Wells D.,Nuvia Limited |
Herrick A.,Nuvia Limited |
Klepikov A.,Nuclear Technology Safety Center |
Yakovlev I.,MAEC Kazatomprom |
And 2 more authors.
Proceedings of the International Conference on Radioactive Waste Management and Environmental Remediation, ICEM | Year: 2011
Kazakhstan's BN-350 fast reactor was shut down in 1999 and is in the process of being decommissioned in preparation for Safestore. A key achievement during 2010 was the removal of the known inventory of spent nuclear fuel (SNF) from the reactor to off-site secure storage. As a complementary activity, surveys of areas of the fuel discharge route where it was considered possible for fuel pins or fuel residues to have collected over many years of operation were also arranged to confirm that no significant amounts of fuel of remained at the plant. This paper reports on the remote radiation and visual (video and still photograph) surveys undertaken on the following areas of the BN-350 spent fuel route: The waste storage repository or vault underneath the Post-Irradiation Examination Hot Cell in which non-destructive and destructive examination of irradiated fuels was undertaken throughout the operating life of the reactor. The fuel transfer and washing cells within which irradiated fuel sub-assemblies were processed and residual sodium coolant removed. The fuel storage ponds. Man access to several areas (particularly the waste vault) was not possible due to very high radiation levels from stored βγ-active wastes or residual contamination and in these cases specially engineered remote camera and radiation detector deployment systems were developed and used. In other areas, such as the ponds, limited man access was possible under prepared and controlled conditions. The survey results, together with associated radiochemical and radiation dose rate analysis, demonstrated that, despite several recorded handling incidents during Hot Cell operations, the maximum estimated amounts of nuclear material which could remain at BN-350 were sufficiently low to give no major safeguards concerns. The data also provides key information to guide decommissioning and dismantling planning in the future. Copyright © 2011 by ASME.
Kulsartov T.,RAS Research Center Kurchatov Institute |
Tazhibayeva I.,RAS Research Center Kurchatov Institute |
Gordienko Yu.,RAS Research Center Kurchatov Institute |
Chikhray E.,SRI ETP Al Farabi KazNSU |
And 3 more authors.
Fusion Science and Technology | Year: 2011
Lithium-based oxide ceramics are considered as the candidate materials for solid breeders of future fusion reactors' blankets. Breeder's goal is effective, safe and reliable production of tritium as a result of lithiumneutron reactions. Main candidates as a breeder material are Li2O, Li 4SiO4, Li2TiO3 and Li 2ZrO3, which are able to keep their physical-chemical properties despite of lithium burn-up. Lithium metatitanate Li 2TiO3 attracts the great attention due to its chemical stability and high speed of tritium release under low temperatures (from 200 to 400°C). This paper contains the results of the studies on tritium and helium release from the samples of irradiated lithium ceramics Li2TiO 3.
Adsley I.,Nuvia Limited |
Tur Y.,Nuclear Technology Safety Center |
Wells D.,Nuvia Limited
Proceedings of the International Conference on Radioactive Waste Management and Environmental Remediation, ICEM | Year: 2013
The paper relates to the determination of the amount of nuclear material contained in a closed, concrete lined vault at the Aktau fast breeder reactor in Kazakhstan. This material had been disposed into the vault after examination in an experimental hot cell directly above the vault. In order to comply with IAEA Safeguards requirements it was necessary to determine the total quantities of nuclear materials - enriched uranium and plutonium - that were held with Kazakhstan. Although it was possible to determine the inventory of all of the accessible nuclear material - the quantity remaining in the vault was unknown. As part of the Global Threat Reduction Programme the UK Government funded a project to determine the inventory of these nuclear materials in this vault. This involved drilling three penetrations through the concrete lined roof of the vault; this enabled the placement of lights and a camera into the vault through two penetrations; while the third penetration enabled a lightweight manipulator arm to be introduced into the vault. This was used to provide a detailed 3D mapping of the dose rate within the vault and it also enabled the collection of samples for radionuclide analysis. The deconvolution of the 3D dose rate profile within the vault enabled the determination of the gamma emitting source distribution on the floor and walls of the vault. The samples were analysed to determine the fingerprint of those radionuclides producing the gamma dose - namely 137Cs and 60Co - to the nuclear materials. The combination of the dose rate source terms on the surfaces of the vault and the fingerprint then enabled the quantities of nuclear materials to be determined. The project was a major success and enabled the Kazakhstan Government to comply with IAEA Safeguards requirements. It also enabled the UK DECC Ministry to develop a technology of national (and international) use. Finally the technology was well received by IAEA Safeguards as an acceptable methodology for future studies. Copyright © 2013 by ASME.
Kulsartov T.V.,RAS Research Center Kurchatov Institute |
Gordienko Y.N.,RAS Research Center Kurchatov Institute |
Tazhibayeva I.L.,RAS Research Center Kurchatov Institute |
Kenzhin E.A.,Shakarim Semey State University |
And 4 more authors.
Journal of Nuclear Materials | Year: 2013
The results of tritium and helium gas release from lithium ceramics samples Li2TiO3 irradiated at the WWR-K reactor (Almaty, Kazakhstan) and from beryllium samples irradiated at the BN-350 reactor (Aktau, Kazakhstan) and the IVG.1M reactor (Kurchatov, Kazakhstan) are presented. Experimentally obtained thermal desorption (TDS) spectra have shown that the dependence of tritium release from lithium ceramics has a complicated behavior and to a large extent depends on lithium ceramics type. Nevertheless, it was found that the total amount of tritium released from all types of lithium ceramics has the same order of magnitude, equal to about 1011 Bq/kg. It was found that in the temperature range from 523 K to 1373 K the process of tritium release from lithium ceramics involves volume diffusion and thermoactivated tritium release from the accumulation centers generated under irradiation. TDS of beryllium samples enables us to obtain characteristics of tritium and helium release during linear heating, to determine integrated quantities of generated helium and tritium, and to determine parameters of release processes. © 2013 Elsevier B.V. All rights reserved.