Gas Turbine High Temperature Reactor Research Articles

Loading method of Li rods for tritium production using High-Temperature Gas-Cooled reactor for fusion reactors

An initial tritium inventory is required to start fusion DEMOnstration Power Station (DEMO) reactors. However, a method to supply adequate tritium has not been determined yet. Tritium production via 6Li(n,α)T reaction by loading Li rods into the burnable poison (BP) holes of a high-temperature gas-cooled reactor (HTGR) has been proposed to address this problem (Matsuura et al., Nucl. Eng. Des. 243 (2012) 95 - 101). In previous preliminary studies, Li rods loaded in all BP holes were assumed to have the same design. This study evaluated whether the performance of Li rods can be improved for future optimization by adjusting the Li rod arrangement and the amount of Li compounds in them. The amount of tritium produced for gas turbine high-temperature reactor 300 (GTHTR300) was evaluated, while the total amount of Li compounds was maintained and the amount of loaded Li compounds changed depending on the layers and fuel regions. The maximum amount of tritium produced did not increase during the evaluations when reactor feasibility was satisfied. This implies that it is possible to reduce the number of Li rods while maintaining the amount of tritium produced for optimization, thereby reducing the costs of manufacturing Li rods and tritium recovery. GTHTR300 can produce 800 g of tritium in 360 days of operation using 2160Li rods. The results showed that the same amount of tritium could be produced by loading 720Li rods with the same number of fuel blocks. In addition, the effective multiplication factor, burn up, and power density of GTHTR300 were not significantly influenced during the operation.

Graded Approach Establishment for the HTGR Maintenance Activities Using Modified Fuzzy FMEA & Expert Judgement Methodology

Grading is an important step of the Nuclear Power Plant (NPP) operation & maintenance activities. However, there are several grading difficulties for the High Temperature Gas Reactor (HTGR) as well as the other type of NPPs causing by the lack of operational experiences and availability of the reliability data. Failure Mode & Effects Analysis (FMEA) is one of the mature techniques that are commonly used to solve such kinds of difficulties. Nevertheless, traditional FMEA has several issues and possibly become an obstacle in the grading process. The modified FMEA by utilizing expert judgment elicitation techniques combined with the fuzzy logic theory is proposed to solve those issues. As a study practice, the proposed methodology is applied by examining Japanese’s HTGR, Gas Turbine High Temperature Reactor 300 for Cogeneration (GTHTR300C) design carefully. This study establishing good practice especially for the future advanced NPP maintenance activities development.

Effect of nuclear heat caused by the 6Li(n,α)T reaction on tritium containment performance of tritium production module in High-Temperature Gas-Cooled reactor for fusion reactors

Tritium is required for research and development activities for the deuterium–tritium (DT) fusion reactor and fueling the DEMOnstration Power Station (DEMO). However, tritium is a very rare nuclide and must be produced artificially. Tritium production by loading Li compounds (Li rods) into burnable poison holes of a high-temperature gas-cooled reactor (HTGR) has been proposed (H. Matsuura, et al., Nucl. Eng. Des. 243 (2012) 95–101.). Al2O3 and Zr are used to prevent tritium leaks. Nuclear reaction heat caused by the nuclear reaction (e.g., 6Li(n,α)T reaction) can cause a spatial temperature profile in the Li rods and may change its tritium containment performance, because Al2O3 and Zr performance strongly depend on these temperatures. The effect of nuclear reaction heat by the 6Li(n,α)T reaction on the tritium containment performance of the Li rods was evaluated by simulation. The temperatures of the Li rods for the high-temperature engineering test reactor (HTTR) and gas turbine high-temperature reactor 300 (GTHTR300) increased by 36 K and 46 K, and the leaked tritium decreased by 32% and 37% via nuclear reaction heat, respectively.

The T-containment properties of a Zr-containing Li rod in a high-temperature gas-cooled reactor as a T production device for fusion reactors

The production of tritium (T) using high-temperature gas-cooled reactors (HTGRs) has been studied for a prior engineering research with T handling and initial T possession in demonstration fusion reactors. Stable containment of T in Li-loading rods during HTGR operation is a critical issue. This study investigates this for an irradiation test to examine T-containment performance in Li-loading rods and develops an analytical model of evaluating the amount of T outflow to a He coolant. The hydrogen absorption characteristics, including the deterioration of the hydrogen absorption speed after Zr has sufficiently absorbed the hydrogen, is experimentally measured assuming an HTGR setting. We present an analytical model of evaluating the T outflow from a Li rod and, on the basis of this model, estimate the total T outflow, assuming the presence of a gas-turbine high-temperature reactor of 300 MWe with a nominal capacity and a high-temperature engineering test reactor. It is demonstrated that, by loading a sufficient amount of Zr into the Li rod, the T outflow can be suppressed to less than a small percent of the total T produced during 360 days of reactor operation.

