A nonlinear adaptive sliding mode control strategy for modular high-temperature gas-cooled reactors

The modular high temperature gas-cooled reactor (MHTGR) is a typical small modular reactor (SMR) that offers simpler, standardized and safer modular design by being factory built, requiring smaller initial capital investment, and having a shorter construction period. Thanks to its small size, the MHTGRs could be beneficial in providing electric power to remote areas that are deficient in transmission or distribution and in generating local power for large population centers. Based on the multi-modular operation scheme, the inherent safety feature of the MHTGRs can be applicable to large nuclear plants of any desired power rating. The MHTGR-based nuclear steam supplying system (NSSS) is constituted by an MHTGR, a side-by-side arranged helical-coil once-through steam generator (OTSG) and some connecting pipes. Due to the side-by-side arrangement, there is a tight coupling effect between the MHTGR and OTSG. Moreover, there always exists the parameter perturbation of the NSSSs. Thus, it is meaningful to study the model-free coordinated control of MHTGR-based NSSSs for safe, stable, robust and efficient operation. In this paper, a new model-free coordinated control strategy that regulates the nuclear power, MHTGR outlet helium temperature and OTSG outlet overheated steam temperature by properly adjusting the control rod position, helium flowrate and feed-water flowrate is established for the MHTGR-based NSSSs. Sufficient conditions for the globally asymptotic closed-loop stability is given. Finally, numerical simulation results in the cases of large range power decrease and increase illustrate the satisfactory performance of this newly-developed model-free coordinated NSSS control law.

Study on the DLOFC and PLOFC accidents of the 200 MWe pebble-bed modular high temperature gas-cooled reactor with TINTE and SPECTRA codes

The modular high-temperature gas-cooled nuclear reactor (MHTGR) is seen as one of the best candidates for the next generation of nuclear power plants. The once through steam generator (OTSG) is a crucial element for any MHTGR power plants with steam cycle. In order to guarantee the efficient and safe operation of the power plant, the outlet steam temperature of the OTSG should be well controlled. In this paper, a saturated output feedback dissipation control (SOFDC) law is presented for regulating the outlet steam temperature of OTSGs, and the corresponding robustness analysis is also given. In the numerical study, the SOFDC is applied to the outlet steam temperature control for the OTSG of Chinese high temperature gas-cooled reactor pebble-bed module (HTR-PM) project. Numerical simulation results show not only feasibility but also good control performance of this regulator, and the influence of its parameters to the regulation performance is also discussed.

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

With the capability of burning a variety of fossil fuels, giving high thermal efficiency, and operating with low emissions, the gas turbine is becoming a major prime-mover for a wide spectrum of applications. Almost three decades ago two experimental projects were undertaken in which gas turbines were actually operated with heat from nuclear reactors. In retrospect, these systems were ahead of their time in terms of technology readiness, and prospects of the practical coupling of a gas turbine with a nuclear heat source towards the realization of a high efficiency, pollutant free, dry-cooled power plant has remained a long-term goal, which has been periodically studied in the last twenty years. Technology advancements in both high temperature gas-cooled reactors, and gas turbines now make the concept of a nuclear gas turbine plant realizable. Two possible plant concepts are highlighted in this paper, (1) a direct cycle system involving the integration of a closed-cycle helium gas turbine with a modular high temperature gas cooled reactor (MHTGR), and (2) the utilization of a conventional and proven combined cycle gas turbine, again with the MHTGR, but now involving the use of secondary (helium) and tertiary (air) loops. The open cycle system is more equipment intensive and places demanding requirements on the very high temperature heat exchangers, but has the merit of being able to utilize a conventional combined cycle turbo-generator set. In this paper both power plant concepts are put into perspective in terms of categorizing the most suitable applications, highlighting their major features and characteristics, and identifying the technology requirements. The author would like to dedicate this paper to the late Professor Karl Bammert who actively supported deployment of the closed-cycle gas turbine for several decades with a variety of heat sources including fossil, solar, and nuclear systems.

The High Flux Isotope Reactor (HFIR) Removable Beryllium-19 and 20 (HRB-19 and 20) irradiation test, in addition to the HRB-21 test and others, will provide fuel performance and fission product transport data for the validation of models of the Modular High Temperature Gas-cooled Reactor (MHTGR) in accordance with the MHTGR Plant Fuel/Fission Product Technology Development Plan. These models are utilized in safety-related calculations and hence the data is safety-related. These tests are planned primarily to provide data to support satisfaction of the requirements of the Design Data Need (DDN) for qualifying the fuel performance under normal MHTGR reactor operating conditions and for fuel performance under MHTGR off normal conditions as well as the conditions listed in the verification and validation plan for radionuclide control methods. This document provides a discussion of test specifications. 11 refs., 4 figs., 3 tabs.

The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab.

