Pressure Tube Heavy Water Reactors Research Articles

Physics Assessment of the Impact of Modified End Pellets on Axial Power Peaking for Advanced/Nonconventional Uranium-Based Fuels in Pressure Tube Heavy Water Reactors

Abstract Axial power peaking is a phenomenon with safety implications for pressure tube heavy water reactors (PT-HWRs). Since PT-HWRs use shorter (∼50 cm) bundles, there are small axial gaps, which expose the ends of the fuel elements to more neutron flux, and therefore results in higher power density levels occurring in the ends of the fuel elements. Power peaking has the potential to cause fuel damage and failure, if the local linear element rating (LER) exceeds 57 kW/m, and may be of greater concern for advanced, higher burnup fuels. Earlier studies have been done using three-dimensional mcnp models of a PT-HWR fuel bundle with slightly enriched uranium; they demonstrated that ThO2 could be used to reduce axial power peaking in fresh fuel. This result was achieved by replacing some of the UO2 with ThO2 in the last 3 cm of fuel pellets at each end of a fuel bundle. This study extends the previous work by performing 3D neutronics and burnup calculations using serpent, to evaluate how power peaking changes with burnup. In addition, alternative dilution materials (such as depleted UO2, ZrO2, and MgO) were also evaluated. It was found that axial power peaking can be significantly reduced by using the ThO2 dilution material for fresh fuel, while ZrO2 or MgO are even more effective at higher burnup levels. Dilution materials have little impact (less than 2%) on the exit burnup of the fuel.

Review of Methods and Results for Reactor Physics Analysis of Thorium-Based Fuels From Irradiation Experiments Conducted in the National Research Universal Reactor

Abstract The development of advanced fuels and fuel cycles for conventional Generation III+ reactors, advanced reactors (such as Generation IV), and small modular reactors will help ensure the long-term sustainability and safety of the use of nuclear energy. Thorium-based fuels are an example of an advanced fuel that could augment and extend uranium resources. As part of previous efforts in Canada to investigate the use of thorium-based fuels for potential use in pressure-tube heavy-water reactors and other technologies, irradiation experiments with various thorium-containing fuels had been conducted in the National Research Universal (NRU) reactor at the Chalk River Laboratories. The NRU reactor physics computational tools and methods, developed initially for conventional enriched uranium fuel, had also been extended for analysis of thorium-based fuel experiments, with most satisfactory results. In this study, a number of physics and operation factors have been analyzed to quantify their effects on the production and burnup of 233U in thorium-based fuel, as well as to improve both the analysis and the use of thorium fuels and fuel cycles.

Study of natural uranium utilization in a heavy water moderated molten salt reactor

Heavy water moderated molten salt reactor (HWMSR) is a newly proposed molten salt reactor. It adopts heavy water as the moderator and molten salt dissolved with actinide elements as the fuel and coolant as well. These unique features endow HWMSR a high neutron economy as performed in CANDU (CANada Deuterium Uranium) pressure tube heavy water reactor and a capability to implement the processes of online refueling and reprocessing as done in the conventional MSRs, making natural uranium (NU) utilization and together with U-233 production become feasible. The obtained results in this study demonstrate that the fuel utilization efficiency of HWMSR with NU fuel cycle can achieve 1%, slightly better than that of the CANDU reactor at the same physical size and power level. Meanwhile, around 2662 kg U-233, which can start 3 new Th–U fuel cycle HWMSR cores, can be produced over 20 years’ operation by online refueling thorium and extracting Pa-233.

Investigation of Critical Heat Flux for Plutonium-Based Mixed Oxide Advanced Fuel Bundle Design

Abstract The critical heat flux (CHF) performance of an advanced plutonium-based mixed oxide (MOX) fuel for potential use in a pressure tube heavy water reactor (PT-HWR) has been studied experimentally at Canadian Nuclear Laboratories with an electrically heated string simulator of 43-element fuel bundles. The fuel simulator has a uniform axial power profile and a radial power profile that is representative of the plutonium-based MOX fuel. The measurements of CHF caused by liquid-film dryout were made in the MR-3 heat transfer loop facility using R-134a refrigerant as the working fluid. The test matrix included system pressures from 1.47 to 2.11 MPa, mass flow rates from 12.7 to 14.7 kg/s and inlet temperatures from 31 to 59 °C, which are representative of the water-equivalent reactor operating conditions of 9 to 12.5 MPa pressure, 13.5 to 21.3 kg/s mass flow rate and the desired inlet subcoolings. Compared to conventional natural uranium fuel, the radial power profile of a MOX fuel exhibits a steeper and less even distribution across the fuel element rings, with a relatively higher power in the outer ring. It was found that CHF values of the MOX fuel are significantly lower than those of the natural uranium fuel. Based on the experimental data, a correlation has been derived to account for the effect of radial power profile on CHF. This correlation can be used to evaluate the relative CHF values of advanced or nonconventional fuel designs with radial power profiles deviating from that of natural uranium fuel.

