Modular High Temperature Gas-cooled Reactor Research Articles

Passivity-based control of fluid flow networks with capacitance

Fluid flow networks (FFNs) widely exist in industrial energy systems tightly related with the industrial production and residents living, such as district heating networks (DHN), mine-ventilation systems and gas transportation systems. The FFN operation is to drive the flowrates and pressures to their desired values, and the flowrate-pressure coordinated control of FFN is crucial in providing safe, stable and efficient operation. Although fluid capacitance has the same importance as fluid resistance and fluid inductance in giving dynamical characteristics of FFN, fluid capacitance is still not fully considered in the flowrate and pressure control design of FFN. In this paper, a nonlinear dynamic model of general FFN with fluid resistance, inductance and capacitance is first given. By using this nonlinear model, it is shown that the general FFN dynamics is strictly passive, and the corresponding generalized Hamiltonian form (GHF) is also given. Based on the passivity of FFN, a decentralized adaptive flowrate-pressure coordinated control is newly proposed, taking a simple proportional-integral (PI) form while providing globally asymptotic regulation for network flowrates and pressures. This FFN control method is applied to design the flowrate-pressure controller of an auxiliary startup circuit (ASC) equipped to a modular high temperature gas-cooled reactor (mHTGR), and simulation results in the cases of heating-up and cooling-down show the feasibility of control design.

Equilibrium core modeling of a pebble bed reactor similar to the Xe-100 with SCALE

As the nuclear industry moves towards licensing and constructing advanced reactors, new attention has been focused on the advanced reactor designs that have past operational experience, such as pebble-bed high-temperature gas-cooled reactors (PB-HTGRs). Pebble-bed reactor designs have many advantages, such as their higher operating temperatures and online refueling capabilities. However, high-fidelity computational modeling of pebble-bed reactor designs, from reactor startup to operation at equilibrium, is more challenging compared to conventionally fueled reactors due to the continuous movement of the fuel pebbles through the reactor during operation. In previous work at Oak Ridge National Laboratory (ORNL), the SCALE Leap-In method for Cores at Equilibrium (SLICE) was developed around tools within the SCALE code system. This iterative method can effectively generate pebble-bed reactor zone-wise fuel inventories at equilibrium core operation within a reasonable computational time. The objective of this work was to further verify the ORNL SLICE method and to investigate the impact of considering temperature profiles during the application of the method. The SLICE method was applied to a modular high-temperature gas-cooled reactor design based upon publicly available design specifications of the Xe-100 pebble-bed reactor. Upon comparing the results from the SLICE method to published literature, the differences in the eigenvalue keffective were on the order of several hundred pcm (percent millirho). To investigate one possible cause of these differences, a study looking at the sensitivity of the full-core equilibrium keffective and discharge nuclide inventory to temperature was performed by developing equilibrium cores of two additional temperature profiles. From this temperature study, differences on the order of hundreds of pcm for the full-core equilibrium keffective, and up to 15% difference for the discharge inventories were found. These results indicated the strong dependence on temperature that needs to be considered for future work in equilibrium modeling of PB-HTGRs.

