Canadian Supercritical Water-cooled Reactor Research Articles

Construction of the Dynamic Model and Control System for the Canadian Supercritical Water–Cooled Reactor Power Plant

Canada has proposed the supercritical water–cooled reactor (SCWR) concept as one of the Generation IV nuclear reactors. In the SCWR power plant, the supercritical water is heated in the reactor and then flows to the turbine directly. Therefore, knowledge of the dynamic behaviors of the system is necessary for the stable operation of the power plant. There is still a lack of study on the control system for the proposed SCWR power plant. In this study, a dynamic model for the entire SCWR power plant is constructed that includes the reactor, turbine, condenser, and feedwater pump. Based on the model, the open-loop characteristics of the system when subjected to perturbations in the inputs are analyzed. Subsequently, a feedback control strategy is adopted to regulate the outputs of the system when there are disturbances. The evaluation of the performance of the control system shows that the proposed control system can return the plant back to the operating conditions effectively.

Construction of a feedback control system based on CFD simulations for the 64-element Canadian SCWR

The knowledge of the dynamic behaviors of the supercritical water in the fuel bundle of the nuclear reactor is necessary for the design of the control system. For this purpose, the transient full-scale computational fluid dynamics (CFD) simulations are used to obtain the dynamic input and output data sets of the 64-element Canadian supercritical water-cooled reactor (SCWR) in this study. Subsequently, the linear dynamic models of the reactor are obtained by the system identification technique and validated by the comparison with the results from CFD simulations. Based on the linear dynamic models, a feedback control system consists of three PI/PID controllers is constructed. The objective is to regulate the system back to the design point when the reactor is subjected to disturbances. Therefore, the performance of the designed feedback control system is also evaluated in this work. It shows that the designed control system can timely minimize the deviations and return the reactor back to the desired operating condition.

Numerical simulation of the fluid flow and heat transfer of the supercritical water in the 64-element Canadian SCWR fuel bundle

ABSTRACT Canada has participated in the Generation IV nuclear energy systems focusing on the Supercritical Water-cooled Reactor (SCWR) system. The 64-element fuel rods channel is used in the Canadian SCWR. There are few publicly available experimental studies for the flow in the fuel channels with multiple fuel rods. To date, the CFD simulation reported in open literature for rod bundle flow in the 64-element SCWR is relatively scarce. The cladding surface temperature is of key importance in assessing the safety of the reactor. Since the inhomogeneities in the bundle cross-section can present complex flow phenomena, a CFD study can provide substantial insight into the flow physics. In this work, the full-scale CFD simulation of the supercritical water flow in the 64-element rod bundle was performed. The results suggest the possibility of the presence of gap vortices in the flow subchannels. Higher streamwise velocities and normal Reynolds stresses always exist at the center subchannel regions. The circumferential cladding surface temperature distribution is extremely non-uniform and there is a large difference between the maximum cladding surface temperatures for different fuel rods.

Joining of Zirconium Alloys to Nickel Bearing Alloys for In-Core Components

Abstract Zirconium alloys are well-positioned as the material of choice for nuclear in-core structures since they have a low neutron absorption cross section and possess good material properties under elevated neutron flux and high-temperature operating conditions. The Canadian supercritical water-cooled reactor (SCWR) concept pressure tube includes a transition from zirconium-tin alloy (Excel) to a yet to be selected stainless steel to allow it to be seal welded to the inlet plenum. Thus, a method to join zirconium and its alloys to ferrous and nickel-bearing alloys is desired. To this end, a series of experiments were carried out to demonstrate the feasibility of this method. A joint approaching the strength of the parent material was achieved using both rotary friction welding and co-extrusion methods, although the joints exhibited lower ductility than that of the parent material. Microstructural examination of the joint revealed a ∼200 nm to ∼1 μm intermetallic layer at the transition zone, which was the location of failure in the tensile specimen.

