Research and Development Roadmaps for Liquid Metal Cooled Fast Reactors – Draft for Public Comment

During the last decade the European activities in the field of nuclear fission research include the design of fast reactors cooled by liquid metals. Within this framework, the Fast REactor NEutronics/Thermal-hydraulICs (FRENETIC) code is being developed at Politecnico di Torino over the last few years. It implements a full-core coupled neutronic/thermalhydraulic model of a liquid-metal-cooled fast reactor as relevant for two of the six options currently under study within the framework of the Generation-IV International Forum, namely the lead-cooled fast reactors and the sodium-cooled fast reactors. The code validation process involves the participation in a coordinated research project of the International Atomic Energy Agency, aiming at testing different computational tools against the shutdown heat removal tests performed many years ago in the sodium-cooled Experimental Breeder Reactor-II (EBR-II) at Argonne National Laboratory, USA. In this paper, results of the FRENETIC analysis of one of the transients considered in the project, the unprotected EBR-II shutdown heat removal test SHRT-45R, are presented and compared to the measurements, providing the first validation of the coupled neutronic/thermal-hydraulic features of the FRENETIC code.

Solar power has innate issues with weather, grid demand and time of day, which can be mitigated through use of thermal energy storage for concentrating solar power (CSP). Nuclear reactors, including lead-cooled fast reactors (LFRs), can adjust power output according to demand; but with high fixed costs and low operating costs, there may not be sufficient economic incentive to make this worthwhile. We investigate potential synergies through coupling CSP and LFR together in a single supercritical CO2 Brayton cycle and/or using the same thermal energy storage. Combining these cycles allows for the LFR to thermally charge the salt storage in the CSP cycle during low-demand periods to be dispatched when grid demand increases. The LFR/CSP coupling into one cycle is modeled to find the preferred location of the LFR heat exchanger, CSP heat exchanger, sCO2-to-salt heat exchanger (C2S), turbines, and recuperators within the supercritical CO2 Brayton cycle. Three cycle configurations have been studied: two-cycle configuration, which uses CSP and LFR heat for dedicated turbocompressors, has the highest efficiencies but with less component synergies; a combined cycle with CSP and LFR heat sources in parallel is the simplest with the lowest efficiencies; and a combined cycle with separate high-temperature recuperators for both the CSP and LFR is a compromise between efficiency and component synergies. Additionally, four thermal energy storage charging techniques are studied: the turbine positioned before C2S, requiring a high LFR outlet temperature for viability; the turbine after the C2S, reducing turbine inlet temperature and therefore power; the turbine parallel to the C2S producing moderate efficiency; and a dedicated circulator loop. While all configurations have pros and cons, use of a single cycle offers component synergies with limited efficiency penalty. Using a turbine in parallel with the C2S heat exchanger is feasible but results in a low charging efficiency, while a dedicated circulator loop offers flexibility and near-perfect heat storage efficiency but increasing cost with additional cycle components.

