Hugues W. Bonin

Release of iodine and noble gas fission products from defected fuel elements during reactor shutdown and start-up

This paper provides data from a single defected fuel element operating in-reactor are used to develop a physically based model for describing the increased release of iodine and noble gas fission products to the primary coolant following a reactor shutdown and start-up. Iodines are only released on reactor shutdown, whereas both species are observed on start-up. a dependence on the decay constant of {lambda} {sup {minus}3/2} is typically observed for these transient releases, indicative of a source of release from fuel cracking with little holdup due to transport in the fuel to-sheath gap.

Monte Carlo Simulation of the LEU-Fueled SLOWPOKE-2 Nuclear Reactor Using MCNP 4A

The availability of the Monte Carlo-based code MCNP 4A has made possible the simulation of the low-enriched uranium (LEU)-fueled SLOWPOKE-2 reactor using a probabilistic approach The reactor core and its surrounding pool can be modeled in three dimensions with numerous details included in the representation. Significant improvement from previous modeling attempts was obtained with the MCNP 4A simulation, with the discrepancy between the calculated and experimental values of the excess reactivity at 20°C reduced to only 0.2 mk, The analysis suggests the error of the MCNP 4A-calculated excess reactivity as between I and 2 mk. The SLOWPOKE-2 reactor was then simulated with its single control rod at various degrees of insertion in the core. The reactivity worth of the rod was calculated as 7.85 mk, only 2.4 mk above the measured value. MCNP was then used for predicting the temperature effects on the excess reactivity. Although the inherent safety of the SLOWPOKE-2 reactor was confirmed in the simulation, the temperature dependence of the excess reactivity, could not however be accurately predicted, due for the most part to the lack of appropriate cross-section libraries available at the time of this work. The potential qf MCNP 4A is nevertheless clearly demonstrated for the simulation of the LEU-fueled SLOWPOKE-2 reactor, once the missing cross sections become available for the low temperatures at which the reactor operates.

Application of polymers for the long-term storage and disposal of low- and intermediate-level radioactive waste

Abstract Research carried out at the Royal Military College of Canada on the effects of mixed fields of radiation on high polymer adhesives and composite materials has shown that some polymers are quite resistant to radiation and could well serve in the fabrication of radioactive-waste disposal containers. A research program was launched to investigate the possibilities of using advanced polymers and polymer-based composites for high-level radioactive waste management on one hand and for intermediate- and low-level radioactive waste disposal on the other hand. Research was thus conducted in parallel on both fronts, and the findings for the later phase are presented. Thermoplastic polymers were studied for this application because they are superior materials, having the advantage over metals of not corroding and of displaying high resistance to chemical aggression. The experimental methods used in this research focused on determining the effects of radiation on the properties of the materials considered: polypropylene, nylon 66, polycarbonate, and polyurethane, with and without glass fiber reinforcement. The method involved submitting injection-molded tensile test bars to the mixed radiation field generated by the SLOWPOKE-2 nuclear reactor at the Royal Military College of Canada to accumulate doses ranging from 0.5 to 3.0 MGy. The physical, mechanical, and chemical effects of the various radiation doses on the materials were measured from density, tensile, differential scanning calorimetry, and scanning electron microscopy tests. For each polymer, the test results evidenced predominant cross-linking of the polymeric chains severed by radiation. This was evident from observed changes in the mechanical and chemical properties of the polymers, typical of cross-linking. The mechanical changes observed included an overall increase in density, an increase in Young’s modulus, a decrease in strain at break, and only minor changes in strength. The chemical changes included differences in chemical transition temperatures characteristic of radiation damage. All the changes in these properties are characteristic of the cross-linking phenomenon. For the glass-fiber-reinforced polymers, the results of the tests evidenced minor radiation degradation at the fiber/matrix interfaces. Based on these results, any of the investigated polymers could potentially be used for disposal containers due to their abilities to adequately resist radiation. This allowed proceeding one step further into determining a potential design framework for containers for the long-term storage and disposal of low- and intermediate-level radioactive waste.