Preliminary Evaluation of the LASSO Method for Prediction of the Relative Power Density Distribution in Mixed Oxide (Pu, DU)O2 Fuel Pellets

Abstract Predicting the power distribution within nuclear fuel is essential for predicting reactor fuel performance, since power distributions can impact pellet temperature distributions and fission product transport and migration. Analytical expressions for radial power distribution in fuel pellets were sought using lattice physics calculations to generate data and a machine learning technique to find representative expressions. Analytical approximations can be useful in nuclear fuel performance codes, such as element simulation and stresses (ELESTRES)/ element simulation code in a loss of coolant accident (ELOCA) for providing very rapid predictions of power distributions with reduced computational effort and memory requirements, relative to using an embedded or coupled neutron transport/burnup reactor physics code. Radial power distributions were calculated a priori using lattice physics codes to model mixed oxide (MOX) 37-element fuel bundles in pressure tube heavy water reactors. Such advanced fuels are of interest for future fuel cycles. Several datasets were generated with different amounts of PuO2 and variable neutron energy spectrum. Results of preliminary studies with the least absolute shrinkage and selection operator (LASSO) regression machine learning method have obtained analytical fitting functions with a mean maximum relative error (MRE) of 0.056 and a maximum MRE of 0.152 on the test set. However, using LASSO to estimate the coefficients of a physically motivated modified Bessel plus an exponential function, results in a lower MRE (mean MRE 0.041 and maximum MRE 0.11) on the same test set. Further potential improvements in both the curve fit and the machine learning methods are discussed.

Initial Exploratory Reactor Physics Assessment of Nonconventional Fuel Concepts for Very Compact Small Modular Reactors Using Hydroxides as Coolants and/or Moderators

In this study, lattice physics calculations were carried out to evaluate the reactor physics characteristics of different advanced fuel lattices cooled with 7LiOH/NaOH or FLiBe and moderated externally by graphite and various types of metal hydroxides, such as 7LiOH, 7LiOD, Mg(OD)2, and ZrE(OD)4. The lithium in these compounds is enriched to 99.995 at. % 7Li/Li. Such lattice fuel concepts could be used in compact, thermal-spectrum, high-temperature (700°C) small modular reactors (SMRs). For an SMR with a bare core size of diameter = height = 163.3 cm, there are several lattice design concepts identified that could achieve modest power densities (up to 18 MW/m3) that are higher than those found typically in high-temperature gas cooled reactors (~ 2 to 10 MW/m3) [IAEA Technical Document 1382 (2019); Report PNR-131-20110914, Delft University, Netherlands (2011)], although lower than those found typically in SMRs based on light water reactor technology (for example, the NuScale SMR has a volumetric power density of ~47 MW/m3) [Proc. PBNC 2018, p. 270 (2018)]. In addition, there are lattice designs identified for the fixed core size that could achieve high fuel burnup (up to 126 MWd/kg), long core lifetimes (up to 24 years before refueling), very good fissile utilization (up to 640 MWd/kg-fissile), and very good relative uranium utilization (up to 44% of that achieved with a conventional pressure-tube heavy water reactor using natural uranium fuel). The best lattice concept found to maximize fuel burnup with 7LiOH/NaOH coolant was an 18-cm-pitch lattice with ZrE(OD)4 external moderator (126.5 MWd/kg). The best lattice concept for FLiBe coolant was a 16-cm-pitch lattice with 7LiOH external moderator (125.99 MWd/kg). Although it is recognized that there are numerous and challenging technical issues to be resolved, particularly with corrosion and materials science, the potential use of hydroxides as coolants and/or external moderators could lead to very important performance improvements for very small and compact SMRs.

