600-MW_(e)high-temperature gas-cooled reactor nuclear power plant HTR-PM600

The HTR-PM600 high-temperature gas-cooled reactor nuclear power plant is based on the technology of the high-temperature gas-cooled reactor pebble-bed module(HTR-PM)demonstration project.It utilizes proven HTR-PM reactor and steam generator modules with a thermal power of 250 MW_(th)and power generation of approximately 100 MW_(e)per module.Six modules in parallel,connected to a steam turbine,form a 600-MW_(e)nuclear power plant.In addition,its system configuration in the nuclear island is identical to that of the HTR-PM in which the technical risks are minimized.Under this principle,the HTR-PM600 achieves the same level of inherent safety as the HTR-PM.The concept of a ventilated lowpressure containment(VLPC)is unchanged;however,a large circular VLPC accommodating all six reactor modules is adopted rather than the previous small-cavity-type VLPC,which contains only one module,as defined for the HTR-PM.The layout of the nuclear island and its associated systems refer to single-unit pressurized water reactor(PWR)practices.With this layout,the HTR-PM600achieves a volume size of the nuclear island that is comparable to a domestic PWR of the same power level.This will be a GenerationⅣnuclear energy technology that is economically competitive.

Adaptive output-feedback power-level control for modular high temperature gas-cooled reactors

Small modular reactors(SMRs) are beneficial in providing electricity power safely and viable for specific applications such as seawater desalination and heat production. Due to its inherent safety feature, the modular high temperature gas-cooled reactor(MHTGR) is considered as one of the best candidates for SMR-based nuclear power plants. Since its dynamics presents high nonlinearity and parameter uncertainty, it is necessary to develop adaptive power-level control, which is beneficial to safe, stable, and efficient operation of MHTGR and is easy to be implemented. In this paper, based on the physically-based control design approach, an adaptive outputfeedback power-level control is proposed for MHTGRs. This control can guarantee globally bounded closedloop stability and has a simple form. Numerical simulation results show the correctness of the theoretical analysis and satisfactory regulation performance of this control.

In-Vessel Melt Retention of Pressurized Water Reactors: Historical Review and Future Research Needs

After the first concrete was poured on December 9, 2012 at the Shidao Bay site in Rongcheng, Shandong Province, China, the construction of the reactor building for the world＇s first high-temperature gas- cooled reactor pebble-bed module （HTR-PM） demonstration power plant was completed in June, 2015. Installation of the main equipment then began, and the power plant is currently progressing well to- ward connecting to the grid at the end of 2017. The thermal power of a single HTR-PM reactor module is 250 MW,h, the helium temperatures at the reactor core inlet/outlet are 250/750 ℃, and a steam of 13.25 MPa/567 ~C is produced at the steam generator outlet. Two HTR-PM reactor modules are connect- ed to a steam turbine to form a 210 MW nuclear power plant. Due to China＇s industrial capability, we were able to overcome great difficulties, manufacture first-of-a-kind equipment, and realize series ma- jor technological innovations. We have achieved successful results in many aspects, including planning and implementing R＆D, establishing an industrial partnership, manufacturing equipment, fuel produc- tion, licensing, site preparation, and balancing safety and economics; these obtained experiences may also be referenced by the global nuclear community.

Nusselt number correlation for turbulent heat transfer of helium-xenon gas mixtures

A gas-cooled nuclear reactor combined with a Brayton cycle shows promise as a technology for highpower space nuclear power systems.Generally,a helium-xenon gas mixture with a molecular weight of14.5-40.0 g/mol is adopted as the working fluid to reduce the mass and volume of the turbomachinery.The Prandtl number for helium-xenon mixtures with this recommended mixing ratio may be as low as 0.2.As the convective heat transfer is closely related to the Prandtl number,different heat transfer correlations are often needed for fluids with various Prandtl numbers.Previous studies have established heat transfer correlations for fluids with medium-high Prandtl numbers(such as air and water)and extremely lowPrandtl fluids(such as liquid metals);however,these correlations cannot be directly recommended for such helium-xenon mixtures without verification.This study initially assessed the applicability of existing Nusselt number correlations,finding that the selected correlations are unsuitable for helium-xenon mixtures.To establish a more general heat transfer correlation,a theoretical derivation was conducted using the turbulent boundary layer theory.Numerical simulations of turbulent heat transfer for helium-xenon mixtures were carried out using Ansys Fluent.Based on simulated results,the parameters in the derived heat transfer correlation are determined.It is found that calculations using the new correlation were in good agreement with the experimental data,verifying its applicability to the turbulent heat transfer for helium-xenon mixtures.The effect of variable gas properties on turbulent heat transfer was also analyzed,and a modified heat transfer correlation with the temperature ratio was established.Based on the working conditions adopted in this study,the numerical error of the property-variable heat transfer correlation was almost within 10%.

