Author : International Atomic Energy Agency
Publisher : IAEA Tecdoc Series No. 1790
Page : 0 pages
File Size : 14,43 MB
Release : 2016
Category : Technology & Engineering
ISBN : 9789201040169
Graphite is widely used in the nuclear industry and in research facilities and this has led to increasing amounts of irradiated graphite residing in temporary storage facilities pending disposal. This publication arises from a coordinated research project (CRP) on the processing of irradiated graphite to meet acceptance criteria for waste disposal.
Author : International Atomic Energy Agency
Publisher : IAEA
Page : 88 pages
File Size : 12,13 MB
Release : 2006
Category : Business & Economics
ISBN :
Graphite has been used as a moderator and reflector of neutrons in more than 100nuclear power plants and in many research and plutonium-production reactors. It is usedprimarily as a neutron reflector or neutron moderator, although graphite is also used for other features of reactor cores, such as fuel sleeves. Many of the graphite-moderated reactors are now quite old, with some already shutdown. Therefore radioactive graphite dismantling and the management of radioactive graphite waste are becoming an increasingly important issue for a number of IAEA Member States. This report provides a comprehensive discussion of radioactive graphite waste characterization, handling, conditioning and disposal throughout the operating and decommissioning life cycle.
Author : Michail Georgioy Lasithiotakis
Publisher :
Page : 189 pages
File Size : 49,19 MB
Release : 2012
Category :
ISBN :
The cores of early UK graphite moderated research and production nuclear fission reactors operated at temperatures below 150°C. Due to this low temperature their core graphite contains significant amounts of stored (Wigner) energy that may be released by heating the graphite above the irradiation temperature. This exothermic behavior has lead to a number of decommissioning issues which are related to long term "safe-storage", reactor core dismantling, graphite waste packaging and the final disposal of this irradiated graphite waste. The release of stored energy can be modeled using kinetic models. These models rely on empirical data obtained either from graphite samples irradiated in Material Test Reactors (MTR) or data obtained from small samples obtained from the reactors themselves. Data from these experiments is used to derive activation energies and characteristic functions used in kinetic models. This present research involved the development of an understanding of the different grades of graphite, relating the accumulation of stored energy to reactor irradiation history and an investigation of historic stored energy data. The release of stored energy under various conditions applicable to decommissioning has been conducted using thermal analysis techniques such as Differential Scanning Calorimetry (DSC). Kinetic models were developed, validated and applied, suitable for the study of stored energy release in irradiated graphite components. A potentially valid method was developed, for determining the stored energy content of graphite components and the kinetics of energy release. Another parameter investigated in this study was dedicated in the simulation of irradiation damage using ion irradiation. Ion bombardment of small graphite samples is a convenient method of simulating fast neutron irradiation damage. In order to gain confidence that irradiation damage due to ion irradiation is a good model for neutron irradiation damage the properties and microstructure of various grades of ion irradiated nuclear graphite were also investigated. Raman Spectroscopy was employed to compare the effects of ion bombardment with the reported effects of neutron irradiation on the content of the defects. The changes of the of defect content with thermal annealing of the ion irradiated graphite have been compared with the annealing of neutron irradiated nuclear graphite.
Author : International Atomic Energy Agency
Publisher :
Page : 30 pages
File Size : 16,65 MB
Release : 2010
Category : Gas cooled reactors
ISBN : 9789201066107
This is the proceedings of an international conference on progress and solutions for radioactive graphite waste management. The main objectives of the meeting were to promote exchange of information among IAEA Member States? representatives and to discuss new trends and innovative developments in the management of radioactive graphite. These developments will have an impact on new graphite-moderated reactor systems as well as on new reactor programmes in the 21st century, where design for decommissioning is very important. In the summary the reader will find evaluations of the relative merits of various options.
