Transport Of Radioactive Waste Research Articles

Evidence of nuclide migration from high-level radioactive waste glass via geogas

Characteristics of high-level radioactive waste transport in glass as gases were studied. A transmission electron microscope was used to observe the morphology of the substance after penetrating the high-level radioactive waste glass and the bentonite soil column, and the composition of the substance was analyzed using electron probe energy spectroscopy. The results show that the radionuclides in the high-level radioactive waste glass were transported out of the glass, the bentonite was penetrated by the geogas, and the transported substance occurred as nanometer-size particles and particle aggregates.

Development of novel composite materials containing rice bran wax and waste polyethylene for neutron shielding applications

Neutrons are utilized in diverse fields, such as physics, biology, chemistry, geology, agriculture, medicine, mining, military, and space research. Neutrons interacting with human tissues and cells can cause harm to soft tissues, such as the cornea, and alter cell functionality, leading to the cessation of cell proliferation and resulting in cell/tisssue damage or cancers. Effective protective materials are necessary to enable the safe use of neutrons in these applications. In this study, several samples of novel types of composites were prepared by blending rice bran wax (RBX) with recycled polyethylene (rPE) at five different ratios ranging from 0% to 50%. The RBX in the composites enhances their neutron absorption capacities and while the rPE in the composites reduce their costs. Several essential fast neutron shielding parameters, such as half-value layer, effective removal cross-section, mean free path, and neutron transmission ratio, were calculated with Monte Carlo simulation GEANT4 code. Absorbed dose amount by the samples were determined using a241Am-9Be neutron source. The theoretical and experimental values of the shielding capacities of the novel composites were compared with those of paraffin and traditional concrete. Compared with the reference samples, the new RBX composite samples had higher neutron radiation absorption capacities. The functional groups of the RBX and rPE were identified using Fourier transform infrared (FTIR) spectrophotometry (Bruker Vertex 70v FTIR spectrometer with the attenuated total reflection (ATR) module) in the spectral area of 400–4000 cm−1 with 32 scans. The thermal chareteristics of the samples were determined with the differential scanning calorimeter Mettler-Toledo/DSC1/700. The developed composite materials are useable as protective material against neutron radiation in radiation settings, such as radiotherapy rooms, within and outside nuclear power plants, as well as in settings involving the storage or transportation of radioactive waste.

Effect of groove angle on the residual stress and deformation of nuclear waste container cylinder body welded by plasma arc welding

AbstractAs an important barrier for high‐level radioactive waste transportation and deep burial disposal, the welding quality of nuclear waste container is particularly important. At present, nuclear waste container is usually manufactured by plasma arc welding. Therefore, in order to manufacture nuclear waste container parts with high quality, based on the finite element simulation software simufact welding, this paper uses a combination of numerical simulation and experimental verification to simulate the temperature field, stress field and deformation of the longitudinal weld of 316L nuclear waste container cylinder body by selecting different welding groove angles. The test results show that the groove angle of the V‐groove increases, the residual stress distribution and size increase, and the deformation increases. The stress value of the I‐shaped groove is between the V‐shaped groove angle of 30° and 45°, but the deformation is the smallest. In practical engineering applications, residual stress and deformation need to be considered comprehensive. Reasonable groove angle plays an important role in controlling residual stress and deformation after welding.

Package Performance Study: A Historical Perspective and Planning for a Path Forward

