Release Of Radionuclides Research Articles

Leaching model of radionuclides in metal-organic framework particles

Metal-organic frameworks (MOFs) have been used to sequester radionuclides and seal them inside of porous scaffolds using postsynthetic modification procedures. Experiments show that certain Zr-MOF with different capping linkers significantly affects the radionuclide release kinetics. In this work, we developed a leaching model of radionuclides in Zr-MOF particles. The model assumes that uranyl species occupy two energetically favored sites: the metal node and the MOF pores. For a given overall concentration of uranyl species, the partitions of uranyl species at the metal nodes and within the pores are determined by their chemical potentials. The model also considers the effect of particle surface and concentration on chemical potentials and diffusivity. The effects of spatial and structural dependent chemical potentials and diffusivity as well as particle sizes on leaching kinetics are investigated with the model. Predicted and measured uranyl leaching kinetics in Zr-MOF particles under batch experiments are compared. The results demonstrate the model’s capability for exploring the mechanisms of leaching and provide guidance for material design.

Study on combustible gas generation and radionuclide release during underwater handling of the AM reactor spent fuel

Introduction. The paper addresses studies on the accumulation of combustible gases during underwater handling simulations for the leaky spent nuclear fuel from the AM reactor. Two fuel compositions were studied- uranium-molybdenum dispersed in magnesium and uranium carbide dispersed in calcium. Methods. The 137Cs release rate was measured during underwater storage of the uranium-molybdenum fuel. The kinetics of hydrogen release for both fuels and methane release for the carbide SNF were obtained. The kinetics approximate most with exponential dependences that formally correspond to first-order chemical reactions. A contribution of radiolytic hydrogen to the gases generated during the experiments was estimated. It was demonstrated that the determining source of the gases is the chemical interaction between the spent fuel and the water. The experiment with the uranium-molybdenum fuel demonstrated a pronounced passivation effect of the chemical processes on the fuel surface due to insoluble corrosion products. For the carbide SNF, an incubation period of about 20 hours was observed followed by an intensive release of hydrogen and methane. Results. The obtained results were subject to a comparative analysis against publications on the behavior of the fuel components in water. Conclusion. The findings can be applied to justify fire and explosion safety of underwater handling techniques for the damaged spent nuclear fuel with the considered fuel compositions (the spent fuel from reactors AM, AMB, EGP-6, etc.), e.g., to justify underwater preparations of the AMB spent fuel for reprocessing.

PRELIMINARY DEVELOPMENT OF RADIONUCLIDES RELEASE OF INDIVIDUAL DOSE CODE PROGRAM FOR RADIATION MONITORING PURPOSES

Environmental radiation monitoring is one of the important efforts in protecting society and the environment from radiation hazards, both natural and artificial. The presence of three nuclear research reactors and plans to build a nuclear power plant reactor prompted Indonesia to prepare a radiation monitoring system for safety and security (SPRKK). The goal of the study is to provide an appropriate method for developing radiation monitoring system to support the development of nuclear power plant in the near future. For this preliminary study, the author developed a code program using Gaussian distribution model approach for predicting radionuclide release and individual dose acceptancy by human being within 16 wind directions sectors and up to 50 km distance. The model includes estimation of source term from the nuclear installation, release of radionuclides source into air following Gaussian diffusion model, some of the release deposit to the land and entering human being through inhalation, direct external exposure, and resuspension, and predicted its accepted individual dose. This model has been widely used in various code program such as SimPact and PC-Cosyma. For this study, the model will be validated using SimPact code program. The model has been successfully developed with less than 5% deviation. Further study will be done by evaluating the model with real measuring data from research reactor installation and prepare for interfacing with real time radiation data acquisition and monitoring as part of radiation monitoring system during normal and accident condition.

