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11	Development and assessment of CFD models including a supplemental program code for analyzing buoyancy-driven flows through BWR fuel assemblies in SFP complete LOCA scenarios Artnak, Edward Joseph 31 January 2013 (has links) This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-of-coolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based control-volume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy. / text CFD CFD Model Buoyancy driven flow Buoyancy flow FLUENT FLUENT CFD SNL Sandia National Laboratories SFP LOCA Spent fuel pool loss of coolant accident SFP loss of coolant BWR fuel assembly CAD Model SolidWorks SFP Model SFP BWR fuel assembly NRC MATLAB MATLAB SFP Model 9x9 BWR fuel assembly MELCOR MELCOR SFP 250 million cells Hydraulic flow loss Loss coefficients Buoyancy flows MELCOR calibration Pressure loss BWR pressure loss SFP airflow SNL fuel assembly
12	Development of a computational method for determining gamma energy escape from calorimeters at CLAB Mehic, Amela January 2022 (has links) Swedish Nuclear Fuel and Waste Management Company- SKB has conducted measurement campaigns at the Swedish central interim storage facility for spent nuclear fuel- CLAB over the years, extending from year 2003 to 2019 where the gamma energy escape was acquired. At CLAB the spent nuclear fuel assembly is inserted into the calorimeter; device intended to measure temperature increase due to decay heat from the fuel assembly. The calorimetric construction is surrounded by water the medium in which the temperature deviations occur and thus are also measured by the calorimeter. However, there is some leakage of gamma energy from the calorimetric construction and does not contribute to the heating of the water. Therefore, only considering the calorimetric measurements is not enough to estimate the total decay heat in the fuel assembly since these measurements fail to account for the gamma escape. Measurements of gamma energy escape acquired over the years at CLAB were observed to have some tendencies that where questionable, mainly some stochastic behavior indicating that their uncertainty was profound. In the scope of the thesis a computational method was developed to calculate the gamma energy escape and thus assist in determining which measurements to discard. Combination of two programs were used one being Spent Nuclear fuel- SNF and the second Monte Carlo N particle Simulator- MCNP, to obtain the computational gamma energy escape for different fuel assemblies and cooling times- CTs. It was established that the escape had a range between 1-3,5% and that it had a dependency on CT, fuel assembly type and operational history. Calculated radial exponential decay coefficient for fuel assemblies of the medium; water had also a clear dependency on CT where values of the coefficient increased over CT. Normalized gamma energy distribution over a rotation around the fuel assembly was calculated and it showed that the assembly tended to have the highest radiation coming from its corner rods. The verification of the computational gamma escape results with corresponding measurements yielded that the agreement was quite good for the earlier measurement campaigns. However, deviation became evident after the 2007 campaign where the calculated values were underestimated compared to the measured. Nuclear physics Calorimeter SNF MCNP Dose rate CLAB Gamma energy Fuel assembly Form function Attenuation coefficient Gamma spectroscopy Physical Sciences Fysik Other Engineering and Technologies Annan teknik
13	Förbättrade effektmarginaler med radiell anrikningsfördelning för PWR-bränsle / Improved peaking factors with radial enrichment distribution för PWR fuel assembly Åkerman, Mattias January 2016 (has links) In recent years, the enthalpy raise hot channel factor limit has decreased significantly due to the power upgrade of Ringhals 4 and the use of shielding fuel assemblies. The shielding fuel assemblies task are to reduce the neutron leakage to the reactor vessel and in that way extend the reactor lifetime. This is achieved by replacing a few fuel rods with steel rods. Experiences from the last fuel cycles show that the core design procedure has been hampered because of this and that it’s hard to stay under the design limit. A way to overcome this problem and to improve the fuel economy is to introduce the use of radial enrichment distribution in the fuel assembly. This master thesis shows, through a case study of three fuel cycles at Ringhals 4, that the internal peaking factor can be improved by roughly 2–3 % and that the maximum enthalpy raise hot channel factor can be improved by about 2.0–2.5 % if the fuel assemblies contain three different levels of enrichments instead of currently one. This can be achieved without any noticeable decrease in cycle length. / Genom en fallstudie av tre driftcykler för Ringhals 4 visar den här rapporten att max FΔH under cykeln kan sänkas med 2,0–2,5 % om bränsleknippena radiellt anrikningsoptimeras med minst tre delanrikningar. Totalt under cykeln kan FΔH sänkas med upp till 4 %. Om radiell anrikningsoptimering införs för Vattenfalls PWR:er skulle arbetet med att designa härdarna förenklas och utrymme ges för att ladda reaktorerna på ett mer ekonomiskt sätt. Ringhals 4 PWR fuel rod fuel assembly shielding fuel assembly reactor core reactor cycle burnable absorber peaking factor internal peaking factor enthalpy raise hot channel factor radial enrichment distribution enrichment optimization CASMO SIMULATE XIMAGE Ringhals 4 PWR bränslestav bränsleknippe skärmande knippen reaktorhärd reaktorcykel brännbar absorbator intern effektformfaktor radiell effektformfaktor effektmarginal radiell anrikningsfördelning anrikningsoptimering bränsledesign härddesign CASMO SIMULATE XIMAGE
14	Studium radiačního poškození nádoby reaktoru VVER-440 jaderné elektrárny Dukovany / Radiation damage of VVER-440 based Dukovany NPP reactor pressure vessel investigation Říha, Tomáš January 2011 (has links) This master‘s thesis deals with radiation damage of reactor pressure vessels, specifically of NPP Dukovany Unit No. 3. In general damage mechanisms of reactor steels are described and possibilities of monitoring of material degradation and its recovery used at NPP’s all over the world are mentioned as well. A practical part of the thesis is focused on interpretation of analyses carried out with the assistance of MOBY DICK code. The ground of these analyses is a neutron fluence value development on different locations of RPV for the whole life of operation up to 24th cycle. The analyses results are put into context with performed in-service inspections. The thesis follows up with neutron fluence computation for the future cycles containing new types of nuclear fuel up to 34th cycle. The outcome of practical part of the master‘s thesis is a comparison between new types of nuclear fuel with respect to radiation damage of RPV’s.
15	Návrh inspekčního sloupu pro kontroly stavu použitého jaderného paliva / Design of equipment of spent nuclear fuel assemblies Šimek, Ondřej January 2018 (has links) The diploma thesis aim to the design of equipment for ŠKODA JS a.s., which is part of a new inspection stand (N-SIO). This equipment is an inspection column that provides the possibility to inspect spent fuel assemblies at the operation of the Temelín nuclear power plant. This master thesis is also a summary of the whole design of the new inspection stand and a description of the individual inspection components and devices. One of the parts of the thesis is also a basic strength analysis and a drawing of the main assembly of inspection equipment.
16	Model palivového souboru tlakovodního reaktoru západní koncepce / PWR fuel assembly model Cekl, Jakub January 2018 (has links) PWR, fuel assembly, benchmark, burnup, lattice, SCALE, Polaris, validation, reactivity

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