High Temperature Gas-cooled Reactor System Research Articles

Adaptive active fault-tolerant dynamic surface load following controller for a modular high-temperature gas-cooled reactor

The control of nuclear reactors is one of the most challenging requirements in performing the load following operation mode; however, most of the previous studies either do not consider the adverse effects of the lumped uncertainty or rely on the assumption of possessing healthy actuators without any faults. Besides, the recently published backstepping load following controllers may result in the problem of “explosion of complexity” due to repeatedly differentiating the virtual control laws. To this end, this paper provides an adaptive active fault-tolerant dynamic surface control strategy for the load following problem of a modular high-temperature gas-cooled reactor suffering from lumped uncertainty and actuator faults. The main contributions of this paper are threefold. (i) a radial basis function neural network is introduced to real-timely learn the lumped uncertainty that contains model uncertainties, parameter variations, external disturbances, and unmeasured states; (ii) a dynamic surface control technique is employed to cope with the problem of “explosion of complexity” induced by traditional backstepping approach; and (iii) an active fault-tolerant controller, incorporated with a fault detection observer and a fault estimator, is developed such that the reliability and safety of the modular high-temperature gas-cooled reactor system could be enhanced. With Lyapunov stability theory, both the stability of the closed-loop system and the error convergence are proved. At the end, comparative simulation studies and analysis using quantitative performance indices indicate the proposed control strategy’s superiority in dealing with lumped uncertainty and actuator faults.

Present status of JAEA’s R&D toward HTGR deployment

Japan Atomic Energy Agency (JAEA) has been promoting the research and development (R&D) of High Temperature Gas-cooled Reactor (HTGR). HTGR having superior safety features can supply high temperature heat above 900 °C and it makes possible to use nuclear energy to heat source for various industries as well as to power generation. Promotion of R&D on HTGR and hydrogen production system is stated in Japanese policies, e,g, The 6th Strategic Energy Plan and The Green Growth Strategy Through Achieving Carbon Neutrality in 2050. R&D on reactor technologies is carried out by using High Temperature engineering Test Reactor (HTTR) which is the first HTGR in Japan. On July 30th 2021, the HTTR was resumed without significant reinforcements. On January 28th 2022, a safety demonstration test under the OECD/NEA LOFC project was carried out. JAEA is promoting R&D on a carbon-free hydrogen production by thermochemical water splitting Iodine-Sulfur process (IS process). Hydrogen production test for 150 h with 0.03 m3/h-H2 was successfully performed in January 2019. JAEA conducts design study for various HTGR systems toward commercialization. The major one is GTHTR300C for hydrogen and electric power cogeneration. JAEA has launched a new test program about demonstration of hydrogen production by the HTTR. Steam methane reforming hydrogen production system was selected for the first demonstration by 2030. The other carbon-free hydrogen production method will be demonstrated, later.

Water-jacket reactor cavity cooling system concept to mitigate severe accident consequence of high temperature gas-cooled reactor

Abstract The High Temperature Gas-Cooled Reactor (HTGR), as a Gen-IV reactor, adopts a Reactor Cavity Cooling System (RCCS) to remove core decay heat after a reactor trip. During a postulated accident such as a pipe break in the reactor coolant system, the core decay heat can be properly removed by a RCCS to ensure the integrity of the fuel, reactor pressure vessel (RPV), and concrete silo. In this study we are more interested in the impact of a failure beyond that of the design basis of the RCCS functions, particularly the case in which natural convection of external air is not established. Such a case could happen when a chimney is collapsed by an external event. Against a severe disaster in which air convection is completely lost, a mixed air-water RCCS concept having a water-jacket system is suggested in order to mitigate the significant overheating of the RPV. In order to quantify the effectiveness of the water-jacket RCCS functions, the Depressurized Conduction Cooldown event (DCC) of a HTGR system with a thermal power of 350 MW is analyzed using the GAMMA+ system transient analysis code. As the RCCS air-cooling function is impaired, the water-jacket system increases its role to take part in the heat removal. From the analysis results of a DCC event with complete failure of the RCCS air-cooling, it is confirmed that, compared to the case without the water-jacket system, the maximum temperature of the reactor vessel can be reduced by about 400 °C thus ensuring its integrity. Therefore, the water-jacket system can be utilized as a backup system to mitigate the consequences of a severe accident such as the collapse of the chimney of the RCCS.

