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.

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.

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.

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.

Two Phase Flow Stability in the HTR-10 Steam Generator

A 10MW High Temperature Gas Cooled Reactor (HTR-10) designed by the Institute of Nuclear Energy Technology (INET) is now being constructed. The steam generator (SG) in the HTR-10 is one of the most important components for reactor safety. The thermal-hydraulic performance of the SG was investigated. A full scale HTR-10 Steam Generator Two Tube Engineering Model Test Facility (SGTM-10) was installed and tested at INET. This paper describes the SGTM-10 thermal hydraulic experimental system in detail. The SGTM-10 simulates the actual thermal and structural parameters of the HTR-10. The SGTM-10 includes three separated loops: the primary helium loop, the secondary water loop, and the tertiary cooling water loop. Two parallel tubes are arranged in the test assembly. The main experimental equipment is shown in the paper. Expermental results are given illustrating the effects of the outlet pressures, the heating power, and the inlet subcooling.