Supercritical Water-cooled Reactor Concept Research Articles

Initial Multidisciplinary Study of Oxidized Chromium-Coated Zirconium Alloy for Fuel Cladding of SCW-SMR Concept: Weight-Gain and Thermal Conductivity Measurements and Coating Cost Evaluation

One of the challenges of small modular reactors (SMRs) in comparison with large reactors is the greater difficulty in achieving high burnups in smaller cores. With greater neutron leakage through the periphery, a key factor is the neutron economy of the fuel cladding. However, all large supercritical water-cooled reactor (SCWR) concepts have employed neutron-absorbing stainless steels and nickel-based alloys in order to meet all the requirements in terms of corrosion and thermalhydraulics. In order to achieve higher burnups and extend the time between refueling in a SCW-SMR, the use of chromium-coated zirconium alloy as a potential fuel cladding candidate has been explored. Chromium coatings up to a few micrometers thick have shown improved oxidation resistance of zirconium-based claddings under operating conditions relevant to SCWR concepts. In this study, Zr-2.5Nb alloy (UNS R60904) from pressure tube samples was coated using a physical vapor-deposition (PVD) method. Oxidation tests were performed on coated samples at 500 °C and approximately 25 MPa in a refreshed autoclave. The effects of the oxide on heat transfer and hydraulic resistance are also discussed in this study. Last, but not least, this study evaluates the coating cost of the fuel cladding with chromium in a vacuum plasma spray process.

ПРОБЛЕМЫ ФИЗИКИ И ТЕПЛОГИДРАВЛИКИ В РЕАКТОРАХ, ОХЛАЖДАЕМЫХ ВОДОЙ СВЕРХКРИТИЧЕСКОГО ДАВЛЕНИЯ

The supercritical Water-cooled Reactor (SCWR) was adopted as one of the promising IV-generation reactors within the framework of the international forum “Generation-IV” (MFP). More than 45 organizations in 16 countries with advanced nuclear power are developing SCWR concept proposals for this program. The SCWR concept is based on the implementation of a direct-flow single-circuit scheme of a nuclear power plant, cooled by SCP water. The introduction of this type of nuclear power plant will increase the efficiency up to 45 %, increase the fuel reproduction coefficient, reduce metal consumption and construction volumes, and improve economic and environmental performance. Countries participating in the SCWR MFP consider the development of a reactor with a thermal neutron spectrum and uranium fuel as a priority task, but in the subsequent stages, with increasing problems with the storage of spent nuclear fuel (SNF) and small actinides (SA), it is possible to switch to a reactor with a fast neutron spectrum, MOX fuel and a closed fuel cycle (CFC). Within the framework of the MFP, various versions of SCWR are being developed differing in the parameters of the coolant and its circulation schemes in the core. Groups have been created to study the issues of physics, thermohydraulics, heat transfer, materials, personnel training. Water-cooled reactors research carried out during ~15 years in A.I. Leypunsky Institute for Physics and Power Engineering (IPPE), OKB “Gidropress”, NRC “Kurchatov Institute” with supercritical thermal and fast neutron spectra, it seems more promising to develop a reactor with fast spectrum of neutrons. For ~10 years, IPPE and OKB “Gidropress” have been working together on the VVER-SKD concept project - a single-circuit RC with a coolant SCP with a fast-resonance neutron spectrum with a capacity of Ne = 1700 MW. This rector is recognized as a prospect for the development of VVER technology with the possibility of using uranium fuel and switching in the future to MOX-based fuel (U-Pu-Th) and to SNF. When developing VVER-SKD, it is necessary to solve a complex of scientific and technical problems: development and verification of calculation codes of improved estimation for neutron physics, hydrodynamics and water heat transfer of SCP in fuel assemblies (FA) of the core and throughout the reactor; development of fuel elements and FA structures, justification of their operability; analysis of reactor stability under transient and emergency conditions; selection of heat-resistant structural materials for fuel rods and reactor vessel with high corrosion resistance; justification and development of optimal water-chemical regime, etc. Some of these problems are investigated in bench and loop tests, but to solve most of them and justify the technology for subsequent licensing, it is necessary to create an experimental test reactor. In relation to the VVER-SKD reactor Ne = 1700 MW, the paper presents the results of calculations of fuel cycles with MOX and nitride fuels, justifies the use of NPS in the ZTC, discusses the problems of heat exchange and thermal hydraulics, and suggests a test reactor option.

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.

РАЗВИТИЕ НАПРАВЛЕНИЯ SCWR ОТ КОНЦЕПЦИИ ДО ТЕСТОВОГО РЕАКТОРА

The nuclear reactor cooled by supercritical pressure water - SCWR (Supercritical Water-cooled Reactor) is adopted as one of the promising reactors of the IV-generation. SCWR conceptual proposals are developed by more than 45 organizations in 16 countries with developed nuclear energy. The SCWR concept is based on the implementation of a single-circuit direct-flow scheme of nuclear power plants cooled by SKD water. The introduction of reactors of this type will increase the efficiency up to 45 %, increase the coefficient of fuel reproduction, reduce metal consumption and construction volumes, improve economic and environmental performance. Countries participating in the MFP in the direction of SCWR consider the development of a reactor with a thermal spectrum of neutrons and uranium fuel to be a priority, but in the subsequent stages, with an increase in problems with the storage of spent nuclear fuel (SNF) and Junior actinides (MA), it is possible to move to a reactor with a fast neutron spectrum, MOX fuel and a closed fuel cycle (ZTC). For ~10 years, the SSC RF - FEI and OKB "Gidropress" have been working together on the conceptual design of VVER-SKD - single-circuit RP with SCD coolant with a fast-resonance neutron spectrum with a capacity of Ne=1700 MW. This rector is recognized as the prospect of development of VVER technology with the possibility of using uranium fuel and the transition in the future to MOX-based fuel (U-Pu-Th) and to ZTZ. Rosatom recognizes this area as an innovative one, and has signed systemic agreements on Russia's participation in the work of the MFP in the SCWR direction. The paper presents the results of computational studies on the main version of the high-power reactor, as well as the test N=30 MW.