Multiphysics Analysis of PLOFC and DLOFC Accidents in the HTGR-TR

High-temperature, gas-cooled reactors (HTGRs) are promising advanced reactor designs. Leveraging passive safety features, higher core outlet temperatures, previous commercial operational experience, and potential coupling with nonelectrical systems, HTGRs present many advantages over traditional light water reactors and other advanced reactor designs. The High-Temperature, Gas-Cooled Test Reactor (HTGR-TR) is a point design developed by Idaho National Laboratory as a response to the need for the expansion of U.S. test reactor capabilities. The HTGR-TR was designed to serve as a test reactor for Generation-IV small modular prismatic block HTGRs and to provide irradiation data, two crucial contributions to the development of advanced reactors. As in any reactor design, it is necessary to understand the behavior of the reactor during accident scenarios to determine the overall safety of the design. Pressurized loss of forced cooling (PLOFC) and depressurized loss of forced cooling (DLOFC) are two accidents that can occur in HTGRs and that present challenges to decay heat removal. In order to understand the behavior of the fuel during these accidents, the resulting mechanical behavior and fission product migration in the fuel are evaluated. Using the HTGR-TR design, which utilizes tristructural isotropic (TRISO) fuel, a multiphysics analysis during the PLOFC and DLOFC transients was performed. This study utilizes neutronics, thermal hydraulics, and fuel performance modeling to analyze the behavior of TRISO particles during steady-state operation and the described transients. This analysis aims to characterize the hoop stress and strain during the transients, as well as predict fission product migration and radiological release. Due to the startup conditions of the HTGR-TR, the silicon carbide (SiC) layer for both of the transients and steady-state conditions are initially in compression for both the hoop stress and strain. The DLOFC transient at the beginning of cycle was the most challenging overall to the fuel, having the closest to tensile behavior and the closest to tensile strain. Because the integrity of the SiC layer was not challenged during any of the transients, the release of 90Sr and 137Cs was found to be negligible. The PLOFC transient at the end of cycle resulted in the largest 110mAg release and subsequent radioactivity increase. However, the increase in radioactivity was minor when compared to typical operating conditions due to the relatively large diffusivity of 110mAg though SiC at normal operating temperatures. Highlights include Neutronics and thermal-hydraulic modeling generating steady-state and transient conditions for use in fuel performance analysis.2. Detailed BISON analysis demonstrating the excellent performance of TRISO fuel during PLOFC and DLOFC transients.3. Fuel behavior showing negligible fission product release.

Discrete-time integral terminal sliding mode load following controller coupled with disturbance observer for a modular high-temperature gas-cooled reactor

With the appearance of digital computers and microprocessor technologies, the control systems used in nuclear reactors continue to transform from analog devices to digital ones in recent years. A heavy challenge is that applying the pre-existing intelligent load following control algorithms developed in the continuous-time domain to the current digital devices of the nuclear reactors may yield the so-called digital chattering effects, and even trigger system instability. To this end, this paper presents a discrete-time integral terminal sliding mode controller (DTITSMC) coupled with a discrete-time disturbance observer (DTDO) for the load following of a modular high-temperature gas-cooled reactor (MHTGR). The common continuous-time MHTGR model is first transformed into a discrete-time form with consideration of model uncertainties, exogenous disturbances, and discretization errors. Then, on the basis of this discrete-time MHTGR model, a DTDO that can estimate the lumped disturbances, composed of model uncertainties, exogenous disturbances, discretization errors, and unmeasured system states, is developed, subsequently being combined with the designed DTITSMC scheme in order to enhance the system robustness. The proposed DTITSMC scheme coupled with the DTDO is novel in that it can be directly implemented on the digital control system of the MHTGR, while at the same time providing an improved load following control performance, owing to the integration of the state-of-the-art sliding mode control techniques, backstepping control techniques, and the disturbance feedforward compensation supported by the disturbance observer techniques. Both the estimation error convergence of the DTDO and the overall system stability are theoretically proved by the Lyapunov stability theory in this paper. Finally, the feasibility and superiority of the proposed DTITSMC scheme coupled with the DTDO over the previous discrete-time linear sliding mode controller (DTLSMC) and discrete-time proportional–integral–derivative controller (DTPIDC) are further confirmed by simulation studies, where in the absence of lumped disturbances, the value of the accumulative absolute load error (AALE) produced by the proposed DTSMPPC scheme coupled with the DTDO is merely 0.0575% and 0.2675% to that of the previous DTLSMC and the optimized DTPIDC, respectively, while in the presence of lumped disturbances, the value of the AALE produced by the proposed DTSMPPC scheme coupled with the DTDO is merely 0.006% and 0.1952% to that of the previous DTLSMC and the optimized DTPIDC, respectively.

