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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.

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.

When X-ray energy above 8 MV is used, photoneutrons are generated by the photonuclear reaction, which activates the components of linear accelerator (linac). Safely managing the radioactive material, when disposing linac or replacing components, is difficult, as the standards for the radioactive material management are not clear in Korea. We surveyed the management status of radioactive components occurred from medical linacs in Korea. And we also measured the activation of each part of the discarded Elekta linac using a survey meter and portable High Purity Germanium (HPGe) detector. We found that most medical institutions did not perform radiation measurements when disposing of radioactive components. The radioactive material was either stored within the institution or collected by the manufacturer. The surface dose rate measurements showed that the parts with high surface dose rates were target, primary collimator, and multileaf collimator (MLC). 60Co nuclide was detected in most parts, whereas for the target, 60Co and 184Re nuclides were detected. Results suggest that most institutions in Korea did not have the regulations for disposing radioactive waste from linac or the management procedures and standards were unclear. Further studies are underway to evaluate short-lived radionuclides and to lay the foundation for radioactive waste management from medical linacs.

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.

Presented in this paper is a methodology for combining a Bayesian statistical approach with Data Quality Objectives (a structured decision-making method) to provide increased levels of confidence in analytical data when approaching a waste boundary. Development of sampling and analysis plans for the characterisation of radioactive waste often use a simple, one pass statistical approach as underpinning for the sampling schedule. Using a Bayesian statistical approach introduces the concept of Prior information giving an adaptive sample strategy based on previous knowledge. This aligns more closely with the iterative approach demanded of the most commonly used structured decision-making tool in this area (Data Quality Objectives) and the potential to provide a more fully underpinned justification than the more traditional statistical approach. The approach described has been developed in a UK regulatory context but is translated to a waste stream from the Fukushima Daiichi Nuclear Power Station to demonstrate how the methodology can be applied in this context to support decision making regarding the ultimate disposal option for radioactive waste in a more global context.

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.

In the geological repository of radioactive nuclear waste, anaerobic corrosion can generate hydrogen, and may conservatively lead to the production of hydrogen-air layer. The accumulated hydrogen may cause a hazardous flame propagation resulting from any potential ignition sources. This study numerically investigates the processes of ignition and flame propagation in the layered mixture. Simple geometry was chosen to represent the geological repository, and reactive flow simulations were performed with different ignition power, energy, and locations. The simulation results revealed the effects of power and energy of ignition source, which were also analyzed theoretically. The mechanism of layered flame propagation was suggested, which includes three stages: propagation into the hydrogen area, downward propagation due to the product gas, and horizontal propagation along the top wall. To investigate the effect of the ignition source location, simulations with eight different positions were performed, and the boundary of hazardous ignition area was identified. The simulation results were also explained through scaling analysis. This study evaluates the potential risk of the accumulated hydrogen in geological repository, and illustrates the layered flame propagation in related ignition scenarios.

As NPPs continue to operate, liquid waste continues to be generated, and containers are needed to store and transport them at low cost and high capacity. To transport and store liquid phase very low-level radioactive waste (VLLW), a container is designed by considering related regulations. The design was constructed based on the existing container design, which easily transports and stores liquid waste. The radiation shielding calculation was performed according to the composition change of barium sulfate (BaSO4) using the Monte Carlo N-Particle (MCNP) code. High-density polyethylene (HDPE) without mixing the additional BaSO4, represented the maximum dose of 1.03 mSv/hr (

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.

Tritium extraction from radioactively contaminated cement mortar samples was performed using heating and liquid scintillation counting methods. Tritiated water molecules (HTO) can be present in contaminated water along with water molecules (H2O). Water is one of the primary constituents of cement mortar dough. Therefore, if tritium is present in cement mortar, the buildings and structures using this cement mortar would be contaminated by tritium. The radioactivity level of the materials in the environment exposed to tritium contamination should be determined for their disposal in accordance with the criteria of low-level radioactive waste disposal facility. For our experiments, the cement mortar samples were heated at different temperature conditions using a high-temperature combustion furnace, and the extracted tritium was collected into a 0.1 M nitric acid solution, which was then mixed with a liquid scintillator to be analyzed in a liquid scintillation counter (LSC). The tritium extraction rate from the cement mortar sample was calculated to be 90.91% and 98.54% corresponding to 9 h of heating at temperatures of 200 °C and 400 °C, respectively. The tritium extraction rate was close to 100% at 400 °C, although the bulk of cement mortar sample was contaminated by tritium.

