[en] Pressure-tube heavy water reactors (PT-HWR) are advantageous for implementing plutonium/thorium fuels because of their online refuelling capability and high neutron economy. The use of annular 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 in fuel management. The bundle concept modeled was a 35-element fuel bundle made with a mixture of reactor grade PuO₂ (~67 wt% fissile) and ThO₂, with a central zirconia rod to reduce coolant void reactivity. Eight annular heterogeneous seed-blanket core concepts with plutonium/thorium-based fuels in a 700 MWe-class PT-HWR were analyzed, using a once-through thorium cycle. Blanket region(s) represented approximately 50% of the total fuel volume. There were 1-4 different blanket regions and 1-4 different seed regions. The seed fuel tested was 3 wt% or 4 wt% PuO₂, whereas the blanket fuel tested was 1 wt% or 2 wt% PuO₂, mixed with ThO₂. For comparison, 2 homogeneous reactor cores with either 3 wt% PuO₂ fuel or 4 wt% PuO₂ fuel were also analyzed. For a number of the core concepts investigated, the fissile utilization was up to 30% higher than what is achieved in a PT-HWR using natural uranium fuel bundles. It was also found among the various core concepts that up to 67% of the Pu was consumed, up to 43% of the energy was produced from thorium, and up to 363 kg/year of fissile uranium (mainly ²³³U) was produced in the discharged fuel. (author)

[en] Thorium is anticipated to play a potentially significant role in the world's future energy supply as the international nuclear power industry moves towards the implementation of more sustainable nuclear technologies and advanced fuel cycles. Canadian Nuclear Laboratories (CNL), formerly Atomic Energy of Canada Limited, has been investigating thorium-fuelled reactor concepts and developing thoria (thorium dioxide) fuel technology for more than 55 years, complimenting international experience in the development of thorium-based fuel cycles. Although there is a strong foundation based on past experience, gaps exist in the science and technology (S&T) required to implement the use of thorium-based fuels on an industrial scale. In this paper, progress in thoria fuel S&T is reviewed and research and development needs for the deployment of thorium-based fuel cycles (with the focus on the use of thoria, ThO₂) are identified from a Canadian perspective. CNL plans to address known S&T gaps are also discussed. (author)

[en] Code-to-code comparisons of lattice physics calculations were made for a series of fuels that could potentially be used in a conventional 700-MWe class Pressure Tube Heavy Water Reactor to improve the sustainability of the fuel cycle. Studies were performed for natural uranium, slightly enriched uranium and thorium-based fuels containing low enriched uranium, reactor grade plutonium, or ²³³UO₂ as the initial fissile driver. The collision probabilities lattice code WIMS-AECL was compared to the stochastic code MCNP using the ENDF/B-VII.0 nuclear data library. Specific parameters that were studied between models include k-infinity, coolant void reactivity, 89-group cell averaged fluence, and ring-by-ring linear element ratings. The calculations performed have demonstrated that physics parameters estimated by WIMS-AECL are consistent with MCNP, especially for fuel where the main fissile component is uranium-based. (author)

[en] Preliminary, conceptual studies have been performed previously using deterministic lattice physics (WIMS-AECL) and core physics codes (RFSP) to estimate performance and safety characteristics of various thorium-based fuels and uranium-based fuels augmented by small amounts of thorium for use in pressure tube heavy-water reactors (PT-HWRs). To confirm the validity of the results, the WIMS-AECL/RFSP results are compared against predictions made with the stochastic neutron transport code MCNP. This paper describes the development of a method for setting up an MCNP core model of at PT-HWR for comparison with WIMS-AECL/ RFSP results, using a core with 37-element natural uranium fuel bundles as a test case for sensitivity studies. These studies included evaluating the sensitivity of the bias of the effective neutron multiplication factor (k_eff), a source convergence study, uncertainties correction with multiple independent simulations, the impact of irradiation map binning methods, and the impact of reflector models. A Python-based software scripting tool was developed to automate the creation, execution, and post-processing of reactor physics data from the MCNP models. The software tool and algorithm for creating an MCNP core model using data from the WIMS-AECL and RFSP models are described in this paper. Based on the preliminary evaluations of the simulation parameters with the base model, reactor physics analyses were performed for PT-HWR cores with thorium-based fuels in a 35-element bundle type. Code-to-code results demonstrate good agreement between MCNP and RFSP, giving confidence in the method developed and its applicability to other fuels and core types. (author)