Energy economic evaluation of solar and nuclear driven steam methane reforming processes

Changing the high temperature heat supply of energy intensive industries from today’s mostly fossil sources to nuclear or renewable sources offers the opportunity of massive reductions of greenhouse gas emissions. Even though the high temperature gas-cooled reactor (HTGR) as well as the solar tower is technically capable to supply heat at the required high temperature level, it is uncertain whether these technologies can compete with the fossil energy supply. Therefore, the aim of this paper is the assessment of the economic competitiveness of process heat supply by HTGR or solar tower to fossil fired processes using steam methane reforming as industrial example process. Applying a self-developed optimization model, energy economic analyses are conducted for nuclear, solar and fossil energy supply systems. The results are benchmarked with the Hydrogen Economic Evaluation Program (HEEP) of the International Atomic Energy Agency (IAEA) and compared to literature values. The analyses show that hydrogen generation costs of the solar tower system are higher than those of the HTGR system. Both technologies benefit from the support of a fossil heater, as the hydrogen generation costs can be significantly reduced even with a small amount of burned natural gas. Nevertheless, a steam methane reforming process run by HTGR or by solar tower is not yet competitive to the natural gas fired hydrogen production.

A New Equivalence Theory Method for Treating Doubly Heterogeneous Fuel—I: Theory

A new methodology has been developed to treat resonance self-shielding in doubly heterogeneous very high temperature gas-cooled reactor systems in which the fuel compact region of a reactor lattice consists of small fuel grains dispersed in a graphite matrix. The new method first homogenizes the fuel grain and matrix materials using an analytically derived disadvantage factor from a two-region problem with equivalence theory and intermediate resonance method. The disadvantage factor accounts for spatial self-shielding effects inside each grain within the framework of an infinite array of grains. Then the homogenized fuel compact is self-shielded using a Bondarenko method to account for interactions between the fuel compact regions in the fuel lattice. In the final form of the equations for actual implementations, the double-heterogeneity effects are accounted for by simply using a modified definition of a background cross section, which includes geometry parameters and cross sections for both the grain and fuel compact regions. With the new method, the doubly heterogeneous resonance self-shielding effect can be treated easily even with legacy codes programmed only for a singly heterogeneous system by simple modifications in the background cross section for resonance integral interpolations. This paper presents a detailed derivation of the new method and a sensitivity study of double-heterogeneity parameters introduced during the derivation. The implementation of the method and verification results for various test cases are presented in the companion paper.

Preliminary studies of compact Brayton cycle performance for Small Modular High Temperature Gas-cooled Reactor system

A High Temperature Gas-cooled Reactor (HTGR) system was generally designed and planned to be constructed in a medium size reactor (around 600MWe power output) in the past. In this paper, authors are exploring the potential of very small modular type HTGR with a special attention given to the power conversion system. Since HTGR is relatively less challenging to achieve passive safety while easy to adopt an air-cooled Brayton cycle for a power conversion system, HTGR can be suitable for a small modular reactor application. However, as the size of the power system decreases the consisting component performance usually degrades and the final effect on the total system performance was not seriously studied before for a small size HTGR system. 20MWth reactor was chosen for this study to investigate how helium Brayton cycle performance can be affected by the component level performance and how the optimal operating condition shifts when the system size reduces. The discussion of component design that can deliver assumed performance for the cycle estimation will be briefly presented as well. Furthermore, a supercritical CO2 cycle option will be compared to the helium Brayton cycle to show that the S-CO2 cycle can be a good alternative for a very small scale system.