Conceptual design study of a high performance commercial HTGR for early introduction

Conceptual design study of a commercial High Temperature Gas-cooled Reactor (HTGR) for early introduction has been performed based on the cumulated experience in design, construction, and operation of the High Temperature engineering Test Reactor (HTTR) and design of the commercial Gas Turbine High Temperature Reactor 300 (GTHTR300). The power output is 165MWt and the inlet and outlet coolant temperatures are 325 °C and 750 °C, respectively, to provide steam for industrial utilization. However, given a requirement for the reactor pressure vessel to be smaller even that of the 30 MWt HTTR, several challenging technical problems have to be dealt with to arrive in a high performance core design that provides extended fuel burnup, prolonged refueling period, improved fuel refueling scheme, improved fuel element and so on from the HTTR.

Process of Air Ingress during a Depressurization Accident of GTHTR300

A depressurization accident is the design-basis accidents of a gas turbine high temperature reactor, GTHTR300, which is JAEA’s design and one of the Very-High-Temperature Reactors (VHTR). When a primary pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. Therefore, it is important to know a mixing process of different kinds of gases in the stable and unstable density stratified fluid layer. In order to predict or analyze the air ingress phenomena during the depressurization accident, we have conducted an experiment to obtain the mixing process of two component gases and the characteristics of natural circulation. The experimental apparatus consists of a storage tank and a reverse U-shaped vertical rectangular passage. One side wall of the high temperature side vertical passage is heated and the other side wall is cooled. The other experimental apparatus consists of a cylindrical double coaxial vessel and a horizontal double coaxial pipe. The outside of the double coaxial vessel is cooled and the inside is heated. The results obtained in this study are as follows. When the primary pipe is connected at the bottom of the reactor pressure vessel, onset time of natural circulation of air is affected by not only molecular diffusion but also localized natural convection. When the wall temperature difference is large, onset time of natural circulation of air is strongly affected by natural convection rather than molecular diffusion. When the primary pipe is connected at the side of the reactor pressure vessel, air will enter the bottom space in the reactor pressure vessel by counter-current flow at the coaxial double pipe break part immediately. Afterward, air will enter the reactor core by localized natural convection and molecular diffusion.

Uranium-based TRU multi-recycling with thermal neutron HTGR to reduce environmental burden and threat of nuclear proliferation

ABSTRACT To reduce environmental burden and threat of nuclear proliferation, multi-recycling fuel cycle with high temperature gas-cooled reactor has been investigated. Those problems are solved by incinerating trans-uranium (TRU) nuclides, which is composed of plutonium and minor actinoid, and there is concept to realize TRU incineration by multi-recycling with fast breeder reactor. In this study, multi-recycling is realized even with a thermal reactor by feeding fissile uranium from outside of the fuel cycle instead of breeding fissile nuclide. In this fuel cycle, recovered uranium and natural uranium are enriched and mixed with recovered TRU to fabricate fresh fuels. The fuel cycle was designed for a gas turbine high temperature reactor (GTHTR300). Reprocessing is assumed as existing reprocessing with four-group partitioning technology. As a result, the TRU nuclides excluding neptunium can be recycled by the proposed cycle. The duration of potential toxicity decaying to natural uranium level can be reduced to approximately 300 years, and the footprint of repository for high-level waste can be reduced by 99.7% compared with the standard case. Surplus plutonium is not generated by this cycle. Moreover, incineration of TRU from light water reactor cycle can be performed in this cycle.

Conceptual design of the iodine–sulfur process flowsheet with more than 50% thermal efficiency for hydrogen production

A conceptual design of a practical large-scale plant of the thermochemical water splitting the iodine–sulfur (IS) process flowsheet was carried out as a heat application of Japan Atomic Energy Agency’s commercial Gas Turbine High Temperature Reactor 300 MW for Cogeneration (GTHTR300C) plant design. Innovative techniques proposed by JAEA were applied for improvement of hydrogen production thermal efficiency; depressurized flash concentration of H2SO4 using waste heat from Bunsen reaction, prevention of H2SO4 vaporization from a H2SO4 distillation column by introduction of H2SO4 solution from the 2nd flash bottom, and I2 condensation heat recovery by direct contact heat exchange in an HI distillation column. A simulation of material and heat balance was made using PRO/II, a commercial chemical process simulator. The result demonstrated that hydrogen of about 31,900 Nm3/h would be produced by 170 MW heat from the GTHTR300C. A process thermal efficiency of 50.2% would be achievable with incorporation of the innovative techniques and the following high performance components expected in future R&D; an electro-electrodialysis cell stack, a reverse osmosis membrane, a HI decomposition reactor incorporated with a H2 permselective membrane, and heat exchangers.