The topic of spent nuclear fuel (SNF) transportation often draws public attention and concerns regarding safety, and with large-scale shipping campaigns expected in the future, public interest is anticipated to substantially increase. In 1999, the U.S. Nuclear Regulatory Commission (NRC) began an initiative called the Package Performance Study (PPS) that initially focused on SNF transportation cask responses to severe transportation accidents. A public participation approach was to be used to help guide the scope and parameters of the study, which was also to consider using full-scale physical testing of a SNF transportation cask, where appropriate. In 2010, the NRC decided not to go forward with physical testing as proposed. In 2006, the National Academy of Sciences (NAS) conducted a study on SNF and high-level radioactive waste transportation in the United States. The NAS study recommended that full-scale cask testing, as well as other accepted methodologies, continue to be used as part of the package performance evaluation. In 2012, the Blue Ribbon Commission on America’s Nuclear Future (BRC) recommended conducting the PPS with a full-scale rail transportation cask for the purpose of building public trust and confidence in the safety of SNF transport. In 2021, the U.S. Department of Energy (DOE) restarted a consent-based siting process, focusing on how to site federal facilities for the interim consolidated storage of SNF using a consent-based approach. Large-scale SNF transportation would be necessary for any such interim storage facility. As the planning activities begin for future large-scale shipping campaigns, the DOE is considering options for conducting a DOE-led PPS to help build public trust and confidence in the transportation of SNF. Therefore, the DOE is gathering information related to the previous NRC-led PPS efforts, as well as applicable international experience with the testing of SNF packages. This paper discusses previous NRC efforts, as well as the current status of a DOE-led PPS, which is in the early planning stages.

Li2B4O7–Bi2O3–ZrO4–CaWO4 glass system for neutron protection in neutron applications

In this study, new glass samples with a chemical composition of 40Li2B4O7+xBi2O3+(55–x)ZrO4+5CaWO4 (x = 10, 20, 30, 40, and 50 wt%) were fabricated using the melt quenching-annealing technique. Important neutron shielding parameters, such as the effective neutron removal cross-section, half-value layer, and neutron transmission factor, of all the new glass samples were calculated using the GEANT4 Monte Carlo simulation code. In addition, neutron dose measurements were performed using an 241Am–Be fast neutron source with an activity of 74 GBq and an average neutron energy of 4.5 MeV and a BF3 gaseous neutron detector. The results were compared with those of conventional shielding materials to determine the neutron absorption capabilities of the new glass materials. All the new glass samples were found to have excellent neutron shielding abilities. In particular, the GB5 sample had a high absorption capacity compared to the other samples. These new glass samples may be used as shielding materials in various radiation applications, such as generation of nuclear energy, radiation medicine, and the transport and storage of radioactive wastes.

Деякі аспекти використання сталей під час позареакторного зберігання та транспортування відпрацьованого ядерного палива. (Огляд)

In the world, the nuclear industry is considered an ideal option for an environmentally friendly source of energy generation, based on the separation of the radioactive nuclide of the chemical element 235U. When generating electricity from nuclear power plants, the main disadvantage is the generation of radioactive waste. After spent fuel is removed from a nuclear reactor, it is placed in special containers containing absorption and retention elements. It is obvious that the storage and transportation of spent nuclear fuel plays an important role in the development and overall safety of the nuclear industry, and high-quality neutron-absorbing materials are the basis for the successful manufacture of reliable structures and containers. The development of modern neutron absorption materials ensures reliability, safety, storage duration, as well as reducing the cost of logistics operations associated with the transportation of radioactive waste. Based on the results of the analysis of scientific and technical sources, modern ideas about materials for the manufacture of containers, which are used for storage and transportation of nuclear fuel waste, are summarized. The advantages and disadvantages of the main materials used in the disposal of spent nuclear fuel are considered. It has been established that, from an economic point of view, the most rational material for storing and transporting radioactive waste is ferritic stainless steel with a high boron content, which, in addition to absorption abilities, is characterized by other properties required for this type of product. Since boron has a low ability to dissolve in both austenite and ferrite, this leads to the formation of borides of various types, which contribute to the embrittlement of the structural material. At the same time, the distribution of borides over the volume of a metal product is determined not only by a set of properties, but also by the protective properties of the material, as well as the manufacturability of the final product from it. Despite the fact that boron-containing steels of the ferritic class have been known for a long time, issues related to the processes of structure formation, increasing technological and operational properties still remain insufficiently studied and controversial. There is also no information on the influence of heat treatment on the formation of the structure and mechanical properties of finished products made of ferritic stainless steels. Successful solution of these scientific and technical problems will ensure the production of modern high-quality neutron-absorbing containers for off-reactor storage and transportation of spent nuclear fuel.