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.

Establishment of control technology of the HTTR and future test plan

The high-temperature engineering test reactor (HTTR) has carried out several demonstration tests to obtain basic data for the design of a future high-temperature gas-cooled reactor (HTGR). In addition, for commercial HTGRs, it also should be demonstrated that the HTGR system can supply stable heat to heat utilization system for the long term. To demonstrate the capability of HTGRs of stable heat supply, the operational experience of the HTTR would be useful. The control characteristics of the HTTR are evaluated by the result of the long-term high-temperature operation performed in 2010. As the result, it is confirmed that control technology of the HTTR is enough to achieve the control response with reasonable damped characteristics and the stable normal operation. In addition, a future possibility of the thermal-load fluctuation and lost test is examined. The data would contribute design of heat utilization system because heat utilization system is required to be a non-safety system. For the future HTGR system design, the requirement, method and result of determination of control constants are discussed. The behavior of control system is described with the characteristics of the HTGR during the past test using HTTR. In addition, this paper proposes a future test plan for demonstration of the reactor stability and control technology against thermal load fluctuation for future HTGR with a heat utilization system design.

A Small-Sized HTGR System Design for Multiple Heat Applications for Developing Countries

Japan Atomic Energy Agency has conducted a conceptual design of a 50 MWt small-sized high temperature gas cooled reactor (HTGR) for multiple heat applications, named HTR50S, with the reactor outlet coolant temperature of 750°C and 900°C. It is first-of-a-kind of the commercial plant or a demonstration plant of a small-sized HTGR system to be deployed in developing countries in the 2020s. The design concept of HTR50S is to satisfy the user requirements for multipurpose heat applications such as the district heating and process heat supply based on the steam turbine system and the demonstration of the power generation by helium gas turbine and the hydrogen production using the water splitting iodine-sulfur process, to upgrade its performance compared to that of HTTR without significant R&D utilizing the knowledge obtained by the HTTR design and operation, and to fulfill the high level of safety by utilizing the inherent features of HTGR and a passive decay heat removal system. The evaluation of technical feasibility shows that all design targets were satisfied by the design of each system and the preliminary safety analysis. This paper describes the conceptual design and the preliminary safety analysis of HTR50S.

Solution of multiple circuits of steam cycle HTR system

In order to analyze the dynamic operation performance and safety characteristics of the steam cycle high temperature gas cooled reactor (HTR) systems, it is necessary to find the solution of the whole HTR systems with all coupled circuits, including the primary circuit, the secondary circuit, and the residual heat removal system (RHRS). Considering that those circuits have their own individual fluidity and characteristics, some existing code packages for independent circuits themselves have been developed, for example THEMRIX and TINTE code for the primary circuit of the pebble bed reactor, BLAST for once through steam generator. To solve the coupled steam cycle HTR systems, a feasible way is to develop coupling method to integrate these independent code packages. This paper presents several coupling methods, e.g. the equivalent component method between the primary circuit and steam generator which reflect the close coupling relationship, the overlapping domain decomposition method between the primary circuit and the passive RHRS which reflects the loose coupling relationship. Through this way, the whole steam cycle HTR system with multiple circuits can be easily and efficiently solved by integration of several existing code packages. Based on this methodology, a code package TINTE–BLAST–RHRS was developed. Using this code package, some operation performance of HTR–PM was analyzed, such as the start-up process of the plant, and the depressurized loss of forced cooling accident when different number of residual heat removal trains is operated.

