Literatura académica sobre el tema "Captur of gaseous radionuclides"

The cyclotron is a device used to create radioactive atoms with a short half-life (radioactive isotopes) that can be utilised for research and medical imaging. When nuclear and radiation facilities are utilized, serviced, or decommissioned, radioactive waste is produced. The amount of radioactive waste produced is greatly decreased by good operating procedures. Iodine-123, Technetium-99m, Iodine-131, Gallium-67 Thallium-201 and fluorine-18 fluorodeoxyglucose are among the radionuclides utilised in medicine. The most widely used gaseous/aerosol radionuclides are (aerosolized) technetium-99m, xenon-133, and krypton-81m. The use of radionuclides (radioactive element) for industrial process control and instrumentation, medical diagnostic and therapeutic purposes, as well as numerous uses in research, education, agriculture, geological exploration, construction, and other human endeavors, results in radioactive waste. These applications generate a variety of radioactive waste, which can come from sealed sources and be in solid, liquid, or gaseous form. If the trash containing considerable amounts of radionuclides is not handled properly, there may be serious concerns to both the environment and human health. Due to the wide variety of waste kinds addressed, special consideration must be paid to safety concerns and regulatory management. This article will examine the fundamental procedures for managing radioactive waste in compliance with the regulatory agencies like AERB (Atomic Energy Regulatory Board) and IAEA (International Atomic Energy Agency).

Molten salt reactor operation will necessitate circulation of a cover gas to remove certain evolved fission products and maintain an inert atmosphere. The cover gas leaving the reactor core is expected to contain both noble and non-noble gases, aerosols, volatile species, tritium, and radionuclides and their daughters. To remove these radioactive gases, it is necessary to develop a robust off-gas system, along with novel sensors to monitor the gas stream and the treatment system performance. In this study, a metal organic framework (MOF) was engineered for the capture of Xe, a major contributor to the off-gas source term. The engineered MOF column was tested with a laser-induced breakdown spectroscopy (LIBS) sensor for noble gas monitoring. The LIBS sensor was used to monitor breakthrough tests with various Xe, Kr, and Ar mixtures to determine the Xe selectivity of the MOF column. This study offers an initial demonstration of the feasibility of monitoring off-gas treatment systems using a LIBS sensor to aid in the development of new capture systems for molten salt reactors.

The paper presents the results from a thermodynamic analysis of the behavior of Sr and Ca radionuclides in the process of heating radioactive graphite in an air atmosphere. The TERRA software package was used for the thermodynamic analysis in a temperature range of 300 to 3600 K to determine the possible composition of the ionized, gaseous and condensed phases. It has been found that strontium is in the form of condensed SrCl2(c) and gaseous SrCl2 in a temperature range of 300 to 1600 K, and in the form of gaseous SrCl2, SrO, SrCl and Sr and ionized SrCl+, Sr+ and SrO+ when the temperature is increased from 1600 to 3600 K. Calcium is in the form of condensed CaCl2(c), CaUO4(c), CaO(c) and gaseous CaCl2 in the temperature interval between 300 and 2100 K, and in the form of gaseous Ca, CaCl and CaO and ionized Ca+, CaO+ and CaCl+ when the temperature is increased from 2100 to 3600 K. The paper determines the key reactions within individual phases and among condensed, gaseous and ionized phases. The equilibrium constants of their reactions have been calculated. Based on the results obtained, dependence plots are presented for the Sr and Ca radionuclide distribution by phases.

Abstract. In the case of severe accident with loss of containment in a nuclear plant, radionuclides are released into the atmosphere in the form of both gases and aerosol particles (Baklanov and Sørensen, 2001). The analysis of radioactive aerosol scavenged by rain after the Chernobyl accident highlights certain differences between the modelling studies and the environmental measurements. Part of these discrepancies can probably be attributed to uncertainties in the efficiencies used to calculate aerosol particle collection by raindrops, particularly drops with a diameter larger than one millimetre. In order to address the issue of these uncertainties, an experimental study was performed to close the gaps still existing for this key microphysical parameter. In this paper, attention is first focused on the efficiency with which aerosol particles in the accumulation mode are collected by raindrops with a diameter of 2 mm. The collection efficiencies measured for aerosol particle in the sub-micron range are quantitatively consistent with previous theoretical model developed by Beard (1974) and thus highlight the major role of rear capture in the submicron range.