Radiological Risk Assessment of Vehicle Transport Accidents Associated with Consumer Products Containing Naturally Occurring Radioactive Materials

Natural and artificial ionizing radiation can be harmful to human health when they come into contact with people and the environment. Transport of naturally occurring radioactive materials (NORMs) and consumer products containing NORM in the public domain is inevitable owing to their potential applications. This study evaluates the dose and risk to the public from transport accidents of NORM and consumer products. Radiological and physical data were obtained from previous literature. The median and maximum values of radioactivity concentration were applied to consumer products and NORM data, which serve as an input. An external dose rate at 1 m from a transported shipment was calculated using Microshield® Pro version 12.11 code, which serves as input to RADTRAN 6 code. Based on developed transport accident scenarios, a RADTRAN 6 code was used to estimate collective dose and risk. The sensitivity analysis was conducted by considering the variation of release, aerosol, and respirable fractions of radionuclides at 0.1%, 1%, 10%, and 100% from the transported shipment during an accident, respectively. The results of dose and risk to the general public because of the damage of the shipment container following a fire accident are below the annual regulatory limits of 1 man-Sv recommended by IAEA transport regulation of 2018. The sensitivity results of all NORMs and associated consumer products are also below the regulatory limits. Therefore, radiological safety can be ensured in the event of a transport accident involving the transit of NORM and consumer products containing NORM.

An Alternative Conceptual Model for the Spent Nuclear Fuel–Water Interaction in Deep Geologic Disposal Conditions

For the long-term safety assessment of direct disposal of spent nuclear fuel in deep geologic repositories, knowledge on the radionuclide release rate from the UO2 matrix is essential. This work provides a conceptual model to explain the results of leaching experiments involving used nuclear fuel or simulant materials in confirmed reducing conditions. Key elements of this model are: direct effect of radiation from radiolytic species (including defects and excited states) in the solid and in the first water layers in contact with its surface; and excess H2 may be produced due to processes occurring at the surface of the spent fuel and in confined water volumes, which may also play a role in keeping the spent fuel surface in a reduced state. The implication is that the fractional radionuclide release rate used in most long-term safety assessments (10−7 year−1) is over estimated because it assumes that there is net UO2 oxidation caused by radiolysis, in contrast with the alternative conceptual model presented here. Furthermore, conventional water radiolysis models and radiation chemical yields published in the literature are not directly applicable to a heterogeneous system such as the spent fuel–water interface. Suggestions are provided for future work to develop more reliable models for the long-term safety assessment of spent nuclear fuel disposal.

Leaching Behavior of Cesium, Strontium, Cobalt, and Europium from Immobilized Cement Matrix

Numerous low-level and intermediate-level radioactive wastes are generated from the decommissioning processes of nuclear power plants; these wastes are immobilized to prevent the release of radionuclides under disposal conditions. In this study, we investigated the leaching behavior of Cs, Sr, Co, and Eu, which are common in immobilized decommissioning wastes. Ordinary Portland cement (OPC) was used as an immobilization agent. During the test, leaching of the nuclides occurred in the order of Cs, Sr, Co, and Eu and decreased over time. The results showed that 41.4% of the total Cs leached over 90 days, although the other elements leached in quantities less than 1.5%. CaCO3 was precipitated by the release of cementitious materials, indicating carbonation of the leachate. The leachability indexes in all cases exceeded the acceptable criteria (>6). The results of the present study suggest that OPC can be effectively used as a binding material to immobilize nuclides (Cs, Sr, Co, and Eu) contained in decommissioning wastes.

The Method for Assessing the Conjugacy of Ocular Drifts

This chapter is an overview of main processes and effects in cementitious materials which occur under the exposure of ionizing radiation. The information provided should be useful for regulators and waste managers since it shows the possibility of extending the application of such materials in immobilization of not only intermediate-level waste, but also some types of high-level waste.Cementitious materials are known to be used in nuclear industry for a long time to construct facilities exposed to high irradiation doses (pools for spent nuclear fuel, nuclear research and power reactors, high-energy particle accelerators, hot cells, high-level waste dry casks, medical facilities using nuclear systems and X-rays, etc.). These materials are convenient for immobilization of different types of liquid and solid radioactive waste. However, currently cements can be only used as matrices for low- and intermediate-level waste. This is due to the doubts about their stability toward higher radiation doses. The main results of radiation damage are typically mechanical weakening, release of hydrogen gas, and cracking that can lead to the destruction of the matrix and enhanced release of radionuclides in contact with groundwater. In this chapter, we have overviewed main effects of ionization radiation on cementitious materials and on applicability of such materials as immobilizing materials suitable for radioactive waste of higher activities.