The Effect Assessment for Fast Reactor Fuel Cycle Deployment

After the Fukushima Dai-ichi Nuclear Power Station accident, the future role and perspectives of the nuclear energy and nuclear fuel cycle is being called into question in Japan. Regarding fast reactor (FR) and related fuel cycle system, the significance of the Research, Development, Demonstration, and Deployment (RDD&D) of them is also being argued. In this paper the socio-economic effect of the FR cycle deployment was assessed by a combination system of two energy economy models, which are a recursive dynamic Computable General Equilibrium (CGE) model based on GTAP-E, i.e. an energy environmental version of the “Global Trade and Analysis Project (GTAP)” model, and a dynamic optimization type energy system model called “Dynamic New Earth 21 (DNE21)” model to clarify the significance of the RDD&D.Some energy systems, mainly renewable energy likes land/offshore wind, residential/mega solar and battery for power system stabilization as well as innovative nuclear systems under development (i.e. high temperature gas-cooled reactor systems, and fusion reactor) were added newly to the latest version of energy economic model for this assessment. Besides that, it is possible to treat mass balance more precisely in comparison to the previous version. Future population and Gross Domestic Product (GDP) base growth rate, energy demand/consumption amount, and carbon footprint were important data to describe future global economic system with the recursive dynamic GTAP-E model. Moreover, potentials to recover resources, innovative technologies, etc. as well as capital costs for power sources, and other energy characteristics were considered to optimize energy supply profiles from the viewpoint of long-term energy system cost minimization through the DNE21 model.The results showed FRs with high economic potential are deployed on a massive scale in Asia and other countries. In addition to that, the combined analyses of both energy economic models show the promising economic effect by the plentiful inexpensive electricity supply from FRs. In particular, the GDP growth addition in Japan compensates expected RDD&D costs after the several decades from FR deployment. Besides calculating GDP after the commercial plants deployment, the socio-economic effects from RDD&D activities were also estimated. The RDD&D of FR is still cost-effective even under the present situation as long as nuclear energy is used continuously.

Advanced High Temperature Gas-Cooled Reactor Systems

Three systems have been proposed for advanced high-temperature gas-cooled reactors: a supercritical carbon dioxide (S-CO2) gas turbine power conversion system, a new microchannel heat exchanger (MCHE), and a once-through-then-out (OTTO) refueling scheme with burnable poison (BP) loading. A S-CO2 gas turbine cycle attains higher cycle efficiency than a He gas turbine cycle because of reduced compression work around the critical point of CO2. Considering temperature reduction at the turbine inlet by 30°C through intermediate heat exchange, the S-CO2 indirect cycle achieves an efficiency of 53.8% at a turbine inlet temperature of 820°C and a turbine inlet pressure of 20 MPa. This cycle efficiency value is higher by 4.5% than that (49.3%) of a He direct cycle at a turbine inlet temperature of 850°C and 7 MPa. A new MCHE has been proposed as an intermediate heat exchanger between the primary cooling He loop and the secondary S-CO2 gas turbine power conversion system and as recuperators of the S-CO2 gas turbine power conversion system. This MCHE has discontinuous “S-shaped” fins providing flow channels resembling sine curves. Its pressure drop is one-sixth that of a conventional MCHE with a zigzag flow channel configuration, but it has the same high heat transfer performance. The pressure drop reduction is ascribed to suppression of recirculation flows and eddies that appear around bend corners of the zigzag flow channels in the conventional MCHE. An optimal BP loading in an OTTO refueling scheme eliminates the shortcoming of its excessively high axial power peaking factor, reducing the power peaking factor from 4.44 to about 1.7, and inheriting advantages over the multipass scheme because it obviates reloading in addition to fuel handling and integrity checking systems. Because of the power peaking factor reduction, the maximum fuel temperatures are lower than the maximum permissible values of 1250°C for normal operation and 1600°C during a depressurization accident.

