Xing L. Yan

Evaluation of high temperature gas reactor for demanding cogeneration load follow

Modular nuclear reactor systems are being developed around the world for new missions among which is cogeneration for industries and remote areas. Like existing fossil energy counterpart in these markets, a nuclear plant would need to demonstrate the feasibility of load follow including (1) the reliability to generate power and heat simultaneously and alone and (2) the flexibility to vary cogeneration rates concurrent to demand changes. This article reports the results of JAEAs evaluation on the high temperature gas reactor (HTGR) to perform these duties. The evaluation results in a plant design based on the materials and design codes developed with JAEAs operating test reactor and from additional equipment validation programs. The 600 MWt-HTGR plant generates electricity efficiently by gas turbine and 900°C heat by a topping heater. The heater couples via a heat transport loop to industrial facility that consumes the high temperature heat to yield heat product such as hydrogen fuel, steel, or chemical. Original control methods are proposed to automate transition between the load duties. Equipment challenges are addressed for severe operation conditions. Performance limits of cogeneration load following are quantified from the plant system simulation to a range of bounding events including a loss of either load and a rapid peaking of electricity.

A STUDY OF AIR INGRESS AND ITS PREVENTION IN HTGR

Abstract A rupture of the primary piping in the helium-cooled and graphite-moderated high-temperature gas-cooled reactor (HTGR) represents a design-basis event that should not result in significant safety consequences. In such a loss-of-coolant event, the reactor would be shut down inherently, and the decay heat would be removed passively with the ultimate reactor temperature rise being less than the design limit. Still, an important concern for reactor safety continues to be graphite oxidation damage to the fuel and core should a major air ingress take place through the breached primary pressure boundary. Two major cases of air ingress are studied. The first case results from the rupture of a control rod or refuel access standpipe atop the reactor pressure vessel (RPV). To rule out the possibility of such a standpipe rupture, a design change is proposed in the vessel top structure. The feasibility of the modified vessel local structure is evaluated. The second case of air ingress results from the rupture of one or more main coolant pipes on the lower body of the RPV. Experiment and analysis are performed to understand the multiphased air ingress phenomena in the depressurized reactor. Accordingly, a new passive mechanism of sustained counter air diffusion is proposed and shown to be effective in preventing major air ingress through natural circulation in the reactor. The results of the present study are expected to enhance the HTGR safety and economics.

Experiments and validation analyses of HTTR on loss of forced cooling under 30% reactor power

In a safety demonstration test involving the loss of both reactor reactivity control and core cooling, the high-temperature engineering test reactor (HTTR) demonstrates spontaneous stabilization of the reactor power. The test and analytical results of tripping one or two out of three gas circulators without reactor scram have already been reported. Moreover, the pre-analytical result of tripping all three gas circulators without reactor scram has been presented. On the other hand, the test and analytical results of tripping all three gas circulators without reactor scram are shown in this paper. About experiments, at an initial reactor power of 30% (9 MW), when all three gas circulators were tripped without reactor scram to reduce the coolant flow rate to zero, the fuel temperature did not show a large increase because the large heat capacity of the graphite core could absorb heat from the fuel in a short period. Moreover, the decay heat could be transferred through the graphite core and the reactor pressure vessel (RPV), emitted by thermal radiation from its outer surface and removed to the active vessel cooling system; therefore, the core at 9 MW was never exposed to the danger of a core melt, and the reactor power was stabilized spontaneously. About analyses, the reactivity performance is important for predicting the converging level of reactor power that affects the fuel temperature during a loss of forced cooling (LOFC) without reactor scram. With regard to thermal hydraulics, the performances of graphite heat conduction in the reactor core and thermal radiation from the RPV surface to the reactor cavity cooling system are crucial for predicting the temperature behavior of the fuel and RPV in the LOFC condition. It was confirmed that reactor kinetics coupled with heat transfer could be applied to reactor safety and accident analysis based on the comparison between the experiments and the analyses.

Conceptual Design of Small-Sized HTGR System for Steam Supply and Electricity Generation (HTR50S)

Japan Atomic Energy Agency (JAEA) has started a conceptual design of a small-sized HTGR for steam supply and electricity generation (HTR50S) to deploy the high temperature gas cooled reactor (HTGR) in developing countries at an early date (i.e., in the 2030s). Its reactor power is 50MWt and the reactor outlet temperature is 750°C. It is a first-of-kind of the commercial plant or a demonstration plant of a small-sized HTGR system for steam supply to the industries and the district heating, and electricity generation using a steam turbine. The design philosophy of the HTR50S is to upgrade the performance from the Japanese first HTGR (HTTR) and to reduce the cost for the commercialization by utilizing the knowledge obtained by the HTTR operation and the design of an advanced commercial plant of 600 MWt-class Very High Temperature Reactor (GTHTR300 series). The major specifications of the HTR50S were determined based on its design philosophy. And the targets of the technology demonstration using the HTR50S for the future commercial small-sized HTGR were identified. The system design of HTR50S was performed to offer the capability of electricity generation, cogeneration of electricity and steam for a district heating and industries. The market potential for the small-sized HTGR in the developing countries was evaluated for the application of the electricity, process heat, district heating and pure water production. It was confirmed that there is enough market potential for the small-sized HTGR in the developing countries. This paper described the major specification and system design of the HTR50S and the market potential for the small-sized HTGR in the developing countries.© 2011 ASME

