Nuclear cogeneration with high temperature reactors

There is growing interest in the use of advanced reactor systems for powering industrial processes which could significantly help to reduce CO 2 emissions in the global energy system. However, there has been limited consideration into the role nuclear power would play in meeting current and future industry heat demand, especially with respect to the advantages and disadvantages nuclear power offers relative to other competing low-carbon technologies, such as Carbon Capture and Storage (CCS). In this study, the current market needs for high temperature heat are considered based on UK industry requirements and work carried out in other studies regarding how industrial demand could change in the future. How these heat demands could be met via different nuclear reactor systems is also presented. Using this information, it was found that the industrial heat demands for temperature in the range of 500 ∘ C to 1000 ∘ C are relatively low. Whilst High Temperature Gas-cooled Reactors (HTGRs), Very High Temperature Reactors (VHTRs), Gas-cooled Fast Reactors (GFRs) and Molten Salt Reactors (MSRs) have an advantage in terms of capability to achieve higher temperatures (>500 ∘ C), their relative benefit over Liquid Metal-cooled Fast Reactors (LMFRs) and Light Water Reactors (LWRs) is actually smaller than previous studies indicate. This is because, as is shown here, major parts of the heat demand could be served by almost all reactor types. Alternative (non-nuclear) means to meet industrial heat demands and the indirect application of nuclear power, in particular via producing hydrogen, are also considered. As hydrogen is a relatively poor energy carrier, current trends indicate that the use of low-carbon derived hydrogen is likely to be limited to certain applications and there is a focus in this study on the emerging demands for hydrogen.

VHTR-Based Systems for Autonomous Co-Generation Applications

Japan Atomic Energy Agency (JAEA) has started a conceptual design of a small-sized HTGR for steam supply and electricity generation (HTR50S) to deploy the high temperature gas cooled reactor (HTGR) in developing countries at an early date (i.e., in the 2030s). Its reactor power is 50MWt and the reactor outlet temperature is 750°C. It is a first-of-kind of the commercial plant or a demonstration plant of a small-sized HTGR system for steam supply to the industries and the district heating, and electricity generation using a steam turbine. The design philosophy of the HTR50S is to upgrade the performance from the Japanese first HTGR (HTTR) and to reduce the cost for the commercialization by utilizing the knowledge obtained by the HTTR operation and the design of an advanced commercial plant of 600 MWt-class Very High Temperature Reactor (GTHTR300 series). The major specifications of the HTR50S were determined based on its design philosophy. And the targets of the technology demonstration using the HTR50S for the future commercial small-sized HTGR were identified. The system design of HTR50S was performed to offer the capability of electricity generation, cogeneration of electricity and steam for a district heating and industries. The market potential for the small-sized HTGR in the developing countries was evaluated for the application of the electricity, process heat, district heating and pure water production. It was confirmed that there is enough market potential for the small-sized HTGR in the developing countries. This paper described the major specification and system design of the HTR50S and the market potential for the small-sized HTGR in the developing countries.

VHTR-based systems for autonomous co-generation applications

This paper presents an analysis of transuranium nuclide (TRU)-fueled very high temperature reactor (VHTR) systems focusing on applications requiring an extended operation with minimized control and no refueling (single-batch mode). As an example of such applications, international deployment opportunities for grid-appropriate VHTR systems could be mentioned addressing demands for electricity, industrial heat, and co-generation in those regions where minimized servicing is desirable for various reasons. The study is performed for the hexagonal block core concept within the framework of the ongoing U.S. DOE Nuclear Energy Research Initiative project on utilization of higher actinides (TRUs and partitioned minor actinides (MAs)) as a fuel component for extended-life VHTRs. The up-to-date analysis has shown reasonable reactivity swings, core life limits with respect to fast fluences, and criticality.