Разработка и применение интегрированной системы математических моделей переноса радионуклидов после гипотетической аварии с целью минимизации радиоэкологических последствий

An integrated system of mathematic models is developed and implemented. The system is aimed at predicting the spread of the radioactive materials in the Arctic waters from a complex source distributed in space and time, formed by an emergency release of radionuclides from a nuclear-powered facility. Such approach allows taking into account various mechanisms of radionuclide transfer in arbitrary combinations. In addition to customary considered atmospheric and marine advection-diffusion processes with sedimentation on the underlying surface, it takes into consideration other mechanisms. Among them are particle sedimentation to the sea bottom with bottom capture, reverse process of washing-out from the bottom sediments. Specially attended is the Arctic-specific mechanism of particle ice-binding in the sea ice, drift of the frozen particles with ice, and their return to marine environment in result of ice thawing. The latter process may result in the appearance of the radioactive source at the large distance from the initial source and long time after the release event. The integrated model complex will provide the most realistic picture of the radioactive trace spread. It will sure be the effective tool for minimizing the emergency negative impact on the population and environment. The article a stage of long-term work that is currently ongoing.

Machine learning-enabled prediction of chemical durability of A2B2O7 pyrochlore and fluorite

Pyrochlore-structure type and its derivative in a general formula A2B2O7 (A = rare earth elements and actinides; B = Ti, Sn, Zr, Hf, Pb, Si, etc.) display excellent structural flexibility and rich crystal chemistry as promising nuclear waste form materials capable of immobilizing actinides and fission products. It is essential to understand these materials’ chemical durability and element release of radionuclides in order to evaluate their performance in near-field environment. However, it is a formidable grand technological challenge to experimentally perform durability testing across hundreds of thousands of possibilities resulting from their extreme compositional complexities due to cation substitutions at both A and B-sites. In this work, we demonstrate a machine learning approach to determine the key materials parameters and structural characteristics governing the leaching behaviors from a small set of selected compositions as model systems, enabling a science-based prediction of their chemical durability that can be extended to a wide range of chemical compositions. The combination of four key structural characteristics and materials parameters, including ionic radius size difference rA-B, ionic potential difference EB-A, electronegativity difference χB-A, and lattice parameter α, creates features an optimized prediction of the chemical durability. Two machine learning models, linear regression and Kernel ridge regression models, are trained on the randomly-split training dataset derived from the experimentally-determined elemental release rates, and subsequently tested on the testing dataset. The predicted leaching rates from both machine learning models show an excellent agreement with the experimental data, demonstrating the feasibility of rapidly evaluating the material properties of new compositions. These results highlight the immense potential of synergizing informatics through machine learning-based models and well-controlled experiments of selected model systems to accelerate materials design and discovery with optimized compositions and performance of promising materials for effective nuclear waste management.

Development of operational intervention levels for events with release of radionuclides

An essential part of a protection strategy for radiological emergencies is the development of national dose criteria and of operational intervention levels (OILs) to decide about protective measures for all ten scenarios Germany is preparing. For the process of planning and implementing such protection strategies as required by the German Radiation Protection Law the Federal Ministry BMU has commissioned the German Radiation Protection Commission (SSK) to recommend dose criteria and OILs for emergency response measures. OILs link a chosen dose criterion for a protective action with a suitable measurement of the contamination situation such as ambient dose rate (μSv h−1), contamination level on surfaces (Bq cm−2) or activity content (Bq g−1, Bq cm−3). This link should adequately model the exposure of persons during a defined exposure period (e.g. seven days, one year) caused by the measured contamination. Dose calculations to quantify OILs should apply assumptions and parameter values that are in tendency realistic and not unduly conservative. OILs have been developed for the following emergency response actions based on radiation measurements: Sheltering on the basis of dose rate (μSv h−1) and contamination level(Bq cm−2). Evacuation on the basis of dose rate (μSv h−1) and contamination level(Bq cm−2). Establishing a radiological hazard area to implement access andcontamination control on the basis of dose rate (μSv h−1) and contaminationlevel (Bq cm−2). Contamination control and possibly decontamination of persons and objects(items, goods, vehicles, etc) based on contamination level (Bq cm−2). A set of precautionary early actions: warning the population not toconsume freshly contaminated food and agricultural measures to reducefood contamination based on dose rate (μSv h−1). Application of maximum permitted levels of radioactive contamination offood and feed (Bq kg−1) according to Euratom Regulation.

