Canadian Supercritical Water-cooled Reactor Research Articles

Various Design Aspects of the Canadian Supercritical Water-Cooled Reactor Core

The Canadian Supercritical Water-Cooled Reactor (SCWR) is a 1200 MW(e) channel-type nuclear reactor. The reactor core includes 336 vertical pressurized fuel channels immersed in a low-pressure heavy water moderator and calandria vessel containment. The supercritical water (SCW) coolant flows into the fuel channels through a common inlet plenum and exits through a common outlet header. One of the main features of the Canadian SCWR concept is the high-pressure (25 MPa) and high-temperature (350°C at the inlet, 625°C at the outlet) operating conditions that result in an estimated thermal efficiency of 48%. This is significantly higher than the thermal efficiency of the present light water reactors, which is about 33%. This paper presents a description of the Canadian SCWR core design concept; various numerical analyses performed to understand the temperature, flow, and stress distributions of various core components; and how the analyses results provided input for improved concept development.

Assessment of Candidate Fuel Cladding Alloys for the Canadian Supercritical Water-Cooled Reactor Concept

Selecting and qualifying a fuel cladding material for the Canadian supercritical water-cooled reactor (SCWR) concept remains the most significant materials challenge to be overcome. The peak cladding temperature in the Canadian SCWR concept is predicted to be as high as 800°C. While advanced materials show promise for future deployment, currently, the best options available are austenitic stainless steels and nickel-based alloys. Many of these alloys were extensively studied for use as fuel cladding materials in the 1960s, as part of programs to develop nuclear superheated steam reactors. After extensive out-of-pile testing and consideration of the existing data, five alloys (347 SS, 310 SS, Alloy 800H, Alloy 625, and Alloy 214) were selected for more detailed assessment using a combination of literature surveys and targeted testing to fill in major knowledge gaps. Wherever possible, performance criteria were developed for key materials properties. This paper summarizes the methodology used for the assessment and presents the key results, which show that 310 SS, Alloy 800H, and Alloy 625 would all be expected to give acceptable performance in the Canadian SCWR concept.

Assessment of Computational Tools in Support of Heat-Transfer Correlation Development for Fuel Assembly of Canadian Supercritical Water-Cooled Reactor

Experimental data and correlations are not available for the fuel-assembly concept of the Canadian supercritical water-cooled reactor (SCWR). To facilitate the safety analyses, a strategy for developing a heat-transfer correlation has been established for the fuel-assembly concept at supercritical pressure conditions. It is based on an analytical approach using a computational fluid dynamics (CFD) tool and the ASSERT subchannel code to establish the heat transfer in supercritical pressure flow. Prior to the application, the CFD tool was assessed against experimental heat transfer data at the pseudocritical region obtained with bundle subassemblies to identify the appropriate turbulence model for use. Beyond the pseudocritical region, where the normal heat transfer behavior is anticipated, the ASSERT subchannel code also was assessed with appropriate closure relationships. Detailed information on the supporting experiments and the assessment results of the computational tools are presented.

Microstructure Instability of Candidate Fuel Cladding Alloys: Corrosion and Stress Corrosion Cracking Implications

This paper addresses some of the overarching aspects of microstructure instability expected from both high temperature and radiation exposure that could affect the corrosion and stress corrosion cracking (SCC) resistance of the candidate austenitic Fe-Cr-Ni alloys being considered for the fuel cladding of the Canadian supercritical water-cooled reactor (SCWR) concept. An overview of the microstructure instability expected by both exposures is presented prior to turning the focus onto the implications of such instability on the corrosion and SCC resistance. Results from testing conducted using pre-treated (thermally-aged) Type 310S stainless steel to shed some light on this important issue are included to help identify the outstanding corrosion resistance assessment needs.

Effect of thermal treatment on the corrosion resistance of Type 316L stainless steel exposed in supercritical water

There are still unknown aspects about the growth mechanism of oxide scales formed on candidate stainless steel fuel cladding materials during exposure in supercritical water (SCW) under the conditions relevant to the Canadian supercritical water-cooled reactor (SCWR). The tendency for intermetallic precipitates to form within the grains and on grain boundaries during prolonged exposure at high temperatures represents an unknown factor to corrosion resistance, since they tend to bind alloyed Cr. The objective of this study was to better understand the extent to which intermetallic precipitates affects the mode and extent of corrosion in SCW. Type 316L stainless steel, used as a model Fe–Cr–Ni–Mo alloy, was exposed to 25MPa SCW at 550°C for 500h in a static autoclave for this purpose. Mechanically-abraded samples were tested in the mill-annealed (MA) and a thermally-treated (TT) condition. The thermal treatment was conducted at 815°C for 1000h to precipitate the carbide (M23C6), chi (χ), laves (η) and sigma (σ) phases. It was found that although relatively large intermetallic precipitates formed at the scale/alloy interface locally affected the oxide scale formation, their discontinuous formation did not affect the short-term overall apparent corrosion resistance.