Physical Phenomena in Nuclear Thermal Hydraulics and Current Status

116 nuclear Thermal-Hydraulic Phenomena T-HP are identified in the present paper, following documents issued during the last three decades by the Committee on the Safety of Nuclear Installations of Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA/CSNI) and by the International Atomic Energy Agency (IAEA). The derived T-HP list includes consideration of experiments performed in Separate Effect Test (SET) and Integral Effect Test (IET) facilities relevant to reactor coolant system and containment of Water Cooled Nuclear Reactors (WCNR). We consider a dozen WCNR types: Pressurized Water Reactors (PWR), Boiling Water Reactors (BWR), Russian reactors (VVER-440, VVER-1000 and RBMK), pressure tube heavy water reactors by Canada (CANDU) and India (PHWR) and so-called ‘advanced’ reactors (e.g. AP-1000 and APR-1400 designed in US and Korea, respectively). We envisage a variety of applications for the T-HP list. Four of the phenomena are helpful to characterize the current state of art in nuclear thermal-hydraulics: Counter Current Flow Limitation (CCFL), Critical Heat Flux (CHF), reflood and Two-Phase Critical Flow (TPCF). Furthermore, the T-HP identification contributes to addressing the scaling issue, performing uncertainty evaluations, developing constitutive equations and ‘special models’ in codes and prioritizing the research.

Reactor Physics Assessment of Potential Feasibility of Using Advanced, Nonconventional Fuels in a Pressure Tube Heavy Water Reactor to Destroy Americium and Curium

Lattice physics and core physics studies have been carried out to investigate the reactor physics feasibility of destroying americium (Am) and curium (Cm) using special target fuel bundles in blanket fuel channels in a heterogeneous seed-blanket pressure tube heavy water reactor (PT-HWR) core fueled primarily with natural uranium. Results indicate that it should be feasible to achieve net-zero production of Am in a single PT-HWR core using 10 to 16 dedicated blanket channels containing Am-based target bundles while only one dedicated blanket channel would be required for achieving net-zero production of Cm. While the use of target blanket fuel bundles with fuel elements made of Am or Cm mixed with thorium (Th) in oxide form ((Am,Th)O2, (Cm,Th)O2) is expected to be suitable for transmutation purposes, the use of fuel elements made of pure americium oxide, especially those in the form of AmO1.55, may not be suitable for transmutation purposes because of potential issues with fuel melting under high-power operations or postulated accident scenarios. The potential to achieve net-zero production of Am and Cm in a single thermal-spectrum reactor, such as a PT-HWR, could help eliminate the need to build and qualify a deep geological repository (DGR) capable of storing minor actinides for a long time (>1 million years). At the very least, the size and/or number of DGRs required for storing radioactive waste could be reduced significantly. Thus, destroying Am and Cm in PT-HWRs could be regarded as a viable solution to the perceived problem of nuclear waste and may help improve public acceptance of the use of nuclear energy. In addition, it may be possible to apply a similar approach for destroying MAs in other Generation III+ (Gen-III+)/Generation IV (Gen-IV)/small modular reactor (SMR) technologies.

Sensitivity Studies to Assess the Impact of Geometry and Operating/Boundary Condition Perturbations on Thermal-Hydraulic Behavior of Advanced Fuel Channels in Pressure Tube Heavy Water Reactors with Uranium and Thorium-Based Fuels

This paper reports the results of subchannel thermal-hydraulic studies (using the ASSERT-PV code) of the effects of variations and uncertainties in operating/boundary conditions and geometry on the predictions of pressure drop, dryout power, and dryout location for two types of advanced, nonconventional fuels in a pressure tube heavy water reactor (PT-HWR) fuel channel with 12 fuel bundles. The fuel bundles tested include a 37-element fuel bundle made with SEUO2 (1.2 wt% 235U/U), with a central fuel element made of ThO2, and 35-element fuel bundle made with (LEU,Th)O2, using 5 wt% 235U/U low-enriched uranium (LEU), 50 wt% LEUO2, and 50 wt% ThO2. Results indicate that for a range of flow conditions, the dryout power for the thorium-based 35-element fuel bundle is 10% to 26% higher than that for the uranium-based 37-element fuel bundle. Variation/uncertainty in the pressure tube diameter has the most significant impact on the pressure drop, dryout power, and dryout location. Results from these studies may have implications for the operations of PT-HWRs with advanced fuels, and further modifications may be desirable to further enhance thermal-hydraulic margins.