A sensitivity study of the factors affecting the risks associated with the disposal of spent nuclear fuel in a geological disposal facility in a clay environment

The disposal route for the UK's inventory of spent nuclear fuel has yet to be decided upon, however if the UK follows the approach taken by many nations, its spent nuclear fuel will be destined for geological disposal. This paper provides a simplified, approach to the calculation of radiological risk posed to members of the public, associated with the disposal of spent advanced gas-cooled reactor (AGR) nuclear fuel in a geological disposal facility (GDF) located in a hypothetical clay geological setting. Simulating the release of radionuclides at differing levels of complexity shall aid in the development of aspects of the design, construction, operation, decommissioning and post-closure phases of geological disposal and the development of a safety case for geological disposal. This paper presents a high-level, 1-D model built in GoldSim, to provide an initial indication of the radiological risks to the public, the nature of which would require further development into a complex total system model in order to facilitate risk evaluation supplementary to a safety case. The model was validated against benchmark calculations provided by Radioactive Waste Management Ltd. The base case calculations suggest that the predicted risks to the potential exposed groups are well below, approximately 2 orders, the recommended risk guidance level of 1 × 10−6 per year. A number of sensitivity studies were carried out to identify the importance of various factors that could influence the predicted risks. Sensitivity analysis indicated that the most influential sensitivities on the annual risk posed by geological disposal were the depth and rate at which spent nuclear fuel dissolved in contact with groundwater. The pathways available for groundwater flow to the biosphere were also noted to significantly alter the peak risk observed; crucially almost all sensitivities did not increase the peak risk to within 1 order of magnitude below the recommended risk guidance level. Validation for a second hypothetical high strength rock geological setting resulted in higher projected predicted risks to the potential exposed groups although risk levels remained below the recommended risk guidance level. This suggests a clay environment may have favourable characteristics for the final disposal of spent nuclear fuel compared to a high strength rock alternative.

CERN-MEDICIS: A Review Since Commissioning in 2017.

The CERN-MEDICIS (MEDical Isotopes Collected from ISolde) facility has delivered its first radioactive ion beam at CERN (Switzerland) in December 2017 to support the research and development in nuclear medicine using non-conventional radionuclides. Since then, fourteen institutes, including CERN, have joined the collaboration to drive the scientific program of this unique installation and evaluate the needs of the community to improve the research in imaging, diagnostics, radiation therapy and personalized medicine. The facility has been built as an extension of the ISOLDE (Isotope Separator On Line DEvice) facility at CERN. Handling of open radioisotope sources is made possible thanks to its Radiological Controlled Area and laboratory. Targets are being irradiated by the 1.4 GeV proton beam delivered by the CERN Proton Synchrotron Booster (PSB) on a station placed between the High Resolution Separator (HRS) ISOLDE target station and its beam dump. Irradiated target materials are also received from external institutes to undergo mass separation at CERN-MEDICIS. All targets are handled via a remote handling system and exploited on a dedicated isotope separator beamline. To allow for the release and collection of a specific radionuclide of medical interest, each target is heated to temperatures of up to 2,300°C. The created ions are extracted and accelerated to an energy up to 60 kV, and the beam steered through an off-line sector field magnet mass separator. This is followed by the extraction of the radionuclide of interest through mass separation and its subsequent implantation into a collection foil. In addition, the MELISSA (MEDICIS Laser Ion Source Setup At CERN) laser laboratory, in service since April 2019, helps to increase the separation efficiency and the selectivity. After collection, the implanted radionuclides are dispatched to the biomedical research centers, participating in the CERN-MEDICIS collaboration, for Research & Development in imaging or treatment. Since its commissioning, the CERN-MEDICIS facility has provided its partner institutes with non-conventional medical radionuclides such as Tb-149, Tb-152, Tb-155, Sm-153, Tm-165, Tm-167, Er-169, Yb-175, and Ac-225 with a high specific activity. This article provides a review of the achievements and milestones of CERN-MEDICIS since it has produced its first radioactive isotope in December 2017, with a special focus on its most recent operation in 2020.

