Model-Free Coordinated Control for MHTGR-Based Nuclear Steam Supply Systems

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.

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

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 demonstration construction of the Chinese design of modular high temperature gas cooled reactor (MHTGR), named as high-temperature gas-cooled reactor-pebble bed module (HTR-PM) is now under construction. HTR-PM employs a layout of two nuclear steam supply system (NSSS) modules coupled to one steam turbine. The verification and validation (V&V) of HMIs is important for HTR-PM, as the first two-modular design. The V&V program of HMIs for HTR-PM is introduced in this paper. The V&V activities may include the static and dynamic verifications. The basic layout and static HMI usabilities will be verified on the static stage, and the task support based on operation procedures will be evaluated in the dynamic V&V phase. A verification platform is built to make verification and validation of the HMIs. The HMIs of HTR-PM will be improved according to the V&V result.

Costs have been estimated to design, construct, operate and maintain Modular High Temperature Gas-Cooled Reactor (MHTGR) power plants. The cost estimates are based upon the conceptual design of the reference MHTGR plant for which a summary description is provided in the MHTGR Conceptual Design Summary Report. The reference MHTGR plant consists of four 350 MW(t) reactor modules and two turbine generators (4 x 2) which produce a net output of approximately 540 MWe. The reactor modules and all other equipment and facilities necessary for the control of radionuclides are contained on a Nuclear Island (NI). All other plant facilities are part of the Energy Conversion Area (ECA) which is physically separated from the NI. Costs for reference MHTGR plants were developed for the following four cases: a Lead plant consisting of the first commercial plant conforming to the reference 4 x 2 plant design including the licensing-related costs from Final Design Approval (FDA) through to completion of the design certification process, approximate costs were also developed for the design licensing and technology development necessary to obtain the FDA; a Replica plant conforming to the certified design that follows the first commercial plant; a Target plant which is the Nth-Of-A-Kind (NOAK) equilibrium plant conforming to the certified design; and a large Target plant consisting of two power blocks with each power block conforming to the certified design.

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.

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.

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

The short-term thermal response of the Modular High-Temperature Gas-Cooled Reactor is analyzed for a range of flow and reactivity transients. These include loss of forced circulation without scram, moisture ingress, spurious withdrawal of a control rod group, hypothetical large and rapid positive reactivity insertion, and a rapid core cooling event. The coupled heat transfer-neutron kinetics model is also described.

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.

Phenomena-based thermal-hydraulic modeling requirements for systems analysis of a Modular High Temperature Gas-Cooled Reactor

Stratification of oxygen in Modular High Temperature Gas-Cooled Reactors (MHTGRs) presents a significant problem in the event of a reactor blowdown. The current study aims to quantify the mixing and stratification phenomena in a specific flow path of a 1/28 prototype of an MHTGR using bulk mixing parameters. CFD simulations are carried out in ANSYS Fluent, and results are analysed to obtain quantitative measures of spatiotemporal distribution of oxygen mass concentration in the reactor cavities.

Passivity-based control of fluid flow networks with capacitance

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

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.