Nuclear power plants could potentially be deployed in a type of nuclear hybrid energy system (NHES) in which their power is used primarily to drive an industrial process but can be diverted to meet demands for electricity when needed. The purpose of this study is to analyze the effects of deploying NHESs as reserve power for the transmission grid in Ontario on the overall Canadian fuel cycle. In this scenario, the fuel cycle demands of 2 high-temperature gas-cooled reactor (HTGR) concepts are analyzed with respect to costs, resource consumption, and enrichment requirements. One HTGR concept is a 30 MW-thermal (MWth) reactor that is based on the UBattery concept, and the other is the Xe-100, which is a 200 MWth reactor. Calculations indicate that such a deployment of HTGRs would have a substantial effect on the fuel cycle in Canada. In particular, NU and enrichment demands would be greatly affected. Beginning this HTGR deployment in the year 2030 would more than double the annual NU demands in Canada, and deplete the uranium resources with extraction costs of <$80/kgU by the year 2142. The uranium enrichment demands of this fleet would be >35% of the US capacity for uranium enrichment.
CANDU fuel bundles experience plastic deformations over time, and the horizontal configuration of the bundle in a crept pressure tube (PT) causes coolant to bypass the sagged lower half of the bundle. Bundle segments where the flow is limited may become more susceptible to dryout due to reactor aging. A finite element model of a 37-element fuel bundle was constructed using the commercial finite element software ANSYS to study the mechanical deformation behaviour of the bundle to maintain a coolable geometry. The main focus was on the contact between the fuel elements and between the fuel elements and PT. The complexity of the model due to all the contact pairs necessitated the use of high-powered computing hardware. Contact was demonstrated between the appendages, and sensitivity of the deformation to different boundary conditions (BC) was investigated. In particular, the radial position where the elements were welded to the endplate significantly impacted the magnitude of the element bowing. Expanding the PT up to 8% diametral creep demonstrated the proper functioning of the spacer pads (SP) and bearing pads in preventing sheath-to-sheath contact at the midplane and sheath-toPT contact. However, the quarter plane was deemed to be the critical region due to the lack of SPs preventing excessive element bowing. This work has successfully illustrated the deformation of a CANDU fuel bundle, with contact, and its similarity with the bow profiles when compared with post-irradiation examination results and bundle heat-up tests.
Discharged CANDU fuel is stored under water in irradiated fuel bays (IFBs) to remove their decay heat. If the fuel is exposed to air, a self-sustaining reaction could result when the Zircaloy-4 sheathing reaches temperatures sufficient for a breakaway oxidation. To predict when the transition occurs, a 2-D fuel bundle cross-section model in air was developed using the COMSOL Multiphysics®platform. Breakaway was predicted to occur at its earliest within 2.6 hours for a range of recently discharged bundle powers. It was concluded due to the time required for heat up and cracking of the oxide layer, sufficient margin exists for operators to intervene before a passively cooled, isolated bundle undergoes breakaway. To examine the effect of multiple bundles, a 3-D model based on a quarter of a stand-alone spent fuel rack was developed to calculate the steady-state temperature and mass fluxes of air. The model provided a lower bound for the ambient temperatures because the flow resistance of the bundle was not considered. The correct incorporation of flow resistance is a necessary step before conclusions could be made about the safety of IFBs. However, the analysis using a Computational Fluid Dynamics model for a 0.5 MW fuel rack, indicated that the maximum temperature of the air within the rack was 642 K and located at the centre of the outlet. This result is encouraging to support the safety of IFBs, as the temperature is well below the 873 K, which is approximately the minimum required for a breakaway reaction.
Incoloy 800H is one of several candidates for a fuel cladding material in super-critical water nuclear reactor concepts. The objective of this work is to determine the effect of the gas tungsten arc welding (GTAW) process on the microstructure and resulting tensile properties of Incoloy 800H tubes. In this work, GTAW was used to join Incoloy 800H. During welding, the weld thermal cycle produces differently featured heat-affected zone and fusion zone (FZ) microstructures. Microstructural examination revealed that weld-characteristic columnar and equiaxed dendritic structures were formed in the FZ. In comparison with the optimum heat input, both increase and decrease of heat input led to the formation of more columnar dendritic structures in the FZ. The chemical element distribution analysis using scanning electron microscopy/energy dispersive X-ray spectroscopy showed the segregation of Ti in the form of Ti-rich carbides and nitrides; other elements did not display any obvious segregation. Tensile test results revealed that Incoloy 800H alloy welds exhibit an excellent combination of strength and ductility almost equal to the base metal (BM) at the optimum and higher than optimum heat input conditions with full penetration. The welding process has no obvious effect on the microhardness across the whole welding zone. The refinement of grain size and morphology in the FZ can contribute to the improvement in the mechanical properties. As a result, the Incoloy 800H weldment shows the comparable mechanical properties to the BM.
Fuel failures are always a cause of concern in any nuclear reactors as it increases the manrem consumption of radiation workers. Although performance of the fuels in pressurized heavy water reactors is good, but still fuel failures occur occasionally. Post irradiation examination (PIE) of the failed fuel elements indicates internal hydriding, not deuteriding, as a major cause for the failures, although secondary deuteriding occurs and, in a few cases, failures are associated with defects in the end plug weld. The sources of hydrogen are either fuel pellets or the clad or the graphite coating. Restriction has been imposed on maximum content of total hydrogen in the fuel element to 1 mg to prevent hydriding of the Zircaloy clad tube. Accidental pick up of hydrogen occurs, which could lead to failure of the fuel bundles. Experimental investigations have been conducted to understand the individual effect of iodine and accidental pick up of moisture on the microstructure of Zircaloy-4 end cap welded samples with graphite coating. Results indicate that severe hydriding in Zircaloy-4 samples due to the existence of internal moisture in presence of graphite under service condition may result in fuel failure and justifies the findings of PIE.