Reactor Physics Analysis Assessment of Feasibility of Using Advanced, Nonconventional Fuels in a Pressure Tube Heavy Water Reactor to Destroy Long-Lived Fission Products

Lattice physics and core physics studies have been carried out to investigate the feasibility of destroying long-lived fission products (LLFPs) using special target fuel bundles in blanket fuel channels in a seed-blanket core in a pressure tube heavy water reactor (PT-HWR) fueled primarily with natural uranium. Results indicate that it should be feasible to achieve net zero production of LLFPs such as 79Se and 129I using one to two dedicated blanket channels containing LLFP target bundles. With 60 blanket channels, the net production rate of 99Tc or 126Sn could be reduced by 75% or more. Further design modifications may be able to achieve net zero production for most LLFPs, with the exception of 135Cs, which would require isotopic separation.

Steady-State Subchannel Thermal-Hydraulic Assessment of Advanced Uranium-Based and Thorium-Based Fuel Bundle Concepts for Potential Use in Pressure Tube Heavy Water Reactors

The use of advanced uranium-based and thorium-based fuel bundles in a 700-MW(electric)–class pressure tube heavy water reactor (PT-HWR) has the potential for improved performance characteristics with higher burnup, higher fissile fuel utilization, and lower coolant void reactivity while also extracting the energy potential in thorium. In this study, thermal-hydraulic subchannel analyses were performed for a single, high-power (6.5 MW), 12-bundle fuel channel at typical reactor operating conditions for 14 different PT-HWR lattice/core concepts using various types of advanced uranium-based and thorium-based fuels in 37-element and 35-element fuel bundle design concepts. Fuel bundle radial power distributions for fresh fuel at zero burnup were used in the thermal-hydraulic calculations, as a bounding case, along with axial power distributions that are representative of those that may be found in a high-power fuel channel in a PT-HWR core at near-equilibrium refueling conditions. The fuel bundle radial power distributions and fuel channel axial power distributions were determined from previous lattice physics and core physics studies. Based on the subchannel thermal-hydraulic analyses, the LC-05b/CC-04 BUNDLE-37-mod concept and the LC-12b/CC-08 BUNDLE-35 concept are recommended as the best candidates for further full-core system thermal-hydraulic transient analyses, based on critical heat flux and void fraction performance factors. BUNDLE-37 concept LC-01/CC-00 is also recommended as the reference case for future analysis.

Transient System Thermal-Hydraulic Assessment of Advanced Uranium- and Thorium-Based Fuel Bundle Concepts for Potential Use in Pressure Tube Heavy Water Reactors—I: Two-Channel Analyses

The use of advanced uranium-based and thorium-based fuel bundles in pressure tube heavy water reactors (PT-HWRs) has the potential to improve the utilization of uranium resources while also providing improvements in performance and safety characteristics of PT-HWRs. Previous lattice physics and core physics studies have demonstrated the feasibility of using such advanced fuels; however, thermal-hydraulic (T-H) studies are required to confirm that these advanced fuels will have adequate T-H safety margins. Preliminary system T-H transient simulations have been performed for a 700-MW(electric)–class PT-HWR in a postulated loss-of-coolant accident (LOCA) using the CATHENA code. One purpose of this work was to demonstrate that such simulations of a PT-HWR with advanced fuels could be set up and executed successfully in a CATHENA transient simulation model. The other purpose was to evaluate the peak fuel sheath and fuel centerline temperatures in two designated fuel channels containing advanced uranium-based or thorium-based fuel during a LOCA transient event. In the CATHENA simulation models, a PT-HWR core is fueled with conventional 37-element natural uranium fuel bundles in 378 out of 380 fuel channels while two designated fuel channels, the channel with the highest total power and the channel containing the bundle with the highest power level, are filled with various types of advanced fuels. Results indicate that setting up these models is feasible and that the predicted peak fuel centerline temperatures and peak sheath temperatures for the advanced fuel channels are well below the fuel melting points and the rapid oxidation temperature for the Zircaloy-4 sheath/clad (~1200°C), respectively. These preliminary results provide confidence that the advanced fuels will likely have adequate T-H safety margins in a transient LOCA event in a PT-HWR. These results set the stage for more detailed and comprehensive system T-H models of PT-HWRs fueled entirely with advanced uranium-based or thorium-based fuels.