[en] The purpose of this study is to determine the effect of the initial power level during a rod ejection accident (REA) on the ejected rod worth and the resulting energy deposition in the fuel. The model used is for the hot zero power (HZP) conditions at the end of a typical fuel cycle for the Three Mile Island Unit 1 pressurized water reactor. PARCS, a transient, three-dimensional, two-group neutron nodal diffusion code, coupled with its own thermal-hydraulics model, is used to perform both steady-state and transient simulations. The worth of an ejected control rod is affected by both power level, and the positions of control banks. As the power level is increased, the worth of a single central control rod tends to drop due to thermal-hydraulic feedback and control bank removal, both of which flatten the radial neutron flux and power distributions. Although the peak fuel pellet enthalpy rise during an REA will be greater for a given ejected rod worth at elevated initial power levels, it is more likely the HZP condition will cause a greater net energy deposition because an ejected rod will have the highest worth at HZP. Thus, the HZP condition can be considered the most conservative in a safety evaluation. (author)

[en] Harsh conditions associated with high-temperature, high-pressure supercritical-water coolant in the Canadian supercritical water-cooled reactor (SCWR) necessitate the use of stainless steel or nickel-based alloys as fuel cladding material rather than more conventional zirconium-based alloys. In-core neutron irradiation is expected to produce helium in nickel-bearing alloys through the 2-step ⁵⁸Ni(n,γ)⁵⁹Ni(n,α)⁵⁶Fe reaction. Increases in displacements per atom (DPA) associated with this reaction will also occur. It is important to quantify neutron-irradiation induced helium production and DPA to assess the consequences for fuel sheath material integrity. In this paper, neutron-flux induced helium production and displacement damage are estimated from lattice physics calculations for 5 candidate SCWR fuel clad materials, alloys 214, 625, 800 H, 347, and zirconium-modified 310 stainless steel, and compared for 2 fuel options, plutonium-thorium dioxide, (Pu-Th)O₂, and low enriched uranium-thorium dioxide, (LEU-Th)O₂. It is found that helium production in the fuel clad material is proportional to nickel content in the fuel clad and is greater, (by a factor of approximately 2) for the (LEU-Th)O₂ fuel. DPA is found to be approximately independent of the composition of fuel clad with only a small variation (∼5%) between fuel materials. (author)

[en] Prior validation studies of 28-element natural uranium (28-NU) CANDU R-type fuel bundles using the WIMS-IST lattice physics code had demonstrated a bias in the calculation of the coolant void reactivity (CVR) of approximately +0.5 to +0.6 mk (1 mk =100 pcm = 0.001 Δk/k). However, these validation studies were performed using experimental data for 28-element bundles with pressure tubes that were smaller than standard CANDU-type pressure tubes, giving a smaller coolant volume, and a modified neutron energy spectrum. Validation studies performed with 37-element and 43-element fuel bundles with a CANDU-type lattice pitch and pressure tube had shown a CVR bias of ∼ 1.7 to 1.9 mk. It was believed that the CVR bias for a 28-element bundle would be closer to this range of values if a standard CANDU pressure tube diameter were used The objective of this study was to confirm this hypothesis, that using a larger CANDU-standard pressure tube would give a larger CVR bias for a 28-NU fuel bundle, as computed by WIMS-IST in comparison to experimental measurements of critical buckling. Thus, new critical-height and flux-map measurements were performed in substitution experiments in the ZED-2 research reactor to determine the pure critical lattice buckling for 28-element fuel with standard-size CANDU pressure tubes. The derived buckling from these experiments were used in WIMS-IST transport calculations to determine the effective multiplication factors for cooled and voided lattices and hence the bias in the CVR. Calculation results demonstrated that the CVR bias for the 28-NU was ∼ 1.7 mk ± 0.42 mk, which is consistent with the results for 37-element and 43-element CANDU-type lattices. (authors)

