Dissertations / Theses: 'Storage of spent nuclear fuel'

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ROMANATO, LUIZ S. "Armazenagem de combustivel nuclear queimado." reponame:Repositório Institucional do IPEN, 2005. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11204.

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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares, IPEN/CNEN-SP

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ROMANATO, LUIZ S. "Estudo de um casco nacional e sua instalacao para armazenagem seca de combustivel nuclear queimado gerado em reatores PWR." reponame:Repositório Institucional do IPEN, 2009. http://repositorio.ipen.br:8080/xmlui/handle/123456789/9476.

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Tese (Doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

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Hartnick, Megan Donna. "Evaluation of nuclear spent fuel dry storage casks and storage facility designs." Master's thesis, University of Cape Town, 2017. http://hdl.handle.net/11427/25279.

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Koeberg Nuclear Power Station (KNPS) is the only nuclear power station in Africa and it stores its spent nuclear fuel (SNF) onsite in the spent fuel pool (SFP). Additional aged SNF assemblies are stored in dry storage casks in a facility located on the KNPS site. This minor research dissertation aims at evaluating various dry storage cask found in open literature. The dissertation provides an overview of cask types, heat transfer, radiation shielding and storage facility types. Specific criteria are required in the selection of casks and the storage facility to house the casks on site. The selection criteria for casks and the storage facility were determined and technically evaluated in this dissertation. The selected casks were evaluated in terms of SNF criticality, radiation shielding, decay heat removal and heat transfer. Other aspects also determined by calculation were the seismic stability of casks and the cask footprint. The results obtained show the relationship of the spent fuel (SF) packing density between the different casks. Different shielding materials are used in the casks and it aided the heat transfer process to take place with some casks having additional features which included cooling fins and air vents for adequate cooling of the SNF. Through these some trends could be identified which could be used in the selection or design of new storage casks. Recommendations for further study are to evaluate a greater range of casks to verify and improve upon the relationship of evaluated parameters that were shown in the technical evaluation. These casks should all have similar means of maintaining sub-criticality, shielding and heat removal in order to generate comparable results.

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Chen, Xinhui 1966. "Thermal analysis of dry spent fuel transportation and storage casks." Thesis, Massachusetts Institute of Technology, 1996. http://hdl.handle.net/1721.1/38395.

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Khoza, Best. "Physics and engineering aspects of South Africa's proposed dry storage facility for spent nuclear fuel." Master's thesis, Faculty of Engineering and the Built Environment, 2019. https://hdl.handle.net/11427/31697.

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The continual increase in electricity dependence for the advancement of society has led to increased demand in electricity globally. This increased demand, among other things such as global warming interventions and energy security have encouraged the need to diversify electricity generation sources. Civilian use of nuclear power dates back to the 1950s. The United States of America and France are currently leading with the highest nuclear power generation in the world, generating 101 GWe and 63 GWe, respectively. Several countries such as China and the United Arab Emirates have committed to new nuclear build in order to increase their nuclear power generation capacities. Standing against the prospects of growth of the nuclear power industry are technical and nontechnical challenges. These include proliferation risk, safety, high capital costs and high-level waste management. Most spent nuclear fuel from power reactors is currently stored in the spent fuel pools on reactor sites, and some have been reprocessed. It is estimated that about 32% (370 000 tons of Heavy Metal) of the total spent fuel generated from power reactors have been reprocessed up to date. With most of the spent fuel pools filling up, alternative interim and long term disposal of spent nuclear fuel solutions have been under investigation from as early as the 1970s. South Africa has planned an interim dry storage facility for the spent nuclear fuel to be established at the existing Koeberg power station. The interim dry storage facility will make use of HI-STAR 100 multi-purpose casks to store spent nuclear fuel until the country decides on final disposal solution. There are many aspects that are critical to safe, efficient and cost-effective long term storage of spent nuclear fuel. Some of the physics and engineering aspects concerning dry storage facilities are briefly discussed. The aspects presented here are: radiation containment, spent fuel, sub-criticality, decay heat removal, site location aspects, response to seismic events, cask corrosion, transportation infrastructure, operability and monitoring. The study of the three existing dry cask storages from the USA, Hungary and Belgium gives an overview of the dry cask technology in use today. These presentations are based on publicly available reliable information. The proposed dry storage facility at Koeberg will be in the existing power station footprint using the HI-STAR 100 casks. The decision to have the proposed dry storage facility at Koeberg will minimise related licence applications and part of security installations as the site already has some security. The location of the facility in the power station’s footprint also allows for cost-effective and safe transportation of casks from the reactor building to the proposed facility. The modularity aspect of the dry cask storage facility at MV Paks in Hungary should also be employed at Koeberg to allow for more storage. This will cater for additional casks that may need to be stored if more nuclear power plants are procured in the future. South Africa’s air traffic around the Western Cape is not as congested as Belgium’s. There is, therefore, no need for the casks to be housed in concrete buildings like Doel’s. Most of Koeberg’s high-level waste would have had a longer cooling time in the pools compared to the minimum cooling time required for the chosen cask technology. This will provide a conservative, safe approach for Koeberg’s facility. Dry cask storage technology has provided a reliable interim dry storage solution for several countries. Despite uncertainties for long term disposal options, the proposed dry cask storage facility at Koeberg is a suitable interim storage alternative for South Africa to allow continuous operation of the plant. This conclusion is based on the physics and engineering aspects that have been presented in this minor dissertation.

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Hugo, Bruce Robert. "Modeling evaporation from spent nuclear fuel storage pools| A diffusion approach." Thesis, Washington State University, 2016. http://pqdtopen.proquest.com/#viewpdf?dispub=10043059.

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Accurate prediction of evaporative losses from light water reactor nuclear power plant (NPP) spent fuel storage pools (SFPs) is important for activities ranging from sizing of water makeup systems during NPP design to predicting the time available to supply emergency makeup water following severe accidents. Existing correlations for predicting evaporation from water surfaces are only optimized for conditions typical of swimming pools. This new approach modeling evaporation as a diffusion process has yielded an evaporation rate model that provided a better fit of published high temperature evaporation data and measurements from two SFPs than other published evaporation correlations. Insights from treating evaporation as a diffusion process include correcting for the effects of air flow and solutes on evaporation rate. An accurate modeling of the effects of air flow on evaporation rate is required to explain the observed temperature data from the Fukushima Daiichi Unit 4 SFP during the 2011 loss of cooling event; the diffusion model of evaporation provides a significantly better fit to this data than existing evaporation models.

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Fairlie, Ian. "Radioactive waste : international examination of storage and reprocessing of spent fuel." Thesis, Imperial College London, 1997. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.268029.

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Burns, Joe 1966. "On selection and operation of an international interim storage facility for spent nuclear fuel." Thesis, Massachusetts Institute of Technology, 2004. http://hdl.handle.net/1721.1/16642.