Transient System Thermal-Hydraulic Assessment of Advanced Uranium- and Thorium-Based Fuel Bundle Concepts for Potential Use in Pressure Tube Heavy Water Reactors—II: Full-Core Analyses

The use of advanced uranium-based and thorium-based fuel bundles in pressure tube heavy water reactors (PT-HWRs) has the potential to improve the utilization of uranium resources while also providing improvements in performance and safety characteristics of PT-HWRs. Earlier lattice physics and reactor core physics studies have demonstrated the feasibility of using such advanced fuels; however, thermal-hydraulic (T-H) studies are required to confirm that these advanced fuels will have adequate T-H safety margins. Preliminary system T-H transient simulations have been carried out for a 700-MW(electric)–class PT-HWR in a postulated loss-of-coolant accident (LOCA) using the CATHENA code. One purpose of this work was to demonstrate that such simulations of a PT-HWR filled entirely with advanced fuels could be set up and executed successfully in a CATHENA transient simulation model. The other purpose was to evaluate the peak sheath and peak fuel centerline temperatures during a LOCA to perform an analysis that compares the relative performance of each of the proposed advanced fuels. System T-H simulations with CATHENA were performed to model a postulated LOCA event with a 20% inlet header break in a typical 700-MW(electric)–class PT-HWR using two types of advanced uranium-based and thorium-based fuel bundles in modified 37-element and 35-element geometries. Calculations were also performed for a PT-HWR using conventional natural uranium fuel in 37-element fuel bundles for comparison. In the event of a LOCA, there is a drop in the primary circuit pressure. It is assumed that there is a 2-s delay between the signal of the low primary pressure and the tripping of the reactor. When the reactor trips, the shutdown rods are inserted. The reactor trip is followed by the activation of the emergency core cooling system, which occurs 30 s after the LOCA starts, with a trip signal on the boiler crash cooling. Simulation results for the LOCA demonstrated that the peak fuel centerline temperatures (ranging from 1822°C to 2183°C) were several hundred degrees below the expected melting point of UO2 (~2865°C). Simulations also demonstrated that the peak sheath temperatures for the advanced fuel concepts ranged from 1177°C to 1204°C, which are lower than that with conventional NU fuel in 37-element fuel bundles. Thus, the system T-H analysis of the relative results provides confidence in the proposed advanced uranium-based and thorium-based fuel concepts for potential use in PT-HWRs.

Measurements on (233U, Th)O2 fuel by substitution experiments in the ZED-2 critical facility

This paper summarizes the results of experimental measurements on mixed-oxide (233U,Th)O2 fuel bundles in the ZED-2 heavy water critical facility of Canadian Nuclear Laboratories (CNL) in Chalk River, Canada. Pressure-tube heavy water reactors can use such fuel to extract the energy potential of thorium, which helps extend nuclear energy resources and reduces waste. Measurements were made in a series of substitution ZED-2 experiments, with (233U,Th)O2 test fuel placed into vertical channels filled with air or heavy water in a reference lattice of very low-enriched uranium fuel. Fuel was arranged to maximize the importance of the test fuel. Each fuel and coolant configuration in the substitution series was simulated in MCNP5 to verify the measurement of critical moderator height. The precision of the substitution measurements can help support the validation of reactor physics codes for evaluation of reactivity coefficients in reactors with fuel that contains oxides of thorium and thorium-derived uranium.

Recent measurements on plutonium-thorium fuel by substitution experiments in the ZED-2 critical facility

This paper summarizes the results of experimental measurements recently made on heavy water moderated plutonium-thorium (Pu-Th) mixed-oxide fuel bundles in the ZED-2 critical facility of Canadian Nuclear Laboratories. Pu-Th fuel in pressure-tube Heavy water reactors can help extend nuclear energy resources by utilizing plutonium from spent fuel, and extracting the energy potential of thorium. The ZED-2 experiment was a series of substitution measurements, with (Pu,Th)O2 test fuel placed into vertical channels filled with air or heavy water in a reference lattice of very low-enriched uranium (LEU) fuel. Sensitivity to the test fuel was optimized by placement into positions of higher power. MCNP5 simulations of the fuel substitution series are within uncertainty of critical moderator height measurements. The precision of the substitution measurements can help support the validation of reactor physics codes for evaluation of reactivity coefficients in PT-HWRs with fuel that contains a mixture of thorium and plutonium oxides.