Molecular Dynamics of Supercritical Water for Nuclear Data Development

Abstract The Canadian supercritical water-cooled reactor was selected as one of the Generation IV International Forum initiatives for reactor design. It uses supercritical light water as a coolant under operating conditions of 25 MPa (250 bar) and 623–898 K. However, the simulation codes used to assess the performance and safety of such a design depend upon the accuracy of available nuclear data parametrizations, which currently do not include models of light water in the supercritical regime. In this paper, we present a study of supercritical water (SCW) through molecular dynamics simulations. Flexible variants of the TIP4P/2005 and simple point charge models for H2O are assessed to determine their ability to reproduce experimental measurements of SCW properties, and their suitability for the future development of nuclear data parametrizations for thermal neutron scattering from SCW. Planned experiments measuring thermal neutron scattering from SCW to inform nuclear data development are also summarized.

Evaluation of Friction-Factor Correlations at Supercritical Water Conditions in Support of the Canadian SCWR

Abstract Canada is participating in the Generation IV International Forum with the main focus on the pressure-tube-type supercritical water-cooled reactor (SCWR) concept. The Canadian SCWR concept is a heavy-water moderated and light-water-cooled reactor. The R&D framework for the development of the Canadian SCWR fuel-assembly concept includes experiments and analyses, including subchannel code development and applications. This paper focuses on the modeling of the hydraulic resistance under supercritical conditions, with or without the wire-wrap spacers. More specifically, it presents an assessment of three friction factor correlations developed for supercritical conditions. A literature survey of wire-wrap hydraulic models is presented. The assessment of the supercritical friction factor correlations and wire-wrap hydraulic models is carried out using three pressure drop experimental datasets, using the subchannel code ASSERT-PV V3.2m2.

General Corrosion of Chromium-Coated Zirconium- and Titanium-Based Alloys in Supercritical Water at 500 °C

Abstract The 300 MWel small Canadian supercritical water-cooled reactor (SCWR), which is a scaled-down version of the original 1200 MWel concept, has a smaller core, uses low enriched uranium fuel instead of a plutonium–thorium fuel, and features a lower (maximum) cladding temperature of 500 °C. The lower cladding temperature may permit the use of different alloys, including zirconium alloys, which had been ruled out as candidates for the Canadian SCWR, whose cladding temperature may reach 850 °C. The potential to use zirconium alloys is exciting because they have a low neutron cross section, which in turn means that fewer neutrons are lost, and the fuel can be used more efficiently. One advantage, for example,, is that the fuel cycle can be lengthened. In this paper, we report on the results of corrosion experiments used to screen zirconium- and titanium-based alloys as well as corrosion-resistant coating materials such as Cr and Al as potential candidates for fuel cladding in the small Canadian SCWR. These experiments were conducted in a refreshed autoclave in deaerated supercritical water at 500 °C and 23.5 MPa. After exposure, the weight gain was measured, and the oxide thickness and the oxide phases were examined. Of all materials, the coated and uncoated Ti-grade 2 and Ti-grade 5 alloys met our screening qualification criteria, however, Al/Cr-coated zirconium coupons showed notable improvement and will be explored further in future testing.

Analysis of Heat Transfer Characteristics of Canadian SCWR Fuel Assembly Concept

Abstract Canada is participating in the Generation IV (Gen IV) International Forum with a main focus on the pressure-tube-type supercritical water-cooled reactor (SCWR) concept. The subchannel code ASSERT-PV modified for SCWR applications was used to design the SCWR fuel assembly, specifically the fuel bundle. Several assumptions were required to model the fuel assembly, including the perfect insulation of (i) the central flow tube (i.e., no heat transfer through the central tube) and (ii) the pressure tube (i.e., no heat loss to the moderator). These two assumptions were considered as conservative, but they were not analyzed or assessed for their validity or accuracy. ASSERT-PV was upgraded to model the heat loss to the moderator, and an external CATHENA system code model was coupled to ASSERT-PV to model the heat transfer to the central flow tube. This paper describes these additional heat transfer components, and presents an assessment of these two assumptions for their impact on the prediction of maximum fuel cladding temperature.