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.

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...

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.

Low level radioactive liquid waste (LLW) obtained after treatment of Intermediate level radioactive waste (ILW) is highly alkaline and rich in 99Tc as the pertechnetate ion. It is considered as a long-term hazard in nuclear waste disposal. Solvent extraction of Pertechnetate species was performed using crown ether impregnated hollow fibre membrane using nitrobenzene and TBP solvents in different membrane modules. Better D values were obtained with Nitrobenzene however in view of toxicity of nitrobenzene study was also conducted with TBP and TBP nitrobenzene mix solvent. With nitrobenzene solvent almost 95% of 99Tc was separated in four contacts. However with increase in TBP content the percentage extraction is decreasing due to decrease in D value. The results are presented and discussed in the paper.

The efficiency of decontamination of model spent ion exchange resins, contaminated with magnetite and hematite, with mineral acid solutions, and using electro-decontamination, was evaluated. It has been shown that effective hematite dissolution occurs in concentrated mineral acid solutions. However, the use of direct current increases the decontamination efficiency of spent ion exchange resins contaminated with hematite. It is determined that with increasing voltage and acid concentration, the dissolution efficiency of hematite deposits increases and can exceed 99%. It has been shown that hematite dissolution is accompanied by secondary adsorption of radionuclides due to ion exchange, which can be removed with sodium nitrate solutions.

Rare-earth (RE) industries generate a massive amount of radioactive residue containing high thorium concentrations. Due to the fact that thorium is considered a non-economic element, large volume of these RE processed residues are commonly disposed of without treatment. It is essential to study an appropriate treatment that could reduce the volume of waste for final disposition. To this end, this research investigates the applicability of carbon-based adsorbent in separating thorium from aqueous phase sulphate is obtained from the cracking and leaching process of solid rare-earth by-product residue. Adsorption of thorium from the aqueous phase sulphate by carbon-based electrodes was investigated through electrosorption experiments conducted at a duration of 180 minutes with a positive potential variable range of +0.2V to +0.6V (vs. Ag/AgCl). Through this research, the specific capacity obtained was equivalent to 1.0 to 5.14 mg-Th/g-Carbon. Furthermore, electrosorption of thorium ions from aqueous phase sulphate is found to be most favorable at a higher positive potential of +0.6V (vs. Ag/AgCl). This study's findings elucidate the removal of thorium from the rare-earth residue by carbon-based electrodes and simultaneously its potential to reduce disposal waste of untreated residue.

Non-consolidated information base, repetitive work, and data unreliability are some of the issues that may hamper a radioactive waste characterization program. Therefore, an algorithmic software model for the characterization of radioactive waste is proposed to boost waste management in nuclear-based facilities. The data obtained have enabled the construction of a guide code in the pseudotechnical language. The developed system fully covered the characterization stage, and met the needs identified in the radioactive waste management service. Here, we propose systematic documentation of the characterization methods that are employed during the radioactive waste management processes, besides formulating the system’s data recording.

Radioactive contamination induced by nuclear leakage accidents, nuclear and radiation terrorism and other radioactive pollution events needs to be controlled and removed urgently, otherwise it will cause a wider range of pollution diffusion and radioactive risk. Strippable coatings have been employed for radioactive contamination removal for a period of time. Since ALARA™ 1146 paved the way, an ever-growing number of developments that aim to bring higher performance and new pattern decontamination process have been conducted. In this article, the composition and decontamination performance of strippable coatings in recent years are reviewed. Some so-called unconventional and new forms of strippable decontamination materials have also been presented. In addition, the direction of growth potential for strippable coatings in the future is also prospected. These directions include (a) paying more attention to optimize the decontamination performance, (b) enabling the strippable coatings multifunctional, (c) improving the environmental friendliness of strippable coatings and (d) improving the decontamination timeliness of strippable coatings.

The purpose of this research is to evaluate the adsorption kinetics, isotherms and thermodynamic parameters correlating the removal of 60Co2+ radioactive ions from waste water of the primary cooling circuit of the Nuclear Uranium Reactor (Nur / Draria-Algeirs) on Fe(III)-modified Algerian bentonite. Characterization was done using X-ray diffraction, SEM analysis, X-ray Photoelectron Spectroscopy and BET method. The pseudo-second-order kinetic model provided the best fit to experimental data. The two-parameter adsorption model of Langmuir and both of the three-parameter models Langmuir–Freundlich and Toth are the most appropriate for fitting experimental results. The adsorption process was found to be endothermic in nature.

