VHTR-Based Systems for Autonomous Co-Generation Applications

VHTR-based systems for autonomous co-generation applications

Nuclear powered heat pumps for near-term process heat applications

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.

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.

ANTARES: The HTR/VHTR project at Framatome ANP

The very high temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 1173 K (900°C) and operational fuel temperatures above 1523 K (1250°C). The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the high-temperature gas-cooled reactor concepts have sufficiently high temperatures to support process heat applications, such as hydrogen production, tar sands, oil shale, desalination, or cogenerative processes, the VHTR’s higher temperatures can be detrimental to safety if a loss-of-coolant accident occurs and causes the mechanical strength degradation of the supporting graphite in the lower plenum. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion or density-gradient–driven stratified flow phenomena and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gases (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Therefore, there was a need to develop a computer code that can be used for VHTR air ingress-related graphite oxidation analyses. Prior to the start of the Republic of Korea/United States International Nuclear Energy Research Initiative collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate an air ingress phenomenon in the VHTR. Therefore, we have worked for the past 3 yr on developing and validating advanced computational methods for simulating air ingress in the VHTR. The Idaho National Laboratory is developing a system integration model of VHTR and hydrogen production plant. GAMMA code is being considered to be an integrated computer tool to analyze the thermal hydraulics of the coupled plant. Computer models for a high-temperature steam electrolysis (HTSE) process were developed and were implemented in an overall system process optimization code, HYSYS. The HTSE model will be implemented into GAMMA code as the integrated computer tool.This paper describes the governing equations and numerical methods used in GAMMA code and presents a portion of verification of the GAMMA code along with turbomachinery models and HTSE models that will be linked to GAMMA code.

Pebble bed very high temperature reactor (VHTR) has been identified as one of six Generation-IV (Gen-IV) types of reactor which could operate at a high thermal power. The calculation of the temperature in the fuel pebble is a key part of VHTR thermal hydraulics numerical simulation. However, due to the special structure of the VHTR fuel pebble, the temperature calculation involves a multiscale problem. The multiscale heat conduction model includes mesoscale temperature of fuel pebble and microscale temperature of TRISO fuel particles calculation. To deal with the particularity of temperature calculation of the fuel pebble, this paper presents a multiscale heat conduction model based on an open source CFD package OpenFOAM. Firstly, the quasi steady state heat conduction method (QSSHC) and homogeneous layers method (HL) was verified by a simple multiscale model. The results show that the QSSHC method has a good ability of multiscale temperature prediction. Secondly, the mesoscale temperature distribution and the maximum temperature in the microscale of VHTR fuel pebbled are calculated with QSSHC method based on OpenFOAM. This multiscale solver will be couple with other solvers of OpenFOAM, to provide a new perspective of VHTR simulation.

A Very High Temperature Reactor (VHTR) is one of the next generation nuclear reactor systems. From a view point of safety characteristics, a passive cooling system should be designed as the best way of a reactor vessel cooling system (VCS) in the VHTR. Therefore, the gas cooling system with natural circulation is considered as a candidate for the VCS of the VHTR. Japan Atomic Energy Agency (JAEA) is advancing the technology development of the VHTR and is now pursuing design and development of commercial systems such as the 300MWe gas turbine high temperature reactor GTHTR300C (Gas Turbine High Temperature Reactor 300 for Cogeneration). In the VCS of the GTHTR300C, many rectangular flow channels are formed around the reactor pressure vessel (RPV), and a cooling panel utilizing natural convection of air has been proposed. In order to apply the proposed panel to the VCS of the GTHTR300C, it is necessary to clarify the heat transfer and flow characteristics of the proposed channel in the cooling panel. Thus, we carried out an experiment to investigate heat transfer and fluid flow characteristics by natural convection in a vertical rectangular channel heated on one side. Experiments were also carried out to investigate the heat transfer and fluid flow characteristics by natural convection when a porous material with high porosity is inserted into the channel. An experimental apparatus is a vertical rectangular flow channel with a square cross section in which one surface is heated by a rubber heater. Dimensions of the experimental apparatus is 600 mm in height and 50 mm on one side of the square cross section. Air was used as a working fluid and fine copper wire (diameter: 0.5 mm) was used as a porous material. The temperature of the wall surface and gas in the channel were measured by K type thermocouples. The flow velocity distribution was obtained by a PIV method. In this paper, we discuss the heat transfer and fluid flow characteristics of the proposed channel. From the results obtained in the experiment, it was found that the amount of removed heat decreased with increasing of temperature of gas when a copper wire was inserted into the channel with high porosity. This is because the mass flow rate decreased with increasing of viscosity of gas. Since it is expected that the porosity of a porous material will have an optimum value, further studies will be needed.

