Thermal Performance Test of the Hot Gas Duct of10MW High Temperature Reactor Test Modulegh

he thermal performance test of the horizontal coaxial double tube hot gas duct (HGD) with an internal thermal insulation for the 10MW High Temperature Reactor Test Module (HTR10) was conducted on a Helium Test Loop(HETL). The present paper deals with the technical feature of the HETL, the test section and the thermal performance test of the HGD. The HGD test section with a triple tube structure includes an inner heater, a HGD model and a coldhot gas mixer. A counterflow of cold and hot helium gas under the pressure of about 3.0 MPa and the minimum temperature of 100℃ in the annular passage and the maximum of 950℃ in the central tube of the HGD model was formed. The HGD model was undergone 20 times of pressure cycle test under the pressure ranging from 0.1 to 3.4 MPa, 18 times of the temperature cycle test under the temperature ranging from 100 to 950℃ and high temperature (700 to 950℃) helium flow test for a period of more than 350 hours. The effective thermal conductivity (λeff) of the internal insulation of the HGD was investigated experimentally. The relationship of the effective thermal conductivity with the average tmperature of the internal insulation layer is λeff(W/m/℃)=0.3512+0.0003T(℃). The test results indicate that the HGD model has good abilities to resist heat flux from the central tube to the annular passage, temperature variations, and pressure variations.

Material Characterization of Ni Base Alloy for Very High Temperature Reactor

The generation of highly efficient electricity and the production of massive hydrogen are possible using a very high temperature reactor （VHTR） among generation IV nuclear power plants. The structural material for an intermediate heat exchanger （IHX） among numerous components should be endurable at high temperature of up to 950 °C during long-term operation. Impurities inevitably introduced in helium as a coolant facilitate the material degradation by corrosion at high temperature. In the present work, the surface reactions available under controlled impure helium at 950 °C were investigated based on the thermodynamics and the corrosion tests were performed in a temperature range of 850-950 °C during 10-250 h for commercial Alloy 617 as a candidate material for an IHX. Moreover, the mechanical property and microstructure for nickel-based alloys fabricated in laboratory were evaluated as a function of the processing parameters such as hot rolling and heat treatment conditions. From the reaction rate constant obtained from an impure helium control system for a material evaluation, it was predicted that the outer oxide layer thickness, internal oxide depth, and carbide- depleted zone depth reach about 116, 600 and 1000 μm, respectively when Alloy 617 is exposed to an impure helium environment at 950 ~C for 20 years. For Ni-Cr-Co-Mo alloy, subsequent annealing and a combination of cold working and subsequent annealing following solution annealing caused increases in the grain boundary carbide coverage and size. The angular distribution of the grain boundary as well as the carbide distribution was also changed leading to a consequent improvement of the mechanical property at 950 °C in air.

Thorium-Based Fuel Cycles in the Modular High Temperature Reactor

Large stockpiles of civil-grade as well as weapons-grade plutonium have been accumulated in the world from nuclear power or other programs of different countries. One alternative for the management of the plutonium is to incinerate it in the high temperature reactor （HTR）. The thorium-based fuel cycle was studied in the modular HTR to reduce weapons-grade plutonium stockpiles, while producing no additional plutonium or other transuranic elements. Three thorium-uranium fuel cycles were also investigated. The thorium absorption cross sections of the resolved and unresolved resonances were generated using the ZUTDGL code based on existing resonance data. The equilibrium core of the modular HTR was calculated and analyzed by means of the code VSOP＇94. The results show that the modular HTR can incinerate most of the initially loaded plutonium amounting to about 95.3% net 239pu for weapons-grade plutonium and can effectively utilize the uranium and thorium in the thorium-uranium fuel cycles.

The Shandong Shidao Bay 200 MW_e High-Temperature Gas-Cooled Reactor Pebble-Bed Module(HTR-PM) Demonstration Power Plant: An Engineering and Technological Innovation

In 2005, the US passed the Energy Policy Act of 2005 mandating the construction and operation of a high-temperature gas reactor （HTGR） by 2021. This law was passed after a multiyear study by national experts on what future nuclear technologies should be developed. As a result of the Act, the US Congress chose to develop the so-called Next-Generation Nuclear Plant, which was to be an HTGR designed to produce process heat for hydrogen production. Despite high hopes and expectations, the current status is that high temperature reactors have been relegated to completing research programs on advanced fuels, graphite and materials with no plans to build a demonstration plant as required by the US Con- gress in 2005. There are many reasons behind this diminution of HTGR development, including but not limited to insufficient government funding requirements for research, unrealistically high temperature requirements for the reactor, the delay in the need for a ＂hydrogen＂ economy, competition from light water small modular light water reactors, little utility interest in new technologies, very low natural gas prices in the US, and a challenging licensing process in the US for non-water reactors.