Significant progress has been made in the ongoing development of the Argonne National Laboratory (ANL) Plant Dynamics Code (PDC), the ongoing investigation and development of control strategies, and the analysis of system transient behavior for supercritical carbon dioxide (S-CO{sub 2}) Brayton cycles. Several code modifications have been introduced during FY2011 to extend the range of applicability of the PDC and to improve its calculational stability and speed. A new and innovative approach was developed to couple the Plant Dynamics Code for S-CO{sub 2} cycle calculations with SAS4A/SASSYS-1 Liquid Metal Reactor Code System calculations for the transient system level behavior on the reactor side of a Sodium-Cooled Fast Reactor (SFR) or Lead-Cooled Fast Reactor (LFR). The new code system allows use of the full capabilities of both codes such that whole-plant transients can now be simulated without additional user interaction. Several other code modifications, including the introduction of compressor surge control, a new approach for determining the solution time step for efficient computational speed, an updated treatment of S-CO{sub 2} cycle flow mergers and splits, a modified enthalpy equation to improve the treatment of negative flow, and a revised solution of the reactor heat exchanger (RHX) equations coupling the S-CO{sub 2} cycle to the reactor, were introduced to the PDC in FY2011. All of these modifications have improved the code computational stability and computational speed, while not significantly affecting the results of transient calculations. The improved PDC was used to continue the investigation of S-CO{sub 2} cycle control and transient behavior. The coupled PDC-SAS4A/SASSYS-1 code capability was used to study the dynamic characteristics of a S-CO{sub 2} cycle coupled to a SFR plant. Cycle control was investigated in terms of the ability of the cycle to respond to a linear reduction in the electrical grid demand from 100% to 0% at a rate of 5%/minute. It was determined that utilization of turbine throttling control below 50% load improves the cycle efficiency significantly. Consequently, the cycle control strategy has been updated to include turbine throttle valve control. The new control strategy still relies on inventory control in the 50%-90% load range and turbine bypass for fine and fast generator output adjustments, but it now also includes turbine throttling control in the 0%-50% load range. In an attempt to investigate the feasibility of using the S-CO{sub 2} cycle for normal decay heat removal from the reactor, the cycle control study was extended beyond the investigation of normal load following. It was shown that such operation is possible with the extension of the inventory and the turbine throttling controls. However, the cycle operation in this range is calculated to be so inefficient that energy would need to be supplied from the electrical grid assuming that the generator could be capable of being operated in a motoring mode with an input electrical energy from the grid having a magnitude of about 20% of the nominal plant output electrical power level in order to maintain circulation of the CO{sub 2} in the cycle. The work on investigation of cycle operation at low power level will be continued in the future. In addition to the cycle control study, the coupled PDC-SAS4A/SASSYS-1 code system was also used to simulate thermal transients in the sodium-to-CO{sub 2} heat exchanger. Several possible conditions with the potential to introduce significant changes to the heat exchanger temperatures were identified and simulated. The conditions range from reactor scram and primary sodium pump failure or intermediate sodium pump failure on the reactor side to pipe breaks and valve malfunctions on the S-CO{sub 2} side. It was found that the maximum possible rate of the heat exchanger wall temperature change for the particular heat exchanger design assumed is limited to {+-}7 C/s for less than 10 seconds. Modeling in the Plant Dynamics Code has been compared with available data from the Sandia National Laboratories (SNL) small-scale S-CO{sub 2} Brayton cycle demonstration that is being assembled in a phased approach currently at Barber-Nichols Inc. and at SNL in the future. The available data was obtained with an earlier configuration of the S-CO{sub 2} loop involving only a single-turbo-alternator-compressor (TAC) instead of two TACs, a single low temperature recuperator (LTR) instead of both a LTR and a high temperature recuperator (HTR), and fewer than the later to be installed full set of electric heaters. Due to the absence of the full heating capability as well as the lack of a high temperature recuperator providing additional recuperation, the temperature conditions obtained with the loop are too low for the loop conditions to be prototypical of the S-CO{sub 2} cycle.

The stepwise conversion of Russian nuclear power industry to a two-component nuclear energy system (NES) with thermal and fast reactors in a single closed NFC allows a number of deferred system problems of modern power industry to be solved, such as accumulation of spent fuel from thermal reactors and spent fuel repatriated from foreign Russian-designed NPPs; inefficient use of highly limited raw uranium inventories; recycling of minor actinides and other long-lived high-level waste accumulated in the course of thermal reactors' operation. The paper examines the fundamental capabilities of two-component NES, which appear due to the availability of commercial fast neutron sodium reactors. Due to the synergistic development of thermal and fast reactors, it is expected to achieve an economic effect associated with introduction of new fuel services in the nuclear energy system, such as production of plutonium and improvement of its isotopic composition for thermal reactors, transmutation of "external" minor actinides, production of isotopes for various purposes, production of hydrogen, etc. The expansion of Russia's export opportunities is also being considered, it can be done through fuel supply for both power units built in compliance with Russian technologies and foreign power units with MOX fuel. In this approach, it is important to analyze the economy of the whole system, where fast reactors, due to their inherent surplus of neutrons, perform the functions required to the entire NES, rather than to analyze the technical and economic indicators of individual reactors. The article makes a preliminary technical and economic assessment of such system functions provided by fast sodium reactors. It is shown that fast and thermal reactors operating together in a two-component nuclear energy system, with the same electrical power and similar technical and economic indicators, can significantly reduce the necessary infrastructure costs and/or generate additional income, thereby reducing the specific levelized cost of electricity production.

