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.

Numerical investigations of thermal mixing performance of a hot gas mixing structure in high-temperature gas-cooled reactor

A numerical simulation study was performed to clarify the thermal mixing characteristics of coolant in the core bottom structure of the high-temperature gas-cooled reactor(HTR). The flow field and temperature field in the hot gas chamber and the hot gas duct of the HTR were obtained based on the commercial computational fluid dynamics(CFD) program. The numerical simulation results showed that the helium flow with different temperatures in the hot gas mixing chamber and the hot gas duct mixed intensively, and the mixing rate of the temperature in the outlet of the hot gas duct reached 98 %. The results indicated many large-scale swirling flow structures and strong turbulence in the hot gas mixing chamber and the entrance of the hot gas duct, which were responsible for the excellent thermal mixing of the hot gas chamber and the hot gas duct. The calculated results showed that the temperature mixing rate of the hot gas chamber decreased only marginally with increasing Reynolds number.

Design of the material performance test apparatus for high temperature gas-cooled reactor

Most materials can be easily corroded or ineffective in carbonaceous atmospheres at high temperatures in the reactor core of the high temperature gas-cooled reactor(HTGR).To solve the problem,a material performance test apparatus was built to provide reliable materials and technical support for relevant experiments of the HTGR.The apparatus uses a center high-purity graphite heater and surrounding thermal insulating layers made of carbon fiber felt to form a strong carbon reducing atmosphere inside the apparatus.Specially designed tungsten rhenium thermocouples which can endure high temperatures in carbonaceous atmospheres are used to control the temperature field.A typical experimental process was analyzed in the paper,which lasted 76 hours including seven stages.Experimental results showed the test apparatus could completely simulate the carbon reduction atmosphere and high temperature environment the same as that confronted in the real reactor and the performance of screened materials had been successfully tested and verified.Test temperature in the apparatus could be elevated up to 1600oC,which covered the whole temperature range of the normal operation and accident condition of HTGR and could fully meet the test requirements of materials used in the reactor.

Analysis on thermophoretic deposit of fine particle on water wall of 10 MW high temperature gas-cooled reactor

The water wall is an important part of the passive natural circulation residual heat removal system in a high temperature gas-cooled reactor. The maximum temperatures of the pressure shell and the water wall are calcu- lated using annular vertical closed cavity model. Fine particles can deposit on the water wall due to the thermophore- sis effect. This deposit can affect heat transfer. The thermophoretic deposit efficiency is calculated by using Batch and Shen’s formula fitted for both laminar flow and turbulent flow. The calculated results indicate that natural convection is turbulent in the closed cavity. The transient thermophoretic deposit efficiency rises with the increase of the pressure shell’s temperature. Its maximum value is 14%.

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.

Study on neutronics design of ordered-pebble-bed fluoride-salt- cooled high-temperature experimental reactor

This paper presents a neutronics design of a 10 MW ordered-pebble-bed fluoride-salt-cooled high-temperature experimental reactor. Through delicate layout, a core with ordered arranged pebble bed can be formed,which can keep core stability and meet the space requirements for thermal hydraulics and neutronics measurements.Overall, objectives of the core include inherent safety and sufficient excess reactivity providing 120 effective full power days for experiments. Considering the requirements above, the reactive control system is designed to consist of 16 control rods distributed in the graphite reflector. Combining the large control rods worth about 18000–20000 pcm, molten salt drain supplementary means(-6980 to -3651 pcm) and negative temperature coefficient(-6.32 to -3.80 pcm/K) feedback of the whole core, the reactor can realize sufficient shutdown margin and safety under steady state. Besides, some main physical properties, such as reactivity control, neutron spectrum and flux, power density distribution, and reactivity coefficient,have been calculated and analyzed in this study. In addition, some special problems in molten salt coolant are also considered, including ~6Li depletion and tritium production.

Framework analysis of fluoride salt-cooled high temperature reactor probabilistic safety assessment

Probabilistic safety assessment(PSA) is important in nuclear safety review and analysis. Because the design and physics of the fluoride salt-cooled high temperature reactor(FHR) differ greatly from the pressurized water reactor(PWR), the methods and steps of PSA in FHR should be studied. The high-temperature gascooled reactor(HTR-PM) and sodium-cooled fast reactors have built the PSA framework, and the framework to finish the PSA analysis. The FHR is compared with the PWR, HTR-PM and sodium-cooled fast reactors from the physics, design and safety. The PSA framework of FHR is discussed. In the FHR, the fuel and coolant combination provides large thermal margins to fuel damage(hundreds of degrees centigrade). The tristructuralisotropic(TRISO) as the fuel is independent in FHR core and its failure is limited for the core. The core damage in Level 1 PSA is of lower frequency. Levels 1 and 2 PSA are combined in the FHR PSA analysis. The initiating events analysis is the beginning, and the source term analysis and the release types are the target. Finally, Level3 PSA is done.

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.