Control of Canadian once-through direct cycle supercritical water-cooled reactors

Canadian supercritical water-cooled reactor (SCWR) can be modelled as a Multiple-input Multiple-output (MIMO) system. It has a high power-to-flow ratio, strong cross-coupling and high degree of nonlinearity in its dynamic characteristics. Among the outputs, the steam temperature is strongly affected by the reactor power and the most challenging to control. It is difficult to adopt a traditional control system design methodology to obtain a control system with satisfactory performance. In this paper, feedforward control is applied to reduce the effect on steam temperature from the reactor power. Single-input Single-output (SISO) feedback controllers are synthesized in the frequency domain. Using the feedforward controller, the steam temperature variation due to disturbances at the reactor power has been significantly suppressed. The control system can effectively maintain the overall system stability and regulate the plant around a specified operating condition. To deal with the nonlinearities, gain scheduling control strategy is adopted. Different sets of controllers combined by smooth weight functions are used for the plant at different load conditions. The proposed control strategies have been evaluated under various operating scenarios. Simulation results show that satisfactory performance can successfully achieved by the designed control system.

Heat transfer effectiveness for cooling of Canadian SCWR fuel assembly under the LOCA/LOECC scenario

The effectiveness of heat transfer from the fuel rods to the moderator in a Canadian, pressure-tube type, Super-Critical Water-cooled Reactor (SCWR) has been assessed for the postulated loss-of-coolant accident (LOCA) coupled with the total loss of emergency core cooling (LOECC) event using the safety analysis code “SCTRAN”. A radiation heat-transfer model, including a two-dimensional heat conduction solution scheme, has been incorporated into the SCTRAN code for this assessment. The assessment result shows that the combined radiation heat transfer from the fuel rods to the fuel channel and natural convective heat transfer from the fuel rods to the steam are effective in removing decay heat for the LOCA/LOECC scenario. Maximum cladding temperatures of fuel rods in inner and outer rings of the fuel assembly in the highest-power channel are predicted at 1278°C and 1192°C respectively, which are lower than the melting point of the modified stainless-steel cladding. Sensitivity analyses of several analytical parameters on the predicted maximum cladding temperatures have been performed.

Nuclear data sensitivity and uncertainty for the Canadian supercritical water-cooled reactor II: Full core analysis

Uncertainties in nuclear data are a fundamental source of uncertainty in reactor physics calculations. To determine their contribution to uncertainties in calculated reactor physics parameters, a nuclear data sensitivity and uncertainty study is performed on the Canadian supercritical water reactor (SCWR) concept. The nuclear data uncertainty contributions to the neutron multiplication factor keff are 6.31mk for the SCWR at the beginning of cycle (BOC) and 6.99mk at the end of cycle (EOC). Both of these uncertainties have a statistical uncertainty of 0.02mk. The nuclear data uncertainty contributions to Coolant Void Reactivity (CVR) are 1.0mk and 0.9mk for BOC and EOC, respectively, both with statistical uncertainties of 0.1mk. The nuclear data uncertainty contributions to other reactivity parameters range from as low as 3% of to as high as ten times the values of the reactivity coefficients. The largest contributors to the uncertainties in the reactor physics parameters are Pu-239, Th-232, H-2, and isotopes of zirconium.

Stability of flashing-driven natural circulation in a passive moderator cooling system for Canadian SCWR

This paper presents an examination of the instability mechanisms in a Passive Moderator Cooling System for the Canadian SCWR (Supercritical Water-cooled Reactor). The passive system is being developed at AECL using a flashing-driven natural circulation loop. Unstable intermittent and sinusoidal oscillations were identified from experimental data of the flashing-driven natural circulation passive moderator cooling system. The oscillation periods were correlated with the boiling delay time. A stability map for a flashing-driven two-phase natural circulation loop was established on the dimensionless plane with Subcooling number and Phase Change number. It was observed that there is thermal non-equilibrium in the single-phase and two-phase oscillation stages of the flashing-driven natural circulation.