Homogenization of bentonite upon saturation: Density and pressure fields

Bentonite structures are a core element of engineered barriers for geological radioactive waste repositories, preventing radionuclide release into the biosphere. These are often assemblies of elements with different densities (bricks, pellets, powder), or may contain regions of different density due to the building process (technological voids, powder segregation). The performance of these barriers is largely dependent on this material reaching a final state of homogeneous hydraulic and mechanical properties upon water intake. In order to understand the coupled processes of swelling and water transport in bentonite at play at the laboratory scale, an experimental approach was designed to characterize in detail the evolution of the state of samples saturated under isochoric conditions. Firstly, three-dimensional fields of water content, solid density and displacement of a heterogeneous powder sample submitted to unidirectional hydration were measured using X-ray microtomography, as a function of time. These fields allowed to gain insight into the phenomenology of non-monotonic water and clay transport resulting from the heterogeneous and coupled evolutions of the permeability, mechanical and friction properties. Then, samples with controlled heterogeneities such as pellets, powder, technological voids or compacted blocks of different densities - but with similar average properties - were saturated in isochoric conditions with a large set of local stress measurements. The different swelling pressure average evolutions and variability, as well as post-mortem measurements provided rich data on the effect of the initial density heterogeneities on the saturation process and the final state as well as insights into the role of wall friction. In particular, hydration kinetics were shown to depend on a complex manner on the initial heterogeneities. Regions first in contact with water were shown to exhibit a non-monotonic hydromecanical response and the development of macroporosity.

Characterization of expanded vermiculite to use as natural barrier in repository

Brazilian National Nuclear Energy Commission – CNEN – is responsible to build a repository for the low and intermediate--level radioactive wastes generated in Brazil. In accordance of Brazilian legislation it will be a near surface one. In the repository, treated waste will be stored, which will be surrounded by engineering and natural barriers, in order to prevent the release of radionuclides, during the required time to their decay to the limit levels. The focus of this study is to characterize the expanded vermiculite, which can be used as a natural barriers, over and under the engineering barriers and also as a backfill. They have the function of limiting water infiltration and stabilizing the disposal system, filling the empty spaces, controlling the gases and delaying the release of radionuclides. In this paper is presented the characterization of an expanded vermiculite. The tests to determine its physical chemical properties were based on a clay characterization Protocol. The granulometric analysis was performed using a sieving method. The moisture content was established by the gravimetric method of weight loss, and the specific surface area by the BET method. The helium ultrapicnometer method was used to determine the density. The microstructure was analyzed by X-ray diffraction and scanning electron microscopy. The results indicate that the studied vermiculite would be efficient as a natural barrier of the Brazilian surface repository, as the it presented a stable structure.

Study of cesium sorption by a Brazilian bentonite

Engineered and natural barriers compose the disposal system of near surface repository to store low and intermediate level radioactive wastes. Waterproofing barriers are constituted of clays, such as the bentonite, in order to avoid or to limit the release of radionuclides to with consequences to the environment and humans. Characteristics such as high cation exchange capacity, high adsorption capacity and high sealing capacity, are taken into consideration for choosing this material. The migration of radionuclides through protection barriers can occur in a variety of ways, such as surface water infiltration, groundwater intrusion, among others. Diffusion is the process which governs the contaminant transport through soil barriers. Adsorption is one of the processes that could be considered in the diffusion mechanism, and sorption isotherms are obtained from values measured in the batch-adsorption experiments. Therefore, such experiments were carried out in order to estimate the cesium sorption by the bentonite. The quantitative analysis was performed using atomic absorption and the energy dispersive X-ray fluorescence spectrometry. With the purpose to calculate the true value of cesium sorption, a mass-balance was made considering all the steps of the batch experiment, and it was found a 6% loss in the whole process. The quantity of cesium sorbed by the bentonite was 66.7 mg.g-1. At the moment, eight additional experiments are being performed, using solutions with different cesium concentrations, leading to all the necessary data for the isotherm.