[en] Rockford Technology Associates, Inc. (RTA) has been doing experiments on the Dense Plasma focus (DPF) device at the Fusion Studies Laboratory of the University of Illinois. This DPF consists of four racks of five 2-μF capacitors whose charge is switched onto the inner electrode of a plasma focus by four Trigatron spark gaps. The stored energy is 12.5 kJ at 25 kV. The bank is usually discharged in a static fill of H₂ at ∼ 6 torr. Preliminary experiments aimed at exploring the potential of the DPF device as a magnetoplasmadynamic (MPD) thruster and as an x-ray source for lithography have investigated various alternative ways of injecting gas between the electrodes. One of those approaches consists of injecting gas from the tip of the inner electrode at a steady rate. In this operation, the DPF chamber pressure was held constant by running the vacuum pump at full throttle. This operation simulated simultaneous pulsed injection at the base insulator and electrode tip. Hydrogen was fed through a 1/16th-inch hole at a flow rate of ∼ 90 cm/s. Pulsing was then performed at 23 kV, and the corresponding variations of the current were observed using a Rogowski coil. It is found that the plasma collapses into a pinch at the same time as in conventional experiments using a static fill. The singularity in the current waveform is slightly smaller with tip injection, but its size and shape are easily reproducible. Further details and comparison of this operation with conventional pulsing will be presented

[en] This paper summarizes calculations of MCNP5 and WIMS-AECL/RFSP compared against measurements in coolant void substitution experiments in the ZED-2 critical facility with CANFLEX R-LEU/RU (Low Enriched Uranium, Recovered Uranium) reference fuels and CANFLEX-LVRF (Low Void Reactivity Fuel) test fuel, and H₂O/air coolants. Both codes are tested for the prediction of the change in reactivity with complete voiding of all fuel channels, and that for a checkerboard voiding pattern. Understanding these phenomena is important for the ACR-1000 R reactor. Comparisons are also made for the prediction of the axial and radial neutron flux distributions, as measured by copper foil activation. The experimental data for these comparisons were obtained from critical mixed lattice / substitution experiments in AECL's ZED-2 critical facility using CANFLEX-LEU/RU and CANFLEX-LVRF fuel in a 24-cm square lattice pitch at 25 degrees C. Substitution analyses were performed to isolate the properties (buckling, bare critical lattice dimensions) of the CANFLEX-LVRF fuel. This data was then used to further test the lattice physics codes. These comparisons establish biases/uncertainties and errors in the calculation of k_eff, coolant void reactivity, checkerboard coolant void reactivity, and flux distributions. Results show small to modest biases in void reactivity and very good agreement for flux distributions. The importance of boundary conditions and the modeling of un-moderated fuel in the critical experiments are demonstrated. This comparison study provides data that supports code validation and gives good confidence in the reactor physics tools used in the design and safety analysis of the ACR-1000 reactor. (authors)

[en] Inertial Electrostatic Confinement (IEC) is a unique approach to fusion and plasma energy systems that was conceptualized in the 1960s (Hirsch 1967) and has been the focus of recent development in the 1990s (Miley et al. 1995a). In the interests of space power and propulsion systems, conceptual rocket design studies (Bussard and Jameson 1994, Miley et al. 1995b) using the IEC have predicted excellent performance for a variety of space missions, since the power unit avoids the use of magnets and heavy drives resulting in a very high, specific impulse compared to other fusion systems. In their recent survey of prior conceptual design studies of fusion rockets, Williams and Borowski (1997) found that the Bussard IEC conceptual study (the ''QED'' engine) offered a thrust-to-weight ratio of 10 milli-g's, a factor of five higher than conventional magnetic confinement concepts and even slightly above anti-proton micro fission/fusion designs. Thus there is considerable motivation to study IEC concepts for eventual space applications. However, the physics feasibility of the IEC still requires experimental demonstration, and an expanded data base is needed to insure that a power unit can in fact be built