A PRELIMINARY CATHENA THERMALHYDRAULIC MODEL OF THE CANADIAN SCWR FOR SAFETY ANALYSIS

The supercritical water-cooled reactor (SCWR) is one of six reactor concepts under development in the Generation-IV International Forum (GIF). As a member of GIF, Canada is developing a pressure-tube type SCWR, which has the potential to fulfill all major GIF goals on enhanced safety, sustainability, economics, and proliferation resistance. The system thermalhydraulics code CATHENA will be used in the safety analyses for the Canadian SCWR. Based on the current conceptual design of the Canadian SCWR, a CATHENA idealization has been developed. This model includes all 336 fuel channels with a detailed model of heat transfer in the reactor core. Also modeled are the main pumps, inlet plenum, outlet plenum, turbines, and heavy water moderator. In this paper, the CATHENA idealization of the Canadian SCWR conceptual design is described. Simulation results for steady-state normal operations are also presented for the current Canadian SCWR conceptual design.

Review of R&D for supercritical water cooled reactors

The Supercritical Water-Cooled Reactor (SCWR) is a high temperature, high pressure water-cooled reactor that operates above the thermodynamic critical point (374 °C, 22.1 MPa) of water. In general terms, the conceptual designs of SCWRs can be grouped into two main categories: pressure vessel concepts proposed first by Japan and more recently by a Euratom partnership, and pressure tube concepts proposed by Canada, generically called the Canadian SCWR. Other than the specifics of the core design, these concepts have many similar features, like outlet pressure and temperatures, steam cycle options, materials, or heat transfer characteristics. Therefore, the R&D needs for each reactor type are common, which enables collaborative research to be pursued. The paper provides an overview on research and development performed so far on the SCWR within the Generation IV International Forum.

Decoupling Control of Canadian Supercritical Water-Cooled Reactors

The Canadian supercritical water-cooled reactor (SCWR) can be modeled as a multiple-input multiple-output system. It has a high power-to-flow ratio, strong cross coupling, and a high degree of nonlinearity in its dynamic characteristics. Because of the existence of strong cross coupling among system inputs and outputs, it is difficult for a traditional control system design methodology to produce a satisfactory control system. In this paper, the direct Nyquist array method is used first to decouple the system into a diagonally dominant form via a precompensator. After decoupling the system successfully, three single-input single-output dynamic compensators are synthesized in the frequency domain. By using the precompensator, the temperature variation because of disturbances at the reactor power and pressure is significantly reduced. The control system can effectively maintain the overall system stability and regulate the plant around a specified operating condition. To deal with the nonlinearities, a control strategy based on gain scheduling is adopted. Different sets of controllers are used for the plant at different load conditions. The proposed control strategies have been evaluated under various operating scenarios. The robustness of the controller with respect to operating condition changes is also investigated. It is shown that the decoupling control can effectively reduce the cross coupling inherent in the Canadian SCWR, and gain scheduling control can successfully achieve satisfactory performance for different operating conditions.

Assessment of CFD for the Canadian SCWR bundle with wire wraps

Within the Generation-IV (Gen-IV) International Forum, Atomic Energy of Canada Limited (AECL) is leading the effort in developing a conceptual design for the Canadian supercritical water-cooled reactor (SCWR). AECL proposed a new fuel bundle design with two rings of fuel elements placed between the central flow tube and the pressure tube. In the current paper, both the bare-rod and wire-wrapped bundle geometries were tested for the proposed fuel bundle design at 25 MPa using STAR-CCM + CFD code to predict the sheath temperature variation along the axial direction. SST k–ω and Reynolds-stress based turbulence models were compared with each other to assess their capability in predicting wall-to-fluid heat transfer in supercritical flows including the heat transfer deterioration (HTD) phenomenon. Using the turbulence model identified from the sensitivity analysis, the mesh refinement for the bare-rod geometry was performed in accordance with the ASME CFD numerical accuracy guidelines. Following these assessments, the recommended CFD model was then applied to simulate the Canadian SCWR fuel bundle with wire wraps. Predictions for the velocity and temperature variation in bare-rod and wire-wrapped geometries were compared. As a result of this study, it was found that wire wraps reduces the maximum temperature variation significantly for the current SCWR bundle design.