Transient Analysis of Depressurized Loss-of-Forced-Circulation Accident Without Scram in High-Temperature Gas-Cooled Reactor

The transient response of the high-temperature gas-cooled reactor (HTGR) to depressurized loss of forced circulation combined with failure of all reactor trip systems, a beyond-design-basis accident, is analyzed for an extended period of time during which no active core cooling is resumed. The characteristic behavior of the reactor during the long-term conduction cooldown event is found to be shaped by several parameters that are usually not considered in the safety design of the HTGR. For example, while the Doppler effect is usually relied upon to provide inherent shutdown of the reactor, the reactivity coefficient of temperature of the graphite moderator is found to be a critical parameter for determining the final settling temperature of the fuel following the recriticality. Furthermore, this study finds that the peak fuel temperature reached during this event is correlated strongly even to the initial core operating temperature prior to the initiation of the transient event. These and other results of this study are expected to provide useful input to the development of enhanced safety design guidelines for commercial HTGRs in the aftermath of the Fukushima accident.

Status and Future Development for Nuclear Cogeneration System GTHTR300C

GTHTR300C is a small modular reactor based on a 600 MWt high temperature gas reactor (HTGR) and intended for a number of cogeneration applications such as process heat supply, hydrogen production, steelmaking, desalination in addition to power generation. The basic design has been completed by JAEA together with Japanese heavy industries. The reactor design and key plant technologies have been validated through test reactor and equipment verification. Future development includes demonstration programs to be performed on a 50 MWt system HTR50S. The demonstration programs are implemented in three steps. In the first step, a base commercial plant for heat and power is to be constructed of the same fuel proven in JAEA’s successful 950°C, 30 MWt HTGR test reactor and a conventional steam turbine such that the construction can readily proceed without major development requirement and risk. Beginning in the second step, a new fuel presently being developed at JAEA is expected to be available. With this fuel, the core outlet temperature is raised to 900°C for purpose of demonstrating more efficient gas turbine power generation and high temperature heat supply. Added in the final step is a thermochemical process to demonstrate nuclear-heated hydrogen production via water decomposition. A licensing approach to coupling high temperature industrial process to nuclear reactor will be developed. The designs of GTHTR300C and HTR50S will be presented and the demonstration programs will be described.Copyright

Control Strategies for VHTR Gas-Turbine System With Dry Cooling

An original control strategy for very high temperature reactor (VHTR) gas-turbine system with dry cooling against ambient air temperature fluctuation was established in order to enable the freedom of site selection wherever desired without significant drawbacks on the performance. First, the operability of power conversion system and degradation of power generation efficiency were examined considering not only the thermodynamics but also the mechanical efficiency of compressor based on detailed performance map derived from experimental data. Second, control simulations for large ambient temperature fluctuations were conducted by system analysis code with the built-in control strategy. In addition, the sensitivity of power generation efficiency for typical steam cycle with dry cooling to ambient air temperature changes was assessed for the comparison. It was shown that the design goal can be effectively met simply by monitoring and controlling a few of key operating parameters such as reactor outlet temperature, primary coolant pressure. Furthermore, distinctive advantages of the VHTR gas-turbine system over nuclear power plant employing Rankine cycle was demonstrated when installing in inland area.Copyright

Experimental Test Plan of Air Ingress for HTGR

Primary pipe rupture is an important design base accident in the high temperature gas cooled reactor (HTGR). When a primary pipe of the HTGR ruptures, helium coolant gas in the reactor blows out into the reactor confinement structure and the reactor primary system depressurizes. After the pressures of the reactor and the confinement equalizes, air is expected to enter the reactor core from the breach. Consequently, the core graphite structures may be oxidized by the air and the complicated natural convection of multi component gas mixtures with chemical reactions would take place inside the reactor. Hence, it is necessary to investigate the air ingress process, the natural convection of multi component gas mixtures in order to understand the effect and develop mitigation of the air ingress. JAEA has performed analysis and fundamental experiments about air ingress from the rupture of one or more main coolant pipes on the lower body of the RPV. These studies showed the air ingress phenomena in the depressurized reactor and proposed a new passive mechanism of sustained counter air diffusion (SCAD) that has been shown effective in preventing major air ingress through natural circulation in the reactor. In the present plan, JAEA will construct an experimental reactor mockup including reactor core, the SCAD system, pressure vessel, coaxial pipe and so on. The core is made of graphite or ceramics and heated by electric heaters to allow for test operation up to 1200°C. Present status of these activities will be presented. Based on the analytical results and know-how obtained through the bench test, a 1/8 scale air ingress mockup test, which intends to simulate the accident condition of GTHTR300, is being planned with a conceptual design as the next step of the air ingress experiment evaluation in JAEA. In the design of mockup experimental facility, it is important to reproduce flow phenomena in a reasonable scale from the viewpoint of construction cost. We designed the internal structure to reproduce mixing performance of multi-component flow involving ingress phenomena especially in the guillotine breaks of primary coolant pipe. Complex flow pattern with gas oxidized chemical interaction in the graphite porous structure of the HTGR core will be characterized. Preliminary analytical results especially natural circulation flow patterns induced by density and concentration difference obtained with a CFD model agreed well with that measured by bench experiments, which showed natural circulation pattern in a simplified reactor.Copyright