The Westinghouse lead-cooled fast reactor (LFR) is a next-generation nuclear power plant whose primary mission is to reduce front-end capital costs and generate flexible and cost-competitive electricity for global markets, while offering mission versatility and satisfying the highest safety and sustainability standards. The LFR is an economical choice for carbon-free power generation to meet the ambitious goals of reducing carbon emissions. The plant utilizes passive safety systems for high reliability and public safety. The LFR testing program aims to fill the technology gaps in the key materials, components, and systems of the LFR plant. It provides demonstrations of LFR engineering that reduce the uncertainty of the LFR design and experimental data for the verification and validation of the modeling and simulation computer codes. With the support of the LFR phenomena identification and ranking table, the phenomena important to plant safety with insufficient states of knowledge have been identified and a Westinghouse testing plan developed to address them. The United Kingdom Advanced Modular Reactor Program Phase 2 was purposed to advance the level of design maturity toward eventual regulatory approval and deployment. Westinghouse and its partners have developed eight state-of-the-art test facilities within the program. Among them, the Passive Heat Removal Facility, Versatile Lead Facility, Lead Water Interaction Facility, and the Lead Freezing Facility are thermal-hydraulic testing facilities, with the other four facilities dedicated to testing lead corrosion and material mechanical behavior in lead. The design, development, and testing of the four thermal-hydraulic facilities are presented in this paper. A variety of computer codes ranging from system-level analysis, component-level analysis, and high-resolution computational fluid dynamics analysis, are utilized to support the development of the facilities. The analyses provide operational and accidental conditions in the LFR to identify major testing conditions and inform the facility design and testing matrix. With a significant amount of experimental data being generated with these highly instrumented test facilities, the computer codes and models will be benchmarked against the experimental results to improve their validation and verification status.

This paper presents an analysis of three new heat-transfer correlations developed for supercritical carbon dioxide (CO2) flowing in vertical bare tubes. A large set of experimental data was obtained at Chalk River Laboratories (CRL) AECL. Heat-transfer tests were performed in upward flow of CO2 inside 8-mm ID vertical Inconel-600 tube with a 2.208-m heated length. Data points were collected at outlet pressures ranging from 7.4 to 8.8 MPa, mass fluxes from 900 to 3000 kg/m2s, inlet fluid temperatures from 20 to 40°C, and heat fluxes from 15 to 615 kW/m2; and for several combinations of wall and bulk-fluid temperatures that were below, at, or above the pseudocritical temperature. The objective of the present experimental research is to obtain reference dataset on heat transfer in supercritical CO2 and improve our fundamental knowledge of the heat-transfer processes and handling of supercritical fluids. In general, heat-transfer process to a supercritical fluid is difficult to model, especially, when a fluid passes through the pseudocritical region, as there are very rapid variations in thermophysical properties of the fluid. Thus, it is important to investigate supercritical-fluid behaviour within these conditions. In general, supercritical carbon dioxide was and is used as a modelling fluid instead of supercritical water due to its lower critical parameters compared to those of water. Also, supercritical carbon dioxide is proposed to be used as a working fluid in the Brayton gas-turbine cycle as a secondary power cycle for some of the Generation-IV nuclear-reactor concepts such as a Sodium-cooled Fast Reactor (SFR), Lead-cooled Fast Reactor (LFR) and Molten-Salt-cooled Reactor (MSR). In addition, supercritical carbon dioxide was proposed to be used in advanced air-conditioning and geothermal systems. Previous studies have shown that existing correlations deviate significantly from experimental Heat Transfer Coefficient (HTC) values, especially, within the pseudocritical range. Moreover, the majority of correlations were mainly developed for supercritical water, and our latest results indicate that they cannot be directly applied to supercritical CO2 with the same accuracy as for water. Therefore, new empirical correlations to predict HTC values were developed based on the supercritical CO2 dataset. These correlations calculate HTC values with an accuracy of ±30% (wall temperatures with accuracy of ±20%) for the analyzed dataset.