The monazite phase, Ce0.9Ca0.1PO4, was prepared and explored as a matrix for the nuclear waste immobilization. Ce0.9Ca0.1PO4 and 20 wt.% simulated waste loaded Ce0.9Ca0.1PO4 were prepared using solution chemistry route and sintered at 1200 °C. Various characterizations such as XRD, Raman, SEM, TGA, UV–Vis and FTIR analysis were carried out. XRD analysis on bare monazite confirms the formation of single phase materials and simulated waste loaded monazite reveals the formation of cubic structured Pd and ZrP2O7 as minor phases present in the system. XPS results on Ce0.9Ca0.1PO4 revealed shift in Ce-3d peak position compared to the reported CePO4 confirming the influence of Ca2+. The UV–Vis result shows the charge compensation existing in the materials which is essential for the accommodation of higher valence cations. FTIR and Raman results of waste loaded Ce0.9Ca0.1PO4 showed peak shift and broadening due to the incorporation of various elements in the crystal lattice. Chemical durability studies on 20 wt.% simulated waste loaded Ce0.9Ca0.1PO4 showed Mo leaching which could be avoided by preparing monazite—IPG glass–ceramic composite. The results suggest that, Ce0.9Ca0.1PO4—IPG glass–ceramic composite can be a versatile matrix for the immobilization of high level radioactive waste.

A novel pretreatment method was used to ameliorate the treatment effect of iodine contained radioactive waste. The pretreatment sample (15 w%) held the highest amorphous index (0.50) at 600 °C, and iodine element was in a state of uniform distribution. The density of the sample could increase to 2.95 gcm3 by using the novel method, which was higher than other similar studies. The leaching result of iodine was as low as 1.53 × 106 gm2d1 after 7 days treatment, which indicated that the solidified body has good stability in a long run.

International audience; This study investigates novel Portland cement-based mortars, developed for hydrogen/tritium trapping and radioactive waste immobilization. They incorporate a γ-MnO2/Ag2O getter powder at 10% wt (i.e. 6.8 to 7.1% vol). Compared to former patented materials, no drying or heat treatment of the materials is needed.Complementarily to X Ray Diffractometry (XRD) and thermo-gravimetry analysis (TGA), 27Al and 29Si Magic Angle Spinning Nuclear Magnetic Resonance (MAS NMR) evidences that the structure of the Portland cement solids (i.e. the C-A-S-H) is not impacted by the presence of getter, even after several month curing. Yet, Scanning Electron Microscopy coupled to Energy Dispersive Spectroscopy (SEM+EDS) shows that calcium is significantly present at the surface of the getter grains; this could affect trapping efficiency.However, after gamma irradiation with cumulated doses of 491 or 997 kGy (i.e. 23 to 46 days at 900 Gy.h−1), all mortars made with getter have a hydrogen trapping efficiency of between 77 and 96%, when compared to the irradiation of mortars made with a non-trapping γ-MnO2 powder, or to the irradiation of pure water considered alone.This means that the developed mortars display excellent hydrogen trapping efficiency, without any impact on their solid structure.

When total volume control is applied to toxicity in nuclear waste management it becomes a limiting factor for the permittable total operating capacity of nuclear reactors. An alternative conceptual...

While evaluating the secondary impact of transport casks under the oblique drop tests based on the IAEA transport regulation, it is important to find the most severe condition defined with respect ...

Purpose: Consideration of the shortcomings of the Government Decision of the Russian Federation of 19 October 2012 No. 1069 on the classification of radioactive waste. Development of recommendations to change the Regulations of radiation safety of current and future generations with the radioactive waste disposal. Results: There was obtained the shortcomings of the Decree of the RF Government dated 19 October 2012, No. 1069 these deficiencies can lead to overexposure of present and future generations of people to unnecessary costs and mixed results in the classification of radioactive waste. Conclusions: the developed proposals for the government’s decision: 1. Completely rework the requirements for waste include in radioactive waste. 2. To exclude from the solution of waste containing tritium. 3. To exclude from the resolution of solid and liquid wastes generated from implementation of the non-use of nuclear energy for mining and processing mineral and organic raw materials with high content of natural radionuclides.