The mission of the U.S. Department of Energy (DOE), Office of Nuclear Energy is to advance nuclear power in order to meet the nation's energy, environmental, and energy security needs. Advanced high temperature reactor systems such as sodium fast reactors and high and very high temperature gas-cooled reactors are being considered for the next generation of nuclear reactor plant designs. The coolants for these high temperature reactor systems include liquid sodium and helium gas. Supercritical carbon dioxide (sCO₂), a fluid at a temperature and pressure above the supercritical point of CO₂, is currently being investigated by DOE as a working fluid for a nuclear or fossil-heated recompression closed Brayton cycle energy conversion system that operates at 550°C (1022°F) at 200 bar (2900 psi). Higher operating temperatures are envisioned in future developments. All of these design concepts require a highly effective heat exchanger that transfers heat from the nuclear or chemical reactor to the chemical process fluid or the to the power cycle. In the nuclear designs described above, heat is transferred from the primary to the secondary loop via an intermediate heat exchanger (IHX) and then from the intermediate loop to either a working process or a power cycle via a secondary heat exchanger (SHX). The IHX is a component in the primary coolant loop which will be classified as "safety related." The intermediate loop will likely be classified as "not safety related but important to safety." These safety classifications have a direct bearing on heat exchanger design approaches for the IHX and SHX. The very high temperatures being considered for the VHTR will require the use of very high temperature alloys for the IHX and SHX. Material cost considerations alone will dictate that the IHX and SHX be highly effective; that is, provide high heat transfer area in a small volume. This feature must be accompanied by low pressure drop and mechanical reliability and robustness. Classic shell and tube designs will be large and costly, and may only be appropriate in steam generator service in the SHX where boiling inside the tubes occurs. For other energy conversion systems, all of these features can be met in a compact heat exchanger design. This report will examine some of the ASME Code issues that will need to be addressed to allow use of a Code-qualified compact heat exchanger in IHX or SHX nuclear service. Most effort will focus on the IHX, since the safety-related (Class A) design rules are more extensive than those for important-to-safety (Class B) or commercial rules that are relevant to the SHX.

The present work is concerned with heat exchanger development for molten salt service, including the proposed molten salt reactor (MSR), a homogeneous reactor in which the fuel is dissolved in a circulating fluid of molten salt. It is an outgrowth of recent work done under the Next Generation Nuclear Plant (NGNP) program; what the two reactor systems have in common is an inherently safe nuclear plant with a high outlet temperature that is useful for process heat as well as more conventional generation The NGNP program was tasked with investigating the application of a new generation of nuclear power plants to a variety of energy needs. One baseline reactor design for this program is a high temperature, gas-cooled reactor (HTGR), which provides many options for energy use. These might include the conventional Rankine cycle (steam turbine) generation of electricity, but also other methods: for example, Brayton cycle (gas turbine) electrical generation, and the direct use of the high temperatures characteristic of HTGR output for process heat in the chemical industry. Such process heat is currently generated by burning fossil fuels, and is a major contributor to the carbon footprint of the chemical and petrochemical industries. The HTGR, based on graphite fuel elements, can produce very high output temperatures; ideally, temperatures of 900 °C or even greater, which has significant energy advantages. Such temperatures are, of course, at the frontiers of materials limitations, at the upper end of the performance envelope of the metallic materials for which robust construction codes exist, and within the realm of ceramic materials, the fabrication and joining of which, on the scale of large energy systems, are at an earlier stage of development. A considerable amount of work was done in the diffusion welding of materials of interest for HTGR service with alloys such as 617 and 800H. The MSR output temperature is also materials limited, and is projected at about 700 °C. (RR E) A different set of alloys, such as Alloy N and 242, are needed to handle molten salts at this temperature. The diffusion welding development work described here builds on techniques developed during the NGNP work, as applied to these alloys. There is also the matter of dissimilar metal welding, since alloys suitable for salt service are generally not suited for service in gaseous oxidizing environments, and vice versa, and welding is required for the Class I boundaries in these systems, as identified in the relevant ASME codes.