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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004.
Includes bibliographical references.
This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.
Disposal of post-irradiation fuel from nuclear reactors has been an issue for the nuclear industry for many years. Most countries currently have no long-term disposal strategy in place. Therefore, the concept of an intermediate nuclear spent fuel storage facility has been introduced as a method of temporarily storing the spent fuel in a central location until long-term disposal of the spent nuclear fuel is made available. General criteria that can be used to compare potential international sites for an intermediate nuclear spent fuel storage facility have been identified and elucidated. Those criteria were then utilized to compare four potential international intermediate nuclear spent fuel storage facility (IINSFSF) sites. Two of the sites are in Russia (one in the area of the old nuclear city of Krasnoyarsk-26 currently known as Zheleznogorsk and one on Sakhalin Island in the area of the town of Kholmsk), one is in China (in the area of the town of Xilinhot in the Nei Mongol province) and one in Australia (in the area of the city of Meekatharra in Western Australia). Safety and safeguard regulations for nuclear facilities were reviewed and appropriate portions that could be applied to a potential IINSFSF are recommended. An analysis was conducted to determine legal issues pertinent to an IINSFSF and a brief, limited overview of the most important legal issues is presented. The effects that nuclear fuels subjected to higher burnups (than practiced now) will have on dry cask storage was examined and recommendations for storage strategies are proposed.
(cont.) The selected criteria involve the areas of Geological Suitability, Seismic Stability, Land Area Suitability, Site Infrastructure Suitability, Transportation Infrastructure Suitability, Meteorological Suitability, Willingness of the Host Nation and Population Density. Application of the criteria to the suggested sites revealed that Krasnoyarsk - 26 is the best alternative. This is mainly due to the willingness of the host nation of Russia to accept this type of facility. Krasnoyarsk - 26 also rates as the best site with respect to the criteria of geological suitability and seismic suitability. Without consideration for the willingness of the host nation, Meekatharra would be the ideal site. Xilinhot was evaluated as the third best alternative followed by the Sakhalin Island site of Kholmsk. The legal issue that would be of most concern to an IINSFSF would be potential liability. It would be best if the host nation were a signatory of an international treaty limiting the liability of the IINSFSF operator. Of the two major international nuclear liability treaties in existence the one preferable is the Paris Convention. Economics are driving nuclear power plants in the United States to look to implement more highly enriched fuels to achieve higher burnupsHow these higher burnup spent fuels will affect dry cask storage of spent fuels at reactor sites should be examined. To determine this, the decay heat output of higher burnup spent fuels was compared to the storage capacity of a typical dry cask storage system ...
by Joe Burns.
S.M.

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Sommer, Christopher. "Fuel cycle design and analysis of SABR subrcritical advanced burner reactor /." Thesis, Atlanta, Ga. : Georgia Institute of Technology, 2008. http://hdl.handle.net/1853/24720.

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Fortkamp, Jonathan C. "Characterization of the radiation environment for a large area interim spent nuclear fuel storage facility /." The Ohio State University, 1999. http://rave.ohiolink.edu/etdc/view?acc_num=osu1488188894437725.

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Romanato, Luiz Sergio. ""Armazenagem de combustível nuclear queimado"." Universidade de São Paulo, 2005. http://www.teses.usp.br/teses/disponiveis/85/85131/tde-15052006-220434/.

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Quando um país se torna auto-suficiente em uma parte do ciclo nuclear, quanto à produção de combustível que será usado em suas centrais nucleares para a geração de energia, precisa voltar sua atenção para a melhor forma de armazenar este combustível após a sua utilização. A armazenagem do combustível nuclear queimado é uma prática necessária e utilizada nos dias atuais em todo o mundo como temporária, tanto por países que não têm definido o plano de destinação final, isto é, o repositório definitivo, como também por aqueles que já o possuem. Existem dois aspectos principais que envolvem os combustíveis queimados: um referente à armazenagem do combustível nuclear queimado destinado ao reprocessamento e o outro ao que será enviado para deposição final quando o sítio de deposição definitiva estiver definido, corretamente localizado, adequadamente caracterizado quanto aos diversos aspectos técnicos, e licenciado. Este último aspecto pode envolver décadas de estudos por causa das definições técnicas e normativas em um dado país. No Brasil, o interesse está voltado para a armazenagem dos combustíveis queimados que não serão reprocessados. Este trabalho analisa os tipos possíveis de armazenagem, o panorama internacional e a possível proposta para a futura construção de um sítio de armazenagem temporária no país.
When a country becomes self-sufficient in part of the nuclear cycle, as production of fuel that will be used in nuclear power plants for energy generation, it is necessary to pay attention for the best method of storing the spent fuel. Temporary storage of spent nuclear fuel is a necessary practice and is applied nowadays all over the world, so much in countries that have not been defined their plan for a definitive repository, as well for those that already put in practice such storage form. There are two main aspects that involve the spent fuels: one regarding the spent nuclear fuel storage intended to reprocessing and the other in which the spent fuel will be sent for final deposition when the definitive place is defined, correctly located, appropriately characterized as to several technical aspects, and licentiate. This last aspect can involve decades of studies because of the technical and normative definitions at a given country. In Brazil, the interest is linked with the storage of spent fuels that won't be reprocessed. This work analyses possible types of storage, the international panorama and a proposal for future construction of a spent nuclear fuel temporary storage place in the country.

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Solis, Dominic (Dominic R. ). "COMSOL finite-element analysis : residual stress measurement of representative 304L/308L weld in spent fuel storage containers." Thesis, Massachusetts Institute of Technology, 2014. http://hdl.handle.net/1721.1/97965.

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Thesis: S.B., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014.
Cataloged from PDF version of thesis.
Includes bibliographical references (pages 30-31).
The ultimate storage destination for spent nuclear fuel in the United States is currently undecided. Spent fuel will be stored indefinitely in dry cask storage systems typically located on-site at the reactor or at a dedicated independent spent fuel storage installation (ISFSI). Since these canisters were not originally designed or qualified for indefinite storage, there is a need to quantify the length of time they will be viable for storing spent fuel. Stress corrosion cracking (SCC) is a concern in these canisters if they are exposed to an aqueous, chloride-containing film. Canisters are fabricated using a concrete overpacking, along with austenitic stainless steel on the inside which is welded together. One factor that would significantly impact SCC behavior inside these canister welds, if the proper conditions developed such that SCC occurred, is the tensile residual stress profile. As the highest residual stresses are present in the welds and their heat-affected zones (HAZ), it would be useful to investigate their influence by predicting the residual stress profile in the container. These data will support further research into the life expectancy of these canisters and the possible ways in which they might fail due to SCC. Residual stress data for nuclear waste canisters are scarce. Without experimental measurements, initial insight must be attained through computational analysis using finite-element analysis (FEA) packages such as COMSOL. Using a representative 304L/308L weld plate as a model in COMSOL, predicted residual stress shows some agreement with expected trends: high tensile stresses in the weld/ HAZ regions and compressive stresses in the surrounding material. Hardness tests show trends similar to the hardening profiles that were created after the weld simulation. Additionally, the thermal model may offer insight in predicting the HAZ profiles in the weld. While the 2D model is simplified and would benefit from further refinement and validation, preliminary results suggest that FEA could be used for residual stress measurement predictions.
by Dominic Solis.
S.B.

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Sharifi, Brojerdi Fatemeh. "Analysis of Seismic Data Acquired at the Forsmark Site for Storage of Spent Nuclear Fuel, Central Sweden." Doctoral thesis, Uppsala universitet, Geofysik, 2015. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-251621.