Advanced nuclear reactors offer various operational advantages over existing light water reactors but could produce types of spent nuclear fuel (SNF) with a wide variety of forms and characteristics depending on how many different concepts are deployed. Each advanced reactor SNF type potentially poses unique management challenges. New planning efforts will be necessary to anticipate how the management requirements of advanced reactor SNF will affect the deployment of an integrated waste management system. This paper applies a framework of high-level facility deployment milestones to a generic SNF management system, reviewing them together with the advanced reactor SNF characteristics and management requirements. This allows for the investigation of factors that influence facility and system deployment, and ultimately, the identification of challenges facing the deployment of different kinds of SNF management facilities. The back end of the once-through fuel cycle is examined for four advanced reactor system technology types: sodium-cooled fast reactors, high-temperature gas-cooled reactors, liquid-fuel molten salt reactors, and lead-cooled fast reactors. It is observed that milestones earlier in the facility deployment process (e.g., siting and facility design) are more impacted by the uniqueness of advanced reactor SNF characteristics than others (e.g., construction and testing). Ultimately, none of the differences are seen as fundamentally disqualifying in a technical sense; however, they should be considered early, potentially as part of reactor design, to avoid issues in the future.

Developing empirical heat-transfer correlations for supercritical CO2 flowing in vertical bare tubes

5 - Key technologies for future sodium-cooled fast reactors

Understanding and predicting nuclear fuel behavior is the cornerstone for fuel design and optimization, and the safe and economic operation of nuclear reactors. Due to the excellent neutronic economy and safe operational characteristics of lead/lead-bismuth-based liquid metal coolant, lead-cooled fast reactor (LFR) has received much attention and has been regarded as one of the top candidate reactors for Generation IV nuclear power plants. Mixed oxide (MOX) fuel is beneficial to improve fuel cycle utilization and consume weapons-grade plutonium, among which hollow MOX fuel has excellent heat transfer advantages, which can further improve fuel economy and safety, and has good development prospects. In this work, an annular uranium-plutonium mixed oxide (MOX) fuel operating in a liquid lead/lead-bismuth cooled fast reactor is modeled and simulated to predict its behavior under transient and steady-state operation. This model has several modules working in a fully coupled approach based on COMSOL Multiphysics, such as heat transfer, fuel burnup, oxygen redistribution, plutonium redistribution, porosity evolution, fission gas release, JOG growth, and grain size evolution. The modeling results were benchmarked with the existing codes and experimental data, and the results were in satisfactory agreement. The parameters analysis was carried out in this work. Consistent with the solid MOX fuel, the O/M ratio (or deviation from the stoichiometry of oxygen) significantly affects temperature evolution, fission gas migration and release behavior, and plutonium redistribution. Linear power also has a significant influence on fuel performance.

A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. The various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.