The near boiling reactor : Design of a small, inherently safe, nuclear reactor to extend the operational envelope of the victoria-class submarine

The present work aims at initiating the conceptual design of a small nuclear reactor intended to provide sufficient electrical power (˜150 kW) to maintain the “hotel” load of the Victoria-class submarine and extend her operational envelope. The scope of the design is to provide the nuclear reactor system with sufficient inherent safety features as to permit the operation of the nuclear reactor by crews with minimal training for automatic operation. Several constraints provide the framework for carrying out the design work, such as, among others, maintaining the excess reactivity of the reactor at safe values at all times, providing enough fuel and reactivity for meeting operational requirements, and keeping the size of the reactor core and shielding such as to fit within the hull of the existing vessel. The final reactor concept, named the Near Boiling reactor, employs TRISO fuel particles in zirconium-sheathed fuel rods. The reactor is light water moderated and cooled. The core life is specifically designed to coincide with the refit cycle of the Victoria-class submarine. The reactor employs a simple and reliable control and shutdown system that requires little intervention on the part of the submarine’s crew. Also, a kinetic model is developed that demonstrates the inherent safety features of the reactor during several accident scenarios. The low steady-state flux level of the reactor during normal operation results in very low negative reactivity after shutdown and eliminates any reactor dead time. The reactor is designed for automatic unattended control and does not require extensive training for its operators.

CANDU pressurized heavy water reactor thorium-/sup 233/U oxide fuel evaluation based on optimal fuel management

The optimization problem of the in-core fuel management of a thorium-fueled CANDU pressurized heavy water reactor (PHWR) consists of several component actions: the number of fresh fuel bundles inserted in the channels, the choice of the channels to be refueled next, the refueling rate, and the composition of the fresh fuel bundles (the latter relevant to advanced fuel cycles). Several fresh fuel compositions of /sup 232/Th and /sup 233/U were investigated and compared to the self-sufficient equilibrium thorium (SSET) cycle fuel in terms of the objective function of an optimal fuel management problem. This optimization problem consisted of minimizing the total refueling rate at equilibrium with respect to critically and power peaking constraints. The maximum acceptable value of the form factor was equal to 1.20, the form factor defined as the maximum-to-average power density ratio in the reactor core. The reactor core was divided into two refueling zones, each characterized by a uniform refueling rate for its channels.

ON FINITE-DIFFERENCE SOLUTIONS OF THE HEAT EQUATION IN SPHERICAL COORDINATES

The finite-difference solution for the temperature distribution within a sphere exposed to a nonuniform surface heat flux involves special difficulties because of the presence of mathematical singularities. For this reason, the adequacy of some finite-difference representations of the heat diffusion equation is examined. In particular, neglecting the contribution from the term causing the singularity is shown as an accurate and efficient method of treating a singularity in spherical coordinates. A method based on the superposition principle is investigated and found quite suitable for this kind of problem in spherical coordinates.

Variation of Burnable Neutron Absorbers in a Heavy Water–Moderated Fuel Lattice: A Potential to Improve CANDU Reactor Operating Margins

Abstract A CANDU lattice cell has been modeled using the Los Alamos National Laboratory’s MCNP 6 code and Atomic Energy of Canada Limited’s WIMS-AECL 3.1. Models for the CANDU 37-element fuel bundle have included a CANLUB coating, as a carrier for the neutron absorbers. The objective is to improve CANDU reactor operating margins by adding small amounts (~1 g) of neutron absorbers to each fuel element. For CANDU natural uranium fuel bundle design, the results indicate that (a) the fueling transient (due to the xenon-free effect) could be significantly reduced using gadolinium oxide (Gd2O3), with no significant impact on fuel burnup, and (b) the reactivity peak (due to plutonium production) could be reduced using europium oxide (Eu2O3), with minimal impact on fuel burnup. An appropriate mixture of Gd2O3 and Eu2O3 that will improve operation and safety margins while having a minimal impact on fuel burnup is determined. Reactivity and power calculations for various mixtures of Gd2O3 and Eu2O3 are reported here. It is concluded that ~180 mg Gd2O3 and ~1000 mg Eu2O3 (~4.9 ×10−3 wt% per bundle) are sufficient to suppress the refueling transient and lower the axial plutonium peak, with a 0.27% burnup penalty (which is a small impact). Fuel safety and performance are always important topics for a nuclear utility. This approach of a relatively simple application of burnable poisons to existing CANDU natural uranium fuel design offers the benefits of improving fuel utilization and safety margins.