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The Forsmark area, the main study area in this thesis, is located about 140 km north of Stockholm, central Sweden. It belongs to the Paleoproterozoic Svecokarelian orogen and contains several major ductile and brittle deformation zones including the Forsmark, Eckarfjärden and Singö zones. The bedrock between these zones, in general is less deformed and considered suitable for a nuclear waste repository. While several site investigations have already been carried out in the area, this thesis focuses primarily on (i) re-processing some of the existing reflection seismic lines to improve imaging of deeper structures, (ii) acquiring and processing high-resolution reflection and refraction data for better characterization of the near surface geology for the planning of a new access ramp, (iii) studying possible seismic anisotropy from active sources recorded onto sparse three-component receivers and multi-offset-azimuth vertical seismic profiling data (VSP). Reflection seismic surveys are an important component of these investigations. The re-processing helped in improving the deeper parts (1-5 km) of the seismic images and allowing three major deeper reflections to be better characterized, one of which is sub-horizontal while the other two are dipping moderately. These reflections were attributed to originate from either dolerite sills or brittle fault systems. First break traveltime tomography allowed delineating an undulating bedrock-surface topography, which is typical in the Forsmark area. Shallow reflections imaged in 3D, thanks to the acquisition design were compared with existing borehole data and explained by fractured or weak zones in the bedrock. The analysis of seismic anisotropy indicates the presence of shear-wave splitting due to transverse isotropy with a vertical symmetry axis in the uppermost hundreds of meters of crust. Open fractures and joints were interpreted to be responsible for the large delays observed between the transverse and radial components of the shear-wave arrivals, both on surface and VSP data.

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Weirich, Timothy Douglas. "Evaluating the Potential for Atmospheric Corrosion of 304 Stainless Steel Used for Dry Storage of Spent Nuclear Fuel." The Ohio State University, 2019. http://rave.ohiolink.edu/etdc/view?acc_num=osu1557098372186951.

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Mohun, Ritesh. "Raman spectroscopy for the characterization of defective spent nuclear fuels during interim storage in pools." Thesis, Aix-Marseille, 2017. http://www.theses.fr/2017AIXM0288/document.

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Une signature spécifique des dommages d’irradiation dans le dioxyde d’uranium, le combustible nucléaire le plus utilisé, dénommé « triplet de défauts » a été récemment mis en évidence par spectroscopie Raman. Ce travail vise à savoir comment cette signature peut être utilisée afin de caractériser les combustibles nucléaires irradiés qui sont entreposés sous eau. Pour cela, trois études à effets séparés sont menées. Tout d’abord, une expérience d’irradiation aux électrons montre que le triplet de défauts est dû à des interactions balistiques et est associé aux déplacements dans le sous-réseau d’uranium. Après l’irradiation aux électrons, l’échantillon d’UO2 s’oxyde de manière accélérée, ce qui a été attribué à la migration des lacunes d’oxygène créées par l’irradiation vers la surface. Ensuite, la cinétique de formation du triplet de défauts dans de l’UO2 exposé à des environnements inerte (Ar) et réactif (eau aérée) a été mesurée grâce à un dispositif Raman in-situ. Dans tous les cas, la cinétique peut être décrite par un modèle d’impact direct, mais avec des coefficients numériques différents. Enfin, de manière à simuler le combustible irradié industriel en laboratoire, l’étude de différents composés d’oxydes mixtes a montré le rôle du dopage chimique sur la formation du triplet de défauts. Ces informations seront mises à profit dans les études futures de combustibles défectueux entreposés sous eau
A specific signature characteristic of irradiation damages in uranium dioxide, the most used nuclear fuel, referred as « triplet defect bands» has recently been evidenced by Raman Spectroscopy. The objective of this study is to determine how this signature can be used to characterize actual spent nuclear fuel stored in pools. For that purpose, three separate effect studies were carried out. Firstly, an electron irradiation experiment shows that the triplet defect bands are due to ballistic interactions and result from the formation displacements in the uranium sub-lattice. Post electron irradiation, the enhanced oxidation of UO2 samples is observed and attributed to the migration of irradiation induced oxygen vacancies towards the surface. The formation kinetics of the triplet defect bands in UO2 when exposed to an inert (Ar) and a reactive (aerated water) contact medium is then investigated through the use of an in-situ Raman installation. Both kinetics can be fitted using a direct impact model, but with different numerical values. Finally, to simulate actual spent nuclear fuels in laboratory conditions, the study of different mixed oxide compounds shows that chemical doping impacts the apparition of the Raman triplet defect bands. The experimental results obtained in this work will be used as complementary data for the interpretation of Raman results of actual defective spent nuclear fuels stored in pool conditions

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Mičian, Peter. "Bezpečnost skladování paliva ve vodním prostředí." Master's thesis, Vysoké učení technické v Brně. Fakulta elektrotechniky a komunikačních technologií, 2018. http://www.nusl.cz/ntk/nusl-377026.

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This diploma thesis deals with storing the spent nuclear fuel and reviewing its safety. The theoretical part analyzes the processes taking place while the fuel is being used, such as fission, isotopic changes, fission gas release, cracking, swelling and densification of fuel pellet. The thesis is also focused on handling the spent fuel and on the way it makes from the reactor, through the spent fuel pool, the transportation, various kinds of storing, till the reprocessing and final deep geological repository. Furthermore, this part of the thesis briefly discusses computing code MCNP, its main characteristics, input files and using. The practical part of the work is focused on creating the model of the spent fuel pool located next to the nuclear reactor WWER 440/V213. This type was chosen, because it is the most used type of nuclear reactor in Czech Republic and Slovakia. With the help of the code MCNP, the multiplication factor of the main configurations of the fuel in the pool was calculated, and then the required safety regulations to ensure sufficient subcriticality, so its safety, were checked. Next, several analysis were performed using this model. These analyses were concerning the temperature of coolant, fuel and the use of various nuclear data libraries. In the future this model can be used to realize new analyses with new kinds of fuels, materials and data libraries.

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Hlatký, Pavel. "Studium tepelných a fyzikálních vlastností skladovacích kontejnerů pro použité jaderné palivo." Master's thesis, Vysoké učení technické v Brně. Fakulta strojního inženýrství, 2011. http://www.nusl.cz/ntk/nusl-229829.

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This work deals with questions of spent fuel storage casks thermal and physical properties investigation. Foundations of mathematics which are necessary for describing field of temperature are included. The work itself contains calculation methods which are split into two parts. The first one deals with simplified analytic solution and the second part solves the whole problem by the numerical computation.

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RODRIGUES, ANTONIO C. I. "Estudo e projeto de novos cestos com boro para o armazenamento de elementos combustíveis queimados do reator IEA-R1." reponame:Repositório Institucional do IPEN, 2016. http://repositorio.ipen.br:8080/xmlui/handle/123456789/26823.

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O reator de pesquisas IEA-R1 opera em regime de 40 h semanais à potência de 4,5 MW. Nestas condições, os cestos disponíveis para o armazenamento dos elementos combustíveis irradiados possuem menos de metade da sua capacidade inicial. Assim, nestas condições de operação, teremos apenas cerca de seis anos de capacidade para armazenamento. Considerando que a vida útil desejada do IEA-R1 seja de pelo menos mais 20 anos, será necessário aumentar a capacidade de armazenamento de combustível irradiado. Dr. Henrik Grahn, especialista da Agência Internacional de Energia Atômica sobre o armazenamento molhado (em piscinas de estocagem), ao visitar o reator IEA-R1 (setembro/2012) fez algumas recomendações. Entre elas, a concepção e instalação de cestos fabricados com aço inoxidável borado e internamente revestidos com uma película de alumínio, de modo que a corrosão dos elementos combustíveis não ocorresse. Após uma revisão da literatura sobre opções de materiais disponíveis para esse tipo de aplicação chegamos ao BoralcanTM fabricado pela 3M devido suas propriedades. Este trabalho apresenta estudos sobre a análise de criticalidade com o código computacional MCNP-5 utilizando duas bibliotecas americanas de dados nucleares avaliados (ENDF/B-VI e ENDF/B-VII) comparativamente. Estas análises demonstraram a possibilidade de dobrar a capacidade de armazenamento de elementos combustíveis, no mesmo espaço ocupado pelos cestos atuais, atendendo a demanda do reator de pesquisas IEA-R1 e também satisfazendo os requisitos de segurança da Comissão Nacional de Energia Nuclear (CNEN) e da Agência Internacional de Energia Atômica (IAEA).
Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

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Sommer, Christopher Michael. "Subcritical transmutation of spent nuclear fuel." Diss., Georgia Institute of Technology, 2011. http://hdl.handle.net/1853/41205.