A safety assessment was performed on the second-phase disposal facility of the Gyeongju low – and intermediate-level radioactive waste repository using the RESRAD-OFFSITE code. On the basis of the ...

Solidification of radioactive waste containing strontium is a significant procedure in the nuclear waste treatment and disposal. In this research, structure evolution and leaching stability characteristics of aluminum phosphate binder matrix doped with different SrO contents are characterized. The results indicate that the change of SrO content leads to phase transformation of solidified samples. The significance of this research is that aluminum phosphate binder solidified system presents a high load capacity for SrO (SrO content of 50 wt%) as well as good leaching stability (normalized leaching rate 1.01 × 10−4 g m−2 d−1 after 42 days) for solidified sample.

For the safety assessement of deep geological repository for high-level radioactive waste, radionuclides transport behavior in rock systems is necessary to clairify and establish a suitable model. The advection–dispersion experiments(ADEx) was effectively designed to inadequate transport models through a calibration/validation process, and HTO, selenate (VI) and selenite (IV) migration behavior on crushed granite was studied by a high reliable and dynamic column device to obtailned retardation factor (R) and the dispersion coeffcient (D) by fitting experimental breakthrough curves (BTCs) in a various range of flow rates. Moreover, it showed the similar results between the BTCs of HTO and Se(VI) by fitting a equlibrium model due to no retardation effect. In fact, there was a relatively obvious sorption of Se(IV) in BTCs by fitting a retardation factor (R) value higher than 1. In addition, the two sites non-equilibrium model with a lowest root mean square error were applied to realize Se(IV) sorption mechanism on granite.

For industrial package (IP)-type transport containers for radioactive materials, a free drop test should be conducted under regulatory conditions. Owing to various uncertainties observed during the drop test, errors in drop angles inevitably occur. In IP-type metal transport containers in which the container directly impacts onto a rigid target without any shock absorbing materials, the error in the drop angle due to a slight misalignment makes a significant difference from the ideal drop. In particular, in a vertical drop, the error in the drop angle causes a strong secondary impact. In this paper, a numerical method is proposed to estimate the error in the drop angle occurring during the test. To determine this error, an optimization method accompanying a computational drop analysis is proposed, and a surrogate model is introduced to ensure calculation efficiency. Effectiveness of the proposed method is validated by performing the verification and comparison between the test and the analysis applied with the drop angle error.

A methodology for analyzing accidents with radioactive material release with EPZDose code was established. This code assesses doses and it is designed and developed by NTHU (National Tsing Hua University). To confirm the capacity of EPZDose, three postulated accident scenarios Taiwanese NPPs Chinshan (BWR/4) and Maanshan (PWR) are analyzed. All these scenarios are SBO (station blackout) transients because it is assumed that they result in a release of radioactive material. In this study, the source term data for EPZDose are taken from MELCOR or RASCAL calculations. In addition, calculated results of RASCAL code are compared with the results of EPZDose for these scenarios. The comparison show that the EPZDose predictions are consistent with the data of RASCAL. This indicates that the EPZDose has a respectable accuracy in the analysis of radioactive material release accidents.

Expensive permanent storage facilities with massive engineered structures are used traditionally to ensure safe temporary storage of solidified radioactive waste at the NPP sites. Such approach is dictated by the need to comply with the regulatory requirements for limiting the gamma background in the area adjacent to the storage facility. The costs involved in temporary storage of solidified RW can be optimized by using light hangar-type storage facilities. At the same time, the safety of storage, including radiation protection of the personnel, the public and the environment, is undoubtedly ensured through the use of special organizational and engineering solutions. The Novovoronezh NPP, a branch of JSC Concern Rosenergoatom, operates successfully light hangar-type facilities for temporary storage of solidified RW classified as medium-level waste in accordance with OSPORB-99/2009. In the process of operation, a methodology and a method for conditioning and temporary storage of solidified RW were developed to ensure the RW removal for final disposal with no extra process operations and unreasonable costs. A methodology has been developed to assess the radiation situation around storage facilities during temporary storage of RW, as well as a software package for predicting the radiation situation when deciding on the arrangement of the storage facility’s peripheral rows.