Determination and Verification of GISSMO Fracture Properties of Bolts Used in Radioactive Waste Transport Containers.

Transport containers for radioactive materials should withstand drop tests according to the regulations. In order to prevent a loss or dispersal of the internal radioactive materials in the drop tests, a tightening of the lid of the transport container should be maintained. The opening of the lid, due to the drop impact, might cause the dispersion of internal contents or a loss of shielding performance. Thus, it is crucial to predict damage to the fastening bolt and its fracture. In this study, the damage parameters of the fastening bolt were acquired, and its fracture was predicted using the generalized incremental stress state-dependent damage model (GISSMO), a phenomenological damage model. Since the dedicated transport container is large and heavy, various jigs that can simulate the fall of the container were designed, and the accuracy of fracture prediction was verified. Digital image correlation (DIC) was introduced for the accurate measurement of the displacement, and load–displacement data for tensile, shear, and combined loads were successfully acquired. Finally, the load–displacement curve of the finite element analysis (FEA) with GISSMO until the point of the bolt fracture was compared with the curve obtained from the experiment, where a good agreement was observed.

Fabrication of novel neutron shielding materials: Polypropylene composites containing colemanite, tincal and ulexite

Exposure to neutron radiation can cause serious health problems, especially in nuclear power plants and their surroundings, high-flying aircrafts, and healthcare facilities. Conventionally, many neutron-shielding materials are used for protection, but their mechanical and thermal properties limit the use of these materials. Low density, low cost, and flexible neutron shields are becoming increasingly important for these uses, in particular where the weight factor, such as in the aerospace sector, is highly important. For this reason, polypropylene (PP) based composite materials were prepared and their neutron shielding features, such as effective removal cross-sections, half value layer, and mean free path were determined using GEANT4 code. The absorbed dose-measuring experiments were carried out using a241Am-9Be neutron source. Boron minerals such as colemanite, tincal, and ulexite were used in the production of the materials to reduce costs and increase neutron absorption capacity. All the results were compared with the shielding abilities of paraffin and conventional concrete. It was determined that these composites absorb neutron radiation much better than the referenced materials. Colemanite was characterized by X-ray diffraction (XRD) while PP was characterized by Fourier transform infrared (FTIR) spectroscopy and differential scanning calorimetry (DSC). These composites prove to be valuable shielding materials for protection against neutron radiation in nuclear medicine, storage or transportation of radioactive waste, and other nuclear applications.

Experimental and Monte Carlo simulation study on potential new composite materials to moderate neutron-gamma radiation

In this study, 12 different concentrations of shielding materials were developed and produced. They were covered with high temperature resistant (1500 °C) sodium silicate sealing paste. Epoxy resin was produced by adding different percentages of additive materials such as chromium oxide (Cr2O3), lithium (LiF), and nickel oxide (NiO). The GEANT4 and FLUKA codes of the Monte Carlo simulation toolkit were used to determine the mixing ratios. The total macroscopic cross-sections, effective removal cross-sections, mean free path, half value layer, and transmission neutron number were determined for fast neutron radiation using GEANT4 and FLUKA simulation codes. The mass attenuation coefficient, the effective atomic number and half-value layer (HVL) of the samples were calculated using Phy-X/PSD software. The absorbed dose was measured. In this study, an 241Am–Be neutron source with 74 GBq activity and average neutron energy of approximately 4.5 MeV and a BF3 gas detector were used. Both simulation and experimental measurements were compared with paraffin and conventional concrete. The new composite shielding material absorbed radiation much better than the reference materials. This new radiation shielding composite material can be used in nuclear medicine, transport and storage of radioactive waste, nuclear power plants, and as a shielding material for neutron and gamma radiation.