Analysis of Shutdown System Effectiveness in the Canadian Super Critical Water Reactor Using Coupled Thermal Hydraulics and Three-Dimensional Neutron Kinetics

Abstract The Canadian supercritical water-cooled reactor concept features a re-entrant fuel channel wherein coolant first travels down a center flow tube and then up around the fuel elements. Previous work demonstrated that in cases of sudden coolant flow reduction or reversal (such as that which would result from a large pipe break near the core inlet), the coolant density reduction around the fuel has a positive reactivity effect that results in a power excursion. Such a transient is inherently self-terminating since the inevitable density reduction in the center flow tube has a very large negative reactivity effect. Nevertheless, a brief power pulse would ensue. In this work, the possibility of mitigating the power pulse with a fast-acting shutdown system was explored. The shutdown system model, consisting of bottom-inserted neutron absorbing blades and realistic estimates of insertion rates and trip conditions, was added to a full-core coupled spatial neutron kinetics and thermal-hydraulics model. It was demonstrated that such a system can effectively mitigate both the peak magnitude of the power excursion and its duration.

A Thoria and Thorium Uranium Dioxide Nuclear Fuel Performance Model Prototype and Knowledge Gap Assessment

Thorium-based fuel cycles can improve fuel sustainability within the nuclear power industry. The Canadian supercritical water-cooled reactor (SCWR) concept uses this path to achieve the sustainability requirement of the Gen-IV Forum. The study of thorium dioxide/thoria ThO2-based fuel irradiation behavior is significantly less advanced than that of uranium dioxide (UO2) fuel, although ThO2 possesses superior thermal conductivity, thermal expansion, higher melting temperature, and oxidation resistance that may improve both fuel performance and safety. The fuel and sheath modeling tool (FAST), a fuel performance model for UO2 fuel, was developed at the Royal Military College of Canada (RMCC). FAST capability has been extended to include thoria (ThO2), thorium uranium dioxide (Th,U)O2, and thorium plutonium dioxide (Th,Pu)O2 as fuel pellet materials, to aid in designing and performance assessment of Th-based fuels, including SCWR (Th,Pu)O2 fuel. The development and integration of ThO2 and (Th,U)O2 models into the existing FAST model led to the multipellet material FAST (MPM-FAST). Model development was performed in collaboration between RMCC and Canadian Nuclear Laboratories (CNL). This paper presents an outline of the ThO2 and (Th,U)O2 MPM-FAST model, a comparison between modeling results with postirradiation examination (PIE) data from a test conducted at CNL, and an account of the knowledge gap between our ability to model ThO2 and (Th,U)O2 fuel compared to UO2. Results are encouraging when compared to PIE data.

Irradiation Issues and Material Selection for Canadian SCWR Components

The Canadian super critical water-cooled reactor (SCWR) concept requires materials to operate at higher temperatures than current generation III water-cooled reactors. Materials performance after radiation damage is an important design consideration. Materials that are both corrosion resistant and radiation damage tolerant are required. This paper summarizes the operating conditions including temperature, neutron flux, and residence time of in-core Canadian SCWR components. The focus is on the effects of irradiation on in-core components, including those exposed to a high neutron flux in the fuel assembly, the high pressure boundary between coolant and moderator, as well as the low-temperature, low-flux calandria vessel that contains the moderator. Although the extreme conditions and the broad range of SCWR in-core operating conditions present significant materials selection challenges, candidate alloys that can meet the performance requirements under most in-core conditions have been identified. However, for all candidate materials, insufficient data are available to unequivocally ensure acceptable performance and experimental irradiations of candidate core materials will be required. Research programs are to include out-of-pile tests on un-irradiated and irradiated alloys. Ideally, in-flux studies at appropriate temperatures, neutron spectrum, dose rate, duration, and coolant chemistry will be required. Characterization of the microstructure and the mechanical behavior including strength, ductility, swelling, fracture toughness, cracking, and creep on each of the in-core candidate materials will ensure their viability in the Canadian SCWR.

Optimization of the Canadian SCWR Core Using Coupled Three-Dimensional Reactor Physics and Thermal-Hydraulics Calculations