Coordinated control of mHTGR-based nuclear steam supply systems considering cold helium temperature

For industrial energy systems, there commonly exists coupling effect amongst those under-controlled process variables. Coordinated control of energy system enhances the operation performance by fully using the coupling effect positive to stabilization. Nuclear steam supply system (NSSS) module is the central of every nuclear plant, and the coordinated control of NSSS module is important for the stable, reliable and efficient operation of nuclear plants. Currently, the process variables considered in developing NSSS coordinated control strategies (CCSs) are mainly neutron flux, steam temperature as well as hot or average primary coolant temperature. Although cold primary coolant temperature, i.e. the primary coolant temperature at reactor inlet, reflects the heat balance between primary and secondary loops, it still not considered in developing CCS for NSSS modules. The modular high temperature gas-cooled reactor (mHTGR) uses helium as primary coolant and graphite as both moderator and structural material. Due to its inherent safety feature and high temperature, the mHTGR can be applied as a heat source driving industrial processes for power generation, chemical production and etc. To further improve the operation performance of mHTGR-based NSSS module, a passivity-based CCS is newly proposed by considering cold helium temperature. The sufficient condition for asymptotical closed-loop stability is given, and numerical simulation results in the cases of power stepping of 10 % reactor full power (RFP) as well as power ramping of 50%RFP and 5%RFP/min show that the control performance can be effectively improved.

High uranium loading TRISO particle for microreactor applications

Tristructural isotropic (TRISO) fuel is currently being considered for novel high-temperature reactor (HTR) designs that deviate significantly from the prismatic or pebble bed modular high-temperature gas-cooled reactors (MHTGRs) that have been the focus of development efforts for the last four decades. These new designs include microreactors with compact cores that can benefit from increasing the core fissile density beyond that conventionally required in an MHTGR using TRISO fuel to enable longer fuel in-core residence time or higher power density. One means of increasing the fissile density in an HTR core is to adjust the coated particle design to increase the ratio of kernel volume to total particle volume. This would increase the volume of fissile phase in a fuel compact with no increase in particle packing fraction beyond that used in fuel compacts tested by the AGR program. Results show that the end-of-life failure probabilities of these high uranium loading particles and those of the validated reference particles are similarly low (∼2 × 10-5) for burnups of at least 9 % FIMA.

Influence of Cooling Water Parameters on the Thermal Performance of the Secondary Circuit System of a Modular High-Temperature Gas-Cooled Reactor Nuclear Power Plant

This study quantitatively analysed the influence of cooling water parameters on the performance of a modular high-temperature gas-cooled reactor (MHTGR) nuclear power plant (NPP). The secondary circuit system and cold-end system were modelled using EBSILON software, version 16.0. The influence of cooling water inlet temperature and mass flow rate on the thermal performance of the secondary circuit system was analysed over the full power range with the goal of optimising net power. Under 100% rated condition, for each 1 °C increase in cooling water inlet temperature between 10 and 33 °C, the net power and cycle efficiency decreased by 0.67 MW and 0.14%, respectively, whereas the heat consumption rate increased by 28.72 kJ/(kW·h). The optimal cooling water mass flow rates corresponding to cooling water inlet temperatures of 16 °C and 33 °C were obtained. The optimal cooling water mass flow rate decreased nonlinearly with decreasing power levels. At a cooling water inlet temperature of 33 °C, an increase in cooling water mass flow rate from the designed value (7697.61 kg/s) to the optimal value (10,922.14 kg/s) resulted in a 1.03 MW increase in net power. These findings provide guidelines for MHTGR NPP operation optimisation and economic improvement, especially under high-temperature weather conditions.

Investigation of thermal hydraulic behavior of the High Temperature Test Facility's lower plenum via large eddy simulation

