerical and field examples, we find that the application of the RSS-RTM improves the final image by mitigating unavoidable errors in the migration velocity model, which would otherwise result in an unfocused image using the conventional RTM approach. In addition, we determine the importance of proper wavefield separation using 3C recordings, which is the key to ensuring the quality of the final image. With an optimized VSP imaging workflow, we provide an enhanced image for the fractured basement to support the geologic interpretations and development decisions.

After nuclear accidents, people can be contaminated internally via ingestion, inhalation and via intact skin or wounds. The assessment of absorbed, committed doses after internal exposure is based on activity measurement by in vivo or in vitro bioassay. Estimation of dose following internal contamination is dependent on understanding the nature and form of the radionuclide. Direct counting methods that directly measure γ-rays coming from within the body or bioassay methods that measure the amount of radioactive materials in urine or feces are used to estimate the intake, which is required for calculating internal exposure doses. The interpretation of these data in terms of intake and the lifetime committed dose requires knowledge or making assumptions about a number of parameters (time, type of exposure, route of the exposure, physical, biological and chemical characteristics) and their biokinetics inside the body. Radioactive materials incorporated into the body emit radiation within the body. Accumulation in some specific organs may occur depending on the types of radioactive materials. Decorporation therapy is that acceleration of the natural rate of elimination of the contaminant will reduce the amount of radioactivity retained in the body. This article presents an overview of treatment of radiological contamination after internal contamination.

Uranium tailing ponds are a potential major source of radioactive pollution. Solidification treatment can control the diffusion and migration of radioactive elements in uranium tailings to safeguard the surrounding ecological environment. A literature review and field investigation were conducted in this study prior to fabricating 11 solidified uranium tailing samples with different proportions of PVA fiber, basalt fiber, metakaolin, and fly ash, and the weight percentage of uranium tailings in the solidified body is 61.11%. The pore structure, volume resistivity, compressive strength, radon exhalation rate variations, and U(VI) leaching performance of the samples were analyzed. The pore size of the solidified samples is mainly between 1 and 50 nm, the pore volume is between 2.461 and 5.852 × 10−2 cm3/g, the volume resistivity is between 1020.00 and 1937.33 Ω·m, and the compressive strength is between 20.61 and 36.91 MPa. The radon exhalation rate is between 0.0397 and 0.0853 Bq·m−2·s−1. The cumulative leaching fraction of U(VI) is between 2.095 and 2.869 × 10−2 cm, and the uranium immobilization rate is between 83.46 and 85.97%. Based on a comprehensive analysis of the physical and mechanical properties, radon exhalation rates, and U(VI) leaching performance of the solidified samples, the basalt fiber is found to outperform PVA fiber overall. The solidification effect is optimal when 0.6% basalt fiber is added.

Brachytherapy programs within radiation therapy departments are subject to stringent radiation safety requirements in order to ensure the safety of the staff and patients. Training programs often include brachytherapy-specific radiation safety training modules that address the specific risks associated with radioactive sources, emergency procedures, and regulatory requirements specific to the use of radioisotopes. Unlike other uses of radioactive materials, brachytherapy uses sealed sources and therefore under routine operations does not encounter radioactive contaminants. This article presents an unusual clinical situation in which an 125I brachytherapy seed was damaged during routine clinical workflow, resulting in radioactive contamination within the clinical environment. Decisions made at the time of the incident resulted in contamination that spread beyond the initial location. The incident highlighted a shortcoming of the radiation safety program in preparing staff for the possibility of having to deal with unsealed radioactivity. Brachytherapy programs would be strengthened by including training specific to radioactive contamination in their emergency training to equip staff to respond to unexpected damage to the sealed sources.

Detection and remediation of radioactive components have become the focus of worldwide research interest due to the ever-increasing generation of nuclear waste and the concerns on nuclear accidents. Among the numerous radionuclides, uranium and its isotopes receive the most attention because of their high proportion in nuclear waste and long half-life. Herein, a highly luminescent terbium-organic framework, formulated as [Tb4(C29O8H17)2(NO3)4(DMF)4(H2O)4]·4H2O·8.5DMF (YTU-100), with exceptional sensitivity and selectivity toward uranium was successfully prepared. The material exhibits fast adsorption kinetics and moderate sorption capacity. Interestingly, the luminescence intensity variation highly correlates to the amount of adsorbed uranium, which results in a quantitative, accurate, and selective uranium detection manner. The detection limits in deionized water and tap water were determined to be 1.07 and 0.75 ppb, respectively, which are lower than the US Environmental Protection Agency standard of the maximum contamination of uranium in drinking water. YTU-100 offers an alternative approach for building multifunctional MOFs used for simultaneous detection and removal of uranium from aqueous solutions.