Design of HTTR-GT/H2 test plant

The pre-licensing design of an HTGR cogeneration test plant to be coupled to JAEA’s existing test reactor HTTR is discussed. The plant is designed to demonstrate the system of JAEA commercial plant design of Gas Turbine High Temperature Reactor for Cogeneration (GTHTR300C). More specifically the test plant (HTTR-GT/H2) aims to 1) demonstrate licensability of the GTHTR300C system for electricity and hydrogen cogeneration and 2) confirm operation and safety performance of such cogeneration system. With construction planned to be completed around 2025, the test plant is expected to be the first-of-a-kind nuclear system operating on two of the advanced energy conversion systems attractive for the HTGR – closed cycle helium gas turbine for power generation and thermochemical iodine–sulfur water-splitting process for hydrogen production.

HTTR-GT/H2 test plant – System performance evaluation for HTTR gas turbine cogeneration plant

This paper presents the system performance evaluation for HTTR gas turbine cogeneration test plant (HTTR-GT/H2 plant) so as to confirm that the design meets the requirements with respect to the demonstration test objective. Start-up and shut down operation sequences as well as operability of load following operation were investigated. In addition, system dynamic and control analyses for the test plant in the events of loss of generator load and upset of H2 plant were performed. The simulation results presented in the paper show that the test plant is suitable for the test bed to validate control schemes against postulated transients in the commercial Gas Turbine High Temperature Reactor Cogeneration (GTHTR300C). The results also lead us to the conclusion that HTTR-GT/H2 plant can be used to test operational procedures unique to HTGR direct-cycle gas turbine cogeneration.

Reduction on high level radioactive waste volume and geological repository footprint with high burn-up and high thermal efficiency of HTGR

Reduction on volume of High Level radioactive Waste (HLW) and footprint in a geological repository due to high burn-up and high thermal efficiency of High Temperature Gas-cooled Reactor (HTGR) has been investigated. A helium-cooled and graphite-moderated commercial HTGR was designed as a Gas Turbine High Temperature Reactor (GTHTR300), and that has particular features such as significantly high burn-up of approximately 120GWd/t, high thermal efficiency around 50%, and pin-in-block type fuel. The pin-in-block type fuel was employed to reduce processed graphite volume in reprocessing. By applying the feature, effective waste loading method for direct disposal is proposed in this study. By taking into account these feature, the number of HLW canister generations and its repository footprint are evaluated by burn-up fuel composition, thermal calculation and criticality calculation in repository.As a result, it is found that the number of canisters and its repository footprint per electricity generation can be reduced by 60% compared with Light Water Reactor (LWR) representative case for direct disposal because of the higher burn-up, higher thermal efficiency, less TRU generation, and effective waste loading proposed in this study for HTGR. But, the reduced ratios change to 20% and 50% if the long term durability of LWR canister is guaranteed. For disposal with reprocessing, the number of canisters and its repository footprint per electricity generation can be reduced by 30% compared with LWR because of the 30% higher thermal efficiency of HTGR.

Investigation of absorption characteristics for thermal-load fluctuation using HTTR

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.

GTHTR300 cost reduction through design upgrade and cogeneration

Japan Atomic Energy Agency began design and development of the Gas Turbine High Temperature Reactor of 300MWe nominal output (GTHTR300) in 2001. The reactor baseline design completed three years later was based on 850°C core outlet temperature and a direct cycle gas turbine balance of plant. It attained 45.6% net power generation efficiency and 3.5US¢/kWh cost of electricity. The cost was estimated 20% lower than LWR. The latest design upgrade has incorporated several major technological advances made in the past ten years to both reactor and balance of plant. As described in this paper, these advances have enabled raising the design basis reactor core outlet temperature to 950°C and increasing power generating efficiency by nearly 5% point. Further implementation of seawater desalination cogeneration is made through employing a newly-proposed multi-stage flash process. Through efficient waste heat recovery of the reactor gas turbine power conversion cycle, a large cost credit is obtained against the conventionally produced water prices. Together, the design upgrade and the cogeneration are shown to reduce the GTHTR300 cost of electricity to under 2.7US¢/kWh.