Development and production of metal oxide doped glasses for gamma ray and fast neutron shielding

In this study, new glasses were designed and produced that did not consist of toxic lead oxide (PbO). Unlike conventional glass with powdered silicon oxide (SiO2) added [cobalt oxide (CoO), cadmium tungsten oxide (CdWO4), bismuth oxide (Bi2O3), chromium oxide (Cr2O3), and boron oxide (B2O3)], new glass specimens were fabricated via melt-quenching. The total macroscopic cross-sections, mean free path, and transmission number were calculated using GEANT4 Monte Carlo code. The equivalent dose rates were measured for 4.5 MeV fast neutrons. Gamma attenuation experiments were conducted at several energies in a range of 53–383 keV. The experimental and theoretical results were compared with those of construction concrete, paraffin, glasses containing lead, borosilicate, normal window, and soda glasses. The new glass minimized both gamma and neutron radiation leaks that may occur in various fields such as radiation treatment, nuclear power plants, radioactive waste transport and storage, and space and laboratory research.

ECOLOGICAL DETERMINANTS IN THE FORMATION OF THE POPULATION'S ATTITUDE TO THE DEVELOPMENT OF NUCLEAR ENERGY.

Evaluation of the key environmental determinants that shape the response of the social environment tothe activities of nuclear power plants. The poll was conducted in the observation zone (OZ) of the South-Ukrainian NPP (SUNPP).A repetition-free probabilistic selection was used in the population survey. The sample of respondents from the population of the SUNPP OZ was 322 people with the sampling error of 5.4 % and confidence interval of 95.0%. An ordinal 5-point measurement scale was used to assess the quality of life and anxiety, and nominal measurement scaleswere used for other issues. All the age groups of population from 20 to 65 years were taken into account. The sociohygienic, sociological, statistical and mathematical research methods were applied in the process of work (relativevalues, averages, reliability of paired differences according to Student's t test). The level of perception of radiation risk by the population of the NPP OZ was the lowest among respondentsof satellite cities of Rivne NPP (RNPP) and SUNPP compared to the population of rural settlements and cities of theOZ. Storage and shipment of radioactive waste are the determinant among the factors that form an environmentalthreat to the population of OZ as a result of NPP activities. Social factors that determine the assessment of environmental safety of NPP activities are employment at the NPP, place of residence, gender, age, level of education of thepopulation, etc. The «Ch-image» of the ecological catastrophe formed in the population of Ukraine as a result of theChNPP accident , regardless of the place of residence on the territory of Ukraine, influences behavior, formation ofplans and assessments, especially in population living near the ionizing radiation sources. The current legislation governing the relationship between functioning NPPs and the communities inwhich they are located needs to be improved, guided by the basic principle of environmental public health - socialjustice, i.e. the human right to a healthy environment. It means that the settlements of the OZ (communities)receive funds to compensate for risk living in compliance with certain safety rules to maintain health. Legal regulation and constant awareness of population are the main directions for the formation of adequate assessments.

Integral transform analysis of radionuclide transport in variably saturated media using a physical non-equilibrium model: application to solid waste leaching at a uranium mining installation.

The Generalized Integral Transform Technique (GITT) was employed to simulate transient one-dimensional flow in variably saturated porous media, as well as radioactive waste transport within different layers (a solid waste pile, nearby soil, and a granular aquifer) towards the edge of a uranium mining installation under institutional control. Computational codes, written using the Mathematica software system, were implemented and tested for solution of the coupled advection-dispersion equations for an arbitrary number of daughter products within a radioactive chain migrating in saturated and unsaturated soil layers. The computer simulations were verified in great detail against results obtained using the built-in routine NDSolve of the Mathematica platform and the HYDRUS-1D software system. The present work reports the main results for 238U chain radionuclide transport using data extracted from a safety assessment of solid waste repositories at a uranium mining and milling installation in Caetité, state of Bahia, Brazil, operated by INB (Indústrias Nucleares do Brasil). Concentration evolutions of the various radionuclides obtained with the simulations were analyzed for five different cases to explore variations in the infiltration and recharge rates, the effect of assuming physical equilibrium or non-equilibrium transport conditions, and of different initial concentrations of some of the radionuclides.