Nowadays, evaluation of amounts and distributions of radioactive waste is an important preparatory step in the process of nuclear reactor decommissioning. For tentative estimation of radioactive waste, a cell-based rigorous 2 step (R2S) method usually is used; however, a poor resolution caused by the averaged flux and spectrum in a cell is still a great challenge because of leading to underestimated or overestimated results. To overcome the poor resolution, several systems were introduced. Neither system, however, provides any function for evaluation of radioactive waste amount and distribution. Thus, it is additionally required to classify radioactive waste based on the results of activation calculation. In this study, the advanced R2S (AR2S) system was developed. To verify the performance of the system, its results for a verification problem were compared with those of the cell-based R2S method. The results showed good agreement, which is to say, within 2.0% relative error. Also, several characteristics of fine/coarse mesh were analyzed. To demonstrate the performance of the AR2S system, the radioactive waste from the Japan Power Demonstration Reactor (JPDR) was estimated, and the result indicated a high-resolution distribution. Therefore, it is expected that the AR2S system will prove useful for precise evaluation of radioactive waste.

Radioactive wastes are the byproduct of nuclear technology and industrial processes. Unlike any other waste stream, radioactive wastes are hazardous to living beings and the environment. Due to the harmful effect of radioactivity, radioactive wastes are governed and regulated by the government authority for the safety of the people and the environment. According to the waste generation and the intensity of hazardous content, countries classified their radioactive waste and implemented a management policy for each category. These strategies vary for different types of radioactive wastes and the actual radioactive contents. Mostly, at first radioactive wastes are collected and segregated into categories. Some wastes are stored in a temporary storage facility before going to the treatment phase of volume reduction, waste immunization and shielding. The majority of the waste volume after the treatment phase goes directly into long term geological disposal at ground level up to 100 m depth. Highly radioactive spent fuel and reprocessing wastes are currently stored in temporary storage facilities until a long-term disposal solution is built. This study aims to provide an overall review of the radioactive wastes and the management strategies.

The structural stability of a waste package is essential for containing radioactive waste for the long term in a repository. A silo-type disposal facility would require more severe verification for the structural integrity, because of radioactive waste packages staked with several tens of meters and overburdens of crushed rocks and shotcretes. In this study, structural safety was analyzed for a silo-type repository, located approximately 100 m below sea level in Gyeongju, Korea. Finite element simulation was performed to investigate the influence of the loads from the backfilling materials and waste package stacks on the mechanical stress of the disposed of wastes and containers. It was identified that the current design of the waste package and the compressive strength criterion for the solidified waste would not be enough to maintain structural stability. Therefore, an enhanced criterion for the compressive strength of the solidified waste and several reinforced structural designs for the disposal concrete container were proposed to prevent failure of the waste package based on the results of parametric studies.

An economic and mathematical model has been developed to determine the competitiveness of REMIX-A and -C uranium-plutonium fuel obtained by reprocessing spent fuel from thermal reactors. The presented model makes it possible to calculate the cost metrics of nuclear fuel components and the fuel component of the cost of NPP electricity as a function of the cost metrics of key process changeovers. It is shown that as the market quotations for natural uranium and separation work increase, REMIX fuel becomes more competitive and, more importantly, all the more significant, the lower the cost of reprocessing spent fuel. However, even with zero cost of reprocessing the gain from the introduction of REMIX fuel does not exceed 25%. At the same time, its use allows a 20-fold reduction in the mass of radioactive waste going to final disposal, 24% (REMIX-A) to 36% (REMIX-C) savings of natural uranium, and 8–18% of separation work (REMIX-A) as well as 18% (REMIX-A) to 40% (REMIX-C) reduction in the accumulation of depleted uranium hexafluoride.

Radioactive iodine (131I) released from nuclear power plants has been a critical environmental concern for workers. The effective trapping of radioactive iodine isotopes from the off-gas stream generated from nuclear facilities is an important issue in radioactive waste treatment systems evaluation. Numerous studies on retaining methyl iodide (CH3I131) by impregnated activated carbons under the high content of moisture have been extensively studied so far. But there have been no good results on how to remove methyl iodide at high humid conditions up to now. A new challenge is to introduce other promising impregnating chemical agents that are able to uptake enough radioactive methyl iodide under high humid conditions. In order to develop a good removal efficiency to control radioiodine gas generated from a high humid process, activated carbons (ACs) impregnated with triethylene diamine (TEDA) and qinuclidine (QUID) were prepared. In addition, the removal efficiencies of the activated carbons (ACs) under humid conditions up to 95% RH were evaluated by applying the standard method specified in ASTM-D3808. Quinuclidine impregnated activated carbon showed a much higher decontamination factor above 1,000, which is enough to meet the regulation index for the iodine filters in nuclear power plants (NPPs).