A distinctive feature of enterprises for extracting and processing uranium ore is the inevitable pollution by solid, liquid and gaseous waste. The amount of radioactive waste (RW) is most significant in the nuclear fuel cycle. In spite of its relatively low activity it is the major contributor to the formation of radiation hazards to the people and environment. The radioactivity of uranium ores and of their processing waste is due to natural radionuclides of uranium (238U and 235U) and thorium (232Th) radioactive decay chains.

Les isotopes radiotoxiques de l'iode et du ruthénium, tels que 129I, 131I, 103Ru et 106Ru, sont produits en quantité significative durant la fission nucléaire. Après un accident de réacteur nucléaire, ces éléments peuvent être rapidement disséminés dans l’environnement, sous la forme d’espèces gazeuses très volatiles comme l'iode moléculaire (I2) ou le tétraoxyde de ruthénium (RuO4). Afin de limiter la dispersion de ces produits de fission en cas d’accident, des filtres composés de matériaux poreux (zéolites ou charbon actifs) peuvent être employés. Cependant, de tels solides poreux présentent des limitations dans le contexte d'accident nucléaire. En effet, la présence d'espèces empoisonnantes (par exemple NOx, H2O, COx) peut inhiber la capture d’espèces radiotoxiques. De plus, leur relative faible porosité n’est souvent pas adaptée au bon piégeage d’espèces volumineuses comme RuO4. Sur la base de ces limites, une classe récente de matériaux poreux appelés Metal-Organic Frameworks (MOFs) pourrait s’avérer être un substitut efficace. En effet, les composés MOFs sont des matériaux hybrides cristallisés, constitués de clusters inorganiques liés les uns aux autres par des ligands organiques. Cette organisation peu dense offre une porosité importante et des surfaces spécifiques élevées (jusqu'à 7000 m2.g-1), nettement supérieurs à celles des solides poreux habituels. Bien que ces solides aient déjà montré de bonnes capacités pour la capture d’espèces radioactives, très peu de données rapportent leur efficacité pour le piégeage d’espèces gazeuses (notamment RuO4) en conditions accidentelles.Afin de renforcer nos connaissances sur les composés MOFs pour une potentielle utilisation en sureté nucléaire, ce travail de thèse s’est intéressé à leur efficacité pour la capture de I2 et RuO4 volatils dans certaines matrices poreuses modèles (type UiO-66). Nous avons mis en évidence l'importance de la fonctionnalisation du ligand espaceur et du confinement de l’iode au sein du réseau poreux. Ainsi, l’iode créé une interaction forte avec la charpente des MOFs pour former d’autres espèces iodées de type Ix-. Cette transformation a notamment été analysée par spectroscopie RAMAN.Suite à cette première étude, nous avons sélectionné le solide UiO-66_NH2 comme matériau de filtration de référence pour réaliser un essai dans l’installation EPICUR de l’IRSN. Celle-ci permet la manipulation d’iode radioactif (131I) et l’étude de son confinement au sein de la charpente poreuse en conditions accidentelles (radiation, température, vapeur d’eau). Ce travail a nécessité, en amont, d’élaborer un protocole de mise en forme, afin de produire un matériau MOF avec une granulométrie sphérique millimétrique. En parallèle, un travail sur la résistance de ce matériau sous irradiation gamma a également été entrepris, dans l’installation IRMA de l’IRSN. Cette étude a confirmé l’excellente efficacité du UiO-66_NH2 dans le contexte choisi. Enfin, le matériau UiO-66_NH2 a également été le candidat choisi pour la capture de RuO4 gazeux. Les différentes analyses (MET, RMN) ont permis de quantifier le RuO4 au sein des pores et de proposer des mécanismes réactionnels expliquant sa très bonne adsorption
The radiotoxic isotopes of iodine and ruthenium, such as 129I, 131I, 103Ru and 106Ru, are produced in significant quantities during nuclear fission. After a nuclear accident, these elements can be rapidly disseminated in the environment, in the form of highly volatile species such as molecular iodine (I2) or ruthenium tetroxide (RuO4). In order to limit the dispersion of these fission products, in case of a nuclear accident, filters composed by porous materials (zeolites or activated carbon) can be used. However, such porous solids have limitations during a nuclear accident. Indeed, the presence of poisonous species (for example NOx, H2O, COx) can ihhibit the capture of radiotoxic species. In addition, their relatively low porosity is often not suitable for the good trapping of large species such as RuO4. Based on these limitations, a recent class of porous materials called Metal-Organic Frameworks (MOFs) could be an effective substitute. Indeed, MOFs are hybrid materials, composed of inorganic clusters linked to each other by organic ligands. This low-density organization allows high porosity and high specific surface areas (up to 7000 m2.g-1), significantly higher than those of the usual porous solids. Although MOFs have already shown good capacities for capturing radioactive species, very little data exist on their effectiveness for trapping gaseous species (especially RuO4) and under accident conditions.In order to strengthen our knowledge of MOFs for potential use in nuclear safety, this thesis work focused on the effectiveness of some model MOFs for the capture of volatile I2 and RuO4 under accident conditions. We have highlighted the importance of the organic linker functionalization and confinement of iodine in the porous matrix. Thus, iodine creates a strong interaction with the framework of MOFs to form other iodine species of type Ix-. This transformation was notably analyzed by RAMAN spectroscopy.Following this first study, we selected the compound UiO-66_NH2 as reference filtration material to be tested in an IRSN facility called EPICUR. This one allows the manipulation of radioactive iodine (isotope-131) and the study of the confinement of iodine in within the porous framework in accidental conditions (radiation, temperature, steam). This work needs, upstream, to develop a shaping process in order to produce a MOF material with a spherical millimeter particle size. In parallel, an investigation on the resistance of this material under gamma irradiation was also undertaken in IRMA facility at IRSN. This study confirmed the excellent capacity of the solid UiO-66_NH2 in the present context. Finally, UiO-66_NH2 was also the candidate of choice for the capture of gaseous RuO4. The various analyzes (TEM, NMR) made it possible to quantify the RuO4 within the pores and to propose reaction mechanisms explaining its very good capture in UiO-66_NH2

