[en] Precision measurements of the half life and branching ratios for the β⁺ decay of ⁶²Ga were performed at the ISAC facility at TRIUMF. These experiments are part of a program to test the Standard Model of particle physics through high precision superallowed β decay studies. A 4π gas proportional counter was used with a fast tape transport system to perform the half life measurement. The deduced half life was 116.01 ± 0.19 ms, giving a world average of 116.17 ± 0.04 ms. The 8π array of twenty HPGe detectors and the SCEPTAR array of twenty plastic scintillators were employed to detect γ rays and β particles for the branching ratio experiment. The ground state superallowed branching ratio was found to be (99.856 ± 0.019)%. In this experiment, twenty new γ rays and five new energy levels populated in the β decay of ⁶²Ga were observed. (author)

[en] One of the best options for meeting growing global energy needs is nuclear energy, since it is both emissions free and has the potential to be a sustainable energy source. The key areas for the development of future reactors are safety, sustainability, economics and security. Heavy water moderated reactors are an appealing option because of their improved neutron efficiency, which is advantageous from a sustainability standpoint. In an evolution of the current CANDU® reactors, the pressure tube structure and heavy water moderator are retained in an advanced reactor design, cooled by supercritical light water. The use of supercritical light water as the coolant enables a large increase in thermal efficiency and therefore provides improved economic benefits. The use of thorium-based fuel provides improved safety, and a non-proliferative and sustainable fuel cycle. The optimization of the SCWR (Super Critical Water-cooled Reactor) is determined in part through reactor physics calculations, with respect to fuel utilization and safety (e.g. coolant void reactivity). These and other aspects of SCWR physics will be discussed in this paper. (author)

[en] New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen was a 35-element bundle made with a homogeneous mixture of reactor grade PuO₂ (∼67 wt% fissile) and ThO₂, with a central zirconia rod to reduce coolant void reactivity. Several annular and checkerboard-type heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. WIMS-AECL 3.1 was used to perform lattice physics calculations using 2-D, 89-group integral neutron transport theory, while RFSP 3.5.1 was used to perform the core physics and fuel management calculations using 3-D two-group diffusion theory. It was found that the various core concepts can achieve a fissile utilization that is up to 30% higher than is currently achieved in a PT-HWR using natural uranium (NU) fuel bundles. Up to 67% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 363 kg/year of U-233 is produced in the discharged fuel. (author)

[en] One of the key challenges in the development of a CANDU pressure-tube supercritical water-cooled reactor (SCWR) is the selection of materials appropriate for in-core use. Such materials must be able to withstand the high-temperature, corrosive environment, and effects of irradiation encountered in the core, while at the same time minimizing parasitic neutron absorption. Achieving the appropriate balance between reactor physics and materials requirements necessitates knowledge of both materials properties of candidate alloys and their impact on lattice physics. In this paper, lattice physics calculations have been performed for the CANDU-SCWR for several categories of candidate in-core materials. In addition, a simple relation is derived that can be used to estimate the relative influence of in-core materials on lattice reactivity and fuel discharge burnup, based on material chemical composition and density. (author)

[en] Accurate and complete nuclear data are a fundamental requirement for any nuclear reactor model. One major challenge to the modeling of advanced nuclear reactor systems is the lack of sufficient nuclear data for the operating conditions and materials relevant to the advanced systems. The Canadian Supercritical Water-Cooled Reactor (SCWR) is an advanced reactor concept which, like all advanced GEN-IV reactor concepts, differs significantly in operating conditions, fuel composition and non-fuel materials from conventional reactors. The Canadian SCWR is a pressure tube-based reactor with heavy water moderator and light water coolant, intended to operate with a coolant pressure of 25 MPa and temperatures ranging from 350 ^oC (inlet) to 625^oC (outlet), with (Pu,Th)O₂ fuel, using advanced fuel bundle and fuel channel designs. Because of these differences from conventional heavy water (HWR) and light water (LWR) reactors, it is not clear whether presently-used core modeling methods or nuclear data libraries are adequate for SCWR modeling. In this paper, an idealized model of an SCWR fuel channel with fresh fuel is modeled in order to examine the nuclear data contributions to the sensitivity and uncertainties in the neutron multiplication factor, k, and various lattice reactivity coefficients. (author)