Gearbox Scheme in High Temperature Reactor Helium Gas Turbine System

Helium Turbine is used in High Temperature Reactor Helium Gas Turbine (HTR-GT) system, by which the direct helium circulation between the reactor and turbine generator system will come true. Between helium turbine and generator, there is gearbox device which reduces the turbine rotation speed to normal speed required by the generator. Three optional gearbox schemes are discussed. The first is single reduction cylindrical gearbox, which consists of one high speed gear and one low speed gear. Its advantage is simple structure, easy to manufacture, and high reliability, while its disadvantage is large volume and misalignment of input and output axle. The second is planetary gear mechanism with static planet carrier. The third is planetary gear mechanism with static internal gear. The latter two gearbox devices have similar structure. Their advantage is small volume and high reduction gear ratio, while disadvantage are complicated structure, many gears, low reliability and low mechanical efficiency.

The AVR as Small Modular Thorium Very High Temperature Reactor:Experiences-Design-Safety-Fuel Cycle

As a young engineer in the power plant department of Brown Boveri,Dr.Schulten had the idea to design nuclear power stations without major risk.The following requirements must be accomplished:ŸA negative temperature coefficient had to avoid an MCA(Maximum Credible Accident);ŸCeramic materials for core construction and fuel elements;ŸA homogenous mixture of nuclear fuel and graphite had to be able to use uranium and thorium as breeding material;ŸThe produced high temperature heat shall be the basis for production of electricity,drinking water,hydrogen,etc.;ŸA relatively simple plant,which could be operated in developing countries,to cogenerate electricity and heat;ŸHelium used as cooling gas.

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.

Macroscopic Structural Analysis on a 10 kW Class Lab-Scale Process Heat Exchanger Prototype under a High-Temperature Gas Loop Condition

A PHE (Process Heat Exchanger) is a key component in transferring high-temperature heat generated from a VHTR (Very High Temperature Reactor) to a chemical reaction for the massive production of hydrogen. Last year, a 10 kW class lab-scale PHE prototype made of Hastelloy-X was manufactured at the Korea Atomic Energy Research Institute (KAERI), and a performance test of the PHE prototype is currently underway in a small-scale nitrogen gas loop at KAERI. The PHE prototype is composed of two kinds of flow plates: grooves 1.0 mm in diameter machined into the flow plate for the primary coolant, and waved channels bent into the flow plate for the secondary coolant. Inside the 10 kW class lab-scale PHE prototype, twenty flow plates for the primary and secondary coolants are stacked in turn. In this study, to understand the macroscopic structural behavior of the PHE prototype under the steady-state operating condition of the gas loop, high-temperature structural analyses on the 10 kW class lab-scale PHE prototype were performed for two extreme cases: in the event of contacting the flow plates together, and when not contacting them. The analysis results for the extreme cases were also compared.

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.

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.

Conceptual design and safety characteristics of a new multi-mission high flux research reactor

Research reactors with neutron fluxes higher than 10^(14) n cm^(−2) s^(−1) are widely used in nuclear fuel and material irradiation,neutron-based scientific research,and medical and industrial isotope production.Such high flux research reactors are not only important scientific research facilities for the development of nuclear energy but also represent the national comprehensive technical capability.China has several high flux research reactors that do not satisfy the requirements of nuclear energy development.A high flux research reactor has the following features:a compact core arrangement,high power density,plate-type fuel elements,a short refueling cycle,and high coolant velocity in the core.These characteristics make it difficult to simultaneously realize high neutron flux and optimal safety margin.A new multi-mission high flux research reactor was designed by the Institute of Nuclear and New Energy Technology at Tsinghua University in China;the reactor can simul-taneously realize an average neutron flux higher than 2.0×10^(15) n cm^(−2) s^(−1) and fulfill the current safety criterion.This high flux research reactor features advanced design concepts and has sufficient safety margins according to the preliminary safety analysis.Based on the analysis of the station blackout accident,loss of coolant accident,and reactivity accident of a single-control drum rotating out accidently,the maximum temperature of the cladding surface,minimum departure from nucleate boiling ratio,and temperature difference to the onset of nucleate boiling temperature satisfy the design limits.

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.

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.

Simulation of radiation dose distribution and thermal analysis for the bulk shielding of an optimized molten salt reactor

The Chinese Academy of Science has launched a thorium-based molten-salt reactor(TMSR)research project with a mission to research and develop a fission energy system of the fourth generation.The TMSR project intends to construct a liquid fuel molten-salt reactor(TMSR-LF),which uses fluoride salt as both the fuel and coolant,and a solid fuel molten-salt reactor(TMSR-SF),which uses fluoride salt as coolant and TRISO fuel.An optimized 2 MWth TMSR-LF has been designed to solve major technological challenges in the Th-U fuel cycle.Preliminary conceptual shielding design has also been performed to develop bulk shielding.In this study,the radiation dose and temperature distribution of the shielding bulk due to the core were simulated and analyzed by performing Monte Carlo simulations and computational fluid dynamics(CFD)analysis.The MCNP calculated dose rate and neutron and gamma spectra indicate that the total dose rate due to the core at the external surface of the concrete wall was 1.91μSv/h in the radial direction,1.16μSv/h above and 1.33μSv/h below the bulk shielding.All the radiation dose rates due to the core were below the design criteria.Thermal analysis results show that the temperature at the outermost surface of the bulk shielding was 333.86 K,which was below the required limit value.The results indicate that the designed bulk shielding satisfies the radiation shielding requirements for the 2 MWth TMSR-LF.

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.