The project supported the Nuclear Regulatory Commission (NRC) in identifying and evaluating the regulatory implications concerning the control and protection systems proposed for use in the Department of Energy's (DOE) Next-Generation Nuclear Plant (NGNP). The NGNP, using modular high-temperature gas-cooled reactor (HTGR) technology, is to provide commercial industries with electricity and high-temperature process heat for industrial processes such as hydrogen production. Process heat temperatures range from 700 to 950 C, and for the upper range of these operation temperatures, the modular HTGR is sometimes referred to as the Very High Temperature Reactor or VHTR. Initial NGNP designs are for operation in the lower temperature range. The defining safety characteristic of the modular HTGR is that its primary defense against serious accidents is to be achieved through its inherent properties of the fuel and core. Because of its strong negative temperature coefficient of reactivity and the capability of the fuel to withstand high temperatures, fast-acting active safety systems or prompt operator actions should not be required to prevent significant fuel failure and fission product release. The plant is designed such that its inherent features should provide adequate protection despite operational errors or equipment failure. Figure 1 shows an example modular HTGR layout (prismatic core version), where its inlet coolant enters the reactor vessel at the bottom, traversing up the sides to the top plenum, down-flow through an annular core, and exiting from the lower plenum (hot duct). This research provided NRC staff with (a) insights and knowledge about the control and protection systems for the NGNP and VHTR, (b) information on the technologies/approaches under consideration for use in the reactor and process heat applications, (c) guidelines for the design of highly integrated control rooms, (d) consideration for modeling of control and protection system designs for VHTR, and (e) input for developing the bases for possible new regulatory guidance to assist in the review of an NGNP license application. This NRC project also evaluated reactor and process heat application plant simulation models employed in the protection and control system designs for various plant operational modes and accidents, including providing information about the models themselves, and the appropriateness of the application of the models for control and protection system studies. A companion project for the NRC focused on the potential for new instrumentation that would be unique to modular HTGRs, as compared to light-water reactors (LWRs), due to both the higher temperature ranges and the inherent safety features.

ABSTRACTA commercial very high-temperature gas-cooled reactor (VHTR) hydrogen cogeneration system named gas turbine high temperature reactor 300-cogeneration (GTHTR300C) is designed and developed in Japan Atomic Energy Agency (JAEA). Moreover, it has been planned that hydrogen production system and gas turbine system is connect to high-temperature engineering test reactor (HTTR). The reactor system is required to continue a stable and safety operation as well as a stable power supply in the case that thermal-load is fluctuated by the occurrence of abnormal event in the heat utilization system like as the hydrogen production facility. Then, it is necessary to confirm that the thermal-load fluctuation could be absorbed by the reactor system so as to continue the stable and safety operation. The thermal-load fluctuation absorption tests using the HTTR were planned to clarify the absorption characteristics of the HTGR system. However, it is difficult to clarify the phenomenon due to many kinds of fluctuation in nuclear thermal power in the reactor core. Moreover, the actual data regarding how the delay of the temperature response is effective for the reactor system had been gained quantitatively.The thermal-load fluctuation absorption tests without nuclear heating were planned and conducted in JAEA to clarify the absorption characteristic of thermal-load fluctuation mainly by the reactor and by the intermediate heat exchanger (IHX). The absorption characteristics of thermal-load fluctuation can be revealed with sufficient temperature fluctuation. So the tests were conducted with the primary coolant temperature 120 °C which is the start-up temperature of the HTTR. As a result it was revealed that the reactor has the larger absorption capacity of thermal-load fluctuation than the expected one and that the IHX can be contributed to the absorption of the thermal-load fluctuation generated in the heat utilization system in the reactor system. It was confirmed from their results that the reactor and the IHX has effective absorption capacity of the thermal-load fluctuation generated in the heat utilization system. Moreover, the calculation with the safety evaluation code based on RELAP5/MOD3 was performed. It was confirmed that the calculated temperature for the reactor is almost same to the measured one with the new analysis model. On the other hand, it was confirmed that the calculated temperature for the IHX decreased faster than the measured one due to smaller absorption capacity in the calculation model than that in actual one. It can be considered that the calculation for the IHX produces the conservative result. It was summarized that the safety evaluation code can represent the thermal-load fluctuation absorption behavior conservatively.