This report provides an integrated overview of measurements and instrumentation for near-term future high-temperature gas-cooled reactors (HTGRs). Instrumentation technology has undergone revolutionary improvements since the last HTGR was constructed in the United States. This report briefly describes the measurement and communications needs of HTGRs for normal operations, maintenance and inspection, fuel fabrication, and accident response. The report includes a description of modern communications technologies and also provides a potential instrumentation communications architecture designed for deployment at an HTGR. A principal focus for the report is describing new and emerging measurement technologies with high potential to improve operations, maintenance, and accident response for the next generation of HTGRs, known as modular HTGRs, which are designed with passive safety features. Special focus is devoted toward describing the failure modes of the measurement technologies and assessing the technology maturity.

Reactivity accident due to inadvertent withdrawal of the control rod is one kind of the design basis accident for high temperature gas-cooled reactors, which should be analyzed carefully in order to validate the reactor inherent safety properties. Based on the preliminary design of the Chinese Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM) with single module power of 250MW, several cases of reactivity accident has been studied by the help of the software TINTE in the paper, e.g., the first scram signal works or not, the absorber balls (secondary shutdown units) drop or not, and the ATWS situation is also taken into account. The dynamic processes of the important parameters including reactor power, fuel temperature and Xenon concentration are studied and compared in detail between these different cases. The calculating results show that, the decay heat during the reactivity accidents can be removed from the reactor core solely by means of physical processes in a passive way, so that the temperature limits of fuel element and other components are still obeyed, which can effectively keep the integrality of the fuel particles to avoid massive fission products release. This will be helpful to the further detail design of the HTR-PM demonstrating power plant project.

Due to the inherent safety feature, the modular high-temperature gas-cooled nuclear reactor (MHTGR) is proper to be utilized to build a new generation of nuclear power plants. Motivated by the large scale closed-loop stability provided by nonlinear controllers, a nonlinear output feedback power-level control based on both the techniques of feedback dissipation and well-established backstepping is given in this paper. The virtue of this controller, i.e. the ability of globally asymptotic stabilization is based on the inherent zero-state detectability property of the dynamics of the MHTGR. This controller is then simplified to a proportional output feedback controller which guarantees strong stability performance if the uncertainty of control rod dynamics is small enough. This simplified power-level controller has chosen to be the practical implementation strategy. Verification results provided by the hardware-in-loop simulation show the feasibility of this simplified control strategy.

In this paper, a chatter free sliding mode control (SMC) strategy for chaos control is introduced to the nonlinear uncertain chaotic systems in the presence of unknown bounded uncertainties and external disturbances. The sliding mode surface with differential operators is proposed to divert the discontinuous gian function swith term into only one control input, and the control input is smooth and differentiable with respect to time. Based on the Lyapunov stability theory, SMC stability analysis is performed and a theorem serving as designing a chatter free sliding mode control input is also introduced. Finally, numerical simulations are presented to verify the usefulness of the proposed control strategy.

Disturbance observer based adaptive sliding mode control for power tracking of PWRs

High temperature gas-cooled reactor (HTGR) technology can play an important role in the energy future of the United States by extending the use of nuclear energy for non-electricity energy production missions, as well as continuing to provide a considerable base load electric power generation capability. Extending nuclear energy into the industrial and transportation sectors through the coproduction of process heat and electricity provides safe, reliable energy for these sectors in an environmentally responsible manner. The modular HTGR provides a substantial improvement in nuclear plant safety for the protection of the public and the environment, and supports collocation of the HTGRhigh temperature gas-cooled reactor with major industrial facilities. Under U.S. Department of Energy direction since 2006, the Next Generation Nuclear Plant Project at Idaho National Laboratory has been working toward commercializing the HTGR technology. However, a recent decision by the Secretary of Energy to reduce the scope of the Next Generation Nuclear Plant Project to a research and development program, considerable realignment has taken place. This report: (1) summarizes the accomplishments of the Next Generation Nuclear Plant Program from FY2011 through FY2013; (2) lays out the path forward necessary to achieve the ultimate objective of commercializing HTGR technology; and (3) discusses ongoing technical, licensing, and evaluation activities under the realigned Next Generation Nuclear Plant program considered important to preserve the significant investment made by the government to-date and to maintain some progress in meeting the objectives of the Energy Policy Act of 2005 (EPAct2005).

Dynamic matrix control of thermal power for multiple nuclear steam supply system modules

During some of the design basis accidents in Modular High Temperature Gas Cooled Reactors (MHTGRs), the main Heat Transport System (HTS) and the Shutdown Cooling System n removed by the passive Reactor (SCS) are assumed to have failed. Decay heat is the Cavity Cooling System (RCCS) only. If either forced flow cooling system becomes available during such a transient, its restart could significantly reduce the down-time. This report used the THATCH code to examine whether such restart, during a period of elevated core temperatures, can be accomplished within safe limits for fuel and metal component temperatures. If the reactor is scrammed, either system can apparently be restarted at any time, without exceeding any safe limits. However, under unscrammed conditions a restart of forced cooling can lead to recriticality, with fuel and metal temperatures significantly exceeding the safety limits.