Neutronic double heterogeneity effect in particle dispersed type inert matrix fuels

Rock-like oxide (ROX) fuel concept is studied in Japan for effective plutonium burning in light water reactors (LWRs). ROX is a heterogeneous fuel, where Pu containing yttria stabilized zirconia (YSZ) particles are dispersed in spinel matrix, and similar to the high temperature gas cooled reactor (HTR) fuel. The effect of such a ‘double’ heterogeneity (fuel, structure and coolant heterogeneity in reactor core, plus fuel heterogeneity) on HTR neutronic characteristics is important, while the effect was not taken into account in the ROX fueled LWR neutronics calculations. Here, this double heterogeneity effect is estimated for ROX fueled LWR, and compared with the Pu containing YSZ particle fueled HTR. As a result, the heterogeneity effect was negligible in the ROX–LWR system, while it is notable in YSZ–HTR system. The volume fraction of YSZ particle in the fuel region is one of the important parameter to cause the difference.

HTGR reactor physics and fuel cycle studies

The high-temperature gas-cooled reactor (HTGR) appears as a good candidate for the next generation of nuclear power plants. In the “HTR-N” project of the European Union Fifth Framework Program, analyses have been performed on a number of conceptual HTGR designs, derived from reference pebble-bed and hexagonal block-type HTGR types. It is shown that several HTGR concepts are quite promising as systems for the incineration of plutonium and possibly minor actinides. These studies were mainly concerned with the investigation and intercomparison of the plutonium and actinide burning capabilities of a number of HTGR concepts and associated fuel cycles, with emphasis on the use of civil plutonium from spent LWR uranium fuel (first generation Pu) and from spent LWR MOX fuel (second generation Pu). Besides, the “HTR-N” project also included activities concerning the validation of computational tools and the qualification of models. Indeed, it is essential that validated analytical tools are available in the European nuclear community to perform conceptual design studies, industrial calculations (reload calculations and the associated core follow), safety analyses for licensing, etc., for new fuel cycles aiming at plutonium and minor actinide (MA) incineration/transmutation without multi-reprocessing of the discharged fuel. These validation and qualification activities have been centred round the two HTGR systems currently in operation, viz. the HTR-10 and the HTTR. The re-calculation of the HTTR first criticality with a Monte Carlo neutron transport code now yields acceptable correspondence with experimental data. Also calculations by 3D diffusion theory codes yield acceptable results. Special attention, however, has to be given to the modelling of neutron streaming effects. For the HTR-10 the analyses focused on first criticality, temperature coefficients and control rod worth. Also in these studies a good correspondence between calculation and experiment is observed for the 3D diffusion theory codes.

高温ガス炉ガスタ-ビン発電システム動特性解析モデルConan-GTHTRの開発,(I)

The Gas Turbine High Temperature Reactor 300 (GTHTR300) composed of an inherent safe 600MWt reactor and a closed gas turbine power conversion system is a high efficient and economically competitive HTGR to be deployed in 2010s. To analyze the plant dynamics and the thermal hydraulics of the GTHTR300, a new analytical code (Conan-GTHTR) based on 'RELAP5/MOD3' has been developed and applied to heat transfer calculations of the High Temperature Engineering Test Reactor (HTTR) for its verification. The results proved that the new code was available for transient simulations in High Temperature Gas-cooled Reactor systems.

The Development and Application of a Unified Corrosion Model for High-Temperature Gas-Cooled Reactor Systems

The main processes of metallic corrosion in primary circuits of advanced high-temperature gas-cooled reactors (HTGRs) at temperatures above 800/sup 0/C are oxidation, carburization, and decarburization. These are caused by helium impurity traces of H/sub 2/O (causing oxidation and decarburization), CO (causing oxidation and carburization), and CH/sub 4/ (causing carburization). At the very low partial pressures of these impurities, the three processes happen independently, leading to a multitude of corrosion effects, which can be classified in terms of active and passive regimes. In an active regime internal corrosion proceeds rapidlyusually linear with exposure time-thereby severely affecting the structural integrity of the alloy. Passive regimes are characterized by stable oxide layers, which either completely inhibit internal corrosion or limit it to a parabolic dependence with exposure time. These passive and active regimes can be related to absolute partial pressures and partial pressure ratios of the main gaseous impurities, H/sub 2/O, CO, and CH/sub 4/.