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A series of fuel cycle simulations were performed using CEA's reactor physics code ERANOS 2.0 to analyze the transmutation performance of the Subcritical Advanced Burner Reactor (SABR). SABR is a fusion-fission hybrid reactor that combines the leading sodium cooled fast reactor technology with the leading tokamak plasma technology based on ITER physics. Two general fuel cycles were considered for the SABR system. The first fuel cycle is one in which all of the transuranics from light water reactors are burned in SABR. The second fuel cycle is a minor actinide burning fuel cycle in which all of the minor actinides and some of the plutonium produced in light water reactors are burned in SABR, with the excess plutonium being set aside for starting up fast reactors in the future. The minor actinide burning fuel cycle is being considered in European Scenario Studies. The fuel cycles were evaluated on the basis of TRU/MA transmutation rate, power profile, accumulated radiation damage, and decay heat to the repository. Each of the fuel cycles are compared against each other, and the minor actinide burning fuel cycles are compared against the EFIT transmutation system, and a low conversion ratio fast reactor.

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Nimander, Fredrik. "Investigation of Spent Nuclear Fuel Pool Coolability." Thesis, KTH, Reaktorteknologi, 2011. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-42440.

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The natural catastrophe at Fukushima Dai-ichi 2011 enlightened the nuclear community. This master thesis reveals the non-negligible risks regarding the short term storage of spent nuclear fuel. The thesis has also investigated the possibility of using natural circulation of air in a passive safety system to cool the spent nuclear fuel pools. The results where conclusive: The temperature difference between the heated air and ambient air is far too low for natural circulation of air to remove any significant amount of heat from the spent nuclear fuel pool in a worst case scenario. Air, as with any gas, has too low density and a specific heat too low to be able to remove the heat generated by spent nuclear fuel shortly after it has been removed from the reactor core. The author does not deny the possibility of slightly prolonging the boiling with other designs. The author does however suggest other possibilities to prolong cooling with the conclusion that large enough spent fuel pools would constitute the simplest solution.

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Yee, Shannon K. "Nuclear Fuel Cycle Modeling Approaches For Recycling And Transmutation Of Spent Nuclear Fuel." The Ohio State University, 2008. http://rave.ohiolink.edu/etdc/view?acc_num=osu1213905425.

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Шевченко, Валентина Владимировна, Алла Викторовна Дон, and Татьяна Геннадиевна Кононова. "Проблемы современной электроэнергетики, пути ее развития и оценка источников электроэнергии." Thesis, Accent Graphics Communications & Publishing, Canada, 2019. http://repository.kpi.kharkov.ua/handle/KhPI-Press/46945.

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Reese, Drew A. (Drew Amelia). "Dependence of transuranic content in spent fuel on fuel burnup." Thesis, Massachusetts Institute of Technology, 2007. http://hdl.handle.net/1721.1/41692.

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Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007.
Includes bibliographical references (p. 33).
As the increasing demand for nuclear energy results in larger spent fuel volume, implementation of longer fuel cycles incorporating higher burnup are becoming common. Understanding the effect of higher burnup on the spent fuel composition and radioactive properties is essential to ensure that spent fuel receives proper cooling in storage before it is sent to a disposal site or proper treatment and reprocessing if its useful content is to be extracted prior to disposal. Using CASMO-4, a standard Westinghouse 4-loop pressurized water reactor model was created and simulated with a three batch fuel cycle. U-235 enrichment was adjusted to achieve fuel burnups of 30, 50, 70 and 100 MWD per kg of initial uranium. These burnups demanded reload enrichments of 3.15%, 4.63%, 6.26% and 9.01% U-235 w/o respectively. The resultant spent fuel transuranic isotopic compositions were then provided as input into ORIGEN to study the decay behavior of the spent fuel. It was found that when burnup increased from 30 MWD/kg to 100 MWD/kg, the activity more than doubled due to the decreased Pu-241 content and the increased Np-239 presence. More importantly, the activity per MWD significantly decreased despite absolute increases in unit mass. The net result is that the half-life of high burnup fuels is greatly increased in comparison to low burnup fuels for the first decade of life. Beginning from day 14 after shutdown and until 10 years later, the 100 MWD/kg fuel has a half-life of 129 days while the 30 MWD/kg spent fuel has a half life of 5 days. Previous work has suggested that different trends dominate decay behavior from years 10 to 100 years following discharge.
by Drew A. Reese.
S.B.

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Alajo, Ayodeji Babatunde. "Impact of PWR spent fuel variations on TRU-fueled VHTRS." Thesis, [College Station, Tex. : Texas A&M University, 2007. http://hdl.handle.net/1969.1/ETD-TAMU-2556.

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Li, Junyi. "Stability of studtite under spent nuclear fuel repository conditions." Thesis, KTH, Skolan för kemi, bioteknologi och hälsa (CBH), 2019. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-260076.

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Sihm, Kvenangen Karen. "Alternative measuring approaches in gamma scanning on spent nuclear fuel." Thesis, Uppsala universitet, Institutionen för fysik och astronomi, 2007. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-162795.

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In the future, the demand for energy is predicted to grow and more countries plan to utilize nuclear energy as their source of electric energy. This gives rise to many important issues connected to nuclear energy, such as finding methods that can verify that the spent nuclear fuel has been handled safely and used in ordinary power producing cycles as stated by the operators. Gamma ray spectroscopy is one method used for identification and verification of spent nuclear fuel. In the specific gamma ray spectroscopy method called gamma scanning the gamma radiation from the fission products Cs-137, Cs-134 and Eu-154 are measured in a spent fuel assembly. From the results, conclusions can be drawn about the fuels characteristics. This degree project examines the possibilities of using alternative measuring approaches when using the gamma scanning method. The focus is on examining how to increase the quality of the measured data. How to decrease the measuring time as compared with the present measuring strategy, has also been investigated. The main part of the study comprises computer simulations of gamma scanning measurements. The simulations have been validated with actual measurements on spent nuclear fuel at the central interim storage, Clab. The results show that concerning the quality of the measuring data the conventional strategy is preferable, but with other starting positions and with a more optimized equipment. When focusing on the time aspect, the helical measuring strategy can be an option, but this needs further investigation.

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Lambert, Hugues. "Molten salt spectroscopy and electrochemistry for spent nuclear fuel treatment." Thesis, University of Manchester, 2017. https://www.research.manchester.ac.uk/portal/en/theses/molten-salt-spectroscopy-and-electrochemistry-for-spent-nuclear-fuel-treatment(89862aa1-a98d-4b5f-9052-91cc9dd4eda3).html.