A high-fidelity computational fluid dynamics (CFD) analysis was performed using the Large Eddy Simulation (LES) model for the lower plenum of the High–Temperature Test Facility (HTTF), a ¼ scale test facility of the modular high temperature gas-cooled reactor (MHTGR) managed by Oregon State University. In most next–generation nuclear reactors, thermal stress due to thermal striping is one of the risks to be curiously considered. This is also true for HTGRs, especially since the exhaust helium gas temperature is high. In order to evaluate these risks and performance, organizations in the United States led by the OECD NEA are conducting a thermal hydraulic code benchmark for HTGR, and the test facility used for this benchmark is HTTF. HTTF can perform experiments in both normal and accident situations and provide high-quality experimental data. However, it is difficult to provide sufficient data for benchmarking through experiments, and there is a problem with the reliability of CFD analysis results based on Reynolds–averaged Navier–Stokes to analyze thermal hydraulic behavior without verification. To solve this problem, high-fidelity 3-D CFD analysis was performed using the LES model for HTTF. It was also verified that the LES model can properly simulate this jet mixing phenomenon via a unit cell test that provides experimental information. As a result of CFD analysis, the lower the dependency of the sub-grid scale model, the closer to the actual analysis result. In the case of unit cell test CFD analysis and HTTF CFD analysis, the volume-averaged sub-grid scale model dependency was calculated to be 13.0% and 9.16%, respectively. As a result of HTTF analysis, quantitative data of the fluid inside the HTTF lower plenum was provided in this paper. As a result of qualitative analysis, the temperature was highest at the center of the lower plenum, while the temperature fluctuation was highest near the edge of the lower plenum wall. The power spectral density of temperature was analyzed via fast Fourier transform (FFT) for specific points on the center and side of the lower plenum. FFT results did not reveal specific frequency-dominant temperature fluctuations in the center part. It was confirmed that the temperature power spectral density (PSD) at the top increased from the center to the wake. The vortex was visualized using the well-known scalar Q-criterion, and as a result, the closer to the outlet duct, the greater the influence of the mainstream, so that the inflow jet vortex was dissipated and mixed at the top of the lower plenum. Additionally, FFT analysis was performed on the support structure near the corner of the lower plenum with large temperature fluctuations, and as a result, it was confirmed that the temperature fluctuation of the flow did not have a significant effect near the corner wall. In addition, the vortices generated from the lower plenum to the outlet duct were identified in this paper. It is considered that the quantitative and qualitative results presented in this paper will serve as reference data for the benchmark.

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.

Fractional‐order sliding mode load following control via disturbance observer for modular high‐temperature gas‐cooled reactor system with disturbances

AbstractLoad‐following control of the modular high‐temperature gas‐cooled reactor (MHTGR) system is one of the essential control problems in practical applications. In this paper, a fractional‐order sliding mode control strategy via a disturbance observer (FOSMCS‐DO) is presented for load following of the MHTGR system with model uncertainties and external disturbances. First of all, the mathematical model of the MHTGR system is constructed by taking neutronics and thermodynamics as well as disturbances into account. Then, based on the MHTGR model, a DO is designed to estimate the lumped disturbances that are composed of model uncertainties, external disturbances, and unmeasured states. Meanwhile, by adopting the Lyapunov stability theory, a FOSMCS is developed to guarantee all the signals of the closed‐loop load‐following control system tend to be stable, where in the control framework, the adverse effects of the lumped disturbances are alleviated by means of the feedforward compensation and the DO. This implies that the proposed overall control strategy possesses strong robustness against model uncertainties and external disturbances. Finally, comparative simulation results with some existing control approaches are provided to illustrate the superiority of the proposed control strategy.

Developing PCC and DCC integral effects test experiments at the High Temperature Test Facility

Among the Next-Generation Nuclear Plant (NGNP) designs, the High-Temperature Gas-Cooled Reactors (HTGRs) are very attractive, due to their inherent safety features, high power conversion efficiency, and potential of providing high-temperature process heat. To perform a thorough safety study and to license these types of reactors, sufficient information needs to be provided about the phenomena that occur during accident scenarios. While several experimental research efforts have been dedicated in the past to investigate accident scenarios, knowledge gaps still exist in the phenomena characteristic of pressurized and depressurized conduction cooldown (PCC/DCC) transients as well as for normal operation scenarios. This paper summarizes the Oregon State University High Temperature Test Facility (HTTF) test matrix, experimental campaign, and selected tests results. High Temperature Test Facility is a scaled Integral Test Facility (IET) that is capable of mimicking scaled dimensions and operational conditions of the Modular High-Temperature Gas Cooled Reactor (MHTGR). The goal of the High Temperature Test Facility is to provide experimental data on the DCC, PCC and normal operating scenarios of the reference Modular High-Temperature Gas Cooled Reactor design. The DCC, PCC, mixing, heat up and cooldown tests described in this paper were performed at prototypical Modular High-Temperature Gas Cooled Reactor temperatures, scaled initial pressure conditions (∼200 kPa), and thermal power input of less than 70 kW. Presented test data show temperature distributions in the High Temperature Test Facility core, upper plenum, cross duct, or lower plenum. Based on these temperature profiles attempts to investigate stratified flow, natural convection flow, heat up, cooldown and mixing phenomena are made. Furthermore, this paper evaluates the performed test campaign in the light of the Very High Temperature Gas-cooled Reactor Phenomena Identification and Ranking Table (PIRT) and proposes experiments to complement the existing PCC/DCC testing database for the validation of the thermal-hydraulic codes.