Сравнительный анализ стоимости операций по обращению с РАО на российском и международном рынках

The article provides an overview of costs associated with radioactive waste management operations in different countries in the following areas: treatment, containerization, storage, transportation, and disposal of radioactive waste. All data are brought to a comparable form using various economic and mathematical methods in order to make a correct comparative analysis. In conclusion, the results of cost comparison across countries are presented, the possible reasons for their deviations are discussed.

Epoksi Bazlı Metal Ve Metal Oksit Katkılı Yeni Kompozit Nötron Ve Gamma Radyasyon Moderatör Malzemesi

In this study, five different composite shield materials were developed and produced. Epoxy resin based different proportions, materials such as [chromium oxide (Cr2O3), lithium florür (LiF), nickel oxide (NiO), bismuth oxide (Bi2O3), manganese oxide (MnO), copper oxide (CuO), titanium oxide (TiO2), cobalt oxide (CoO), gadolinium oxide (Gd2O3), selit (CaWO4), iron oxide (Fe2O3), lead oxide (PbO)] were used on production. The GEANT4 code of the Monte Carlo simulation program was used in determining mixing ratios. The total macroscopic cross section, mean free path and transmission neutron number were determined for fast neutron radiation by using GEANT4 simulation code. The mass attenuation coefficient and half-value layer (HVL) were calculated for gamma radiation by using WinXCom software. Experimental absorbed dose measurement was carried out and in these measurements 241Am-Be neutron source with 74 GBq activity which average neutron energy is approximately 4.5 MeV and BF3 gas detector were used. Both simulation and experimental measurements were compared with paraffin, conventional concrete. It was found that the new composite shielding material absorbed radiation much better than these reference materials. It has been suggested that this new radiation shielding composite material can be used in areas such as nuclear medicine, transport and storage of radioactive waste, nuclear power plants and as shield material for both neutron and gamma radiation.

Risk Assessment of Accidents During the Transportation of Liquid Radioactive Waste in Multimodal Transport

To enhance the safety of the transport of dangerous goods in tanks, the latter should improve the mechanical (physical) protection to prevent accidental damage to the boilers, such as objects containing dangerous goods directly. Damage to the boiler will almost inevitably leads to such consequences as decompression, the leakage of liquid radioactive waste and radiation pollution of the surrounding area. This article discusses scenarios accidents when overloading with marine railway transport; the explosion of gas cylinders on board the vessel. Calculation formulas are obtained to determine the speed at which the shards is provided by probabilistic the breakout a certain thickness of the tank walls. Received probability distribution function for the specified parameters shards and material barriers, which was determined by conducting a full-scale experiment. The results of theoretical studies, experiments used to improve designs of tanks carrying LRW and recommendations for improving the security of multimodal transport.

Radiological safety assessment for incident-free transportation of radioactive decontamination waste after the Fukushima nuclear power plant accident

Radioactive decontamination waste generated by remediation following the Fukushima nuclear power plant accident was recently transported from its temporary storage sites to interim storage facilities as a pilot test. A transportation plan for the radioactive decontamination waste will be developed based on this pilot transportation experience and radiological safety assessment of pilot transportation. The objective of this study was to assess radiation doses to the public and crew workers released during incident-free pilot transportation. External dose rates around a transportation vehicle were calculated by using the Monte Carlo N-Particle code. Collective doses and maximally exposed individual doses to the public and effective doses to crew workers were calculated by using INTERTRAN. Two transportation routes, Asakawa-machi to Okuma-machi and Iwaki-shi to Okuma-machi, were considered. The maximum radioactivity concentration in the decontamination waste was calculated to be 660 kBqkg?1 which meets the value laid down by the guidelines of Japan's Ministry of the Environment. The collective doses to the public per shipment were 1.9 10?3 person-mSv for the Asakawa route and 2.2 10?4 person-mSv for the Iwaki route. Maximally exposed individual doses to the public were 9.6 10?7 mSv for the Asakawa route and 2.7 10?5 mSv for the Iwaki route. The total effective doses to crew workers were 0.27 mSv for the Asakawa route assuming five shipments per worker and 1.07 mSv for the Iwaki route assuming 45 shipments per worker. The radiation dose levels to the public and workers evaluated in this study were much lower than the annual dose limits for the general public and radiation workers. These study results can be used to develop transportation plans and guidelines for decontamination waste transportation.