[en] The CANDU supercritical water-cooled reactor (CANDU-SCWR) is a pressure tube reactor intended to operate with a coolant pressure of 25 MPa and temperatures ranging between 350°C (core inlet) and 625°C (core outlet). Along the length of a fuel channel, there is a drastic decrease in the coolant density and dielectric constant, which is expected to result in a rapid decrease in the solubility of corrosion products. Therefore, it is anticipated that corrosion product deposition onto the cladding and liner in an SCWR fuel channel will be much greater than in conventional water-cooled reactors operating below the critical point of water. While optimized materials selection and chemistry control strategies may mitigate corrosion and corrosion product deposition to some degree, it may not be possible to completely eliminate corrosion product deposition within SCWR fuel channels. Corrosion product deposition on fuel cladding will have a negative impact on the neutron economy of the CANDU-SCWR because of parasitic absorption of neutrons within the deposited material. In this paper, lattice physics calculations are used to assess the impact of corrosion product deposition on fuel exit burnup, based on corrosion product deposition rates estimated for prototypical SCWR conditions. (author)

[en] Increasing the burnup of reactor fuel can have advantages such as improved utilization of fissile resources, reduced spent fuel volume, and, in the case of reactors with on-power refuelling, decreased operational demand on the fuelling machines. Higher burnups are also desirable for many advanced fuel cycles, such as minor actinide-bearing fuels where high burnups are required to achieve better actinide destruction, and thorium based fuel cycles where increased irradiation time converts and burns more U-233 from Th-233. However, this will impose more challenging operating conditions on the fuel, particularly in the case of on-power refuelling. A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a pressure tube heavy water reactor (PT-HWR) with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (author)

[en] Small modular reactors are defined as reactors with power output of approximately 1.5 MW to 300 MW, with option for grid/non-grid connection, modular approach to construction and deployment and are cost-competitive. Designs of small modular reactors vary globally and include molten salt, high temperature gas, sodium cooled and lead cooled. Characteristics also include advances in safety and efficiency as well as applications in electricity production or other applications. Some of the benefits of small modular reactors are lower costs, enhanced safety, environmental effects, flexibility and energy beyond electricity.

[en] The irradiation of Th-232 breeds fewer of the problematic minor actinides (Np, Am, Cm) than the irradiation of U-238. This characteristic makes thorium an attractive potential matrix for the transmutation of these minor actinides, as these species can be transmuted without the creation of new actinides as is the case with a uranium fuel matrix. Minor actinides are the main contributors to long term decay heat and radiotoxicity of spent fuel, so reducing their concentration can greatly increase the capacity of a long term deep geological repository. Mixing minor actinides with thorium, three times more common in the Earth's crust than natural uranium, has the additional advantage of improving the sustainability of the fuel cycle. In this work, lattice cell calculations have been performed to determine the results of transmuting minor actinides from light water reactor spent fuel in a thorium matrix. 15-year-cooled group-extracted transuranic elements (Np, Pu, Am, Cm) from light water reactor (LWR) spent fuel were used as the fissile component in a thorium-based fuel in a heavy water moderated reactor (HWR). The minor actinide (MA) transmutation rates, spent fuel activity, decay heat and radiotoxicity, are compared with those obtained when the MA were mixed instead with natural uranium and taken to the same burnup. Each bundle contained a central pin containing a burnable neutron absorber whose initial concentration was adjusted to have the same reactivity response (in units of the delayed neutron fraction β) for coolant voiding as standard NU fuel. (author)

[en] The CANDU® supercritical water-cooled reactor (SCWR) is Canada's primary contribution to the Generation IV International Forum (GIF). The goals of GIF include the development of next-generation reactors with enhanced safety, resource sustainability, economic benefit and proliferation resistance. There is great potential for enhancing the sustainability of the nuclear fuel cycle by extending the availability of current resources through the use of thorium fuel cycles. Recent studies of thorium-based fuel cycles in contemporary CANDU reactors demonstrate the possibility for substantial reductions in natural uranium (NU) requirements of the fuel cycle via the recycling of U-233 bred from thorium. As thorium itself is not fissile, neutrons must be provided by adding a fissile material, either within or outside of the thorium based fuel. Various thorium fuel cycles can be categorized by the type and geometry of the added fissile material. The simplest of these fuel cycles are based on homogeneous thorium fuel designs, where the fissile material is mixed uniformly with the fertile thorium. These fuel cycles can be competitive in resource utilization with the best uranium-based fuel cycles, while building up an inventory of U-233 in the spent fuel for possible recycling in thermal reactors. When U 233 is recycled from the spent fuel, thorium-based fuel cycles can provide substantial improvements in the efficiency of energy production from existing fissile resources. In this paper, two homogeneous CANDU-SCWR thorium-based fuel cycles using reactor-grade plutonium as the fissile driver material have been examined. As the CANDU-SCWR reactor concept is still in the early development and design stages, various lattice and channel parameters can be varied to optimize the reactor for a specific fuel type. The impact of varying some of these parameters to optimize for thorium fuel has been studied. In this paper, thorium fuel cycle options are examined and compared with respect to initial Pu driver fuel requirements, U-233 recycling, and exit burnup. (author)