碩士
國立清華大學
原子科學研究所
81
The purpose of this study is to examine the applicabi- lity of real-time monitoring of gaseous radionuclide with a high- purity germanium (HPGe) detector. In the case of normal operations, continuous release of gaseous radio- nuclides of tracer amount of isotopes of Ar,Kr and Xe from reactor is found. The system, consisted of portable HPGe and multichannel analyzer (MCA), is designed for the measurement both in sealed chamber and in outdoors sur- rounded with gaseous radionuclide. Besides, quantitative, qualitative, prompt and on-line measurement can be reached with the movable detectors without air sucking pump and filter paper. The research was divided into two major parts as follows : (1) The HPGe detector was calibrated by using of artificial gaseous Ar-41,Kr-85m and Xe-125,133,135 releas- ed to a sealed chamber. Minimum detectable concentration and possible application field monitoring are discussed. (2) Three various types of HPGe detectors were calibrated , the relationship between detectors' efficiency and the crystal volume was observed. The results show : (1)linear relationship between eff- iciency and energy in log coordinate can be obtained. The detector system is available for the measurement of minimum detectable concentration (MDC) far less than legal maximum permissible concentration (MPC), application to detect the gaseous radionuclide of nuclear disaster and nuclear war.(2) There exists a linear relationship between crystal volume and relative efficiency so that the inter- relation between efficiencys with various detectors can be developed in emergenecy.