Research on manufacture technology of spherical fuel elements by dry-bag isostatic pressing

The steady development of high-temperature gas-cooled reactors(HTRs) has increased the requirements for the production cost and quality of fuel elements. Green fuel element pressing is one of the key steps to increase the production capacity. This paper proposes a proprietary vacuum dry-bag isostatic pressing(DIP) apparatus. The structural change of the matrix graphite powder during the DIP process was examined by analyzing the density change of the matrix graphite spheres with pressure. The soft molding process was simulated using the finite element method. The dimensional changes in the spheres during the pressing, carbonization, and purification stages were explored. The performance of the fuel matrix produced by the DIP method was comprehensively examined. The fuel matrix met the technical requirements and its anisotropy was significantly reduced. The DIP method can significantly improve both the production efficiency and quality of fuel elements. This will play a key role in meeting the huge demand for fuel elements of HTRs and molten salt reactors.

Research on graphite powders used for HTR-PM fuel elements

Different batches of natural graphite powders and electrographite powders were characterized by impurity, degree of graphitization, particle size distribution, specific surface area, and shape characteristics. The graphite balls consist of proper mix-ratio of natural graphite, electrographite and phenolic resin were manufactured and characterized by thermal conductivity, anisotropy of thermal expansion, crush strength, and drop strength. Results show that some types of graphite powders possess very high purity, degree of graphitization, and sound size distribution and apparent density, which can serve for matrix graphite of HTR-PM. The graphite balls manufactured with reasonable mix-ratio of graphite powders and process method show very good properties. It is indicated that the properties of graphite balls can meet the design criterion of HTR-PM. We can provide a powerful candidate material for the future manufacture of HTR-PM fuel elements.

PD Power-Level Control Design for MHTGRs

Due to its inherent safety feature, the modular high temperature gas-cooled reactor (MHTGR) has been seen as one of the best candidates in building next generation nuclear plants (NGNPs). Since the MHTGR dynamics has high nonlinearity, it is necessary to develop nonlinear power-level controller which is not only beneficial to the safe, stable, efficient and autonomous operation of the MHTGR but also easy to be implemented practically. In this paper, based on the concept of shiftedectropy and the physically-based control design approach, it is proved theoretically that the simple proportional-differential (PD) output-feedback power-level control can provide globally asymptotic closed-loop stability. Numerical simulation results verify the theoretical results and show the influence of the controller parameters to the dynamic response.

The Module HTGR Development in China

High Temperature Gas-cooled Reactors are recognized as a representative advanced nuclear system for the future owing to the excellent safety performance,high efficiency,multipurpose uses and hydrogen production.These type reactors are characterized by ceramic coated particle fuel,inert helium as coolant,and graphite used as moderator and reflector in core,which makes the outlet temperature of coolant reaching 950℃ even more.Under the National High Technology Program,the HTR-10 project has been successfully implemented and achieved full power operation in connection with the grid in January of 2003.HTR-10,which is the first module HTR with inherent safety feature around world,has carried out safety demonstration tests simulating the severe accident conditions in 2004.Based on the proven technologies and experience feedback during HTR-10 design,manufacture,construction and operation,a HTR-PM demonstration power plant with 200MWe power capacity sited at Rongcheng of Shandong province has been initiated.

Evaluation of thermophoretic effects on graphite dust coagulation in high-temperature gas-cooled reactors

In high-temperature gas-cooled reactors,graphite dust particles within the reactor core and the heat transfer equipment experience large temperature gradients.Under such conditions,thermophoresis may play an important role in determining aerosol evolution.This study presents a theoretical and numerical analysis of the thermophoretic effects on aerosol coagulation within these reactors.The coagulation rates for Brownian versus thermophoretic coagulation are calculated and compared for various temperature gradients.Our results show that thermophoretic coagulation dominates over Brownian coagulation for large temperature gradients.We defined an enhancement factor to evaluate the role of thermophoretic coagulation under various reactor conditions.The enhancement factor increased dramatically with increasing temperature gradient,decreasing pressure and increasing particle diameter,but was not very sensitive to temperature change.The time evolution of the particle size distribution related to combined Brownian and thermophoretic coagulation was simulated using a log-skew-normal method of moments.The simulation results indicate that aerosol evolution can be strongly accelerated by thermophoretic coagulation under large temperature gradients.