In most countries, the thermal criteria for the engineered barrier system (EBS) is set to below 100 °C due to the possible illitization in the buffer, which will likely be detrimental to the performance and safety of the repository. On the other hand, if the thermal criteria for the EBS increases, the disposal density and the cost-effectiveness for the high-level radioactive wastes will dramatically increase. Thus, fundamentals on the thermal behavior of the buffer under the elevated temperatures is of crucial importance. Yet, the behaviors at the elevated temperatures of the bentonite under groundwater-saturated conditions have not been reported to-date. Here, we have developed an in-situ synchrotron-based method for the thermal behavior study of the buffer under the elevated temperatures (25–250 °C), investigated d-spacings of the montmorillonite in the Korean bentonite (i.e., Ca-type) at dry and KURT (KAERI Underground Research Tunnel) groundwater-saturated conditions (KJ-ii-dry and KJ-ii-wet), and compared the behaviors with that of MX-80 (i.e., Na-type, MX-80-wet). The hydration states analyzed show tri-, bi-, and mono-hydrated at 25, 120, and 250 °C, respectively for KJ-ii-wet, whereas tri-, mono-, and de-hydrated at 25, 150, and 250 °C, respectively for MX-80-wet. The Korean bentonite starts losing the interlayered water at lower temperatures; however, holds them better at higher temperatures as compared with MX-80.

The impact of disposal of mixed oxide (MOX) fuel from light water reactors (LWRs) has been considered from the viewpoint of the thermal properties and the number of the vitrified waste for various ...

How does naming an object affect the way it is or could be managed? This paper examines and compares classification systems for radioactive waste applied by the International Atomic Energy Agency (...

Entities generating low and intermediate-level radioactive waste, including the nuclear power plant (KHNP), Korea Atomic Energy Research Institute, and KEPCO Nuclear Fuel Company, are currently managing the established waste classification criteria and waste tracking system (WTS). However, management thereof remains inefficient; thus, an integrated approach is required. This study examined the establishment of an integrated WTS in Korea based on an analysis of domestic and overseas systems to consolidate the classification criteria of the different entities. Furthermore, codes were assigned to locations where waste is processed and stored to offer scenarios based on movements between these locations.

The radionuclides are extracted from the minerals which make up the rocks. This work takes natural magmatic granite as the matrix and CeO2 as the simulated radioactive waste to explore the maximum solid solubility. After analyzing the phase transition of pure granite with varied temperatures, the solidification was carried out at 1300 ℃. The phase, microstructure, element distribution, hardness and chemical stability of the solidified bodies were studied. The results showed that the maximum solid solubility was up to 8 wt.%. The normal leaching rate of Ce4+ after the first 3 d was about 3.89 × 10–7 g·m−2·d−1.

90Sr radiochemical method using nitric acid for Sr and Ca separation was optimized using design of experiments methodology. Nitric acid concentration and total volume of solution were determined as influencing factors applying Plackett–Burman design. These parameters were then optimized using Box–Behnken experimental design. An optimal nitric acid concentration of 63% and a total solution volume of 26 mL were found to achieve a Sr recovery of 93% with only 5% of Ca for a single precipitation. The developed procedure was applied on real radwaste samples: effluent, sludge and concrete. By comparison of the recovery and the 90Sr activity measurement with the results of the in-house standard method based on fuming nitric acid, the optimized method with concentrated nitric acid was validated. The fuming nitric acid can now be eliminated for 90Sr radioactive waste characterization.

This study assessed the influence of geological depth and formation on activity concentrations of 232Th, 226Ra and 40K in soil and groundwater, and radio-chemical health risks. Preference ranking organisation method for enrichment evaluation and geometrical analysis for interactive aid indicated significant correlation between 232Th and 40K, thus indicating potential similar primordial origin. Deepest depth rocks at 145 m and 148 m constituted of dark coloured silicified schist exhibited minimal activity with potential hydro-geological interactions with groundwater. Age-dependent annual effective ingestion dose for 226Ra and 232Th are higher than the UNSCEAR reference dose, thus the groundwater is unsuitable for long-term consumption.