Conceptual design overview of the ITER WCLL Water Cooling System and supporting thermal-hydraulic analysis

In this paper, the conceptual design of the International Thermonuclear Experimental Reactor (ITER) Water-Cooled Lithium-Lead (WCLL) Test Blanket Module (TBM) Water Cooling System (WCS) from Europe is presented. The system consists of two loops in series. This design feature allows the removal of heat from the TBM box avoiding at the same time the release of radionuclides into the ITER Component Cooling Water System (CCWS), that acts as WCLL Test Blanket System heat sink. For this purpose, the WCS primary loop deals with the direct heat removal from the ITER TBM and the secondary one implements physical separation between the contaminated primary loop coolant and the CCWS. The insertion of an economizer into the primary loop determines the characteristic “eight” shape of the circuit. This choice was done in order to reduce the temperature difference on the intermediate heat exchanger. Hairpin type and steam bubble are the technologies selected for heat exchangers and pressurizers, respectively. Pressure and temperature control systems are foreseen to limit excursions from rated values in normal operational states and abnormal transients.A computational activity was promoted to assess the WCLL-WCS conceptual design, using a modified version of the RELAP5 Mod3.3 system code. A detailed thermal-hydraulic model was developed on the basis of design outcomes. The nodalization scheme includes the TBM, the WCS, a portion of the CCWS and the lithium-lead circuit. The computational campaign involved both the normal operational state and selected abnormal transients. In all the scenarios simulated, the conceptual design has highlighted the capability of operating the system respecting all the thermal-hydraulic requirements. The abnormal transient selected and presented is the loss of flow in the CCWS (loss of heat sink). In these conditions, TBM cooling function has been verified, keeping standard control strategies without any external action.

Study on Release and Migration of Radionuclides under the Small Break Loss of Coolant Accident in a Marine Reactor

Using severe accident analysis program MELCOR, the small break loss of coolant accident (SBLOCA) analysis model was established for a marine reactor. The release and migration of radionuclides were analyzed during a severe accident induced by SBLOCA. The analysis of the hydrogen source term release showed that the maximum hydrogen release amount was 248.567 kg, and the hydrogen release amount accounted for less than 4% of the air volume. So, there would be no danger of hydrogen explosion accidents. The research mainly focused on the behaviors of the release, the transport, the retention, and the final distribution of inert gases represented by Xe, volatile gases represented by CsI, and nonvolatile nuclides represented by Ba. The results showed that the reactor core exposed completely with a lagging by 510 s and the initial release time of nuclides was lagged by 1916 s. The release shares of Xe in the primary circuit system, the containment, and the environment were 0.013%, 0.06%, and 32.71%, respectively. Also, Ba shared 0.016%, 0.0032%, and 3.28%, respectively, and CsI shared 0.0145%, 0.0012%, and 2.845%, respectively.

Radiation-hygienic characteristic of the protected zones of peaceful nuclear explosions in the Arkhangelsk region

