Canadian Supercritical Water-cooled Reactor Concept Research Articles

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.

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.

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.

AN ECONOMIC ANALYSIS OF THE CANADIAN SCWR CONCEPT USING G4-ECONS

A top-down cost estimation of the Canadian supercritical water-cooled reactor (SCWR) concept has been performed using the methodology developed by the Generation-IV International Forum Economic Modelling Working Group. This methodology establishes 2 figures of merit: the total capital investment cost (TCIC) and the levelised unit energy cost (LUEC). The calculations are based on 2007 United States dollars and assume an nth-of-a-kind Canadian SCWR. The analysis considers capital costs, operating and maintenance costs, fuel costs, and decommissioning costs in estimating TCIC and LUEC for the Canadian SCWR concept. Sensitivity analyses of the 2 figures of merit with respect to the uncertainties in the cost assumptions have been carried out to identify key factors affecting the estimates. Limitations of the methodology are discussed.

EVOLUTION OF THE CANADIAN SCWR FUEL-ASSEMBLY CONCEPT AND ASSESSMENT OF THE 64 ELEMENT ASSEMBLY FOR THERMALHYDRAULIC PERFORMANCE

Canada has participated in the Generation IV (Gen-IV) International Forum with a 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 research and development framework for development of the Canadian SCWR fuel-assembly concept includes experiments, analysis, subchannel code development, and applications, as well as coupled neutronic and thermalhydraulic analyses. The evolution of the Canadian SCWR fuel-assembly concept is summarized in this paper to demonstrate the enhanced performance characteristics and to show that the peak fuel cladding temperature is within allowed limits. The peak cladding temperature is a limiting thermalhydraulic parameter for the Canadian SCWR fuel assembly and is calculated using the ASSERT-PV subchannel code under steady state normal operating conditions with axial and radial power profiles corresponding to the beginning and the end of the fuel cycle. The power profiles were obtained from WIMS AECL/RFSP and MCNP physics code calculations. Using the ASSERT-PV subchannel code, the geometry of the Canadian SCWR fuel assembly has been further optimized to reduce the peak fuel cladding temperature for axial and radial power profiles of interest to the Canadian SCWR concept.

Coupled 3D neutron kinetics and thermalhydraulic characteristics of the Canadian supercritical water reactor

The Canadian Supercritical Water-cooled Reactor concept, as an evolution of the CANada Deuterium Uranium (CANDU) reactor, includes both pressure tubes and a low temperature heavy water moderator. The current Pressure Tube type SCWR (PT-SCWR) concept features 64-element fuel assemblies placed within High Efficiency Re-entrant Channels (HERCs) that connect to core inlet and outlet plena. Among current SCWR concepts the PT-SCWR is unique in that the HERC separates multiple coolant and moderator regions, giving rise to coupled neutronic-thermalhydraulic feedbacks beyond those present in CANDU or contemporary Light Water Reactors. The objective of this work was thus to model the coupled neutronic-thermal hydraulic properties of the PT-SCWR to establish the impact of these multiple regions on the core's transient behavior. To that end, the features of the PT-SCWR were first modeled with the neutron transport code DRAGON to create a database of homogenized and condensed cross-sections and thermalhydraulic feedback coefficients. These were used as input to a core-level neutron diffusion model created with the code DONJON. The behavior of the primary heat transport system was modeled with the thermalhydraulic system code CATHENA. A procedure was developed to couple the outputs of DONJON and CATHENA, facilitating three-dimensional spatial neutron kinetics and coupled thermalhydraulic analysis of the PT-SCWR core. Several postulated transients were initiated within the coupled model by changing the core inlet and outlet boundary conditions. Decreasing coolant density around the fuel was demonstrated to produce positive power excursions (i.e., the coolant void reactivity around the fuel was positive), but such power transients were found to be inherently self-terminating as low density coolant inevitably reaches other parts of the HERC geometry (where the void reactivity is highly negative). Nevertheless, the observed power excursions potentially demonstrate the need for fast-acting shutdown system intervention, similar to CANDU designs.

Development of High-Performance Pressure Tube Material for the Canadian SCWR Concept

The Canadian super-critical water-cooled reactor (SCWR) concept is moderated by using heavy water, while the coolant is light water at 25 MPa with an inlet temperature of 625 K and an outlet temperature of 900 K. The fuel assemblies reside in vertical pressure tubes that are the pressure boundary. The pressure tubes are insulated from the fuel assemblies and operate at temperatures near the moderator temperature, at 390 K. The zirconium alloy Excel has been selected as a candidate material for the pressure tube based on favorable properties such as high strength, resistance to radiation-induced diametral strain, and high terminal solid solubility. However, significant future effort will be required to obtain material properties and crack initiation mechanisms at super-critical water (SCW) conditions to verify that annealed Excel is a viable option as a pressure tube material in the Canadian SCWR.

Corrosion Assessment of Candidate Materials for Fuel Cladding in Canadian SCWR

The supercritical water-cooled reactor (SCWR) is an innovative next generation reactor that offers many promising features, but the high-temperature high-pressure coolant introduces unique challenges to the long-term safe and reliable operation of in-core components, in particular the fuel cladding. To achieve high thermal efficiency, the Canadian SCWR concept has a coolant core outlet temperature of 625°C at 25 MPa with a peak cladding temperature as high as 800°C. International and Canadian research programs on corrosion issues in supercritical water have been conducted to support the SCWR concept. This paper provides a brief review of corrosion in supercritical water and summarizes the Canadian corrosion assessment work on potential fuel cladding materials. Five alloys, SS 347H, SS310S, Alloy 800H, Alloy 625 and Alloy 214, have been shown to have sufficient corrosion resistance to be used as the fuel cladding. Additional work, including tests in an in-reactor loop, is needed to confirm that these alloys would work as the fuel cladding in the Canadian SCWR.

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.

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.