Author : J. H. W. Simmons
Publisher : Elsevier
Page : 264 pages
File Size : 47,56 MB
Release : 2013-10-22
Category : Technology & Engineering
ISBN : 1483186490
Nuclear Energy, Volume 102: Radiation Damage in Graphite provides a general account of the effects of irradiation on graphite. This book presents valuable work on the structure of the defects produced in graphite crystals by irradiation. Organized into eight chapters, this volume begins with an overview of the description of the methods of manufacturing graphite and of its physical properties. This text then presents details of the method of setting up a scale of irradiation dose. Other chapters consider the effect of irradiation at a given temperature on a physical property of graphite. This book discusses as well the changes in dimensions produced by irradiation and the effects of irradiation on the mechanical properties of graphite. The final chapter deals with the accumulation of stored energy, which is one of the main problems caused by the irradiation of graphite in nuclear reactors. This book is a valuable resource for physicists and chemical physicists.
Author :
Publisher :
Page : 141 pages
File Size : 35,23 MB
Release : 2016
Category :
ISBN :
Author :
Publisher :
Page : 6 pages
File Size : 14,83 MB
Release : 2014
Category :
ISBN :
The aim of the research presented here was to identify the chemical form of 14C in irradiated graphite. A greater understanding of the chemical form of this longest-lived isotope in irradiated graphite will inform not only management of legacy waste, but also development of next generation gas-cooled reactors. Approximately 250,000 metric tons of irradiated graphite waste exists worldwide, with the largest single quantity originating in the Magnox and AGR reactors of UK. The waste quantity is expected to increase with decommissioning of Generation II reactors and deployment of Generation I gas-cooled, graphite moderated reactors. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 14C, with a half-life of 5730 years.
Author : Kazys Almenas
Publisher :
Page : 198 pages
File Size : 35,29 MB
Release : 1998
Category : Nuclear power plants
ISBN : 9789986492351
Author : Lorraine Mcdermott
Publisher :
Page : 175 pages
File Size : 40,20 MB
Release : 2012
Category :
ISBN :
Once current nuclear reactor operation ceases in the U.K. there will be an estimated 99,000 tonnes of irradiated nuclear graphite waste which may account for up to 30% of any future UK geological ILW disposal facility [1]. In order to make informed decisions of how best to dispose of such large volumes of irradiated graphite (I-graphite) within the UK nuclear programme, it is necessary to understand the nature and migration of isotopes present within the graphite structure. I-graphite has a combination of short and long term isotopes such as 14C, 3H and 36Cl, how these behave prior to and during disposal is of great concern to scientific and regulatory bodies when evaluating present decommissioning options. Various proposed decontamination and immobilisation treatments within the EU Euroatom FP7 CARBOWASTE program have been explored [2, 3]. Experiments have been carried out on UK irradiated British Experimental Pile Zero and Magnox Wylfa graphite in order to remove isotopic content prior to long term storage and to assess the long term leachability of isotopes. Several leaching conditions have been developed to remove 3H and 14C from the irradiated graphite using oxidising and various acidic environments and show mobility of 3H and 14C. Leaching analysis obtained from this research and differences observed under varying leaching conditions will be discussed. Thermal analysis of the samples pre and post leaching has been performed to quantify and validate the 14C and 3H inventory. Finally the research objectives address differences in leachability in the graphite to that of structural and operational variation of the material. Techniques including X-ray Tomography, Scanning Electron Microscopy, Autoradiography and Raman spectroscopy have been examined and show a significant differences in microstructure, isotope distribution and location depending of irradiation history, temperature and graphite source. Ultimately the suitability of the developed chemical treatments will be discussed as whether chemical treatment is a viable option prior to irradiated graphite long term disposal.
Author : Tara Elizabeth Smith
Publisher :
Page : 290 pages
File Size : 39,52 MB
Release : 2012
Category :
ISBN :
Presently, there are 250,000 metric tons of irradiated graphite waste worldwide and with the development of the High Temperature Gas Cooled Reactor this volume is expected to drastically increase indicating the need for a graphite waste management strategy. The most probable management strategy is long-term storage in a deep geological repository; however, the disposal of large irradiated graphite components is unnecessary and uneconomical. Previous characterization of irradiated graphite has indicated the most significant long-term isotope of concern is carbon-14. Most radionuclides can be removed using established purification methods; however, the carbon-14 in irradiated graphite is chemically indistinguishable from carbon-12. Thus purification methods are not applicable. Fachinger et al. (2006) have demonstrated removal of carbon-14 from irradiated graphite using pyrolysis and oxidation. This thesis presents the further refinement of thermal treatment method in order to offer an optimal graphite waste management solution.