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Pyroprocessing, via electrorefining in a molten salt bath, is a promising treatment route for spent nuclear fuel reprocessing. In order to implement such a technology and ensure its safe operation it is vital to develop on-line techniques to understand and monitor the molten salt and its contents. These tools are technically challenging because of the high temperatures and corrosive environment experienced in molten salt media. Electrochemical, spectroscopic and spectroelectrochemical methods were developed and used to study actinide and fission product behaviour in molten LiCl-KCl eutectic. A spectroscopic furnace was designed and supporting methodology developed in order to allow the acquisition of reproducible quantitative data. The apparatus monitored the precipitation of NdCl3 by the addition of Li2CO3 and PrCl3 by the addition of Li2O in LiCl-KCl eutectic. The precipitates formed were identified as the respective LnOCl. In order to probe actinide behaviour in this hygroscopic medium, dry actinides chlorides were synthesised. The oxidation of uranium metal by BiCl3 in LiCl-KCl eutectic yielded UCl3 while neptunium and plutonium were prepared as Cs2AnCl6 via precipitation in concentrated aqueous HCl by addition of CsCl. The molar extinction coefficients for U(III), U(IV), Np(IV) and Pu(III) were obtained in LiCl-KCl eutectic at 450 áμC. The study of the Np(IV)/Np(III) couple via spectroelectrochemical techniques, enabled the determination of the Np(III) molar extinction coefficients. Uranium was studied in LiCl-KCl eutectic using square wave voltammetry, cyclic voltammetry and chronoabsorptometry. The electrochemical techniques benchmarked the results obtained by spectroelectrochemistry. The results from the different techniques were compared to and explained by determining the Gibbs energy and activation energy of U(III) and U(IV). It was concluded that all the mentioned techniques are suitable for the study of high temperature molten chlorides. Because of its capacity to gather numerous data parameters while minimising the number of experiments required and the quantity of material needed, spectroelectrochemical methods were highlighted as the most promising technique for future studies of radionuclides in high temperature melts.

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Delandar, Arash Hosseinzadeh. "Modeling defect structure evolution in spent nuclear fuel container materials." Doctoral thesis, KTH, Materialteknologi, 2017. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-206175.

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Materials intended for disposal of spent nuclear fuel require a particular combination of physical and chemical properties. The driving forces and mechanisms underlying the material’s behavior must be scientifically understood in order to enable modeling at the relevant time- and length-scales. The processes that determine the mechanical behavior of copper canisters and iron inserts, as well as the evolution of their mechanical properties, are strongly dependent on the properties of various defects in the bulk copper and iron alloys. The first part of the present thesis deals with precipitation in the cast iron insert. A nodular cast iron insert will be used as the inner container of the spent nuclear fuel. Precipitation is investigated by computing effective interaction energies for point defect pairs (solute–solute and vacancy–solute) in bcc iron using first-principles calculations. The main considered impurities in the iron matrix include 3sp (Si, P, S) and 3d (Cr, Mn, Ni, Cu) solute elements. By computing interaction energies possibility of formation of different second phase particles such as late blooming phases (LBPs) in the cast iron insert is evaluated. The second part is devoted to the fundamentals of dislocations and their role in plastic deformation of metals. Deformation of single-crystal copper under high strain rates is simulated by employing dislocation dynamics (DD) method to examine the effect of strain rate on mechanical properties as well as dislocation microstructure development. Creep deformation of copper canister at low temperatures is studied. The copper canister will be used in the long-term storage of spent nuclear fuel as the outer shell of the waste package to provide corrosion protection. A glide rate is derived based on the assumption that at low temperatures it is controlled by the climb rate of jogs on the dislocations. Using DD simulation creep deformation of copper at low temperatures is modeled by taking glide but not climb into account. Moreover, effective stresses acting on dislocations are computed using the data extracted from DD simulations.

QC 20170428

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Taniuchi, Hiroaki. "STUDY ON SHIELDING PERFORMANCE OF SPENT FUEL TRANSPORT AND STORAGE PACKAGES." Kyoto University, 1999. http://hdl.handle.net/2433/182372.

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Owen, Paul E. (Paul Edward) 1968. "Waste characteristics of spent nuclear fuel from a pebble bed reactor." Thesis, Massachusetts Institute of Technology, 1999. http://hdl.handle.net/1721.1/9548.

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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1999.
Includes bibliographical references.
A preliminary comparative assessment is made of the spent fuel characteristics and disposal aspects between a high-temperature, gas cooled, reactor with a pebble bed core (PBR) and a pressurized water reactor (PWR). There are three significant differences which impact the disposal characteristics of PBR spent pebble fuel from PWR spent fuel assemblies. Pebble bed fuel has bum-up as high as 100,000 MWD(t)/MTHM and thus, there is significantly less activity and decay heat in the fuel when it is disposed. The large amount of graphite in the waste form leads to a low power density and more waste per unit volume than a typical PWR. Pebble Fuel contains a protective layer of Silicon Carbide. The theoretical spacing of waste packages of spent pebble fuel given its unique characteristics as applied to the conditions of Yucca Mountain is of major concern when determining the cost of disposing of the larger volumes of spent pebble fuel. Graphite is a unique waste form and atypical of waste designated for Yucca Mountain. The interactions of silicon carbide with uranium oxide fuel and its implications to long term storage at the repository are examined. There are three primary conclusions to this thesis. First, the area required to store pebble fuel is less than the area required to store light water reactor spent fuel. Second, graphite has excellent characteristics as a waste form. The waste form of the spent pebble fuel is more robust and will perform better than light water reactor fuel at the United States repository at Yucca Mountain. Third, a secondary phase forms between the layers of silicon carbide and the uranium oxide fuel. The secondary phase retards the release of radionuclides to the environment.
by Paul E. Owen.
S.M.

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Araya, Pablo E. "Design of an experiment that simulates spent nuclear fuel within transport casks." abstract and full text PDF (free order & download UNR users only), 2007. http://0-gateway.proquest.com.innopac.library.unr.edu/openurl?url_ver=Z39.88-2004&rft_val_fmt=info:ofi/fmt:kev:mtx:dissertation&res_dat=xri:pqdiss&rft_dat=xri:pqdiss:1442846.

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Zino, John Frederick. "Analysis of subcritical experiments using fresh and spent research reactor fuel assemblies." Diss., Georgia Institute of Technology, 1999. http://hdl.handle.net/1853/17507.

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Bobolea, Ruxandra. "A Study of Continuous Electrochemical Processing Operation Feasibility for Spent Nuclear Fuel." NCSU, 2009. http://www.lib.ncsu.edu/theses/available/etd-02272009-172349/.

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Several methods of reprocessing are currently available to separate recyclable materials from spent nuclear fuel. Electrochemical processing, also known as pyroprocessing, represents a non-aqueous method of reprocessing that uses high temperature molten-salt based electrochemical technology. This method provides several advantages over conventional aqueous processing with respect to proliferation resistance. With electrochemical processing there is no pure plutonium separation and the presence of large decay heat and high radiation barriers dissuades diversion attempts. As the current electrochemical processing relies on a batch operation, the total throughput of the system is inherently limited and nuclear materials accounting is difficult due to the nonhomogeneous nature of the process. This results in much larger uncertainties in the total amount of material processed compared to the aqueous UREX+ or PUREX processes. Continuous electrochemical processing was considered as a way to address these concerns. The objective of this research was to investigate the feasibility of a continuous electrochemical processing operation to achieve the desired separation performance by using computer based simulation. The conceptual design of the continuous electrochemical processing includes two separate stages in a molten salt medium. First, a pure uranium deposit is collected at a solid cathode during the uranium extraction stage. When the amount of plutonium in electrorefiner becomes comparable or higher than the amount of uranium in the electrorefiner, a liquid cathode is employed to extract both uranium and plutonium in the second stage. In this approach, molten salt, as the material carrier, flows through the electrorefiner while chopped spent fuel is continuously fed into the system. Simulations of electrochemical reactions at the electrode surfaces were based on the kinetic modeling capability of a time-dependent code, REFIN. Based on a screening study performed for the most significant process parameters over a broad range of values, a functional combination of initial uranium and plutonium concentrations at the anode and in the molten salt was determined for continuous operation. This dictated the use of a higher concentration of uranium than plutonium at the anode and a lower concentration of uranium than plutonium in the molten salt. Furthermore, using design of experiment technique for computers, a refinement of initial concentrations was performed to maximize the total throughput and minimize the operational time. The flow velocity profiles and chemical concentration distributions of elements in molten salt have been determined through three dimensional Computational Fluid Dynamics simulations using ANSYS CFX. This approach resulted in the need to evaluate the diffusion layer thickness at the cathode â molten salt interface, an important parameter for the electrochemical process. Computer based simulations of the continuous electrochemical processing concept presented in this study have provided an indication that electrochemical processing could be a viable technology for closing the nuclear fuel cycle.