Simultaneous approach for simulation of a high-temperature gas-cooled reactor

The simulation of a high-temperature gas-cooled reactor pebble-bed module(HTR-PM) plant is discussed.This lumped parameter model has the form of a set differential algebraic equations(DAEs) that include stiff equations to model point neutron kinetics.The nested approach is the most common method to solve DAE,but this approach is very expensive and time-consuming due to inner iterations.This paper deals with an alternative approach in which a simultaneous solution method is used.The DAEs are discretized over a time horizon using collocation on finite elements,and Radau collocation points are applied.The resulting nonlinear algebraic equations can be solved by existing solvers.The discrete algorithm is discussed in detail;both accuracy and stability issues are considered.Finally,the simulation results are presented to validate the efficiency and accuracy of the simultaneous approach that takes much less time than the nested one.

Dynamic simulation of a space gas-cooled reactor power system with a closed Brayton cycle

Space nuclear reactor power(SNRP)using a gas-cooled reactor(GCR)and a closed Brayton cycle(CBC)is the ideal choice for future high-power space missions.To investigate the safety characteristics and develop the control strategies for gas-cooled SNRP,transient models for GCR,energy conversion unit,pipes,heat exchangers,pump and heat pipe radiator are established and a system analysis code is developed in this paper.Then,analyses of several operation conditions are performed using this code.In full-power steady-state operation,the core hot spot of 1293 K occurs near the upper part of the core.If 0.4$reactivity is introduced into the core,the maximum temperature that the fuel can reach is 2059 K,which is 914 K lower than the fuel melting point.The system finally has the ability to achieve a new steady-state with a higher reactor power.When the GCR is shut down in an emergency,the residual heat of the reactor can be removed through the conduction of the core and radiation heat transfer.The results indicate that the designed GCR is inherently safe owing to its negative reactivity feedback and passive decay heat removal.This paper may provide valuable references for safety design and analysis of the gas-cooled SNRP coupled with CBC.

Safety Features of Modular High Temperature Gas-cooled Reactors (MHTGR)

The following design features which satisfy fundamental safety design objectives of an MHTGR are analyzed: (i) inherent safety features to reactivity effect: (ii) passive decay heat removal: and (iii) multiple barriers.Several events have been identified to be the bounding. hypothetical accidents for the MHTGR. The important accident sequences leading to severe accidents are ingress of a large amount of water or air into the core. The analyses of severe accident scenarios have shown that even the harm of fuel element predicted to occur by chmeical reaction after a hypothetical large amount of water ingress into the core or air ingress into the core will not result in major impact on the environment due to the nitegrity of fuel particles remained. Therefore, it would not be necessary to require an emergency plan to evacuate nearby inhabitants.

Experimental and mechanistic study of dispersed micrometer-sized particle resuspension in a square straight duct with rough walls

The resuspension of graphite dust is an important phenomenon in the release of radioactivity and the safety of nuclear reactors during severe accidents.In this study,a visualization experimental platform is constructed to study effects of particle size,flow velocity,and wall roughness on the resuspension characteristics of graphite particles.A statistical model of particle resuspension applicable to monolayer dispersed particles is developed based on the moment equilibrium of the particles and the flow field characteristics,as calculated by the large-eddy simulation framework.The results show that particle resuspension can be divided into short-and long-term resuspension stages.Most particle resuspension occurs during the short-term stage.With increases in flow velocity and particle diameter,the aerodynamic or adhesion force acting on the particles increases,and corresponding particle resuspension fraction increases.The influence of rough walls on particle resuspension is related to both the force on the particles and the arm ratio between the wall morphology and the particle diameter.A comparison with the experimental results demonstrates that the particle resuspension model developed in this study accurately predicts the impact of flow velocity,particle size,and wall roughness on particle resuspension.