Experimental Work on Alloy Development for High-Temperature Gas-Cooled Reactor Systems

In the context of developing high-temperature wrought alloys especially designed for advanced high-temperature gas-cooled reactor systems at temperatures up to 1000°C, a program of investigations has been performed during the past few years. The progress of the work can be traced through several stages: the addition of single minor alloying elements to model Fe-Cr-Ni alloysthe study of multiple minor alloy additions in commercial alloysthe additions of preferred minor alloy combinations to modified commercial alloysthe variation of chromium in experimental nickel-base alloys containing titanium and aluminumthe variation of iron, molybdenum, tungsten, and cobalt in nickel-base alloys containing low chromium, titanium, and aluminum. The first three stages were directed toward the study and improvement of the corrosion resistance of existing alloys, but in the last two stages, radical changes in composition have been investigated. The work so far has culminated in a family of alloys that have excellent microstructural stability and corrosion resistance combined with strengths comparable with those of traditional alloys.

Dual-Purpose Nuclear Power Plants for Military Installations

Preliminary studies conducted by the Facilities Engineering Support Agency indicate that the electrical power and recoverable waste heat from a 50- to 100-MW nuclear energy center could supply the future power and space conditioning requirements of a large military installation. The plant design under study is powered by a high-temperature gas-cooled reactor (HTGR) providing electrical energy and precooler effluent of sufficiently high temperature (200/sup 0/C) for use in a pressurized water district heating network. The military installation was found to be an attractive candidate for utilization of waste heat and electrical power from a central plant because of its size, diversity of energy demand, and operational character. The HTGR system was shown to have an economic advantage over a comparable system using a pressurized water reactor. It is concluded that the nuclear total energy system is technically feasible and capable of serving the utility needs of military installations in the late 1980's.

Experience with the Fort St. Vrain reactor

Public Service Company of Colorado began electric power generation in late 1976 at its Fort St. Vrain Nuclear Generating Station located near Platteville, Colorado, 35 miles north of Denver. The plant utilizes the advanced high temperature gas cooled reactor (HTGR) concept. General Atomic Company designed and constructed the reactor using Sargent & Lundy as architect-engineer and EBASCO as general construction contractor. The research and development associated with this endeavor was undertaken under the former Atomic Energy Commission Power Reactor Demonstration Program. Preliminary low power and intermediate power tests have proved the design concept features and calculations and most of the components. Safe conditions following loss of coolant were confirmed. Release of radioactivity from the fuel was less than predicted up to 70% full power. Modifications and repairs caused delays and included, chiefly: Pelton wheel emergency cooling drives, helium circulator bearings, the pressure-regulating system for water lubricant in circulator bearings, liner cooling lines, removal of moisture from the vessel, electrical cable rerouting, helium seals around control rod drives, and a leak in one steam generator. Core temperature oscillation problems are being analyzed before passing the currently imposed 70% power level. Fort St. Vrain has demonstrated that the concept of a large HTGR system incorporating a prestressed concrete reactor vessel is sound. The station has produced more than 500 million kilowatt hours of electric energy to date.

An eigenvalue sensitivity study of a high temperature gas-cooled reactor system

This correspondence discusses the application of an eigenvalue sensitivity method [1] to a linear, time-invariant model of the high-temperature gas-cooled reactor (HTGR) manufactured by General Atomic. It is shown that for this type of system with distinct eigenvalues, a relation can be obtained between the solutions to the classical sensitivity. equation and the eigenvalue and eigenvector sensitivities. This relation indicates that the eigenvalue and eigenvector sensitivities are more fundamental indicators of parametric effects than the solutions of the sensitivity. equation. A Fortran program (SENSIT) was developed to implement the algorithms required for eigenvalue sensitivity calculations. This program was applied to a large scale (58 state) linear, time invariant model of the HTGR operating under closed-loop control.