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Goode, James Bruce. "Transitioning of spent advanced gas reactor fuel from wet to dry storage." Thesis, University of Leeds, 2017. http://etheses.whiterose.ac.uk/20585/.

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Upon removal from a reactor spent nuclear fuel is placed into water storage for a period of time since water provides both a decay heat sink and radiation shielding. In the UK this fuel has traditionally been reprocessed however the decision has been made that this will cease in 2018 in favour of an open fuel cycle and direct disposal to a degological disposal facility, although this is not expected to be available until 2075. Current experience with pond storage has found that fuel can be safely stored in caustic dosed ponds for at least 25 years and therefore the current plan is to extend water storage until around 2040, however how the fuel will be stored for the remaining 35 years is yet to be decided. One option that is being considered is the use of dry storage which is being used successfully in the USA for the storage of fuel from Light Water Reactors. One of the most important factors when utilising dry stores is an effective drying step and this PhD is aiming to look at this area in particular. Commercial drying methods have been developed for use with Zircaloy and aluminium clad fuels however this PhD intends to develop a method that can be used to dry Stainless Steel clad fuels such as those used in the UK's Advanced Gas Reactors (AGR). Since intact AGR fuel has yet to be examined following storage the first part if this thesis looked at preparing samples for later macro scale testing. Samples were treated to induce defects known to be of concern and analysed with the conclusion being that bound water is of no concern for drying operations and can be disregarded when preparing samples. Since AGR is known to have failed by stress corrosion cracking water trapped within microcracked pins is a likely issue. A section of tube was prepared by compression and boiling in magnesium chloride which micro CT imaging found to have images of a similar size to known failures. This tube was prepared into a test piece for later testing. The second part of the thesis looks at macro scale testing using the test piece described above and a similar test piece with a 300 um pinhole. A multipurpose drying rig capable of vacuum drying and flowed gas drying was constructed. In the first phase of drying tests vacuum drying and flowed gas drying were compared using a test piece with a 300 um pinhole and the impact of different cover gas was assessed. vacuum drying was found to be significantly more effective than flowed gas drying with the type of cover gas having little impact. Most importantly this work guided further improvements to the rig. The second phase of this work carried out more detailed drying tests on vacuum drying and flowed gas drying and utilised the pinholed and cracked test piece. The greatest influencing factor on the drying rate (the rate at which water could be removed from inside the test piece) was found to be the water level inside the test piece with broadly similar rates being found in all conditions and with both test piece when the test piece was full. As the water level within the test piece dropped, vacuum drying quickly became more effective. This was not helped by considering energy usage. There was however a feeling that flowed gas drying was limited due to the pipework size. In the final phase of testing the ability to detect the point which all water was removed (the end point) from the test piece using online instrumentation was assessed. Currently a vacuum rebound test is used which is both time consuming and if flowed gas drying is being used requires additional plant equipment. It was found that dew point measurements appeared to give a relatively clear indication of the end point when for both vacuum drying and flowed gas drying. For vacuum drying only the measurement of mass flow and pressure also gave a reasonable indication of end point.

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Casella, Andrew M. "Modeling of molecular and particulate transport in dry spent nuclear fuel canisters." Diss., Columbia, Mo. : University of Missouri-Columbia, 2007. http://hdl.handle.net/10355/4695.

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Thesis (Ph. D.)--University of Missouri-Columbia, 2007.
The entire dissertation/thesis text is included in the research.pdf file; the official abstract appears in the short.pdf file (which also appears in the research.pdf); a non-technical general description, or public abstract, appears in the public.pdf file. Title from title screen of research.pdf file (viewed on November 26, 2007 Vita. Includes bibliographical references.

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Wantz, Olivier. "A study of in-package nuclear criticality in possible Belgian spent nuclear fuel repository designs." Doctoral thesis, Universite Libre de Bruxelles, 2005. http://hdl.handle.net/2013/ULB-DIPOT:oai:dipot.ulb.ac.be:2013/211019.

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About 60 percent of the electricity production in Belgium originates from nuclear power plants. Belgium owns 7 nuclear pressurized water reactors, which are located in two sites: 4 reactors in Doel and 3 reactors in Tihange. Together they have a capacity of approximately 5900 MWe. All these reactors use classical uranium oxide fuel assemblies. Two of them (Doel3, Tihange2) have also accepted a limited number of mixed (uranium and plutonium) oxide fuel assemblies. These mixed fuel assemblies came from the reprocessing of spent uranium oxide fuel assemblies in La Hague (France). The reprocessing of spent fuel gives birth to vitrified high-level waste, and to different isotopes of uranium and plutonium, which can be used in the manufacture of mixed oxide fuel assemblies. Each country producing radioactive waste must find a solution to dispose them safely. The internationally accepted solution is to dispose high-level radioactive waste in a deep and stable geological layer. This seems to be the most secure and environment-friendly way to get rid of the high-level radioactive waste. One of the few stable geological layers, which could accept radioactive waste in Belgium, is the Boom clay layer. Another possible layer is the Ypresian clay layer, but it is not the reference option for the moment. The Boom clay layer is quite thin (about 100 m thick) and is not at a large depth (about 240 m below the ground surface) at the proposed disposal site, beneath the SCK CEN Nuclear Research Centre in Mol. A large number of studies have already been performed on the Boom clay layer, and on the possibility of building a high-level radioactive waste repository in this geological medium. Since 1993, the Belgian government has promulgated a moratorium on the reprocessing of spent uranium oxide fuels in La Hague. Since then, spent fuel assemblies are considered as waste, and ONDRAF/NIRAS (the Belgium Agency for Radioactive Waste and Enriched Fissile Materials) has thus to deal with them as waste. This rises a number of questions on how to deal with this new kind of waste. A solution is to directly dispose these spent fuel assemblies in containers in a repository, just like the other high-level radioactive waste. This repository would be build in the Boom clay layer at a depth of about 240 m beneath the SCK CEN. One of the questions raised by this new kind of waste is: "could the direct disposal of the spent nuclear fuel assemblies lead to nuclear criticality risks in the future?". Nuclear criticality is the ability of a system to sustain a nuclear fission chain reaction. This question was not a key issue with vitrified high-level waste because these do not include fissile uranium and plutonium isotopes, which could lead to a criticality event. The spent fuel repository will be designed in order to totally avoid the occurrence of a criticality event at the closure time. But in the future history of the repository, external events could possibly affect this. These events could maybe lead to criticality inside the repository, and this has also to be avoided. This work tries to answer this question, and to determine how to avoid a long-term criticality event inside the repository. The only complete research work answering this question has been performed in the U.S. for the Yucca Mountain repository but this design is fully different from the Belgian one studied here: for example, the waste are not only spent fuel waste, and the geological layer is volcanic tuff.