A bentonite buffer is an important component of the engineering barrier of a high-level radioactive waste (HLW) repository, which can delay the infiltration of groundwater and retard the migration of radionuclides. The long-term-coupled thermo-hydro-mechanical (THM) performances and the retardation capacities of composite bentonite buffers are studied in this work. A theoretical coupled THM model is presented, and a finite element code, namely LAGAMINE, is used for the numerical simulation under coupled THM conditions. Six kinds of composite bentonite buffer computational models are considered. The boundary conditions of these models and computational parameters of bentonite are presented. The repository time for HLW is considered to be 10,000 years. The evolution processes of temperature, saturation degree and swelling pressure are presented according to the computational results. The combination forms of blocks and joints have an obvious influence on the swelling pressure in the bentonite buffers. The circumferential and radial joints can significantly reduce the maximum value of the swelling pressure. The influence coefficient for the retardation capacity ψ is proposed for evaluating the retardation capacity of composite bentonite buffers. According to the value of ψ, the six kinds of computational models are ranked in terms of retardation capacity, and three kinds of models are suggested for usage in design. The results of this investigation are helpful for optimising the design of bentonite buffers for HLW repositories.

Diffusion is the dominant process involved in the transport of 129I through geological media; therefore, the evaluation of the experimental results on the diffusion of I− through compacted bentonite was performed by three diffusion models, CC-CC, CC-VC, and VC-VC. Multiple experiments were cross-checked to verify internal consistency and thereby obtain reliable diffusion parameters. The results showed that I− diffusion coefficients ranged from 1.15 × 10–12 to 4.67 × 10–12 m2/s, which were consistent with those of the literature. These techniques could potentially be performed to measure the diffusion coefficients of non-sorbing or weakly sorbing radionuclides onto compacted bentonite in radioactive waste repositories.

In this paper, natural (geological and hydrological) and climatic impact conditions on the influence zone for two different sites of radioactive waste disposals have been studied. Probabilistic approach to assessment of the groundwater vulnerability from radionuclide contamination during disposal of radioactive waste is developed. As to climatic conditions, an effective numerical and analytical methods for annual precipitation rates assessment of rare recurrence have been proposed which allow to take into account uncertainties of rare events as well.

The article analyses the results of application of the radioactive waste activity measurement method based on measuring the energy distribution of gamma radiation near the container at the Novovoronezh NPP. The proposed method is applied to homogeneous radioactive waste and provides a reduction of the characterization costs. The results of an experimental study of the method and proposals for its improvement are presented.

The accident at the Fukushima Daiichi Nuclear Power Plant in 2011 released many radioactive elements into the environment, contaminating a large amount of soil. Currently, 14 million m3 of contamin...

For the vitrification of zeolites used for absorbing radioactive cesium, the composition of the glass in which nepheline forms at the cooling rates of 0.5, 1, 2.5°C min−1 were clarified for the Na2...

The reprocessing of spent nuclear fuel (SNF) of nuclear power plants leads to the generation of radioactive waste (RW), including high level waste (HLW). HLW contains hazardous long-lived fission products and actinides, therefore, the creation of effective methods of handling it is the main task of the final stage of the closed nuclear fuel cycle. At present, HLW is immobilised in Na–B–Si or Na–Al–P glasses, which does not meet the requirements of safety and economic efficiency. In particular, the low content of waste (3–20 wt %) in the glasses worsens the use of the deep storage space for HLW. Over time, the glasses crystallize with increasing solubility of the matrix in groundwater. The solution to the problem of handling HLW is related to the separation of radionuclides onto fractions. Many of these techniques have already been tested on real liquid waste from SNF reprocessing. As a result, fractions of elements can be obtained for immobilisation in optimal matrices with crystalline (mineral-like) phases. Compounds with structures of pyrochlore, zirconolite, monazite, brannerite, perovskite, britholite, murataite, garnet are suitable for incorporating actinides. For Cs and Sr, phases with the structures of hollandite, pollucite, perovskite, and langbeinite are proposed. Glass crystalline materials (glassceramics) are optimal for isolating wastes of the complex composition. Given the methods for the synthesis of HLW matrices and promising technologies for their industrial production are considered.