Preliminary Assessment of Radiological Impact on the Domestic Railroad Transport of High Level Radioactive Waste

국내 원자력발전소의 사용후핵연료는 소내저장시설에 보관되어 있으나 저장시설의 용량 확장이 어려움이 있으며, 연구기관 의 연구로에서 발생하는 고준위 방사성폐기물도 자체 보관중이나 영구적으로 저장할 수 없다. 또한 원전의 해체 시에도 고 준위 방사성폐기물이 발생할 것으로 예상된다. 이에 따라, 본 연구에서는 현재 개발된 사용후핵연료 운반용기를 사용하여 고준위 방사성폐기물을 가상의 관리시설로 철도를 통하여 운반하는 경우에 대하여 작업자 및 운반경로 주변 일반인의 예상 피폭선량을 평가하였으며, 그 결과를 국내 법적기준치와 비교하였다. 또한, 고준위 방사성폐기물의 상하차 작업 시 작업자 와 운반용기 간 거리와 운반사고 시 방사성핵종의 누출율의 변화에 따른 피폭선량의 변화에 따른 피폭선량 추이와 운반에 사용되는 열차의 구성에 따른 운반작업자의 피폭선량 변화를 분석하였다. 본 연구에서 설정한 모든 조건에서의 예상피폭선 량은 국내 법적제한치 이하임을 확인하였다.

Improvement of Radioactive Waste Physical Protection as a Step to Maintain Nuclear Security

The paper considers the procedure of ensuring physical protection of radioactive materials, in particular establishing levels of physical protection and its impact on nuclear security in Ukraine. There is a brief overview of the nature and types of radioactive waste in Ukraine, types of storage facilities, other radioactive waste management facilities, locations of radioactive waste in Ukraine, in particular the Chornobyl exclusion zone, causes of threats and risks. Efficient prevention of these threats requires the state system for physical protection of facilities and installations. Based on experience, paper authors, who review physical protection in construction and modernization of physical protection systems of radioactive waste management facilities, consider the state of physical protection of storage facilities and other installations, their modernization and unsolved problems. The paper also considers the physical protection issue in radioactive waste transport and provides conclusions on the improvement of the regulatory and legal framework of nuclear security in Ukraine.

Between public and private responsibility: The case of inter-state transport of radioactive waste

Between public and private responsibility: The case of inter-state transport of radioactive waste

Assessment of uncertainty in discrete fracture network modeling using probabilistic distribution method.

There have been widespread concerns about solute transport problems in fractured media, e.g. the disposal of high-level radioactive waste in geological fractured rocks. Numerical simulation of particle tracking is gradually being employed to address these issues. Traditional predictions of radioactive waste transport using discrete fracture network (DFN) models often consider one particular realization of the fracture distribution based on fracture statistic features. This significantly underestimates the uncertainty of the risk of radioactive waste deposit evaluation. To adequately assess the uncertainty during the DFN modeling in a potential site for the disposal of high-level radioactive waste, this paper utilized the probabilistic distribution method (PDM). The method was applied to evaluate the risk of nuclear waste deposit in Beishan, China. Moreover, the impact of the number of realizations on the simulation results was analyzed. In particular, the differences between the modeling results of one realization and multiple realizations were demonstrated. Probabilistic distributions of 20 realizations at different times were also obtained. The results showed that the employed PDM can be used to describe the ranges of the contaminant particle transport. The high-possibility contaminated areas near the release point were more concentrated than the farther areas after 5E6 days, which was 25,400 m2.