The content of radionuclide uranium in the water, soil and rice around the decommissioned uranium mine was analyzed, and the radioactive pollution, distribution characteristics, migration pattern and health risks were evaluated. Its radioactive contamination may have potentially harmful biological effects on the ecosystem and food chain. This paper pays more attention to the pollution of radionuclide 238U, 235U around the tailings dam, and makes a comprehensive comparative analysis of the pollution of water, soil and rice at 500 m, 1000 m and even 60 km around the dam. The single factor index method was mainly used to evaluate the pollution degree, and it was found that the water, soil and rice in the area around the uranium field were all polluted to different degrees. Radionuclide uranium pollution is mainly caused by uranium mining, quarry discharge and ecological cycle in the early stage. According to the characteristics of the uranium content in the study area, it is recommended to use the farmland around the tailing dam as a test site to study the radioactivity of different crops, improve wastewater treatment technology, and prevent crops with excessive levels of radioactive elements from entering the market.

Solid waste samples consisting of shredded cellulose, coated with either mesoparticles of metallic salts or dried metal nitrate (lutetium, ytterbium, or depleted uranium) solutions, were generated ...

The results of the dismantling of the active drain system and the decontamination of technological enclosures of the loop installations of the MR reactor, which is being decommissioned at the NRC Kurchatov Institute, are presented. The radioactive contamination of the pipelines of the system and the soil had a complex nuclide composition. Methods for assessing the volume of radioactive waste being removed from the premises are discussed, and criteria for the radioactive contamination of individual building structures and the soil to reach the conditions of the final state of the object as determined by the plan of operations are presented. Remote-controlled machines and dust suppression technologies are used to reduce the effect of the radiation fields on the workers, to protect the workers as they perform their jobs, and to increase the power availability of per worker in the process of dismantling and handling radioactive waste. As a result, criteria for the radiation situation corresponding to the working conditions of level-B personnel were reached. This corresponds to the parameters of the final state of a facility as prescribed by the plan for decommissioning the MP and RFT reactors.

The current study relates to RADAR (RAdio Detection and Ranging) application for level measurement of vitrified radioactive liquid nuclear waste. The vitrification of radioactive liquid waste is carried out in special equipment called ‘Melters’. The study is directed towards the design and frequency modulation used in the level measurement of vitrified waste. More specifically, the RADAR design and frequency used for level measurement in a melter. This level measurement technique can also be used for dynamic vitrification process and can be used to measure the level variations without using any external medium/material and using only electromagnetic waves. Also, this technique is durable and accurate even under the high radioactive environment present inside the melter.

The analysis of 41Ca in concrete generated from the nuclear facilities decommissioning is critical for ensuring the safe management of radioactive waste. An analytical method for the determination of 41Ca in concrete is described. 41Ca is a neutron-activated long radionuclide, and hence, for accurate analysis, it is necessary to completely extract Ca from the concrete sample where it exists as the predominant element. The decomposition methods employed were the acid leaching, microwave digestion, and alkali fusion. A comparison of the results indicated that the alkali fusion is the most suitable way for the separation of Ca from the concrete sample. Several processes of hydroxide and carbonate precipitation were employed to separate 41Ca from interferences. The method relies on the differences in the solubility of the generated products. The behavior of Ca and the interfering elements such as Fe, Ni, Co, Eu, Ba, and Sr is examined at each separation step. The purified 41Ca was measured by a liquid scintillation counter, and the quench curve and counting efficiency were determined by using a certified reference material of known 41Ca activity. The recoveries in this study ranged from 56 to 68%, and the minimum detectable activity was 50 mBq g−1 with 0.5 g of concrete sample.

Evaluating the quantitative damage to rocks through acoustic emission (AE) has become a research focus. Most studies mainly used one or two AE parameters to evaluate the degree of damage, but several AE parameters have been rarely used. In this study, several data-driven models were employed to reflect the combined features of AE parameters. Through uniaxial compression tests, we obtained mechanical and AE-signal data for five granite specimens. The maximum amplitude, hits, counts, rise time, absolute energy, and initiation frequency expressed as the cumulative value were selected as input parameters. The result showed that gradient boosting (GB) was the best model among the support vector regression methods. When GB was applied to the testing data, the root-mean-square error and R between the predicted and actual values were 0.96 and 0.077, respectively. A parameter analysis was performed to capture the parameter significance. The result showed that cumulative absolute energy was the main parameter for damage prediction. Thus, AE has practical applicability in predicting rock damage without conducting mechanical tests. Based on the results, this study will be useful for monitoring the near-field rock mass of nuclear waste repository.