The recent development of molten salt fast reactors has generated a renewed interest in them. As compared to traditional solid fuel fast neutron systems, it has many unique advantages, e.g., lower fissile inventory,no...The recent development of molten salt fast reactors has generated a renewed interest in them. As compared to traditional solid fuel fast neutron systems, it has many unique advantages, e.g., lower fissile inventory,no initial criticality reserve, waste reduction, and a simplified fuel cycle. It has been recognized as an ideal reactor for achieving a closed Th–U cycle. Based on the carrier salt, molten salt fast reactors could be divided into either a molten chloride salt fast reactor(MCFR) or a molten fluoride salt fast reactor(MFFR);to compare their Th–U cycle performance, the neutronic parameters in a breeding and burning(B&B) transition scenario were studied based on similar core geometry and power. The results demonstrated that the required reprocessing rate for an MCFR to achieve self-breeding was lower than that of an MFFR.Moreover, the breeding capability of an MCFR was better than that of an MFFR;at a reprocessing rate of 40 L/day,using LEU and Pu as start-up fissile materials, the doubling time(DT) of an MFFR and MCFR were 88.0 years and 48.0 years, and 16.5 years and 16.2 years, respectively.Besides, an MCFR has lower radio-toxicity due to lower buildup of fission products(FPs) and transuranium(TRU),while an MFFR has a larger, delayed neutron fraction with smaller changes during the entire operation.展开更多
A highly accurate and precise technique for measurement of the 93 Nb(n,n’)93m Nb reaction rate was established for the material surveillance tests,etc.in fast reactors.The self-absorption effect on the measurement of...A highly accurate and precise technique for measurement of the 93 Nb(n,n’)93m Nb reaction rate was established for the material surveillance tests,etc.in fast reactors.The self-absorption effect on the measurement of the characteristic X-rays emitted by 93m Nb was decreased by the dissolution and evaporation to dryness of niobium dosimeter.A highly precise count of the number of 93 Nb atoms was obtained by measuring the niobium solution concentration using inductively coupled plasma mass spectrometry.X-rays of 93m Nb were measured accurately by means of comparing the X-ray intensity of irradiated niobium solution with that of the solution in which stable 93 Nb was added.The difference between both intensities indicates the effect of 182 Ta,which is generated from an impurity tantalum,and the intensity of X-rays from 93m Nb was evaluated.Measurement error of the 93 Nb(n,n’)93m Nb reaction rate was reduced to be less than 4%,which was equivalent to the other reaction rate errors of dosimeters used for Joyo dosimetry.In addition,an advanced technique using Resonance Ionization Mass Spectrometry was proposed for the precise measurement of 93m Nb yield,and 93m Nb will be resonance-ionized selectively by discriminating the hyperfine splitting of the atomic energy levels between 93 Nb and 93m Nb at high resolution.展开更多
The feasibility of rhenium （Re） production by irradiating tungsten （W） metal in a medium size fast reactor was evaluated by using a Monte Carlo code. The fast reactor can produce about 50 kilograms of Re per every...The feasibility of rhenium （Re） production by irradiating tungsten （W） metal in a medium size fast reactor was evaluated by using a Monte Carlo code. The fast reactor can produce about 50 kilograms of Re per every 3 years, which corresponds 10% of Japanese domestic production. The specific activity of Re can be reduced below the exemption level or even the natural Re level if W and osmium is separated after the irradiation. The use of ZrD1.7 moderator reduces the specific activity by half compared to that of ZrH1.7 case, and even the no moderator case is permissible to produce the production of Re which has lower specific reactivity than that of natural Re.展开更多
The purpose of the present study is to develop a methodology to evaluate fuel discharge through the CRGT （control-rod guide tube） during CDAs （core-disruptive accidents） of SFRs （sodium-cooled fast reactors）, si...The purpose of the present study is to develop a methodology to evaluate fuel discharge through the CRGT （control-rod guide tube） during CDAs （core-disruptive accidents） of SFRs （sodium-cooled fast reactors）, since fuel discharge will decrease the core reactivity and CRGTs have a potential to provide an effective discharge path. Fuel discharge contains multi-component fluid dynamics with phase changes, and, in the present study, the SFR safety analysis code SIMMER （Sn, implicit, multifield, multicomponent, Eulerian recriticality） was utilized as a technical basis. First, dominant phenomena affecting fuel discharge through the CRGT are identified based on parametric calculations by the SIMMER code. Next, validations on the code models closely relating to these phenomena were carried out based on experimental data. It was shown that the SIMMER code with some model modifications could reproduce the experimental results appropriately. Through the present study, the evaluation methodology for the molten-fuel discharge through the CRGT was successfully developed.展开更多
Accurate determination of inner wall temperature fluctuations is critical for thermal fatigue assessment in sodiumcooled fast reactors(SFR)piping systems,but remains inaccessible for direct measurement due to extreme ...Accurate determination of inner wall temperature fluctuations is critical for thermal fatigue assessment in sodiumcooled fast reactors(SFR)piping systems,but remains inaccessible for direct measurement due to extreme operational conditions involving high temperature and chemical activity of liquid sodium.To overcome this challenge,this study proposes a self-adaptive Gaussian process regression(GPR)approach.The large eddy simulations(LES)of hot and cold liquid sodium mixing in T-junction pipes are conducted to quantify intense thermal-fluid interactions,revealing that inner wall temperature fluctuations are significantly higher than those at the outer walls.Building on these insights,we develop a self-adaptive GPR approach that integrates tree-structured composite kernel optimization with gradient-based hyperparameter tuning.The resulting approach accurately predicts inner wall temperature fluctuations using only outer wall measurements and corresponding operational parameters,achieving a predictive performance of determination coefficient R^(2)>0.95,and retaining robustness(R^(2)>0.75)even when trained on limited datasets.The proposed self-adaptive GPR approach offers non-intrusive,real-time thermal diagnostics for SFR piping systems,utilizing composite kernels that afford clear physical interpretability.Moreover,it provides a promising tool for safety monitoring in reactor cores,heat exchangers,and other nuclear components requiring high-fidelity thermal transient analysis.展开更多