碩士
國立清華大學
原子科學系
85
In the case of normal operations, continuous release of gaseous radionuclides of trace amount of isotopes of Ar, Kr and Xe form reactor is found. The purpose of this study is to examine the applicability of real- time monitoring of gaseous radionuclides with a high-purity germanium detector (HPGe). The system, consisted of portable HPGe and multichannel analyzer (MCA), is designed for the measured both in sealed chamber and in outdoors surrounded with gaseous radionuclides. Besides, quantitative, qualitative, prompt and on-line measurement can be reached with the portable detectors without air sucking pump and filter paper. The research was divided into two major parts as follows : (1) The HPGe detector was calibrated using artificial gaseous Ar-41 and Xe- 125,133,135 released to a sealed chamber. The efficiency in both sealed chamber and outdoors are formulated from the calibrated volumetric sealed chamber. (2) We can obtain the effective dose equivalent of the gaseous radionuclides using dose rate conversion factors. Detection limit and possible application in field monitoring are discussed. The results show : (1) linear relationship between efficiency and energy in log coordinate be obtained, the detector system is available for measurement of detection limit (DL) far less than legal derived air concentration (DAC), application can be applied to detect the gaseous radionuclides in nuclear accident and nuclear war. (2) The dose rate formulated from the counting rate can provide the rapid information for personal protection and post accident assessment.

Abstract The complicated severe accident phenomena in typical Pressurized Water Reactor (PWR) Generation III may have a strong influence on source term release into environment and radiological consequence. The study on sensitivity analysis is beneficial to the identification of important factors in severe accident source term analysis and the quantification of their impact. ASTEC, the integral code of severe accident analysis developed by IRSN, is used to analyze the sensitivity of key parameters of severe accident source term for typical PWR Generation III, with the simulation of safety systems and source term phenomena, in the representative sequence with fast accident progression, Large Break Loss of Coolant Accident (LBLOCA). With the consideration of the design features of typical PWR Generation III and research status of severe accident source term, the key parameters for sensitivity analysis are identified and selected based on the whole process of radionuclides release, including gaseous iodine mass release fraction from primary circuit to containment, silver iodide reaction, dose rate and pH value in sump, washing effect, etc. The sensitivity is quantified by iodine release mass to containment, which is one of the most dangerous radionuclides due to its threat to environment and human thyroid after inhalation and ingestion. The gaseous iodine mass release fraction from primary circuit to containment, silver iodide reaction and washing effect are presented in results as the major contributors to the variation of severe accident source term evaluation.

In the U.K., High Level Waste from reprocessing operations is vitrified at the Sellafield Waste Vitrification Plant (WVP). A small number of the nuclides present in the waste have the potential to volatilize during vitrification. In order to prevent release of any radionuclides to the environment it is important to understand the mechanisms by which volatilization may occur and to have suitable controls in place. One element of particular concern is ruthenium, formed during the fission of nuclear fuel, which has the potential to form gaseous species such as RuO4 during the vitrification process and whose behavior must therefore be understood in order to underpin the safe operation of WVP.

The Pebble Bed Modular Reactor is being developed in South Africa. Important for PBMR implementation is a viable strategy for waste management. Irradiated graphite from fuel and structural components is too voluminous for practical treatment with traditional higher level waste methods and too radioactive to recycle. To clean the graphite of radionuclides, a two-step process is being pursued: (1) non-carbon radionuclides (activation products, fission products and actinides) are removed on an elemental basis by a chemical or microbial process. (2) 14C requires separation at an isotopic level, which would be impractical with established methods (gaseous diffusion or centrifuge). PBMR is investigating a method of isotope separation using biofractionation. Preliminary experiments indicate that microorganisms do separate radioactive 14C from stable 12C. An aqueous slurry of 14C-spiked, powdered graphite was “fed” to the microbes for 15–18 hours. The microbes initially contained only background levels of 14C, i.e. orders of magnitude less than the slurry. In post-experiment analyses, a sample of the microbes was found to contain approximately twice the amount of 14C present in the bulk slurry material. Experiments are underway to further quantify and verify these results, which indicate distinct microbial processing mechanisms for 14C and 12C. The most current results will be presented.