As part of the Next Generation Nuclear Plant (NGNP) project, an evaluation of a low-temperature heat-pump interface design for a nuclear-driven high-temperature electrolysis (HTE) hydrogen production plant was performed using the UniSim process analysis software. The low-temperature interface design is intended to reduce the interface temperature between the reactor power conversion system and the hydrogen production plant by extracting process heat from the low temperature portion of the power cycle rather than from the high-temperature portion of the cycle as is done with the current Idaho National Laboratory (INL) reference design. The intent of this design change is to mitigate the potential for tritium migration from the reactor core to the hydrogen plant, and reduce the potential for high temperature creep in the interface structures. The UniSim model assumed a 600 MWt Very-High Temperature Reactor (VHTR) operating at a primary system pressure of 7.0 MPa and a reactor outlet temperature of 900°C. The low-temperature heat-pump loop is a water/steam loop that operates between 2.6 MPa and 5.0 MPa. The HTE hydrogen production loop operated at 5 MPa, with plant conditions optimized to maximize plant performance (i.e., 800°C electrolysis operating temperature, area specific resistance (ASR) = 0.4 ohm-cm2, and a current density of 0.25 amps/cm2). An air sweep gas system was used to remove oxygen from the anode side of the electrolyzer. Heat was also recovered from the hydrogen and oxygen product streams to maximize hydrogen production efficiencies. The results of the UniSim analysis showed that the low-temperature interface design was an effective heat-pump concept, transferring 31.5 MWt from the low-temperature leg of the gas turbine power cycle to the HTE process boiler, while consuming 16.0 MWe of compressor power. However, when this concept was compared with the current INL reference direct Brayton cycle design and with a modification of the reference design to simulate an indirect Brayton cycle (both with heat extracted from the high-temperature portion of the power cycle), the latter two concepts had higher overall hydrogen production rates and efficiencies compared to the low-temperature heat-pump concept, but at the expense of higher interface temperatures. Therefore, the ultimate decision on the viability of the low-temperature heat-pump concept involves a tradeoff between the benefits of a lower-temperature interface between the power conversion system and the hydrogen production plant, and the reduced hydrogen production efficiency of the low-temperature heat-pump concept compared to concepts using high-temperature process heat.

This report provides an overview of the mechanical, structural, and neutronic aspects of the Advanced High Temperature Reactor (AHTR) design concept. The AHTR is a design concept for a large output Fluoride salt cooled High-temperature Reactor (FHR) that is being developed to enable evaluation of the technology hurdles remaining to be overcome prior to FHRs becoming an option for commercial reactor deployment. This report documents the incremental AHTR design maturation performed over the past year and is focused on advancing the design concept to a level of a functional, self-consistent system. The reactor concept development remains at a preconceptual level of maturity. While the overall appearance of an AHTR design is anticipated to be similar to the current concept, optimized dimensions will differ from those presented here. The AHTR employs plate type coated particle fuel assemblies with rapid, off-line refueling. Neutronic analysis of the core has confirmed the viability of a 6-month two-batch cycle with 9 wt. % enriched uranium fuel. Refueling is intended to be performed automatically under visual guidance using dedicated robotic manipulators. The report includes a preconceptual design of the manipulators, the fuel transfer system, and the used fuel storage system. The present design intent is for used fuel to be stored inside of containment for at least six months and then transferred to local dry wells for intermediate term, on-site storage. The mechanical and structural concept development effort has included an emphasis on transportation and constructability to minimize construction costs and schedule. The design intent is that all components be factory fabricated into rail transportable modules that are assembled into subsystems at an on-site workshop prior to being lifted into position using a heavy-lift crane in an open-top style construction. While detailed accident identification and response sequence analysis has yet to be performed, the design concept incorporates fully passive responses to all identified design basis or non-very-low frequency beyond design basis accidents as well as multiple levels of radioactive material containment. Key building design elements include (1) below grade siting to minimize vulnerability to aircraft impact, (2) multiple natural circulation decay heat rejection chimneys, (3) seismic base isolation, and (4) decay heat powered back-up electricity generation.

The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this pers-pective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). We obtain the temperature profile distribution in the core for regimes where the coolant flow rate is smaller than recommended in a normal operation. In general, the temperature distributions calculated are consistent with phenomenological behavior. Even without considering the reactivity changes to reduce the reactor power or other safety mechanisms, the maximum temperatures do not exceed the recommended limits for TRISO fuel elements.

Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

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.

To start up a fusion reactor, it is necessary to provide a sufficient amount of tritium from an external device. The fusion DEMO reactor is planned to start up in the 2030s. Herein, methods for supplying the reactor with tritium are discussed. For the initial startup of the fusion reactor, use of a high temperature gas-cooled reactor (HTGR) as a tritium production device has been proposed. So far, the analyses have been focused only on the operation in which fuel is exchanged at stated periods (batch) using the block type HTGR. In this paper, to improve the performance of tritium production, properties of the HTGR are studied from the viewpoint of continuous operation for several conditions. In continuous operation, for example, in the pebble bed type HTGR, it is possible to design an operation that has no time loss for refueling. The pebble bed modular reactor (PBMR) and the gas turbine high temperature reactor of 300 MWe nominal capacity (GTHTR300) are assumed as the calculation and comparison targets, and simulation is made using the continuous-energy Monte Carlo transport code MVP-BURN. It is shown that the continuous operation using the pebble bed type HTGR has almost the same tritium productivity compared with the batch operation using the block type HGTR. The issues for pebble bed type HTGR as a tritium production device are discussed.

A novel approach to fabricating fuel compacts for the next generation nuclear plant (NGNP)