Impact of the TRISO fuel homogenisation approximation in the MHTGR

TRistructural ISOtropic (TRISO) fuels are commonly used in High-Temperature Gas cooled Reactors due to their resistance to high temperatures, irradiation, and oxidation. Reactor physics modelling of a fuel form consisting of millions of layered particles is quite challenging, however. Homogenisation techniques are often used to simplify the fuel form, representing it as a solid material to reduce computational costs. The impact of these approximations on reactor physics parameters was evaluated using OpenMC. Geometries at various scales (i.e. fuel pin, fuel block, full core) were developed based on the Modular High-Temperature Gas-cooled Reactor. It was found that homogenisation results in a significant reduction in simulation time. The Reactivity equivalent Physical Transformation (RPT) method was found to be superior to simple Volume-Weighted Homogenization (VWH). The neutron energy spectra and flux distributions calculated by RPT had insignificant differences to the explicit model. Reactivity coefficients were found to be similar, but diverged at very high temperatures.

Dynamical modeling and coordinated control design of a multimodular nuclear power-hydrogen cogeneration plant

Based on the idea of interconnecting the modular high temperature gas-cooled reactor (mHTGR), steam Rankine cycle, Cu-Cl cycle and high temperature electrolysis (HTE), the schematic design of a multimodular nuclear power-hydrogen cogeneration is proposed. This cogeneration plant can be utilized to balance the net load by redistributing the motive steam between power generation and hydrogen production. The operational flexibility provided by steam redistribution relies heavily on plant-wide coordinated control. Since dynamic modeling is a crucial basis of control design, a lumped-parameter dynamic model of this nuclear cogeneration plant (NCP) is developed and implemented as a numerical simulation program. The steady-state modeling results show that the rated thermal power of this NCP is about 1523 MW, and the maximal values of hydrogen production rate and output electric power are 11.76 metric tons per hour and 647.3 MW, respectively. The open-loop transient analysis shows that the model's dynamic behavior is in good accordance with physical mechanisms. Further, a hierarchical coordinated control scheme is designed for this NCP, providing a load-following control function based upon steam redistribution, and the corresponding validations are also given. For the load decreasing with ramp rate at 8 MW/min and amplitude of 540 MW, the control system guarantees the grid frequency deviation is smaller than 0.2 Hz while the reactor thermal power, as well as the temperatures and pressures of hydrogen production reactions, are maintained around their expected values. The results in this paper show that the mHTGR-based power-hydrogen cogeneration plant is not only a clean source of electricity and hydrogen but also a flexible source for grid frequency stabilization.

Neural network extended state-observer for energy system monitoring

Due to the complexity of industrial energy systems such as the thermal power plants, renewable plants and batteries, energy system monitoring is gaining more and more attentions. Energy system monitoring methods give the estimations of internal process variables or parameters that can be adopted as the possible indicators for health conditions, which not only founds the basis for diagnosis and prognostics but also strengthens the situation awareness of system operators. In this paper, a neural network extended state-observer (NN-ESO) is newly proposed for general energy systems with measurements of key process variables, where the neural network adopted is the multi-layer perception (MLP). The advanced characteristics of this NN-ESO lies in the following three aspects: (1) No prior knowledge about the process dynamics is used in the observer design. (2) The globally bounded estimations of the exterior disturbances and their changing rates in the measurement channels can be provided. (3) The MLP in the NN-ESO is trained and deployed online without any offline training samples. By adopting the estimations of the exterior disturbances and their varying rates as the health indicators, the NN-ESO is applied for the operation monitoring of a modular high temperature gas-cooled reactor which is a typical small modular nuclear reactor. Simulation results show not only the satisfactory estimation performance but also the feasibility of applying NN-ESO for system monitoring.

