Advanced Control and Protection system Design Methods for Modular HTGRs

The U.S Department of Energy (DOE) is conducting research and development (R&D) on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core could be either a prismatic graphite block type core or a pebble bed core. Use of a liquid salt coolant is also being evaluated. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission Demonstrate safe and economical nuclearassisted production of hydrogen and electricity. The DOE laboratories, led by the INL, will perform R&D that will be critical to the success of the NGNP, primarily in the areas of: High temperature gas reactor fuels behavior High temperature materials qualification Design methods development and validation Hydrogen production technologies Energy conversion. The current R&D work is addressing fundamental issues that are relevant to a variety of possible NGNP designs. This document describes the NGNP R&D planned and currently underway in the first three topic areas listed above. The NGNP Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is presented in Section 2, the NGNP Materials R&D Program Plan is presented in Section 3, and the NGNP Design Methods Development and Validation R&D Program is presented in Section 4. The DOE-funded hydrogen production [DOE 2004] and energy conversion technologies programs are described elsewhere.

The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5-year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant (NGNP) project. Because the NRC's interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC).

The licensing experience of the Modular High-Temperature Gas-Cooled Reactor (MHTGR)

A novel approach to fabricating fuel compacts for the next generation nuclear plant (NGNP)

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 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.

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.

VHTR-Based Systems for Autonomous Co-Generation Applications

The United States Department of Energy is promoting the resurgence of nuclear power in the U. S. for both electrical power generation and production of process heat required for industrial processes such as the manufacture of hydrogen for use as a fuel in automobiles. The DOE project is called the next generation nuclear plant (NGNP) and is based on a Generation IV reactor concept called the very high temperature reactor (VHTR), which will use helium as the coolant at temperatures ranging from 450 ºC to perhaps 1000 ºC. While computational fluid dynamics (CFD) has not been used for past safety analysis for nuclear reactors in the U. S., it is being considered for safety analysis for existing and future reactors. It is fully recognized that CFD simulation codes will have to be validated for flow physics reasonably close to actual fluid dynamic conditions expected in normal and accident operational situations. To this end, experimental data have been obtained in a scaled model of a narrow slice of the lower plenum of a prismatic VHTR. The present report presents results of CFD examinations of these data to explore potential issues with the geometry, the initial conditions, the flow dynamics and the data needed to fully specify the inlet and boundary conditions; results for several turbulence models are examined. Issues are addressed and recommendations about the data are made.

The present work is concerned with heat exchanger development for molten salt service, including the proposed molten salt reactor (MSR), a homogeneous reactor in which the fuel is dissolved in a circulating fluid of molten salt. It is an outgrowth of recent work done under the Next Generation Nuclear Plant (NGNP) program; what the two reactor systems have in common is an inherently safe nuclear plant with a high outlet temperature that is useful for process heat as well as more conventional generation The NGNP program was tasked with investigating the application of a new generation of nuclear power plants to a variety of energy needs. One baseline reactor design for this program is a high temperature, gas-cooled reactor (HTGR), which provides many options for energy use. These might include the conventional Rankine cycle (steam turbine) generation of electricity, but also other methods: for example, Brayton cycle (gas turbine) electrical generation, and the direct use of the high temperatures characteristic of HTGR output for process heat in the chemical industry. Such process heat is currently generated by burning fossil fuels, and is a major contributor to the carbon footprint of the chemical and petrochemical industries. The HTGR, based on graphite fuel elements, can produce very high output temperatures; ideally, temperatures of 900 °C or even greater, which has significant energy advantages. Such temperatures are, of course, at the frontiers of materials limitations, at the upper end of the performance envelope of the metallic materials for which robust construction codes exist, and within the realm of ceramic materials, the fabrication and joining of which, on the scale of large energy systems, are at an earlier stage of development. A considerable amount of work was done in the diffusion welding of materials of interest for HTGR service with alloys such as 617 and 800H. The MSR output temperature is also materials limited, and is projected at about 700 °C. (RR E) A different set of alloys, such as Alloy N and 242, are needed to handle molten salts at this temperature. The diffusion welding development work described here builds on techniques developed during the NGNP work, as applied to these alloys. There is also the matter of dissimilar metal welding, since alloys suitable for salt service are generally not suited for service in gaseous oxidizing environments, and vice versa, and welding is required for the Class I boundaries in these systems, as identified in the relevant ASME codes.

VHTR-based systems for autonomous co-generation applications

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

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

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.

Nuclear powered heat pumps for near-term process heat applications