The Canadian supercritical water-cooled reactor (SCWR) design is part of Canada's Generation IV reactor development program. The reactor uses batch fueling, light water above the thermodynamic critical point as a coolant and a heavy water moderator. The design has evolved considerably and is currently at the conceptual design level. As a result of batch fueling, a certain amount of excess reactivity is loaded at the beginning of each fueling cycle. This excess reactivity must be controlled using a combination of burnable neutron poisons in the fuel, moderator poisons, and control blades interspersed in the heavy water moderator. Recent studies have shown that the combination of power density, high coolant temperatures, and reactivity management can lead to high maximum cladding surface temperatures (MCST) and maximum fuel centerline temperatures (MFCLT) in this design. This study focuses on improving both the MCST and the MFCLT through modifications of the conceptual design including changes from a 3 to 4 batch fueling cycle, a slightly shortened fuel cycle (although exit burnup remains the same), axial graded fuel enrichment, fuel-integrated burnable neutron absorbers, lower reactivity control blades, and lower reactor thermal powers as compared to the original conceptual design. The optimal blade positions throughout the fuel cycle were determined so as to minimize the MCST and MFCLT using a genetic algorithm and the reactor physics code PARCS. The final design was analyzed using a fully coupled PARCS-RELAP5/SCDAPSIM/MOD4.0 model to accurately predict the MCST as a function of time during a fueling cycle.

Determination of the optimal few-energy group structure for the Canadian Super Critical Water-cooled Reactor

Most deterministic neutron transport codes implement the multi-group approximation theory to solve the linear Boltzmann equation by which the neutron flux and the interaction cross-sections are averaged over discretized energy groups. In order to further reduce the computational cost for large scale and/or time dependent problems, this multi-group information undergoes additional energy group condensation into the so-called few energy group structure which may be used in a broad range of diffusion based full-core analyses. The accuracy and efficiency of few-group based computations is dependent on the number and the boundaries of the discrete energy group structure. Since the flux spectrum used in the homogenization process may not be known a priori and indeed may evolve in space, with burnup, and during transients, the optimal energy group structure depends on reactor type, design, operating conditions, fuel type, and composition. The Canadian Pressure Tube Super-Critical Water-cooled Reactor (PT-SCWR) is a Generation IV advanced reactor system that uses light water above its thermodynamic critical point as coolant and a plutonium-driven thorium fuel mixture. Considering that the anticipated flux spectrum for such a design deviates significantly from the thermal-neutron dominated CANDU designs, there may be a need for improvements in the number and boundaries of the few-group nuclear data. This paper presents a systematic methodology for the delineation of few-group energy structure for the Canadian PT-SCWR. The methodology used the SCALE (Standardized Computer Analysis for Licensing Evaluation) code package to examine the effect of energy group homogenization and determine the structure which minimizes the sensitivity of the results to energy group partitioning. As such, it utilizes normal and off-normal operating conditions to determine the effect of energy homogenization and determines the minimal group of energy boundaries which can accurately capture the lattice physics phenomena within the lattice cell over a wide range of operating and accident conditions.

Analysis of Control Rod Drop Accidents for the Canadian SCWR Using Coupled 3-Dimensional Neutron Kinetics and Thermal Hydraulics

The Canadian Supercritical Water-cooled Reactor (SCWR), a GEN IV reactor design, is a hybrid design of the well-established CANDU™ and Boiling Water Reactor with water above its thermodynamic critical point. Given the batch fueled design, control rods are used to manage the reactivity throughout the fuel cycle. This paper examines the consequences of a control rod drop accident (CRDA) for the Canadian SCWR. The asymmetry generated by the dropped rod requires an accurate 3-dimensional neutron kinetics calculation coupled to a detailed thermal-hydraulic model. Before simulating the CRDAs, the proper implementation of the 3D reactivity feedback was verified and various sensitivity studies were performed. This work demonstrates that the proposed safety systems for the SCWR core are capable of terminating the CRDA sequence prior to exceeding maximum sheath and centerline temperatures. In one instance involving a rod on the periphery of the core, the proposed trip setpoint (115% FP) was not exceeded and a new steady state was reached. Therefore it is recommended that the design also include provisions for a high-log rate and/or local Neutron Overpower Protection (NOP) trips, similar to existing CANDU designs such that reactor shutdown can be assured for such spatial anomalies.

