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.
Natural thorium contains impurities of numerous isotopes. A study was performed to examine the influence of naturally occurring impurities in thorium-based fuels on a few parameters of interest such as: exit burnup, coolant void reactivity (CVR), fuel temperature coefficients (FTC), and the radiotoxicity of the used fuel. Two different fuel bundle designs were modeled: a 43-element bundle containing 2.25% U-233, and a 35-element bundle containing 1.45% U-233. Naturally occurring thorium fuel impurities were applied to both fuel bundle models at various concentrations, from 0% to 100% of the expected maximum. For burnup-averaged k-infinity (k∞) values of 1.050 and 1.030, exit burnup, burnup-weighted CVR, and burnup-weighted FTC were calculated using the neutron transport code WIMS-AECL, and plotted against fraction of full impurity concentration to determine how the impurity levels affect these reactor physics parameters of interest. For the most-realistic (for CANDU) burnup-averaged k∞ of 1.050, both the inhalation radiotoxicity and the production of U-232 were calculated using the fuel depletion code WOBI. Up to the maximum impurity concentrations considered, no effects on the investigated fuel performance parameters were found to be greater than a few percent.
Over the course of the last several years the Canadian Nuclear Safety Commission (CNSC) has engaged with numerous vendors and potential licenses of small modular reactors (SMR) technology. This paper describes why Canada, and the CNSC, is of such interest to the international SMR community for prelicensing engagement and potential licensing of SMRs. It discusses what an SMR is and what potentially differentiates them from standard nuclear power plants (NPP). Readiness activities for the potential licensing of SMRs are described as well as modifications being made to the CNSC’s existing regulatory framework to facilitate the same, without reducing safety. The role of the CNSC’s discussion paper (DIS-16-04, Small Modular Reactors: Regulatory Strategy, Approaches and Challenges) and how feedback received on it helped confirm the CNSC’s modifications to be undertaken to the regulatory framework, as well as areas requiring further clarity, are highlighted. Finally, The CNSC Vendor Design Review (VDR) process is described as well as its part in ensuring a state of readiness to evaluate a licence application.
The Burnup of Fuel Elements (BURFEL) code system has been used to calculate powers and burnups for experimental fuel irradiated in the National Research Universal Reactor (NRU) loops. BURFEL-calculated burnups, based on the calorimetrically measured loop powers, have been observed to exhibit biases with respect to their chemically measured counterparts. A high-fidelity Monte Carlo N-Particle method involving the NRU full-core model has been used for benchmarking BURFEL calculations, resulting in similar biases attributed to uncertainties in the loop powers. This study provides quantitative insights into the observed BURFEL biases for the purpose of possibly correcting existing loop fuel irradiation data for such biases.
The ready availability of codes such as LAMMPS (Large-scale Atomic/Molecular Massively Parallel Simulator) for molecular dynamics simulations has opened up the realm of atomistic modelling to novice code users with an interest in computational materials modelling but who lack the appropriate theoretical or computational background. As such, there is significant risk of the “user effect” having a negative impact on the quality of results obtained using such codes. Here, we present a “how-to” procedure for equilibrium molecular dynamics-based nuclear fuel thermal conductivity calculations using the Green–Kubo method with an interatomic potential developed by Cooper et al. [ 1 ]. The various steps of the simulation are identified and explained, along with criteria to assess the quality of the intermediate and final results, discussion of some problems that can arise during a simulation, and some inherent limitations of the method. Calculated thermal conductivities for UO2 and ThO2 will be compared with the available experimental data and also with similar thermal conductivity calculations using nonequilibrium molecular dynamics, reported in the open literature.
A computational benchmark, using the deterministic codes WIMS-AECL and WOBI, and the stochastic code SERPENT, is made for burnup calculations of advanced thorium fuels in heavy water moderated reactors. Exit burnups and the concentration of the longer-lived actinides from the deterministic code set of WIMS-AECL and WOBI, which are 2-D, were compared to those from a full 3-D calculation in SERPENT. Results for reactivity vs. time are in general agreement to within a few mk (<1% in overall neutron multiplication) and appear to be systematic. Results for exit burnup were larger, in the 3%–6% range, because small reactivity effects can be amplified here.
Small nuclear reactors can offer safe, reliable, and long-lasting district heating and electrical power generation to remote locations in northern Canada. A conceptual design of an organic-cooled and moderated reactor based upon the SLOWPOKE-2 research reactor is proposed for potential employment in northern Canada. For viability, this design extends the SLOWPOKE-2's power to 1 MWth. An added pump circulates the organic coolant, a partially hydrogenated terphenyl mixture known as HB-40, to facilitate greater heat transfer. The reactor incorporates the same low-enriched uranium dioxide fuel as the SLOWPOKE-2. Reactor control is accomplished through hafnium absorber rods and a movable beryllium reflector. The reactor neutronics are simulated using the deterministic code, WIMS-AECL, and the probabilistic code, MCNP 6. The service life of fuel in this reactor operating at full power exceeds 11 years. The conceptual design has demonstrated negative reactivity coefficients indicating strong potential for inherent safety.
There are many issues and challenges in assessing ecological and human health risk from siting small modular reactors (SMRs) in northern or Arctic regions. Environmental guidance for Canadian nuclear facilities is largely derived from data and models relevant to temperate regions, with no explicit guidance or parameters for northern regions. International Atomic Energy Agency guidance provides some data and parameters for northern regions, but there remains a paucity of data and models. Although wildlife often comprise a major part of northern and Arctic diets, there are few data or transfer parameters for these ecosystems. Data and transport models are available for weapon test and Sellafield/La Hague fission products in northern oceans, but very little is known about circulation or fate and transport in estuarine and coastal areas typical of northern Canada. Baseline data, parameters, and models are needed for key processes and pathways to accurately assess ecological and human health risk.