The main achievements of this work are:

*A first set of in-package criticality scenarios for different design options for a Belgian spent fuel repository in the Boom clay layer.

*A large number of criticality calculations with different parameters (fuel type, fuel burnup, fuel enrichment, distance between the fuel assemblies, distance between the fuel rods, water fraction inside the overpack) for the different design options.

*A preliminary study of the effects of the spent fuel assemblies isotopic evolution with time on the multiplication factor.

*For the first time, a coupling between the in-package criticality scenarios and the criticality calculations has been performed.
Doctorat en sciences appliquées
info:eu-repo/semantics/nonPublished

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Jacobsson, Staffan, Ane Håkansson, Camilla Andersson, Peter Jansson, and Anders Bäcklin. "A Tomographic Method for Verification of the Integrity of Spent Nuclear Fuel." Uppsala universitet, Tillämpad kärnfysik, 1998. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-200297.

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A tomographic method for experimental investigation of the integrity of usedLWR fuel has been developed. It is based on measurements of the gamma radiation fromthe fission products in the fuel rods. A reconstruction code of the algebraic type has beenwritten. The potential of the technique has been examined in extensive simulationsassuming a gamma-ray energy of either 0.66 MeV (137Cs) or 1.27 MeV (154Eu).The resultsof the simulations for BWR fuel indicate that single fuel rods or groups of rods replacedwith water or fresh fuel can be reliably detected independent of their position in the fuelassembly using 137Cs radiation. For PWR fuel the same result is obtained with the exceptionof the most central positions. Here the more penetrable radiation from 154Eu must be used inorder to allow a water channel to be distinguished from a fuel rod. The results of the simulations have been verified experimentally for a 8x8 BWRfuel assembly. Special equipment has been constructed and installed at the interim storageCLAB. The equipment allows the mapping of the radiation field around a fuel assemblywith the aid of a germanium detector fitted with a collimator with a vertical slit. Theintensities measured in 2 520 detector positions were used as input for the reconstructioncode used in the simulations. The results agreed very well with the simulations and revealedsignificantly a position containing a water channel in the central part of the assembly.

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Park, Yongsoo S. M. Massachusetts Institute of Technology. "Improving heat transfer in spent nuclear fuel disposal packages using metallic void fillers." Thesis, Massachusetts Institute of Technology, 2016. http://hdl.handle.net/1721.1/107320.

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Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2016.
Cataloged from student-submitted PDF version of thesis.
Includes bibliographical references (pages 71-75).
Disposal packages containing high heat generating spent nuclear fuels (SNF) require improved heat transfer to keep the peak cladding temperature from going above the tolerance limit. Filling the accessible void spaces between the container and the SNF with a high heat conducting metal is a potential solution. In metal casting, it is well known that a gap forms at the metal-mold interface due to solidification shrinkage and it significantly reduces heat transfer during cooling. This negative heat transfer effect is persistent for a disposal package since the filler stays in the container after solidification. The key to close the gap is to promote metallic bonding by minimizing the oxidation of the container during the required preheating stage of the void filling process. However, direct contact between the container and the molten filler can lead to the growth of intermetallic phases, which can embrittle the container. The contribution of this work is twofold. First, through a down-scaled experiment, it was shown that coating a steel container with Zn and using Zn or Zn-4wt.%Al as a filler and unidirectionally cooling the melt from the bottom successfully suppressed the formation of the gap. Closing the gap increased the effective thermal conductivity of the package by a factor of roughly 6 under the employed experimental conditions. Second, tests showed that using near eutectic Zn-Al and executing the filling process at a temperature below the melting point of Zn suppressed the growth of any intermetallic phases. Specifically, this prevents the growth of Fe-Zn intermetallic phases due to the sufficiently high composition of Al, and it inhibits the dissolution and diffusion of Fe from the container by extending the presence of the ZnO diffusion barrier, which delays the growth of the Fe-Al intermetallic phases.
by Yongsoo Park.
S.M.

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39

Willman, Christofer. "Applications of Gamma Ray Spectroscopy of Spent Nuclear Fuel for Safeguards and Encapsulation." Doctoral thesis, Uppsala : Acta Universitatis Upsaliensis, 2006. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-7116.

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Scott, Mark Robert. "Nuclear forensics: attributing the source of spent fuel used in an RDD event." Texas A&M University, 2005. http://hdl.handle.net/1969.1/2368.

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An RDD attack against the U.S. is something America needs to prepare against. If such an event occurs the ability to quickly identify the source of the radiological material used in an RDD would aid investigators in identifying the perpetrators. Spent fuel is one of the most dangerous possible radiological sources for an RDD. In this work, a forensics methodology was developed and implemented to attribute spent fuel to a source reactor. The specific attributes determined are the spent fuel burnup, age from discharge, reactor type, and initial fuel enrichment. It is shown that by analyzing the post-event material, these attributes can be determined with enough accuracy to be useful for investigators. The burnup can be found within a 5% accuracy, enrichment with a 2% accuracy, and age with a 10% accuracy. Reactor type can be determined if specific nuclides are measured. The methodology developed was implemented into a code call NEMASYS. NEMASYS is easy to use and it takes a minimum amount of time to learn its basic functions. It will process data within a few minutes and provide detailed information about the results and conclusions.

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Hoggett-Jones, Craig. "Modelling and assessment of partitioning and transmutation approaches to spent nuclear fuel management." Thesis, University of Strathclyde, 2001. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.248302.

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42

Martinik, Tomas. "Development of Differential Die-Away Instrument for Characterization of Swedish Spent Nuclear Fuel." Licentiate thesis, Uppsala universitet, Tillämpad kärnfysik, 2016. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-268143.

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The Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project was established in 2009 by the U.S. Departmentof Energy with main objective to investigate, and potentially develop and test new technologies for spent nuclearfuel (SNF) characterization. In Sweden the SNF is currently being considered to be verified and encapsulated in canistersand deposited into a geological repository. The need for an independent instrument for SNF verification by theSwedish operator turned into the collaborative effort with NGSI-SF to develop an instrument for future deployment inSweden.One of the techniques investigated within this project is the differential die-away (DDA) technique, which followingthe theoretical investigation by means of high fidelity Monte Carlo simulations indicated the potential to be applied fordetermining of various spent fuel assembly (SFA) parameters.This work introduces the first deployable DDA instrument which was designed to be used for characterizing ofSwedish SFAs currently stored in the Central Interim Storage Facility for Spent Nuclear Fuel (Clab). All the instrumentcomponents relevant for DDA design functionality were evaluated to ensure reliable operation in Clab. Although mostof the components were tuned with special consideration given to concerns from the operator (The Spent Nuclear FuelandWaste Management Company) , several post-simulation modification of the design were made. These modificationsare described in this work.A complementary study of the detector responses to asymmetrically burned SFAs indicated a different detector responses,depending on which of the four different orientations was used to assay individual SFAs. This study illustratedthe sensitivity of detectors with respect to the SFA orientation if there is a strong burn-up gradient across the SFA andhence a strong asymmetry in isotopic distribution in the SFA. In addition, the study of asymmetry provided the informationon different operational scenarios of the DDA instrument. The DDA instrument may provide general informationabout the complete SFA as well as give local information about certain parts of the SFA.

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Lundqvist, Tobias. "Investigation of Algebraic Reconstruction Techniques for Tomographic Measurements on Spent Nuclear Fuel Assemblies." Thesis, Uppsala universitet, Institutionen för strålningsvetenskap, 2004. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-307832.

