• Nuclear Engineering and Technology
• /
• v.45 no.7
• /
• pp.847-858
• /
• 2013

Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea) in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4-5 $g-U/cm^3$ were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr), additional protective coatings (silicide or nitride), and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

• Journal of Mechanical Science and Technology
• /
• v.15 no.3
• /
• pp.327-338
• /
• 2001

This paper presents the results of a structural analysis to determine design variables such as the inner basket array type, and thicknesses of the outer shell, and lid and bottom of a spent nuclear fuel disposal canister. The canister construction type introduced here is a solid structure with a cast iron insert and a corrosion resistant overpack, which is designed for the spent nuclear fuel disposal in a deep repository in the crystalline bedrock, entailing an evenly distributed load of hydrostatic pressure from the groundwater and high swelling pressure from the bentonite buffer. Hence, the canister must be designed to withstand these high pressure loads. Many design variables may affect the structural strength of the canister. In this study, among those variables, the array type of inner baskets and thicknesses of outer shell and lid and bottom are attempted to be determined through a linear structural analysis. Canister types studied hear are one for the pressurized water reactor (PWR) fuel and another for the Canadian deuterium and uranium reactor (CANDU) fuel.

• Nuclear Engineering and Technology
• /
• v.53 no.9
• /
• pp.2937-2952
• /
• 2021

A typical three-dimensional finite element model for a fuel assembly is established, which is composed of 16 monolithic U-10Mo fuel plates and Al alloy frame. The distribution and evolution results of temperature, displacement and stresses/strains in all the parts are numerically obtained and analyzed with a self-developed code of FUELTM. The simulation results indicate that (1) the out-of-plane displacements of Al alloy side plates are mainly attributed to the bending deformations; (2) enhanced out-of-plane displacements appear in fuel plates adjacent to the outside Al plates, which results from the occurred bending deformations due to the applied constraints of outside Al plates; (3) an intense interaction of fuel foil with the cladding occurs near the foil edge, which appears more heavily in the fuel plates adjacent to the outside Al plates. The maximum first principal stresses in the fuel foil are similar for all the fuel plates and appear near the fuel foil edge; while, the through-thickness creep strains of fuel foil in the fuel plate near the central region of fuel assembly are larger, and the induced creep damage might weaken the fuel skeleton strength and raise the fuel failure risk.

• Journal of the Computational Structural Engineering Institute of Korea
• /
• v.17 no.4
• /
• pp.413-421
• /
• 2004

In this paper, a structural analysis for the pressurized water reactor(PWR) spent nuclear fuel disposal canister which is deposited under the 500m deep underground is carried out to predict the creep deformation of the canister while the underground water and swelling bentonite pressure are applied on the canister. Usually the creep deformation may be caused due to the Pressure and the high heat applied to the canister even though additional external loads are not applied to the canister. These creep deformations depend on the time. In this paper, oかy the underground water and bentonite swelling Pressure are considered for the creep deformation analysis of the canister, because the heat distribution inside canister due the spent fuel is not simple and depends on time. A proper creep function is adopted for the creep analysis. The creep analysis is carried out during $10^8$ seconds. The creep analysis results show that the creep strains are very small and these strains occur usually in the lid and bottom of the canister not in the cast iron insert. A much smaller strain is found in the cast iron insert. Hence, the creep deformation doesn't affect the structural safety of the canister, and also the creep stress which shows the stress relaxation phenomenon doesn't affect the structural safety of the canister.

• Proceedings of the Computational Structural Engineering Institute Conference
• /
• 2000.04b
• /
• pp.77-84
• /
• 2000

In this paper, the stress analysis of the cast iron insert of spent nuclear fuel disposal canister in a deep repository at 500m underground is done for the underground pressure variation. Since the nuclear fuel disposal usually emits much heat and radiation, its careful treatment is required. And so a long term safe repository at a deep bedrock is used. Under this situation, the canister experiences some mechanical external loads such as hydrostatic pressue of underground water, swelling pressure of bentonite, sudden rock movement etc.. Hence, the canister should be designed to withstand these loads. The cast iron insert of the canister mainly supports these loads. Therefore, the stress analysis of the cast iron insert is done to determine the design variables such as the diameter versus length of canister and the number and array type of inner baskets in this paper, The linear static structural analysis is done using the finite element analysis method. And the finite element analysis code, NISA, is used for the computation.