A load following control strategy for Chinese Modular High-Temperature Gas-Cooled Reactor HTR-PM

Chinese Modular High-Temperature Gas-Cooled Reactor HTR-PM is a promising energy source for comprehensive utilization. The original coordinated control strategy of HTR-PM is a “turbine following reactor” strategy, which is not able to perform load following operation. To have more flexible operation capability, this paper presents a load following control strategy for HTR-PM. A balance of plant (BOP) model of HTR-PM is first developed for dynamic simulation, including the once through steam generator (OTSG), turbine generator, condenser, feed water heater, deaerator, feed water pump, valves and piping system. Then the original coordinated control strategy and the proposed load following control strategy are clarified, and model validation, simulation results are documented respectively. For turbine trip without reactor shutdown from 100%PFP transient, the steam dump strategy is improved to meet the safety requirements. Results demonstrate that the proposed load following strategy can meet the requirements of safety operation, which is desirable for participation in peak load regulating of power system.

Preliminary study on HTR-10 operating at higher outlet temperature

High Temperature Gas-cooled Reactor (HTGR), which has well-known safety features and high temperature heat supply capability, is expected to be widely used for heat supply and technology heat utilization including the hydrogen production, and so contributing to the reduction of carbon dioxide emissions in various sectors. The 10 MW High Temperature gas-cooled test Reactor (HTR-10) had been constructed and operated in China as a pilot plant to demonstrate the inherent safety features of the modular HTGR. The first criticality of HTR-10 at air condition was realized on December 1, 2000, and the full power operation for 72 h on January 29, 2003.Supported by Chinese National S&T Major Project, HTGR for hydrogen production are now being studied. The physical and thermal hydraulic design to raise the outlet helium temperature of the HTR-10 reactor core from 700 °C to 850–1000 °C is carried out.In this paper, the preliminary thermal hydraulic design of the HTR-10 with the outlet helium temperature of 950 °C (HTR-10H) is introduced. The power density distribution, the fuel temperature distribution and the reactor pressure vessel (RPV) temperature are studied to identify what need to be focused on next. Besides, the DLOFC accident has been studied to evaluate the safety feature of the HTR-10 operating under higher core temperature and outlet temperature.Due to the lower power and power density, as well as the large heat capacity of the graphite reflector and the carbon brick, the maximal fuel temperature during the DLOFC would not exceed the limitation of 1620 °C (250th meeting of the reactor safety commission, 1990). Nevertheless, the higher fuel temperature during the normal operation should be of concern.

Progress of Establishing the China-Indonesia Joint Laboratory on HTGR

Based on good collaborative intents, the China-Indonesia Joint Laboratory on High Temperature Gas-cooled Reactor (HTGR) was co-established in 2017 by the Ministry of Science and Technology of China and Ministry of Research, Technology, and Higher Education of Indonesia. INET, Tsinghua University of China and BATAN of Indonesia are designated as the implementing institutions of each country. INET and BATAN have sincerely cooperated and worked together to launch the three years’ program (2018–2020) for establishing the joint laboratory funded by the Chinese National Key R&D program. The program focuses on capacity building in the field of modular HTGR technology by means of trainings, seminars, workshops, visits and joint researches. This paper reviews the mutual interests and cooperation base of INET and BATAN on the modular HTGR technology. Supported by both governments, INET and BATAN jointly establish the Joint Laboratory on HTGR and proposed four research tasks, including joint studies and development of modular HTGR technology and its application potentials, joint studies on safety and licensing of modular HTGR technology, joint development of modular HTGR simulation technology, and joint studies on localization potentials. Each task is organized and performed by specific joint working team. Regular seminars, embedded with training workshops and technical visits, have been held twice a year to enhance exchanges and discussions on joint research topics. With great efforts, the tasks gradually reach their objects and generate designs and reports. Joint studies also significantly promote the capacity building for many Indonesia colleagues and earn mutual trust and friendship. Moreover, the program of Joint Laboratory sets a good example for future collaborations under the framework of “Belt and Road” initiative.