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.

Thermal-Hydraulic and Neutronic Analysis of a Reentrant Fuel-Channel Design for Pressure-Channel Supercritical Water-Cooled Reactors

To address the need to develop new nuclear reactors with higher thermal efficiency, a group of countries, including Canada, have initiated an international collaboration to develop the next generation of nuclear reactors called Generation IV. The Generation IV International Forum (GIF) Program has narrowed design options of the nuclear reactors to six concepts, one of which is supercritical water-cooled reactor (SCWR). Among the Generation IV nuclear-reactor concepts, only SCWRs use water as a coolant. The SCWR concept is considered to be an evolution of water-cooled reactors (pressurized water reactors (PWRs), boiling water reactors (BWRs), pressurized heavy water reactors (PHWRs), and light-water, graphite-moderated reactors (LGRs)), which comprise 96% of the current fleet of operating nuclear power reactors and are categorized under Generation II, III, and III+ nuclear reactors. The latter water-cooled reactors have thermal efficiencies of 30–36%, whereas the evolutionary SCWR will have a thermal efficiency of approximately 45–50%. In terms of a pressure boundary, SCWRs are classified into two categories, namely, pressure-vessel (PV) SCWRs and pressure-channel (PCh) SCWRs. A generic pressure-channel SCWR, which is the focus of this paper, operates at a pressure of 25 MPa with inlet and outlet coolant temperatures of 350°C and 625°C, respectively. The high outlet temperature and pressure of the coolant make it possible to improve thermal efficiency. On the other hand, high operating temperature and pressure of the coolant introduce a challenge for material selection and core design. In this view, there are two major issues that need to be addressed for further development of SCWR. First, the reactor core should be designed, which depends on a fuel-channel design. Second, a nuclear fuel and fuel cycle should be selected. Several fuel-channel designs have been proposed for SCWRs. These fuel-channel designs can be classified into two categories: direct-flow and reentrant channel concepts. The objective of this paper is to study thermal-hydraulic and neutronic aspects of a reentrant fuel-channel design. With this objective, a thermal-hydraulic code has been developed in MATLAB, which calculates fuel-centerline-temperature, sheath-temperature, coolant-temperature, and heat-transfer-coefficient profiles. A lattice code and diffusion code were used to determine a power distribution inside the core. Then, heat flux in a channel with the maximum thermal power was used as an input into the thermal-hydraulic code. This paper presents a fuel centerline temperature of a newly designed fuel bundle with UO2 as a reference fuel. The results show that the maximum fuel centerline temperature exceeds the design temperature limits of 1850°C for fuel.

Supercritical Water-Cooled Reactors

Supercritical water-cooled reactor (SCWR) is the watercooled reactor using supercritical pressure water as coolant [1]. It is considered as one of the promising Generation IV reactors, due to its advantages of plant simplification and high thermal efficiency [2, 3]. Several design concepts of SCWRs have been proposed: (a) supercritical water-cooled thermal neutron reactor; (b) supercritical water-cooled fast neutron reactor; (c) supercritical water-cooled mixed neutron spectrum reactor; (d) supercritical water-cooled pebble bed reactor; (e) supercritical heavy-water-cooled reactor. The detailed design parameters of some typical SCWR concepts around the world are summarized in Table 1 [4]. Recently, the use of thorium in the SCWRs has been investigated [5, 6]. The advantages of the SCWRs are shown as follows [1].

Full-core SSCWR calculations applying a fast computational method

The Supercritical Water-Cooled Reactor (SCWR) is a generation IV reactor concept which uses supercritical pressure water as coolant. By now, several versions of these concepts have been developed. A new thermal design called Simplified SCWR (SSCWR) was proposed in Reiss et al. (2010): the most important characteristics of the assembly are partial-length zirconium-hydride rods as extra moderator, axially varying fuel enrichment and a longer active length; also, the core had a smaller diameter compared to the original SCWR concept. This new type of assembly is optimized taking into account the neutronic and thermal-hydraulic properties. A one-pass core consisting of these assemblies is examined. Full-core calculations are computationally very expensive, thus a step-by-step homogenization process – complemented with extensive validation work – is applied using the code system SCALE6.0. This approach enabled a preliminary optimization of the one-pass core resulting in a favourable power distribution. This paper presents the outcome of the optimization of the assembly (fuel pin diameter 8.0 mm, pitch 10.12 mm, P/ D 1.1), the developed fast computational method and the preliminary results of the full-core calculations (average fuel enrichment around 6.0%, radial power peaking below 1.3, maximum coolant temperature below 800 K and maximum cladding temperature below 850 K).

Mixing of cooling water in the mixing chambers of the HPLWR—High Performance Light Water Reactor

The High Performance Light Water Reactor (HPLWR), a supercritical water cooled reactor concept with multiple heat-up steps, requires efficient mixing of the coolant between these steps to minimize hot spots in the core. Analyzing and improving the mixing in the mixing chamber above the core, situated between evaporator and superheater assemblies, and below the core, between the first and second superheater, is one of the challenges in the design process of the HPLWR. Different measures to enhance mixing have been studied with CFD analyses, in which a new design approach has been applied to the upper mixing chamber. It simplifies the complex structures and takes the effects of the disregarded structures into account by introducing source terms into the momentum equations.