Computational Fluid Dynamic Simulations of Heat Transfer From a 2 × 2 Wire-Wrapped Fuel Rod Bundle to Supercritical Pressure Water

Within the Generation-IV International Forum, Canadian Nuclear Laboratories (CNL) led the conceptual fuel bundle design effort for the Canadian supercritical water cooled reactor (SCWR). The proposed fuel rod assembly for the Canadian SCWR design comprised of 64-elements with spacing between elements maintained using the wire-wrap spacers. Experimental data and correlations are not available for the fuel-assembly concept of the Canadian SCWR. To analyze the thermalhydraulic performance of the new bundle design, CNL is using computational fluid dynamics (CFD) as well as the subchannel approach. Simulations of wire-wrapped bundles can benefit from the increased fidelity and resolution of a CFD approach due to its ability to resolve the boundary layer phenomena. Prior to the application, the CFD tool has been assessed against experimental heat transfer data obtained with bundle subassemblies to identify the appropriate turbulence model to use in the analyses. In the present paper, assessment of CFD predictions was made with the wire-wrapped bundle experiments performed at Xi'an Jiaotong University (XJTU) in China. A three-dimensional CFD study of the fluid flow and heat transfer at supercritical pressures for the rod-bundle geometries was performed with the key parameter being the fuel rod wall temperature. This investigation used Reynolds-averaged Navier–Stokes turbulence models with wall functions to investigate the behavior of flow through the wire-wrapped fuel rod bundles with water subjected to a supercritical pressure of 25 MPa. Along with the selection of turbulence models, CFD results were found to be dependent on the value of turbulent Prandtl number used in simulating the experimental test conditions for the wire-wrapped fuel rod configuration. It was found that the CFD simulation tends to overpredict the fuel wall temperature, and the predicted location of peak temperature differs from the measurement by up to 65 deg.

Fuel Assembly Concept of the Canadian Supercritical Water-Cooled Reactor

The Canadian supercritical water-cooled nuclear reactor (SCWR) is a 2540 MWth channel-type SCWR concept that employs 336 fuel channels in the reactor core. Each fuel channel includes a pressure tube that is submerged in a heavy water moderator and contains a removable fuel assembly. The fuel assembly is designed so that all in-core components exposed to high radiation fields (other than the pressure tube) are part of the fuel assembly, which is removed from the reactor core as part of the assembly after three operating cycles. This design feature significantly reduces the likelihood of component failures due to radiation damage. To achieve high (>45%) power conversion efficiency, the Canadian SCWR operates at a supercritical water pressure (25 MPa) and high temperatures (350 °C at the inlet, 625 °C at the outlet). These conditions lead to fuel cladding temperatures close to 800 °C. Because of the reduced material strength at this temperature and higher fission gas production of the fuel, collapsible fuel cladding is selected over internally pressurized cladding. To increase heat transfer and to reduce cladding temperatures, turbulence-inducing wire-wraps are employed on fuel elements. Numerical models have been developed to analyze the thermal-structural behavior of Canadian SCWR fuel at normal and accident conditions. It was found that axial ridging, a possible failure mechanism with collapsed fuel cladding, can be avoided if the cladding thickness is larger than 0.4 mm. Detailed numerical analysis showed that the maximum fuel cladding temperature for the worst-case accident scenario is below the melting point by a small margin. This result was obtained with conservative assumptions, suggesting that the actual margin is greater. Hence, one of the design goals, the exclusion of the possibility of melting of the fuel, which is called the “no-core-melt” concept, seems attainable. However, this needs to be demonstrated more rigorously by removing the conservative assumptions in the analysis and performing supporting experimental work. This paper presents a description of the Canadian SCWR fuel assembly concept, its unique features, the rationale used in the concept development and the results of various numerical analyses demonstrating the performance and characteristics of the Canadian SCWR fuel channel.

Thermal-Hydraulics Program in Support of Canadian SCWR Concept Development

The thermal-hydraulics program in support of the development of the Canadian supercritical water-cooled reactor (SCWR) concept has undergone several phases. It focused on key parameters such as heat transfer, critical flow, and stability of fluids at supercritical pressures. Heat-transfer experiments were performed with tubes, annuli, and bundles in water, carbon dioxide (CO2), or refrigerant flows. Data from these experiments have led to enhancement in understanding of the phenomena, improved prediction methods, and verified analytical tools. In addition, these experiments facilitated the investigation of separate effects on heat transfer (such as geometry, diameter, spacing device, and transient). Chocking flow characteristics were studied experimentally with sharp-edged nozzles of two different sizes of opening. Experimental data have been applied in improving the critical-flow correlation in support of accident analyses. A one-dimensional (1D) analytical model for instability phenomena has been developed and assessed against the latest experimental data for quantifying the prediction capability and applicability.