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.

Oxidation behaviors of wrought nickel-based superalloys in various high temperature environments

Oxidation characteristics of Alloy 617 and Haynes 230 at 900 oC in simulated helium environment,hot steam environment containing H2 as well as in air and pure helium conditions were investigated.Compared to air condition,the oxidation rate of Alloy 617 was not significantly affected in helium and hot steam environments,while Haynes 230 showed lower oxidation rate in helium environment.On the other hand,the oxide morphology and structure of Alloy 617 were strongly affected by the environments,but those of Haynes 230 were less dependent on the environments.For Haynes 230,a Cr2O3 inner layer and a protective MnCr2O4 outer layer were formed in all environments,which contributed to the better oxidation resistance.As the mechanical properties,such as creep and tensile properties,were significantly affected by the oxidation behaviors,surface treatment methods to enhance oxidation resistance of these alloys should be developed.

Long-Term High-Temperature Oxidation of Alloys for Intermediate Heat Exchangers

Alloy 617 is the reference candidate material for high temperature components of gas-cooled reactors, like intermediate heat exchangers. Oxidation tests were performed with two heats of Alloy 617 up to 5000 hours at 950℃ under a simulated helium-cooled reactor environment. Post-treatment examination showed that all materials actually oxidized during the tests with the growth of a surface chromium oxide scale that includes titanium, formation of a carbide-depleted zone underneath the surface, and internal oxidation of aluminum. These oxidation-related phenomena are in good agreement with the data published in the 1980s for Alloy 617 in equivalent testing conditions and were used to assess the alloy corrosion performances. The oxidation kinetics was globally parabolic corresponding to the growth of the external oxide as well as to internal oxidation. In the given test environment, the parabolic rate constants are 0.00090 and 0.00058 mg^2·c^-4m·h^-1 for the two heats of Alloy 617.

Wide Range Neutron Monitoring(WRNM)System in Boiling Water Reactors(A Short Communication&Memorandum)

The WRNM(wide range neutron monitoring)is a newly developed neutron monitoring channel which was initially conceived as a means to meet Regulatory Guide 1.97 requirements for post-accident neutron monitoring.The scope was expanded to include the startup monitoring function with the aim of replacing both the source and IRMs(intermediate range monitors)in BWRs(boiling water reactors).The WRNMs,consisting of a newly designed fixed incore regenerative sensor and new electronics,which include both counting and MSV(mean square voltage)channels,have been tested in several reactors and its capabilities have been confirmed.The channel will cover the neutron flux range from 103 nv to 1.5×103 nv;it has greater than 1 decade overlap between the counting and MSV channels.Because of the regenerative fissile coating the sensor,even though fixed incore,has a life of approximately 6.0 full power years in a 51 kW/L BWR and similar situation has been proposed for newly designed small modular reactor such as BWRX-300 of General Electric Hitachi reactor.

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.

Monte Carlo studies on the burnup measurement for the high temperature gas cooling reactor

Online fuel pebble burnup measurement in a future high temperature gas cooling reactor is proposed for implementation through a high purity germanimn （HPGe） gamma spectrometer. By using KORIGEN software and MCNP Monte Carlo simulations, the single pebble gamma radiations to be recorded in the detector are simulated under different, irradiation histories. A specially developed algorithm is applied to analyze the generated spectra to reconstruct the gamma activity of the ~arCs monitoring nuclide. It is demonstrated that by taking into account the intense interfering peaks, the 137Cs activity in the spent pebbles can be derived with a standard deviation of 3.0% （l（r）. The results support the feasibility of utilizing the HPGe spectrometry in the online determination of the pebble burimp in future modular pebble bed reactors.