Neural network-based adaptive fault-tolerant control for load following of a MHTGR with prescribed performance and CRDM faults

A neural network-based adaptive fault-tolerant control scheme (NN-AFTCS) for load following of a modular high-temperature gas-cooled reactor (MHTGR) system possessing parameter uncertainties, exogenous disturbances, and control rod drive mechanism (CRDM) faults, capable of guaranteeing prescribed performance is developed in this paper. From practical point of view, the mathematical model of the MHTGR system is first described with consideration of total CRDM fault effects together with lumped uncertainties that consist of parameter uncertainties, exogenous disturbances, and unmeasured states, where the unknown values of both time-varying total CRDM fault effects and lumped uncertainties are approximated by utilizing the radial basis function neural network (RBFNN) owing to its online learning capability. After that, with the help of the RBFNN, a fault-tolerant controller is proposed to compensate the adverse effects of the CRDM faults and lumped uncertainties, while at the same time achieving prescribed performance guarantees by invoking error transformation technique. Moreover, the dynamic surface control technique is also employed such that the problem of “explosion of complexity” existing in the conventional backstepping is averted. With Lyapunov stability theorem, it is proved that the load following error in the closed-loop MHTGR system would converge to a small residual set within a finite time. At last, simulation and comparison results confirm that the proposed NN-AFTCS exhibits higher load following accuracy, better CRDM fault-tolerant capability, and stronger robustness against lumped uncertainties than a disturbance observer-sliding mode controller and an active disturbance rejection controller.

Power-pressure coordinated control of modular high temperature gas-cooled reactors

Modular high temperature gas cooled reactor (mHTGR) adopts helium as coolant and graphite as both the moderator and structural material, whose fuel element is built by embedding thousands of TRISO coated particles into a graphite matrix. Since helium is an ideal gas, the dynamics of reactor coolant temperature is tightly coupled with that of primary pressure. To further improve the operation performance, it is meaningful to develop the coordinated control of power-level and primary pressure for mHTGRs. In this paper, a passivity-based power-pressure coordinated control is proposed for mHTGRs, and the sufficient condition for globally asymptotic closed-loop stability is given. This newly-built power-pressure coordinated control law is applied to the mHTGR of six-modular high temperature gas-cooled reactor plant HTR-PM600, and simulation results in the case of power ramping decrease and increase as well as performance comparison not only verify the theoretical result about closed-loop stability but also illustrate the influence of control parameters as well as the stronger performance relative to the classical proportional integral/differential (PI/D) controller.

Reactor Performance and Safety Characteristics of Beryllium-Based Composite Moderators as Replacements for Graphite in mHTGRs

This study evaluates beryllium-based two-phase composite moderators as an alternative to graphite in an evaluation of reactor performance and safety characteristics. Historically, modular high-temperature gas-cooled reactors (mHTGRs) use graphite as a moderator because of its high moderating ratio and reasonable thermal properties; however, graphite has unfavorable properties under irradiation, which can require component replacement and a significant radioactive waste burden. In this assessment, we explore advanced moderators comprised of magnesium oxide (MgO) as the host matrix and beryllium metal and/or beryllium oxide (Be and/or BeO) as the entrained moderating phase. For the reactor performance and thermal-hydraulic safety analysis, the core design model of the General Atomics mHTGR-350 was used to demonstrate the feasibility of a “drop-in” replacement of graphite using the beryllium-based moderators. We employed the neutronics code Serpent to analyze the moderating behavior of the composite moderators with comparisons drawn to graphite. We performed a scoping analysis of accidents for mHTGRs using RELAP to show that these moderators do not present impediments to safety and are expected to stay within temperature limits. Measured thermophysical properties of the composite moderators are used in the thermal-hydraulic assessments. Our analysis reveals that the two-phase composite MgO-matrix beryllium-based moderators are a suitable replacement for graphite.