Design of a neutron gauge for the detection and measurement of water ingression in flat roofs

This paper reports on a portable neutron gauge designed to detect water ingression in flat roofs and to measure with good accuracy the moisture content in the roofing materials. The gauge consists of a small {sup 252}Cf neutron source inserted in a collimator head made of borated paraffin contained in a steel vessel. Neutron detection is performed with a boron trifluoride detector and the associated electronic counting equipment. Experimental testing, calibration, and assessment are done in the laboratory using full-scale models of typical, Canadian-built flat roofs. Several experiments are conducted to determine the sensitivity of the gauge for various controlled water densities in the roofing insulation materials and for a large selection of geometries for the source and the detector with respect to the roof surface.

Neutron Absorbers in CANDU Natural Uranium Fuel Bundles to Improve Operating Margins

Safety margins are particularly tight in natural uranium-fuelled CANDU reactors which are refueled on-power. During on-power refueling, the insertion of xenon-free fresh fuel bundles into the reactor core affects the reactor’s excess reactivity in such a way that this could lead to temporary power derating. It is desirable from a fuel management perspective, and to maintain safety margins to eliminate this xenon-free effect and any other power ripples such as the subsequent plutonium reactivity peak. A redesign of the CANDU NU fuel bundle with an appropriate combination of elements, with some including neutron-absorbers, could well address the issue of the xenon-free initial portion of the bundle’s irradiation and also lower the plutonium-peak that occurs shortly thereafter. This may improve the fuel utilization (by further optimizing the fuelling strategy) and provide improved safety margins (by lowering the maximum channel and bundle powers).The use of neutron-absorbers in fuel design and manufacturing has been a regular practice in Light Water Reactor fuels for more than three decades. In CANDU applications, neutron absorbers have also been considered for the conceptual Advanced CANDU Reactor and the Low Void Reactivity fuel designs, for which the fissile content is made of low enriched uranium (LEU) or MOX fuels. The application to CANDU natural uranium (NU) fuel, however, especially as burnable poisons, is a relative novel approach. The reason for this is that the neutron economy in natural uranium-fuelled CANDU reactors is a prime concern, thus the addition of extra neutron absorbers is generally shunned. In our proposed application of burnable poisons to existing CANDU NU fuel design, because of low excess reactivity for NU fuel, the amount of neutron-absorber is expected to be restricted to small quantities and in a manner whereby the poison effect is restricted to the initial period of excess reactivity of a newly inserted fuel bundle. This implies that the impact on neutron economy would be relatively minimal, but the fuel performance would be significantly improved.Small amounts and appropriate mixtures of neutron absorbers were selected (approximately 500 mg of absorbers in a CANDU fuel bundle having a nominal weight of 24 kg). Preliminary results indicate that the fuelling transient and the subsequent reactivity peak can be lowered to improve the reactor’s operating margins. A parametric study using the Los Alamos National Laboratories’ MCNP 5 and Atomic Energy of Canada Limited’s WIMS-AECL 3.1 codes is presented in this paper. Details of this project and future work are also to be discussed.Copyright

Neutronic Design and Analysis of a Small Nuclear Reactor to Supply District Heating and Electrical Energy to Canadian Forces Bases Located in the Arctic or the Northern Remote Communities

Abstract The present work describes the preliminary design of a 25-MW(thermal) nuclear reactor capable of providing safe and reliable heating and electricity to any Canadian Forces Bases, especially in the Arctic, as well as in comparable civilian applications. The aim of the project is to provide a nuclear reactor system with sufficient inherent safety characteristics as it is intended to run in automatic mode and be monitored by operators with limited experience and training. For the neutronics calculations, the design work of the reactor’s core is carried out using the probabilistic simulation code MCNP 5 along with the Winfrith Improved Multigroup Scheme–Atomic Energy of Canada Limited (WIMS-AECL) deterministic code, Version 3.1, thus permitting a code-to-code comparison of the numerical results. Several design constraints related to coolant temperature and pressure, reactivity control, fuel enrichment, and time between refueling have been considered. The final reactor concept, named the Super Near Boiling 25 reactor (SNB25), provides heat energy dedicated to building and domestic water heating and supplies electricity through an organic Rankine cycle energy conversion plant. SNB25 employs TRISO fuel particles, contained in zirconium-sheathed fuel rods, and is light water cooled and moderated. Complete reactivity control is achieved through simple and reliable mechanical means consisting of 133 control rods and six adjustable radial reflector plates. The optimized reactor core configuration, along with its intrinsic control system, allows for the power plant to operate safely for more than a decade between refuelings from a typical central heating plant or the basement of a multilevel office building. The work also included a preliminary investigation of the nonnuclear part of the energy supply system including heat exchangers and the turbine-driven, electricity-generating system.