Self-acting Afterheat Removal in High Temperature Gas Cooled Reactors

Much more nuclear energy capacity is needed than currently installed to meet the demand of energy and the requirement on environment protection in the next decades. More stringent nuclear safety standards have to be established for future nuclear power plants.The philosophy of a catastrophe free nuclear technology is presented in this paper. The issue of afterheat removal of high temperature gas cooled reactors is handled.It is a striking inherent safety feature of the modular high temperature gas cooled reactor design that the afterheat removal takes place without any active core cooling systems.

A study of function mechanism of hemxamethyl tetra-amine in gelation process of uranium

The UO2 ceramic microspheres are the most important materials in the spherical fuel elements for high temperature reactor (HTR). A process for preparation of UO2 kernels known as total gelation process of uranium (TGU) was developed as the production process of 10 mW HTR at Tsinghua University. The TGU process is based on the traditional sol-gel process, external gelation process and internal gelation process of uranium (EGU and IGU), which implies that the gelation action is initiated both by ammonia out of the gel particles and hemxamethyl tetra-amine (HMTA) inside the gel particles. The gelation behavior and the properties of uranium microspheres were investigated of the solution with and without HMTA. It is observed that good spherical particles can be obtained without HMTA in the sol, which indicates a more controllable and industrialized route will be set up. Contrasts between this route and the traditional EGU were also listed.

Manufacture of HTR Fuel Element

The HTR Fuel Element R & D Program,set in 1987,aims to develop the manufacturetechnology of HTR fuel element and to produce the fuel element for the first core of our 10MW experimental reactor.Now the work on laboratory scale is phased out.In this paper,the fuel element manufacture technology is described and the test results are given.

A review of TRISO-coated particle nuclear fuel performance models

The success of high temperature gas cooled reactor depends upon the safety and quality of the coated particle fuel. The understanding and evaluation of this fuel requires the development of an integrated mechanistic fuel performance model that fully describes the mechanical and physicochemical behavior of the fuel particle under irradiation. In this paper, a review of the analytical capability of some of the existing computer codes for coated particle fuel was performed. These existing models and codes include FZJ model, JAERI model, Stress3 model, ATLAS model, PARFUME model and TIMCOAT model. The theoretic model, methodology, calculation parameters and benchmark of these codes were classified. Based on the failure mechanism of coated particle, the advantage and limits of the models were compared and discussed. The calculated results of the coated particles for China HTR-10 by using some existing code are shown. Finally, problems and challenges in fuel performance modeling were listed.

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.

Application of Finite Difference Method to the Analysis of HTTR Reactor Cavity Natural Convection

An attempt is made to investigate the ability of the revised K FIX code to the solution to natural convection. The new experimental data of the benchmark problem of the IAEA coordinated research program (CRP 3) on heat transport and afterheat removal for GCRs under accident conditions provided by JAERI are used to calculate nitrogen natural convection in the pressurized vessel and air natural convection in the reactor cavity by using this revised code. Based on analysis, a refined mesh is used to solve the differential equations so as to get more detailed and more accurate result. The obtained velocity profiles are consistent with the result of TRIO EF code and the result of Bechtel laboratory. It can be drawn that the revised K FIX code can be used to solve this kind of problems.

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.

Model Tests of Coolant Thermal Mixing in the HTR-10 Hot Gas Plenum

The coolant thermal mixing performance in the hot gas plenum (HGP) in the core bottom reflector of the 10MW high temperature gas-cooled reactor test module (HTR-10) was experimentally investigated on a 1:1.5 scale test model. The experimental results show that the HGP installed with a radial partition static mixer results in excellent thermal mixing of the coolant with a nondimensional temperature mixing degree (ε) value of 94%. Within the Re range from 1. 4 ×105～5. 8×105, the ε value reaches 94% at the outlet of the HGP and 99% at the outlet of the hot gas duct (HGD). There is little influence of the inlet flow rate ratio. Gh/Ge. on the thermal mixing performance in the Gh/Ge range from 0.5～2.0.

HTGR Process Heat Application Study

The 10MW high temperature gas-cooled reactor test module (HTR-10) is currently under construction.One of its objectives is to develop high temperature process heat applications. To realize this target, various high temperature gas-cooled reactor (HTGR) process heat applications have been analyzed. This paper briefly describes the possibilities and experimental schemes for using the HTR-10 for process heat application studies.

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.