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 a...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.展开更多
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(HE...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.展开更多
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...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.展开更多
Since its first successful use in the CP-1 nuclear reactor in 1942,nuclear graphite has played an important role in nucle-ar reactors especially the high temperature gas-cooled type(HTGRs)owing to its outstanding comp...Since its first successful use in the CP-1 nuclear reactor in 1942,nuclear graphite has played an important role in nucle-ar reactors especially the high temperature gas-cooled type(HTGRs)owing to its outstanding comprehensive nuclear properties.As the most promising candidate for generation IV reactors,HTGRs have two main designs,the pebble bed reactor and the prismatic re-actor.In both designs,the graphite acts as the moderator,fuel matrix,and a major core structural component.However,the me-chanical and thermal properties of graphite are generally reduced by the high fluences of neutron irradiation of during reactor opera-tion,making graphite more susceptible to failure after a significant neutron dose.Since the starting raw materials such as the cokes and the subsequent forming method play a critical role in determining the structure and corresponding properties and performance of graphite under irradiation,the judicious selection of high-purity raw materials,forming method,graphitization temperature and any halogen purification are required to obtain the desired properties such as the purity and isotropy.The microstructural and correspond-ing dimensional changes under irradiation are the underlying mechanism for the changes of most thermal and mechanical properties of graphite,and irradiation temperature and neutron fluence play key roles in determining the microstructural and property changes of the graphite.In this paper,the basic requirements of nuclear graphite as a moderator for HTGRs and its manufacturing process are presented.In addition,changes in the mechanical and thermal properties of graphite at different temperatures and under different neutron fluences are elaborated.Furthermore,the current status of nuclear graphite development in China and abroad is discussed,and long-term problems regarding nuclear graphite such as the sustainable and stable supply of cokes as well as the recycling of used material are discussed.This paper is intended to act as a reference for graphite providers who are interested in developing nuclear graphite for potential applications in future commercial Chinese HTGRs.展开更多
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...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.展开更多
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...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.展开更多
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 condit...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.展开更多
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 ...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.展开更多
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...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.展开更多
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 barrier...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.展开更多
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 Mon...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.展开更多
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 e...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.展开更多
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...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.展开更多
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 labo...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.展开更多
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 f...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.展开更多
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 gra...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.展开更多
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 ...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.展开更多
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 hea...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.展开更多
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 tes...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.展开更多
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 te...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.展开更多
基金supported by the Ministry of Education,Science and Technology (MEST) of Korea
文摘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.
文摘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.
文摘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.
文摘Since its first successful use in the CP-1 nuclear reactor in 1942,nuclear graphite has played an important role in nucle-ar reactors especially the high temperature gas-cooled type(HTGRs)owing to its outstanding comprehensive nuclear properties.As the most promising candidate for generation IV reactors,HTGRs have two main designs,the pebble bed reactor and the prismatic re-actor.In both designs,the graphite acts as the moderator,fuel matrix,and a major core structural component.However,the me-chanical and thermal properties of graphite are generally reduced by the high fluences of neutron irradiation of during reactor opera-tion,making graphite more susceptible to failure after a significant neutron dose.Since the starting raw materials such as the cokes and the subsequent forming method play a critical role in determining the structure and corresponding properties and performance of graphite under irradiation,the judicious selection of high-purity raw materials,forming method,graphitization temperature and any halogen purification are required to obtain the desired properties such as the purity and isotropy.The microstructural and correspond-ing dimensional changes under irradiation are the underlying mechanism for the changes of most thermal and mechanical properties of graphite,and irradiation temperature and neutron fluence play key roles in determining the microstructural and property changes of the graphite.In this paper,the basic requirements of nuclear graphite as a moderator for HTGRs and its manufacturing process are presented.In addition,changes in the mechanical and thermal properties of graphite at different temperatures and under different neutron fluences are elaborated.Furthermore,the current status of nuclear graphite development in China and abroad is discussed,and long-term problems regarding nuclear graphite such as the sustainable and stable supply of cokes as well as the recycling of used material are discussed.This paper is intended to act as a reference for graphite providers who are interested in developing nuclear graphite for potential applications in future commercial Chinese HTGRs.
文摘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.
文摘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.
基金supported by the MEST/NRF (Nuclear R&D Program,2005-2004718 and 2009 0083392) of Korea
文摘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.
文摘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.
文摘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.
文摘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.
基金Supported by National Science and Technology Major Project(ZX06901)National Natural Science Foundation of China(10975083,11079025)Tsinghua University Initiative Scientific Research Program
文摘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.
文摘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.
文摘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.
基金the High Technology Research and Development Programme of china
文摘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.
文摘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.
文摘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.
文摘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.
文摘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.
文摘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.
文摘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.