Nuclear data sensitivity and uncertainty for the Canadian supercritical water-cooled reactor

Accurate and complete nuclear data are a fundamental requirement for any nuclear reactor model. Present nuclear data evaluations have been tested and validated against experiments that cover operating conditions and materials that are typical of conventional reactors (e.g. thermal light and heavy water moderated reactors, and some fast reactors). The Canadian supercritical water-cooled reactor (SCWR) is an advanced reactor concept for which, like all advanced GEN-IV reactor concepts, operating conditions and materials differ significantly from conventional reactors. Consequently, current nuclear data evaluations may not be suitable for modeling the Canadian SCWR, either due to assumptions or gaps in the evaluations or gaps in the experimental data on which the evaluations are based. In this paper, a simplified model of an SCWR fuel channel with fresh fuel is analyzed in order to determine the sensitivities and uncertainties of the model to nuclear data. The results are used to determine which isotopes and which reactions make the largest contributions to the sensitivity and uncertainties in key reactor physics parameters such as the neutron multiplication factor, k, and various lattice reactivity coefficients.

Simulation Strategy for the Evaluation of Neutronic Properties of a Canadian SCWR Fuel Channel

The Canadian Supercritical-Water-Cooled Reactor (SCWR) is a vertical pressure tube reactor cooled with supercritical light water and moderated with heavy water. For normal operation, the local conditions of the coolant (density and temperature) and fuel (temperature) vary substantially along the channel. This means that to simulate adequately the behavior of the core under operating conditions or for anticipated accident scenario, expensive 3D transport calculations for a complete fuel channel are required. Here, we propose a simulation strategy that takes into account axial variations of the local conditions and avoids 3D transport calculations. This strategy consists in replacing the 3D simulation by a series of isolated 2D calculations followed by a single 1D simulation. It is shown that this strategy is efficient because the axial coupling along the fuel channel is relatively weak. In addition, the neutronic properties of a channel with axial reflector can be modeled using a simplified 3D transport calculation.

Construction and Analysis of a Dynamic Model for a Canadian Direct-Cycle SCWR for Control System Studies

In this paper, a dynamic model of the Canadian supercritical water-cooled reactor (SCWR) is developed to examine its dynamics for potential control system design and analysis. The model development is based on fundamental mass, energy, and momentum conservation equations of major components within the Canadian SCWR operating at supercritical condition. A full set of nonlinear dynamic equations is first derived, from which linearized models are obtained. The linearized models are validated against the full-order nonlinear models in both time domain and frequency domain. The open-loop dynamic characteristics of the Canadian SCWR are investigated through extensive simulations. Steady-state and dynamic couplings among different inputs and outputs are examined using relative gain array and Nyquist plots, and adequate input-output pairings are identified. Cross-coupling at different operating conditions is also evaluated to illustrate the nonlinear behaviors of the system. The developed dynamic model provides a necessary platform for systematic investigation in the control system design and analysis of the Canadian SCWR.

The Supersafe© Reactor: A Small Modular Pressure Tube SCWR

The SUPERSAFE© Reactor (SSR) is proposed as a small modular version of the Canadian supercritical water-cooled reactor (SCWR). The SCWR is Canada’s primary contribution to the Generation-IV (GEN-IV) International Forum’s (GIF) research and development effort toward the study and eventual deployment of advanced nuclear energy systems. All GEN-IV concepts, including the SCWR, have enhanced safety, improved economics, improved sustainability and enhanced security compared to contemporary reactors. The SUPERSAFE© Reactor (SSR) concept incorporates the enhanced features of the SCWR in a smaller core which could be deployed in areas with sparsely distributed population bases where it is impractical to have a full scale SCWR or large centralized energy grid. An overview of the SSR concept is presented in this work.

Breakage induced by a grinding ball dropped onto a randomly packed particle bed

Within the Generation-IV (Gen-IV) International Forum, Atomic Energy of Canada Limited (AECL) is leading the effort in developing a conceptual design for the Canadian supercritical water-cooled reactor (SCWR). AECL proposed a new fuel bundle design with two rings of fuel elements placed between the central flow tube and the pressure tube. In the current paper, both the bare-rod and wire-wrapped bundle geometries were tested for the proposed fuel bundle design at 25 MPa using STAR-CCM + CFD code to predict the sheath temperature variation along the axial direction. SST k–ω and Reynolds-stress based turbulence models were compared with each other to assess their capability in predicting wall-to-fluid heat transfer in supercritical flows including the heat transfer deterioration (HTD) phenomenon. Using the turbulence model identified from the sensitivity analysis, the mesh refinement for the bare-rod geometry was performed in accordance with the ASME CFD numerical accuracy guidelines. Following these assessments, the recommended CFD model was then applied to simulate the Canadian SCWR fuel bundle with wire wraps. Predictions for the velocity and temperature variation in bare-rod and wire-wrapped geometries were compared. As a result of this study, it was found that wire wraps reduces the maximum temperature variation significantly for the current SCWR bundle design.