Microstructure Study of NiCrAlY and FeCrAlY Exposed to Superheated Steam at 800 °C

Supercritical water-cooled reactor (SCWR) was chosen as Generation IV reactor concept in Canada to utilize Canada's expertise and technical capabilities from past research and designs. The conceptual design of Canadian SCWR has a core outlet temperature of 650 °C at 25 MPa and a peak cladding temperature as high as 800 °C. Corrosion/oxidation resistance is an important factor in material selections and also coating considerations. Most of the reported supercritical water (SCW) test data have been obtained at temperatures up to 700 °C as no autoclave exists that can operate above 700 °C at supercritical pressures and temperatures. Superheated steam (SHS) is used as a surrogate fluid to SCW in this study to evaluate two coating materials, FeCrAlY and NiCrAl, at 800 °C. The results showed that the FeCrAlY became discolored rapidly while NiCrAl still maintained some metallic sheen after 600 h. The weight change results suggest that more oxide formation took place on FeCrAlY than NiCrAl. In particular, grain boundary oxide (Al2O3) formed on FeCrAlY surface upon exposure to steam after 300 h. Further exposure caused more intragranular Al2O3 to form, in addition to magnetite formation on the grain boundary regions. For NiCrAl samples, NiO formed after steam exposure for 300 h. Spinel and (Cr,Al)2O3 were also found after 300 h along with very limited amount of Al2O3. After 600 h, Al2O3 became well developed on NiCrAl and the coverage of spinel and Cr2O3 on the surface reduced.

DEVELOPMENT OF A HYDROGEN MANAGEMENT CONCEPT FOR THE CANADIAN SUPERCRITICAL WATER-COOLED REACTOR

Accidental hydrogen production in nuclear reactors has been a significant focus of nuclear reactor safety for decades. However, since the accident at Fukushima Daiichi nuclear generating station, hydrogen safety in nuclear reactors is a more relevant topic. As new reactor concepts, such as the supercritical water-cooled reactor (SCWR), are designed and developed the risk of unintentional hydrogen generation is not eliminated; however, it can be mitigated in the design.A systematic assessment of the hydrogen risk from both normal and accident conditions in the Canadian SCWR design was performed, in which various techniques to mitigate the hydrogen combustion potential were considered. While the rate of hydrogen generation under normal operating conditions was found to be low when held at supercritical water conditions, conservative estimates suggest that a significant quantity of hydrogen may be produced and released to the containment building in a severe accident. As a result, a hydrogen–oxygen management concept has been proposed to mitigate the hydrogen produced in a severe accident that includes a nitrogen-inerted containment building to reduce the combustion potential of hydrogen and the installation of passive autocatalytic recombiners for oxygen management. This hydrogen–oxygen management concept results in significant design changes and likely significant economic and operational impacts on the Canadian SCWR design.

Robust Control of Once-Through Canadian Supercritical Water-Cooled Reactors

The Canadian SuperCritical Water-cooled Reactor (SCWR) is a once-through pressure tube–type SCWR under development in Canada. It is a multivariable system with strong cross coupling and a high degree of nonlinearity. The outputs are sensitive to disturbances, and the variations in the thermal parameters should be limited to avoid thermal stress to its components. Therefore, designing an adequate control system is challenging. In this paper, robust multivariable feedback control and feedforward control are proposed to design the control system of the Canadian SCWR. Three uncertainty sources are considered: unmodeled uncertainty, linearization uncertainty, and model reduction uncertainty. These uncertainties are evaluated taking into account all aspects affecting the linear dynamic model used in the robust controller synthesis, and the uncertainty bounds are determined to cover the uncertainties. The robust feedback controller is synthesized using the μ-synthesis approach. The feedforward control is added to the robust feedback control to further improve the control performance. It is obtained through disturbance compensation features for a feedforward controller. The control performance of the hybrid control system is evaluated based on the nonlinear simulation by introducing different setpoint changes. The designed control system can stabilize the Canadian SCWR, and the control performance is satisfactory.