<p indent=0mm>After the Fukushima nuclear accident in Japan, the safety requirements of nuclear energy systems are constantly increasing. A series of new nuclear energy systems are under development, such as very high-temperature gas-cooled reactor (VHTR), lead-cooled fast reactor (LFR), gas-cooled fast reactor (GFR) and so on. In order to achieve the goals of higher safety, these nuclear systems have put forward higher requirements for nuclear materials. For example, materials working in VHTR and GFR are supposed to withstand higher temperatures, and LFR requires materials to be more corrosion-resistant. Nuclear materials have to meet these requirements under normal operating conditions, as well as in incidental and accidental conditions. Therefore, it is necessary to optimize the existing material system and develop new high-performance materials to meet the future development needs of advanced nuclear power systems. Compared with traditional nuclear materials, carbide ceramics have better mechanical properties under elevated temperatures, stronger irradiation resistance and better comprehensive thermophysical properties. Up to now, great breakthroughs have been made in the research of carbide ceramic materials and some key materials are going to be used in the new-generation nuclear systems. For example, uranium carbide (UC) is gradually replacing traditional uranium dioxide (UO₂), and silicon carbide (SiC) is acting as the main fission product barrier layer of coated fuel particles in HTGR. Other new carbide materials such as zirconium carbide (ZrC) and MAX phase carbide materials are also becoming a new research hotspot in the field of nuclear materials. Compared with other ceramic materials, the fabrication especially its densification of carbide is generally more difficult. In addition, high purity and good neutron properties are also the basic requirements of carbide ceramic materials for nuclear energy. The vapor deposition process and powder metallurgy process have been used to fabricate carbide ceramics for different components in nuclear reactors. Some new preparation methods have been developed to fabricate fuels and structural materials. For example, gelation progress is applied to prepare UCO microsphere fuel, and nano infiltration and transient eutectic (NITE) process is used in the preparation of SiC_f/SiC composites, which is potential for nuclear claddings. The irradiation-induced degradation of nuclear ceramic materials in the reactor is unavoidable. In order to evaluate the performance changes of materials under irradiation conditions, different irradiation methods such as electron irradiation, ion irradiation, proton irradiation and neutron irradiation are used. Due to its unique microstructure and properties, carbide ceramics exhibit different properties from other nuclear materials under irradiation. The performance changes of main nuclear carbide materials after irradiation are summarized, including microstructure evolution, macroscopic irradiation-induced swelling, changes in mechanical and thermal properties, etc. In general, based on the current research progress, the article reviews the application of carbide ceramics in new nuclear power systems and elaborates their service scenarios. Basic properties, preparation methods and irradiation properties of carbide ceramic materials are introduced in detail. The future development directions and the processing techniques are proposed. In the future, carbide ceramic materials should be further developed, and carbide ceramic materials will find diverse applications in the new generation nuclear systems.

Among the advanced reactor designs of the fourth generation of nuclear reactors, the Lead-cooled Fast Reactor (LFR) stands out as one of six distinct types.The corrosion damage of the steels, it is caused by liquid metal acting on the structural material, which is considered a significant obstacle in the development of the fast reactors concept.The corrosion resistance and chemical compatibility with environment of 9Cr-1Mo, ferritic alloy steel (T91), in condition of the static liquid lead with 1.10 -6 wt.% oxygen at 520 C are analysed in this study.The specimens were exposed for 3500 hours.The specimens tested were the reference T91 and T91 with the coating on base Al and Ti was applied to the second specimen.The coating applied by physical vapor deposition were used for corrosion resistance improvement.The formed oxide scale of reference T91 consists of an outer layer with a Fe-Cr spinel structure and an inner internal oxidation zone (IOZ).Corrosion damage was characterized using electron microscopy techniques, which identified networks of solution-based attack (SBA) intrusion beneath liquid lead wetted surfaces.The damage to the coating depends on their chemical composition and structure of the coating.However, in general, the application of the coating increased the resistance of the surface to corrosion damage.

With the increasing global demand for sustainable energy, the importance of advanced nuclear technologies, such as fourth-generation reactors, has become increasingly prominent. Fourth-generation reactors often use non-water coolants, among which liquid lead and lead–bismuth eutectics (LBEs) are highly promising, making lead-cooled fast reactors (LFRs) a popular area of research internationally. On the basis of extensive analysis and comparison conducted previously, this article summarizes and analyzes the advantages, problems, and differences in lead and LBE as LFR coolants. Overall, both lead and LBE have excellent neutronic characteristics, good heat transfer performance, and chemical inertness, and make LFRs highly efficient in nuclear fuel utilization, inherently safe, and relatively simplified in design. However, both of them corrode materials severely and produce highly toxic 210Po, which are the problems that need to be considered for further engineering development. Moreover, LBE has a lower melting point, which allows a wider temperature range and lower insulation requirements in its design, making it easier to achieve engineering and miniaturization under the same conditions. Lead has a lower cost, is less corrosive to materials, and produces less 210Po, which makes it a more ideal coolant for future development.

Feasibility study of detection of coolant void in liquid metal cooled fast reactors using changes in the neutron spectrum