The article presents data on the radiation situation on the territories of the protected zones of peaceful nuclear explosions «Globus-2», «Agat» and «Rubin-1» based on the results of field studies in 2020. Peaceful nuclear explosions were conducted in the Arkhangelsk region for seismic exploration. The results of the radiation-hygienic state of protected zones and radiation safety in the territories adjacent to peaceful nuclear explosions are analyzed by the main indicators: the values of the ambient dose equivalent of gamma radiation, the identification of gamma-emitting radionuclides in situ by field gamma spectrometry, the specific activity of technogenic radionuclides in the soil and tritium in the water of reservoirs and drinking water sources. An assessment of the sanitary condition of the studied territories of protected zones of peaceful nuclear explosions is given. It was found that the levels of the ambient dose equivalent rate of the gamma radiation at all sites are at the level of fluctuations in the natural regional radiation background in the range of 0.08-0.20 microSv/h. On the territory of the protected zone of the peaceful nuclear explosion «Globus-2» were repeatedly identified areas of minor local soil contamination with 137Cs (the previous study was performed in 2011). Increased values of the ambient equivalent dose rate of gamma radiation (0.20 -0.22 μSv/h) and specific activity of 137Cs in two soil samples (154 and 160 Bq/kg, respectively) were found in these areas. The tritium content in the water of water supply sources does not exceed 5 Bq/kg. These values are typical for global levels of this radionuclide in water bodies. Remains of technological equipment were found on the territories of the protected zones of the peaceful nuclear explosions (especially in the protected zone of the explosion «Agat»), as well as rusty metal pipes and barrels, the remains of wooden structures, etc. Information signs installed on dilapidated concrete foundations at the mouths of combat wells of all surveyed peaceful nuclear explosions have subtle inscriptions. The presence of forest thickets in the places where the investigated peaceful nuclear explosions are conducted and their ignition during forest fires is a potential threat to the release of technogenic radionuclides during depressurization of the mouths of combat wells.

Nuclear Emergency: A Review of Radioactivity in Human Ecosystem

One of the serious impacts on the environment and human health is the proper evaluation of radioactivity from elevated naturally occurring isotopes or those from industrial activities. Radionuclides produced from nuclear explosions and nuclear facilities when released into the environment may reach the human body through the food pathways. Radionuclides from environmental sources have the ability to cause a hazard to human health through food sources. There is the need for radiological estimation for transfers from the vegetative covers through the human ecosystem. The assessment models are subdivided mainly into three categorical states; that is from the soil-plant transfer, the transfer to animals, and spatial models. Radionuclides may enter the human ecosystem through atmospheric releases from nuclear facilities. Freshwater contamination happens via the release of water bodies which may affect aquatic lives. Various assessment models have been developed based on the fallout deposition from the Chernobyl accident, however, the distinct uncertainties involved in propagating this data to a completely different geographical location cannot be ignored. This is a review paper that analysis the Absalom model, its application in system assessment models, and how they can be applied in country-specific data to achieve a realistic model prediction in food safety from radionuclide release and the need for further research.

Aerosol dynamics and dispersion of radioactive particles

Abstract. In the event of a failure of a nuclear power plant with release of radioactive material into the atmosphere, dispersion modelling is used to understand how the released radioactivity is spread. For the dispersion of particles, Lagrangian particle dispersion models (LPDMs) are commonly used, in which model particles, representing the released material, are transported through the atmosphere. These model particles are usually inert and undergo only first-order processes such as dry deposition and simplified wet deposition along the path through the atmosphere. Aerosol dynamic processes including coagulation, condensational growth, chemical interactions, formation of new particles and interaction with new aerosol sources are usually neglected in such models. The objective of this study is to analyse the impact of these advanced aerosol dynamic processes if they were to be included in LPDM simulations for use in radioactive preparedness. In this investigation, a fictitious failure of a nuclear power plant is studied for three geographically and atmospherically different sites. The incident was simulated with a Lagrangian single-trajectory box model with a new simulation for each hour throughout a year to capture seasonal variability of meteorology and variation in the ambient aerosol. (a) We conclude that modelling of wet deposition by incorporating an advanced cloud parameterization is advisable, since it significantly influence simulated levels of airborne and deposited activity including radioactive hotspots, and (b) we show that inclusion of detailed ambient-aerosol dynamics can play a large role in the model result in simulations that adopt a more detailed representation of aerosol–cloud interactions. The results highlight a potential necessity for implementation of more detailed representation of general aerosol dynamic processes into LPDMs in order to cover the full range of possible environmental characteristics that can apply during a release of radionuclides into the atmosphere.