• Journal of Radiation Protection and Research
• /
• v.29 no.1
• /
• pp.49-55
• /
• 2004

This paper presents the mechanical and thermal stress analysis of a disposal canister to provide basic information for dimensioning the canister and configuration of the canister components. The structural stress analysis is carried out using a finite element analysis code, NISA, and focused on the structural strength of the canister against the expected external pressures due to the swelling of the bentonite buffer and the hydrostatic head, and the thermal load build up in the container.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.8 no.3
• /
• pp.207-213
• /
• 2010

The effect of temperature on swelling pressure was observed with a Korean domestic Ca-bentonite which has been considered as a potential buffer material in the engineering barrier of a high level radioactive waste (HLW) disposal system. The Ca-bentonite was compacted to a dry density of 1.6 g/$cm^3$, and then de-ionized water was supplied into it with a constant pressure of 0.69 MPa. The equilibrium swelling pressures were measured with different temperatures of $25^{\circ}C$, $30^{\circ}C$, $40^{\circ}C$, $50^{\circ}C$, $60^{\circ}C$, $70^{\circ}C$, respectively. The Ca-bentonite showed a sufficiently high swelling pressure of 5.3 MPa at room temperatures. Then it was clearly showed that the equilibrium swelling pressure was decreased with an increase of temperature. Interestingly, there were some differences in temperature effect on the equilibrium swelling pressure when the environmental temperature is increasing or decreasing. For further clarifying the swelling behaviour of a Korea domestic Ca-bentonite, the change of a compaction level, and the composition variation of a supplied water would be needed to use in conceptual design of HLW disposal system.

• 한국신재생에너지학회:학술대회논문집
• /
• 2010.06a
• /
• pp.129.2-129.2
• /
• 2010

AEM which were used for solid alkaline fuel cell(SAFC) were prepared by photo polymerization in method pore-filling with various quaternary ammonium cationic monomers and crosslinkers without an amination process. Their specific thermal and chemical properties were characterized through various analyses and the physico-chemical properties of the prepared electrolyte membranes such as swelling behavior, ion exchange capacity and ionic conductivity were also investigated in correlation with the electrolyte composition. The polymer electrolyte membranes prepared in this study have a very wide hydroxyl ion conductivity range of 0.01 - 0.45S/cm depending on the composition ratio of the electrolyte monomer and crosslinking agent used for polymerization. However, the hydroxyl ion conductivity of the membranes was relatively higher at the whole cases than those of commercial products such as A201 membrane of Tokuyama. These pore-filling membranes have also excellent properties such as smaller dimensional affects when swollen in solvents, higher mechanical strength, lowest electrolyte crossover through the membranes, and easier preparation process compared of traditional cast membranes. The prepared membranes were then applied to solid alkaline fuel cell and it was found comparable fuel cell performance to A201 membrane of Tokuyama.

• Journal of Energy Engineering
• /
• v.13 no.3
• /
• pp.181-190
• /
• 2004

Computational models for analyzing the in-reactor behavior of metallic fuel pins under transient conditions in liquid-metal reactors are developed and implemented in the TRAMAC (TRAnsient thermo-Mechanical Analysis Code) for a metal fuel rod under transient operation conditions. Not only the basic models for a fuel rod performance but also some sub-models used for transient condition are installed in TRAMAC. Among the models, a fission gas release model, which takes the multi-bubble size distribution into account to characterize the lenticular bubble shape and the saturation condition on the grain boundary and the cladding deformation model have been developed based mainly on the existing models in the MAC-SIS code. Finally, cladding strains are calculated from the amount of thermal creep, irradiation creep, and irradiation swelling. The cladding strain model in TRAMAC predicts well the absolute magnitudes and gen-eral trends of their predictions compared with those of experimental data. TRAMAC results for the FH-1,2,6 pins are more conservative than experimental data and relatively reasonable than those of FPIN2 code. From the calculation results of TRAMAC, it is apparent that the code is capable of predicting fission gas release, and cladding deformation for LMR metal fuel finder transient operation conditions. The results show that in general, the predictions of TRAMAC agree well with the available irradiation data.

• Nuclear Engineering and Technology
• /
• v.28 no.3
• /
• pp.229-246
• /
• 1996

A mechanistic model has been developed to predict the release behavior of fission gas during steady-state irradiation of LWR UO$_2$ fuel. Under the assumption that UO$_2$ grain surface is composed of fourteen identical circular faces and grain edge bubble can be represented by a triangulated tube around the circumference of three circular grain faces, it introduces the concept of continuous formation of open grain edges tunnels that is proportional to grain edge swelling. In addition, it takes into account the interaction between the gas release from matrix to grain boundary and the reintroduction of gas atoms into the matrix by the irradiation-induced re-solution of grain face bubbles. It also treats analytically the behavior of intragranular, intergranular, and grain edge bubbles under the assumption that both intragranular and intergranular bubbles are uniform in both radius and number density. Comparison of the present model with experimental data shows that the model's prediction produces reasonable agreement for fuel with centerline temperatures of 1000 to 140$0^{\circ}C$, wide scatter band for fuel with centerline temperatures lower than 100$0^{\circ}C$, and underprediction for fuel with centerline temperatures higher than 140$0^{\circ}C$.