A study on transmutation of LLFPs using various types of HTGRs

In order to investigate the potential of high temperature gas-cooled reactors (HTGRs) for transmutation of long-lived fission products (LLFPs), numerical simulation of four types of HTGRs were carried out. In addition to the gas-turbine high temperature reactor system “GTHTR300”, which is the subject of our previous research, a small modular HTGR plant “HTR50S” and two types of plutonium burner HTGRs “Clean Burn with MA” and “Clean Burn without MA” were considered. The simulation results show that an early realization of LLFP transmutation using a compact HTGR may be possible since the HTR50S can transmute fair amount of LLFPs for its thermal output. The Clean Burn with MA can transmute a limited amount of LLFPs. However, an efficient LLFP transmutation using the Clean Burn without MA seems to be convincing as it is able to achieve very high burn-ups and produce LLFP transmutation more than GTHTR300. Based on these results, we propose utilization of variety of HTGRs for LLFP transmutation and storage.

Study on Operation Scenario of Tritium Production for a Fusion Reactor Using a High Temperature Gas-Cooled Reactor

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.

Study on a method for loading a Li compound to produce tritium using high-temperature gas-cooled reactor

Tritium production using high-temperature gas-cooled reactors and its outflow from the region loading Li compound into the helium coolant are estimated when considering the suppression of tritium outflow. A Li rod containing a cylindrical Li compound placed in an Al2O3 cladding tube is assumed as a method for loading Li compound. A gas turbine high-temperature reactor of 300MW electrical nominal capacity (GTHTR300) with 600MW thermal output power is considered and modeled using the continuous-energy Monte Carlo transport code MVP-BURN, where burn-up simulations are carried out. Tritium outflow is estimated from equilibrium solution for the tritium diffusion equation in the cladding tube. A GTHTR300 can produce 400–600g of tritium over a 180-day operation using the chosen method of loading the Li compound while minimizing tritium outflow from the cladding tube. Optimizing tritium production while suppressing tritium outflow is discussed.

GTHTR300—A nuclear power plant design with 50% generating efficiency

GTHTR300 is a gas turbine high temperature reactor power generation plant design. The baseline design reported by Japan Atomic Energy Agency a decade ago attained 45.6% net efficiency. Technological improvements have since been made that make further increase in efficiency practical: first, the cycle parameters are upgraded by utilizing the newly acquired design data including those from component tests. Next, the core design is optimized to raise the reactor outlet coolant temperature from the baseline of 850°C to the level of 950°C demonstrated on the long-term test reactor operation. Both core physics and thermal hydraulics are investigated to demonstrate the corresponding temperature rise is within the design limit so that the existing fuel design can continue to apply. Finally, an advanced type of turbine blade material that has only recently entered in commercial service in aircraft engine is found to be useable for this design to realize a turbine inlet temperature of 950°C without requiring blade cooling.As detailed in this paper, these design improvements result in a nearly 5% gain in overall plant efficiency and enable the GTHTR300 to break the 50% efficiency barrier of nuclear plant while using only the existing technologies. This result is expected to contribute to the early market deployment of high temperature gas-cooled reactor.

Helium leak and chemical impurities control technology in HTTR

Japan Atomic Energy Agency (JAEA) has designed and developed high-temperature gas-cooled reactor (HTGR) hydrogen cogeneration system named gas turbine high-temperature reactor (GTHTR300C) as a commercial HTGR. Helium gas is used as the primary coolant in HTGR. Helium gas is easy to leak, and the primary helium leakage should be controlled tightly from the viewpoint of preventing the release of radioactive materials to the environment. Moreover from the viewpoint of preventing the oxidization of graphite and metallic material, the helium coolant chemistry should be controlled tightly.The primary helium leakage and the helium coolant chemistry during the operation is the major factor in the HTGR for commercialization of HTGR system.This paper shows the design concept and the obtained operational experience on the primary helium leakage control and primary helium impurity control in the high-temperature engineering test reactor (HTTR) of JAEA. Moreover, the future plan to obtain operational experience of these controls for commercialization of HTGR system is shown.

Core configuration of a gas-cooled reactor as a tritium production device for fusion reactor

The performance of a high-temperature gas-cooled reactor as a tritium production device is examined, assuming the compound LiAlO2 as the tritium-producing material. A gas turbine high-temperature reactor of 300MWe nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations are carried out. To load sufficient Li into the core, LiAlO2 is loaded into the removable reflectors that surround the ring-shaped fuel blocks in addition to the burnable poison insertion holes. It is shown that module high-temperature gas-cooled reactors with a total thermal output power of 3GW can produce almost 8kg of tritium in a year.

Evaluation of tritium production rate in a gas-cooled reactor with continuous tritium recovery system for fusion reactors

The performance of a high-temperature gas-cooled reactor as a tritium production with continuous tritium recovery system is examined. A gas turbine high-temperature reactor of 300-MWe (600MW) nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the three-dimensional entire-core region of the GTHTR300 were performed. A Li loading pattern for the continuous tritium recovery system in the gas-cooled reactor is presented. It is shown that module gas-cooled reactors with a total thermal output power of 3GW in all can produce ∼6kg of tritium maximum in a year.