Graphite dust resuspension in an HTR-10 steam generator

Graphite dust has an important effect on the safety of high-temperature gas-cooled reactors（HTR）.The flow field in the steam generator was studied by the computational fluid dynamics（CFD） method,with the results indicating that the friction velocity in the windward and the leeward of the heat transfer tubes is relatively low and is higher at the sides.Further analysis of the resuspension of graphite dust indicates that the resuspension fraction reaches nearly zero for particles with a diameter less than 1 μm,whereas it will increases as the helium velocity in the steam generator increases for particle size larger than 1 μm.Moreover,the resuspension fraction increases as the particle size increases.The results also indicate that resuspension of the particles with sizes larger than 1 μm exhibited obvious differences in different parts of the steam generator.

Graphite dust deposition in the HTR-10 steam generator

Graphite dust has an important effect on the safe operation of a high-temperature gas-cooled reactor （HTR）. The present study analyzes temperature and flow field distributions in the HTR-10 steam genera- tor. The temperature and flow field distributions are then used to study thermophoretic deposition and turbulent deposition. The results show that as the dust diameter increases, the thermophoretic deposition decreases, while the turbulent deposition first decreases and then increases. The thermophoretic deposi- tion is higher at higher reactor powers, with turbulent deposition growing more rapidly at higher reactor power. For small particles, the thermophoretic deposition effect is greater than the turbulent deposition effect, while for large particles, the turbulent deposition effect is dominant.

Research on the Silicon Carbide Layer of Coated Fuel Particles

The 10MW high temperature gas-cooled test reactor (HTR-10) under construction at INET uses whole ceramic fuel elements. The main barrier which prevents fission product release is the SiC layer of the coated fuel particles. Fabrication of high quality SiC layers is one of the key R&D tasks for the HTR-10 fuel element. The SiClayer was deposited on the fuel particles in a 50 mm conical fluidized bed using the CVD (chemical vapour deposition) technique. The density, thickness, strength and elastic modulus of the SiC layer were measured. The microstructure was observed using SEM (scanning electron microscope ). Parameters were established for manufacturing the SiC layer of the coated fuel particles to be used in the HTR-10. It was found that the traditional density measurement by the sink-float method is questionable in the low density region and that the SiC layer may be contaminated by uranium under certain conditions.

Thermochemical Water Splitting for Hydrogen Production Utilizing Nuclear Heat from an HTGR

A very promising technology to achieve a carbon free energy system is to produce hydrogen from water, rather than from fossil fuels. Iodine-sulfur (IS) thermochemical water decomposition is one promising process. The IS process can be used to efficiently produce hydrogen using the high temperature gas-cooled reactor (HTGR) as the energy source supplying gas at 1000℃. This paper describes that dem- onstration experiment for hydrogen production was carried out by an IS process at a laboratory scale. The results confirmed the feasibility of the closed-loop operation for recycling all the reactants besides the water, H2, and O2. Then the membrane technology was developed to enhance the decomposition efficiency. The maximum attainable one-pass conversion rate of HI exceeds 90% by membrane technology, whereas the equilibrium rate is about 20%.

Oxidation Behaviour of the Matrix Materials in Spherical Fuel Elements

The oxidation resistance of the matrix materials is vital to the normal operation of HTGR and is also an important parameter for evaluating the safety response under accidental air or water ingress conditions. The oxidation kinetics of the three matrix material components: natural graphite, artificial graphite and resin carbon. was studied in a flowing gas mixture of oxygen and nitrogen using an auto thermogravimetric system. The results indicate that the artificial graphite has the slowest oxidation rate followed by the natural graphite and then the resin carbon with the highest oxidation rate. Vacuum heat treatment of the natural graphite at 1950℃ decreases the impurities and increases the oxidation activation energy. Differences between the activation energy and the oxidation rate of the resin carbon heat treated at 1950℃ and 1600℃ resulted from changes in the micro-pore texture. and the reduction of impurities.

Feasibility of Burning Civilian Grade Pu in the Modular HTR with Th Fuel Cycle

The Modular High Temperature Gas-Cooled Reactor (HTR) can be used to burn plutonium fuel to reduce Pu stockpiles because of its inherent safety characteristics and ability to burn a variety of fuel mixtures. The equilibrium core is calculated and analyzed for Pu enriched fuel. Fuel spheres with 7g heavy metal including the civilian grade Pu and thorium are loaded into the reactor. An enrichment of 11% is chosen to provide the desired equilibrium core reactivity. The fuel and moderator temperature coefficients are both negative. The maximum fuel element temperature during normal operation and during a loss of coolant accident is less than 1500℃. 92% of 239 Pu will be burnt during nomal operation. Therefore, a thorium fuel cycle in the modular HTR is an effective method for burning civilian grade plutonium.