The high-temperature gas-cooled reactor (HTGR) with a high outlet temperature from 700°C to 800°C to 1000°C is expected to be widely used for heat supply, hydrogen production, steelmaking, seawater desalination, thermal recovery of heavy oil, and coal liquefaction and gasification. The 10-MW High Temperature gas-cooled test Reactor (HTR-10), with an outlet temperature of 700°C, has been constructed and operated in China as a pilot plant to demonstrate the inherent safety features of the modular HTGR. Supported by the Chinese National S&T Major Project and National Key R&D Program of China, some research on HTGR technology with much higher outlet temperatures has been carried out in the Institute of Nuclear and New Energy Technology (INET), Tsinghua University, China, including the preliminary design and analysis of a HTR-10 with an outlet helium temperature of 950°C (HTR-10H). This paper presents the simulation results of the depressurized loss-of-forced cooling (DLOFC) accident of the HTR-10H. This analysis was performed with two different codes, TINTE and SPECTRA, within a cooperation between the Nuclear Research Group of the Netherlands and the INET. In the past, comparisons between TINTE and SPECTRA have also been performed for the Chinese 200-MW(electric) High Temperature gas-cooled Reactor Pebble bed Module (HTR-PM). The calculation results showed that both the fuel temperature and the reactor pressure vessel temperature remained below the acceptable limits during the DLOFC accident. The inherent safety of the HTR-10 can be guaranteed in the accident condition. Both the TINTE results and SPECTRA results for the maximum and average fuel temperature agreed well. In both models, the heat carrying capacity of the Reactor cavity cooling system (RCCS) had been underestimated, especially in the SPECTRA model. The RCCS performance needs to be investigated and the models need to be further improved in the future. The work introduced in this paper shows the feasibility of increasing the outlet helium temperature of the HTR-10 from the current 700°C to 950°C, if further design optimization can effectively reduce the fuel temperature during steady-state operation, or more irradiation tests can be carried out to prove the retention performance of the TRISO particles for the radioactive fission products at temperatures above 1200°C.

KAERI (Korea Atomic Energy Research Institute) has been studying a VHTR (Very High Temperature Reactor) to utilize a nuclear power plant for diverse purposes. The VHTR is designed to operate under high-temperature and pressure condition for its lifetime (40 years or longer). To use the VHTR safely and efficiently, a reliable code to analyse the thermal-hydraulics in the reactor core is in demand. A commercial CFD tool might give an accurate solution, but it still has a burden in terms of the computational speed and computer memory usage. A system code to solve a fluid one dimension and a solid as three dimensions has been developed to investigate the temperature distribution and flow rate in the reactor core of a VHTR. In the present study, the hot spot temperature and temperature gradient in the columns in the reactor core are investigated with combinations of various unit cells.

Facing Global Environmental Change

Recent advances on the use of reprocessed fuels and combined thorium fuel cycles in HTR systems

The development of High-Temperature Gas-cooled Reactors (HTGRs) has a long history, and originated from the Dragon Project in 1959. After the achievement of the Dragon Reactor's criticality in 1964, the Fort-Saint Vrain Reactor in the U.S. and the THTR-300 in Germany were constructed and successfully operated.In Japan the High-Temperature engineering Test Reactor (HTTR) with a thermal output of 30MW is now under construction. Its main objectives are: fundamental or basic studies for high-temperature engineering, development for future HTGR's technology, and research on heat application to industrial purposes.In addition to this widely envisaged use of high-temperature heat, HTGRs with excellentsafety and easy operation features have a possibility of obtaining higher thermal efficiency, because of its high outlet temperature. In this sense, an HTGR system with a high-temperature helium-gas turbine is most promising. The maximum thermal efficiency of this system under the same operational condition of HTTR is expected to be about 45% or more.

Assessment of load-following capability of VHTR cogeneration systems