ABSTRACT Underground nuclear explosions produce noble gases that can migrate to the surface and become detectable by atmospheric monitoring tools. However, it is challenging to predict radionuclide gas migration in the complex engineered and natural subsurface systems due to several issues. These issues include generation of complex fracture networks near an engineered cavity, reactivation of natural fractures, coupled hydro-thermo-mechanical processes for transport of high-pressure gases in a fractured porous rock system. To improve our understanding of these technical challenges, we used a triaxial direct shear test scheme to characterize coupled hydro-mechanical gas transport in fractured porous rocks. In the experiments, we tested nitrogen (an analog noble gas) flow through two types of rocks with distinct petrophysical properties: porous Bandelier tuff and tight Climax stock granite. For each type, we measured the Biot effective stress coefficient, the rock matrix permeability, and the fracture permeability under various stress states, which allowed us to examine the stress-dependency of the Biot coefficient and investigate rock matrix-fracture interactions for transport of pressurized gas. Comparing the experimental results for the two distinct rock types, we observed some important findings for gas transport in fractured rocks. First, rock fracturing does not necessarily increase samples’ gas permeability when the rock matrix is highly porous. Furthermore, gas permeability of intact rocks can show strong exponential stress dependency even when the rock matrix is very tight. Additionally, Biot effective stress coefficients are not necessarily close to unity for highly porous rocks, especially when rocks are subjected to large effective stresses. In summary, the experimental results can be used to improve the physics in high-fidelity numerical modeling to elevate our understanding of radionuclide gas transport in fractured porous rocks after an underground nuclear explosion event. INTRODUCTION Surface radionuclide monitoring is the primary means of determining if an underground explosion is nuclear in nature (Maceira et al., 2017). Following an underground nuclear explosion (UNE), signature noble gas radionuclides, such as Xe-131m, Xe-133, Xe-135, and some krypton radioisotopes, will be produced by nuclear fission (Carrigan et al., 1996; De Geer, 1996; Sun and Carrigan, 2014). They are hard to contain and tend to seep from the underground explosion and migrate to the surface. Surface sampling and detection of these signature gaseous radionuclides, when above some background levels, is a strong indicator of the occurrence of an underground nuclear explosion. By comparison, seismic monitoring cannot definitively discriminate between chemical (for example, TNT) explosions and nuclear events. It is thus important to understand transport of noble gas radionuclides in the subsurface rock strata.

Carbon-14 is a key radionuclide in the assessment of the safety of a geological disposal facility for radioactive waste because of the calculated assessment of the radiological consequences of gaseous carbon-14 bearing species [i]. It may be that such calculations are based on overly conservative assumptions and that better understanding could lead to considerably reduced assessment of the radiological consequences from these wastes. Alternatively, it may be possible to mitigate the impact of these wastes through alternative treatment, packaging or design options. The Radioactive Waste Management Directorate of the UK’s Nuclear Decommissioning Authority (NDA RWMD) has established an integrated project team in which the partners are working together to develop a holistic approach to carbon-14 management in the disposal system [ii]. For a waste stream containing carbon-14 to be an issue: • There must be a significant inventory of carbon-14 in the waste stream; AND • That waste stream has to generate carbon-14 bearing gas; AND • A bulk gas phase has to entrain the carbon-14 bearing gas: AND • These gases must migrate through the engineered barriers in significant quantities; AND • These gases must migrate through the overlying geological environment (either as a distinct gas phase or as dissolved gas); AND • These gases must interact with materials in the biosphere (i.e. plants) in a manner that leads to significant doses and risks to exposed groups or potentially exposed groups. The project team has developed and used this “AND” approach to structure and prioritise the technical work and break the problem down in a manageable way. We have also used it to develop our approach to considering alternative treatment, packaging and design options. For example, it may be possible to pre-treat some wastes to remove some of the inventory or to segregate other wastes so that they are removed from any bulk gas phase which might facilitate migration through the geosphere. Initially, the project team has undertaken a six month programme of work to examine the current understanding of these aspects and has captured this in the Phase 1 report [ii], in a modelling basis spreadsheet and in scoping assessments, which help us better understand the potential significance of carbon-14. Using the current modelling basis, but ignoring any potential benefits from the geosphere in retarding or preventing gas from reaching the surface, the calculated release of carbon-14 is dominated by: corrosion of irradiated reactive metals (in the operational and early post-closure time frame); corrosion of irradiated stainless steel and leaching of irradiated graphite (in the longer term). The Phase 1 work has shown that there is considerable scope for reducing the calculated radiological consequence for these wastes and a roadmap has been developed for a second Phase of work.