The National Research Universal (NRU) reactor is a major research facility that provides a beam of slow neutrons with a minimum of gamma rays and other types of radiation for experimental purposes. The thermal column consists of 5 graphite radial sections separated with an air gap for cooling. The graphite components require continuous monitoring to ascertain that temperatures are controlled within safe margins. Wall temperatures of the graphite sections are obtained via thermocouples affixed to the column walls. The safety margins for operation of the thermal column are driven by the temperatures of the closest radial section to the reactor core (HG1). Although, most of the thermocouples in HG1 are no longer functional, the thermocouples are functional in the adjacent graphite section (HG2). This study relied on the historical data of the graphite temperatures over a few years to develop an empirical correlation that relates temperatures in HG1 to those of HG2. The correlation sets limits on the functional thermocouples in HG2 to ensure HG1 remains within the prescribed limits (149–232 °C). Correlations were developed using statistical analysis of the historical data. A control band of approximately 40 °C for HG2 with confidence levels of 68% and 95%, respectively, were established.
Knowledge of the isotopic composition of a nuclear reactor core is important for accurate core-follow and reload analysis. In the McMaster Nuclear Reactor, fuel depletion estimates are based upon a semi-empirical calculation using flux-wire measurements. These estimates are used to plan and guide fuelling operations. To further support operations, an OSCAR-4 model is being developed. To evaluate the performance of the OSCAR-4 code for this application, 2 points of comparison, considering the period between 2007 and 2010, are presented: (i) the multiplication factor keff and (ii) U-235 fuel inventory. The latter is compared with a simple first-order semi-empirical calculation. The calculation of keff for the last operational 3 months yields 0.997 ± 0.002 (vs. 1.000 for an operating reactor), and differences in both core-average inventory and the maximum standard fuel assembly inventories estimates are found to be 5.7% and 7.5%, respectively.
Small modular reactors (SMRs) are considered as potential energy sources in remote areas. SMR technologies that are under prelicensing vendor’s design review by Canadian Nuclear Safety Commission i...
A fuel survey was carried out at all operating CANDU reactors in Canada in 5 sites (Bruce A, Bruce B, Darlington, Pickering, and Point Lepreau). The survey used the National Fire Protection Association 557 combination method for the fire zones that contain fire safe shutdown analysis equipment. A fire zone group list for the sites was developed to combine fire zones with similar functions; 38 fire zone groups were produced from this exercise. The results of the survey show that the average fuel load density for all 1230 fire zones is 170.1 MJ/m(2), and the average fuel load is 79 183 MJ. The maximum fuel load density is 1319 MJ/m(2), and the maximum fuel load is 2 785 404 MJ. High-energy arcing faults risk was found in 254 fire zones out of the 1230 fire zones. Electric fault is the highest ignition source risk present in all 1230 fire zones.
The Nuclear Energy Agency (NEA) is a specialized agency within the Organization for Economic Co-operation and Development (OECD). The International Fire Data Exchange Project (OECD FIRE) was designed by the NEA to encourage multilateral co-operation in the collection and analysis of data relating to fire events in nuclear power plants. We used Python advanced software to analyze the data related to CANDU reactor plants in Canada from the OECD FIRE Database, while providing weighting factors/percentage tables to be used in CANDU Fire probabilistic risk assessment analysis. We also used 5 different time-series methods to predict future potential fires in CANDU reactors, compared the results from different methods, and identified the best method to predict future fires in CANDU power plants.
Fire Probabilistic Risk Assessment (PRA) is being introduced to the fire protection engineering practice both locally and worldwide. The commercial nuclear power industry has also experiencing the impact of this new approach. This paper examines the work performed to assess the relative accuracy of fire models for CANDU nuclear power plant (NPP) applications. The Canadian NPP uses some portions of NUREG/CR-6850 in performing fire PRA. Canadian fire ignition frequencies have been provided by International Fire Data Exchange Project. The CANDU Fire PRA Model can quantitatively evaluate plant damage states and core damage frequencies. This model will assist fire engineers in performing CANDU Fire PRA analysis, by recognizing vulnerabilities related to fire events and will contribute to further improvement of the Canadian NPPs’ safety.
Canada has attracted specific interest from developers of nonwater-cooled small modular reactor (SMR) technologies, including concepts based on high-temperature gas-cooled reactors (HTGRs). It is anticipated that some research and development (R&D) will be necessary to support safety analysis and licensing of these reactors in Canada. The Phenomena Identification and Ranking Table (PIRT) process is a formalized method in which a panel of experts identifies which physical phenomena are most relevant to the reactor safety analysis and how well understood these phenomena are. The PIRT process is thus a tool to assess current knowledge levels and (or) predictive capabilities of models, thus providing direction to a focused R&D program. This paper summarizes the results of a PIRT process performed by a panel of experts at Canadian Nuclear Laboratories for a limiting or "worst-case" accident scenario at a generic HTGR-type SMR. Suggestions are given regarding the highest priority R&D items to support severe accidents analysis of these reactors.
A machine-learning algorithm has been implemented by use of a neural network as a preliminary study on the applicability of this method to special nuclear materials detection. The algorithm predicts the presence of the 238U isotope when learning from a gamma spectrum data measured with a high-purity germanium detector from a sample of depleted uranium. In this work, both a fully connected neural network and a convolutional neural network have been implemented, and the performance of different configurations of the network has been studied. The use of convolutional network showed better performance over the fully connected network, with cost function and success rate values supporting a better prediction while avoiding overfitting. Furthermore, implemented network features such as filtering, max-pooling, dropout regularization, and momentum optimization also showed improved prediction performance.