Advances of the AC-DC code, a coupled computational tool to perform thermalhydraulic modeling of fuel bundles with annular fuel elements

A new computational tool, the AC-DC code, is presented that couples the two well-established two-phase thermalhydraulic codes ASSERT (a subchannel thermalhydraulics code) and CATHENA (a system thermalhydraulics code). The AC-DC code can be used to model, evaluate, and optimize the thermalhydraulic performance characteristics for advanced fuel bundle concepts involving the use of annular-type fuel elements that are cooled both internally and externally. Such innovative concepts may be potentially implemented in various research and power reactors, including small modular reactors (SMRs), Generation-IV systems, and pressure tube heavy water reactors (PT-HWRs). In this study, AC-DC code is used to simulate the performance of proposed 24-element and 18-element internally-cooled annular fuel (ICAF) bundles in single-phase and two-phase flows. Such ICAF bundle concepts could potentially be implemented in PT-HWRs, in conjunction with the use of high-burnup and advanced fuels, such as (Pu, U)O2 (MOX), thorium-based fuels, metal alloys, nitride fuels, silicide fuels and others. The use of ICAF-type fuels have the potential advantage of reduced fuel temperatures, and enhanced thermalhydraulic margins. The AC-DC code simulation results show excellent agreement with available single-phase experimental results. Two-phase simulations highlight the usefulness of the coupled code to help optimize the performance of proposed ICAF concepts. The results show encouraging performance characteristics for ICAF concepts even using relatively conservative assumptions.

Coupling of reactor transient simulations via the SALOME platform

Canadian Nuclear Laboratories has developed a toolset of coupled reactor transient simulation codes for pressure tube heavy water reactors. This paper presents the first demonstration of the toolset’s capabilities, using a hypothetical 20% reactor inlet header break (critical break) scenario. This scenario induces strong feedback between thermalhydraulic and reactor physics phenomena, owing to a positive coolant voiding coefficient. The scenario was simulated using the advanced thermalhydraulics code ARIANT and the reactor physics code NESTLE-C, coupled together using the SALOME platform. This paper summarizes recent advances in ARIANT, the coupling methodology, and the results of the transient simulation.

MODELING STUDIES AND CODE-TO-CODE COMPARISONS FOR PRESSURE TUBE HEAVY WATER REACTOR CORES

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 (keff), 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.

Physics characteristics of internally cooled annular fuel for potential application in pressure tube heavy water reactors

This paper summarizes the results of exploratory lattice physics studies of alternative, advanced fuel bundle concepts that could potentially be implemented in pressure tube heavy water reactors (PT-HWRs). The lattice physics code WIMS-AECL was used to analyze the physics performance and operational characteristics of an 18-element and a 12-element internally cooled annular fuel (ICAF) fuel bundle, made with (LEU,Th)O2 fuel, with both low-burnup and high-burnup options. Such fuel bundles with annular fuel elements may be able to operate at higher bundle power levels and with higher linear element (LER) ratings than fuel bundles with conventional solid cylindrical fuel elements. In addition, the use of thorium mixed with LEU can help extend uranium resources, exploit the energy potential in thorium, and also reduce the production of plutonium and minor actinides, due to the smaller fraction of 238U in the fuel. The influence of improvements in the neutron economy of these lattices was also investigated, by incorporating higher-purity heavy water moderator and coolant (99.90 at.% D2O) in the models, along with enriched zirconium (95 wt% 90Zr/Zr) for the zirconium alloys used in the structural components. The results were compared with those for a more conventional 37-element PT-HWR fuel bundle using natural uranium (NU) fuel. Results show that annular fuels could be very attractive, being able to achieve higher burnup, comparable or better fissile utilization, reduced coolant void reactivity, and comparable or more negative fuel temperature coefficients.

Optimization studies for planning of substitution experiments in the ZED-2 critical facility with plutonium-thorium test fuel

Reactivity data from experiments in the Zero Energy Deuterium (ZED-2) critical facility enable advanced pressure-tube heavy water reactor (PT-HWR) fuel design concepts to be tested. Experiments on a limited quantity of test fuel show an increased relative uncertainty, due to the influence of the larger reference lattice, but new substitution analysis methods increase the flexibility of fuel substitution configurations. MCNP computer simulations are used to optimize the configurations for better sensitivity to the test fuel, which is found to enhance the precision substantially. Conceptual designs with thorium-based fuel cycles have been proposed to improve resource utilization. A plan has been developed for a substitution experiment to be conducted on plutonium-thorium test fuel in a smaller reference lattice of slightly enriched uranium fuel, and more of the test fuel in regions of higher importance. The findings suggest an opportunity to glean better data from optimization of new substitution measurements on other experimental fuel types that are already available to the ZED-2 critical facility.