Processes occurring when struvite (potassium-magnesium and ammonium–magnesium phosphates) matrices are formed in high-salt boron-containing model solutions simulating the evaporated still bottoms from a nuclear power plant with PWR reactors were examined. Relations were obtained, describing how the strength of the matrices depends on the content of salts in a matrix, and the main phases responsible for the strength were determined. It was shown that the matrices synthesized at the stoichiometric ratio between the reagents, MgO and NH4H2PO4, fail to satisfy the requirements to cemented radioactive wastes as regards the strength and leaching of β-radionuclides. Introduction of a 10–20% excess of MgO during the synthesis of the matrices makes it possible to significantly improve the strength of the matrices and diminish the degree of 137Cs leaching. Matrices were obtained containing up to 17–17.5 wt % salts, which is 1.7–2.5 times larger than that in portland cement matrices. These matrices fully satisfy the requirements to cemented radioactive wastes.

The products of oil shale processing are of interest as sorbents for the purification of liquid radioactive waste (LRW). The work studies the adsorption and textural characteristics of samples of heat-treated oil shale by the method of low-temperature adsorption–desorption of nitrogen and sorption of radionuclides from aqueous solutions. It is shown that with an increase in the processing temperature of oil shale, their sorption capacity increases during the extraction of 137Cs and 60Co radionuclides from aqueous solutions, while the processing temperature does not affect the sorption of 85Sr. This effect of the treatment temperature on the oil shale sorption parameters is associated with the removal and carbonization of the organic constituent, as well as the different contribution of ion exchange and adsorption processes to the adsorption of 137Cs, 60Co, and 85Sr radionuclides. Therewith the distribution coefficients of radionuclides reaches 103–104 cm3/g.

Buffer/backfill materials for radioactive waste disposal sites consist of pure bentonite or bentonite-rock mixtures. In this study, the batch test method was used to obtain the sorption characteristics of important radionuclides such as Cs, Sr and Co on buffer/backfill materials; i. e., mixing Wyoming MX-80 bentonite or local Taiwanese Zhi-Shin bentonite with possible host rock (argillite and granite) in different proportions (0∼100%). The distribution coefficients (Kd) for Cs, Sr and Co were obtained from the experiments. The distribution coefficient for the bentonite-rock mixtures were found, with more than 50% of mixing proportion of bentonite to argillite or granite, to have very similar values to that of pure bentonite. Furthermore, it was clearly found that the sorption of Cs, Sr and Co to bentonite-rock mixtures is decreased as ionic strength of the liquid phase is increased from 0.001M to 1M for NaCl solutions. According to the experimental results, in synthetic groundwater, it is quite convenient and helpful to assess the distribution coefficients (Kd) of Cs, Sr and Co for buffer/backfill materials using batch sorption experiments with bentonite-rock mixtures of fixed mixing proportions.

Rationally designed new materials for the selective detection and adsorption of

In the present work, radium (226Ra), thorium (232Th), potassium (40 K), and cesium (137Cs) (Bq/kg) was measured for 24 soil samples of some districts of Al-Doura city in Baghdad governorate. The gamma spectrometry method with NaI (Tl) detector was used for radiometric measurements. The average values of specific activity for 226Ra, 232Th, 40K and 137Cs in soil samples were 38.03, 42.48, 16.34 and 2.76 Bq/kg, respectively. The radiation indicators were measured and the average values recorded were 100 Bq/kg, 44.6 nGy/h, 0.27 Bq/kg, 0.373 Bq/kg, 0.219 mSv/y, 0.055 mSv/y and 0.689 Bq/kg for the parameters of radium equivalent activity (Raeq), the absorbed dose rate (D), the external hazard index (Hex), the internal risk index (Hin), and the annual effective equivalent dose for the indoor (In Eff.), outdoor annual effective dose equivalent (Out Eff.) and representative gamma index (Iγr), respectively. All the health hazard indices were well below their recommended limits, except in one soil sampling site (AL-Eskan site) which were found to be 0.495 Hex and 0.668 Hin.