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A non-destructive tomographic measurement technique for application on nuclear fuel assemblies has beendeveloped at the Uppsala University. Using this technique, the rod-by-rod distribution of selectedradioactive isotopes is determined experimentally. In the present work, the numerical technique to reconstruct the activity distribution inside the fuelassemblies has been analyzed. Three iterative reconstruction algorithms have been investigated, ART(Additive Reconstruction Technique), ML (Maximum Likelihood) and ASIRT (Additive SimultaneousIterative Reconstruction Technique). It was found that the ART algorithm is too sensitive to data points where the gamma-ray intensityis low, while ASIRT handles it in the best manner. Furthermore, ASIRT appears to be the most stablealgorithm and produces the best agreement to theoretical data.

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Lundkvist, Niklas. "AMS on the actinides in spent nuclear fuel : a study on a technique for inventory measurements." Thesis, Uppsala universitet, Tillämpad kärnfysik, 2010. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-130011.

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This report is concerned with the question whether Accelerator mass spectrometry (AMS) is asuitable technique for measuring actinide inventory in spent nuclear fuel, and if it is better thanpresent techniques for these measurements. AMS is a kind of Mass spectrometry (MS) and has alot of applications where radio carbon dating is one of the most common. AMS has been used formaking measurements on actinides before but mostly from traces in bioassay that could have beenin contact with weapon plutonium, and in bioassay near enrichment plants and reprocessingplants. It is shown in this report that AMS is more sensitive in low level measurements than thecurrent technique for spent nuclear fuel. ICP-MS is the current technique in use for inventorymeasurements on nuclear fuel at Swedish Nuclear Fuel and Waste Management Company (SKB).ICP-MS is also a kind of MS technique which is well-tried for inventory measurements on spentnuclear fuel. The difference in sensitivity ranges in levels of magnitude depending on whichisotope that is interesting for measurements. The lower detection limits for AMS is about 105-107atoms which makes it possible to use samples from nuclear fuel that is in the order of 10-10-10-16gto achieve the lower detection limit. The recommendation from this report is to make studies ifAMS also is an economical and efficiently suitable technique for future use on the actinideinventory in spent nuclear fuel.
Denna rapport handlar om huruvida Accelerator masspektrometri (AMS) är en lämplig teknik förmätning av aktinidinventeriet i använt kärnbränsle. Rapporten går också igenom om AMS är bättreän nuvarande tekniker för dessa mätningar. AMS är en typ av masspektrometri (MS) och har enmängd användningsområden, kol-14 metoden är en av de vanligaste. AMS har också ofta använtsför att göra mätningar på aktinidinnehåll i biomassa som kan ha varit i kontakt medvapenplutonium, och i närheten av anrikningsanläggningar och upparbetningsanläggningar. Detvisas i rapporten att AMS är en mer känslig metod än de nuvarande teknikerna som används förmätningar på aktinidinventariet i använt kärnbränsle. ICP-MS är den aktuella teknik som användsför mätningar på aktinidinventariet i använt kärnbränsle vid Svenska Kärnbränslehantering AB(SKB). ICP-MS är också en typ av MS teknik. MS är väl beprövad för mätningar av inventariet påanvänt kärnbränsle. Skillnaden i känslighet varierar i flera storleksordningar beroende på vilkenisotop som är intressant för mätningarna. Den lägre detektionsgränsen för AMS är cirka 105-107atomer, vilket gör det möjligt att använda prover från kärnbränsle som är i storleksordningen 10-10-10-16g för att uppnå den lägre detektionsgränsen. Rekommendationen från denna rapport är attgöra undersökningar om AMS också är ekonomiskt lönsam och tillräckligt effektiv teknik förframtida bruk inom mätningar av aktinidinventariet i använt kärnbränsle.

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45

Larsson, Cecilia. "Upgrade and validation of PHX2MCNP for criticality analysis calculations for spent fuel storage pools." Thesis, Uppsala University, Applied Nuclear Physics, 2010. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-113572.

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A few years ago Westinghouse started the development of a new method for criticality calculations for spent nuclear fuel storage pools called “PHOENIX-to–MCNP” (PHX2MCNP). PHX2MCNP transfers burn-up data from the code PHOENIX to use in MCNP in order to calculate the criticality. This thesis describes a work with the purpose to further validate the new method first by validating the software MCNP5 at higher water temperatures than room temperature and, in a second step, continue the development of the method by adding a new feature to the old script. Finally two studies were made to examine the effect from decay time on criticality and to study the possibility to limit the number of transferred isotopes used in the calculations.

MCNP was validated against 31 experiments and a statistical evaluation of the results was done. The evaluation showed no correlation between the water temperature of the pool and the criticality. This proved that MCNP5 can be used in criticality calculations in storage pools at higher water temperature.

The new version of the PHX2MCNP script is called PHX2MCNP version 2 and has the capability to distribute the burnable absorber gadolinium into several radial zones in one pin. The decay time study showed that the maximum criticality occurs immediately after the takeout from the reactor as expected.

The last study, done to evaluate the possibility to limit the isotopes transferred from PHOENIX to MCNP showed that Case A, a case with the smallest number of isotopes, is conservative for all sections of the fuel element. Case A, which contains only some of the actinides and the strongest absorber of the burnable absorbers gadolinium 155, could therefore be used in future calculations.

Finally, the need for further validation of the method is discussed.

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46

Szakaly, Frank Joseph. "Assessment of uranium-free nitride fuels for spent fuel transmutation in fast reactor systems." Thesis, Texas A&M University, 2003. http://hdl.handle.net/1969.1/31.

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The purpose of this work is to investigate the implementation of nitride fuels containing little or no uranium in a fast-spectrum nuclear reactor to reduce the amount of plutonium and minor actinides in spent nuclear fuel destined for the Yucca Mountain Repository. A two tier recycling strategy is proposed. Thermal spectrum transmutation systems converted from the existing LWR fleet were modeled for the first tier, and the Japanese fast reactor MONJU was used for the fast-spectrum transmutation. The modeling was performed with the Monteburns code. Transmutation performance was investigated as well as delayed neutron fraction, heat generation rates, and radioactivity of the spent material in the short and long term for the different transmutation fuel cycles. A two-tier recycling strategy incorporating fast and thermal transmutation with uranium-free nitride fuel was shown to reduce the long-term heat generation rates and radioactivity of the spent nuclear fuel inventory.

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47

Lovett, Phyllis Maria. "An experiment to simulate the heat transfer properties of a dry, horizontal spent nuclear fuel assembly." Thesis, Massachusetts Institute of Technology, 1991. http://hdl.handle.net/1721.1/17294.

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Thesis (M.S.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1991.
Science hard copy bound in 1 v.
Includes bibliographical references (leaves 163-165).
by Phyllis Maria Lovett.
M.S.

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48

Burdo, James. "Monte Carlo Characterization of PWR Spent Fuel Assemblies to Determine the Detectability of Pin Diversion." University of Cincinnati / OhioLINK, 2010. http://rave.ohiolink.edu/etdc/view?acc_num=ucin1267546076.

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49

Foster, Jack Warren. "Development and implementation of a response-function concept for spent nuclear fuel cask analysis." Thesis, Georgia Institute of Technology, 1993. http://hdl.handle.net/1853/17275.

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50

Lundström, Tim. "Radiation chemistry of aqueous solutions related to nuclear reactor systems and spent fuel management /." Linköping : Univ, 2003. http://www.bibl.liu.se/liupubl/disp/disp2003/tek840s.pdf.

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Dissertations / Theses on the topic 'Storage of spent nuclear fuel'

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