The accurate modeling of depletion,intricately tied to the solution of the neutron transport equation,is crucial for the design,analysis,and licensing of nuclear reactors and their fuel cycles.This paper introduces a ...The accurate modeling of depletion,intricately tied to the solution of the neutron transport equation,is crucial for the design,analysis,and licensing of nuclear reactors and their fuel cycles.This paper introduces a novel multi-group Monte-Carlo depletion calculation approach.Multi-group cross-sections(MGXS)are derived from both 3D whole-core model and 2D fuel subassembly model using the continuous-energy Monte-Carlo method.Core calculations employ the multi-group Monte-Carlo method,accommodating both homogeneous and specific local heterogeneous geometries.The proposed method has been validated against the MET-1000 metal-fueled fast reactors,using both the OECD/NEA benchmark and a new refueling benchmark introduced in this paper.Our findings suggest that microscopic MGXS,produced via the Monte-Carlo method,are viable for fast reactor depletion analyses.Furthermore,the locally heterogeneous model with angular-dependent MGXS offers robust predictions for core reactivity,control rod value,sodium void value,Doppler constants,power distribution,and concentration levels.展开更多
In this study,a multi-physics and multi-scale coupling program,Fluent/KMC-sub/NDK,was developed based on the user-defined functions(UDF)of Fluent,in which the KMC-sub-code is a sub-channel thermal-hydraulic code and t...In this study,a multi-physics and multi-scale coupling program,Fluent/KMC-sub/NDK,was developed based on the user-defined functions(UDF)of Fluent,in which the KMC-sub-code is a sub-channel thermal-hydraulic code and the NDK code is a neutron diffusion code.The coupling program framework adopts the"master-slave"mode,in which Fluent is the master program while NDK and KMC-sub are coupled internally and compiled into the dynamic link library(DLL)as slave codes.The domain decomposition method was adopted,in which the reactor core was simulated by NDK and KMC-sub,while the rest of the primary loop was simulated using Fluent.A simulation of the reactor shutdown process of M2LFR-1000 was carried out using the coupling program,and the code-to-code verification was performed with ATHLET,demonstrating a good agreement,with absolute deviation was smaller than 0.2%.The results show an obvious thermal stratification phenomenon during the shutdown process,which occurs 10 s after shutdown,and the change in thermal stratification phenomena is also captured by the coupling program.At the same time,the change in the neutron flux density distribution of the reactor was also obtained.展开更多
Owing to the inherent instability of the natural circulation system,flow instability can easily occur during the operation of a natural circulation lead-cooled fast reactor,especially during the startup phase.A compre...Owing to the inherent instability of the natural circulation system,flow instability can easily occur during the operation of a natural circulation lead-cooled fast reactor,especially during the startup phase.A comprehensive startup scheme for SNCLFR-100,including primary and secondary circuits,is proposed in this paper.It references existing more mature startup schemes in various reactor types.It additionally considers the restriction conditions on the power increase in other schemes and the characteristics of lead-based coolant.On this basis,the multi-scale coupling code ATHLET-OpenFOAM was used to study the flow instability in the startup phase under different power-step amplitudes and power duration times.The results showed that obvious flow instability phenomena were found in the different startup schemes,such as the short-term backflow phenomenon of the core at the initial time of the startup.Moreover,an obvious increase in the flow rate and temperature to the peak value at the later stage of a continuous power rise was observed,as well as continuous oscillations before reaching a steady state.It was determined that the scheme with smaller power-step amplitude and a longer power duration time requires more time to start the reactor.Nevertheless,it will be more conducive to the safe and stable startup of the reactor.展开更多
In the core of a molten salt fast reactor(MSFR),heavy metal fuel and fission products can be dissolved in a molten fluoride salt to form a eutectic mixture that acts as both fuel and coolant.Fission energy is released...In the core of a molten salt fast reactor(MSFR),heavy metal fuel and fission products can be dissolved in a molten fluoride salt to form a eutectic mixture that acts as both fuel and coolant.Fission energy is released from the fuel salt and transferred to the second loop by fuel salt circulation.Therefore,the MSFR is characterized by strong interaction between the neutronics and the thermal hydraulics.Moreover,recirculation flow occurs,and nuclear heat is accumulated near the fertile blanket,which significantly affects both the flow and the temperature fields in the core.In this work,to further optimize the conceptual geometric design of the MSFR,three geometries of the core and fertile blanket are proposed,and the thermal-hydraulic characteristics,including the three-dimensional flow and temperature fields of the fuel and fertile salts,are simulated and analyzed using a coupling scheme between the open source codes OpenMC and OpenFOAM.The numerical results indicate that a flatter core temperature distribution can be obtained and the hot spot and flow stagnation zones that appear in the upper and lower parts of the core center near the reflector can be eliminated by curving both the top and bottom walls of the core.Moreover,eight cooling loops with a total flow rate of0.0555 m3 s-1 ensur an acceptable temperature distribusure an acceptable temperature distribution in the fertile blanket.展开更多
A molten chloride salt fast reactor(MCFR)is well suited to fuel breeding and the transmutation of transuranium(TRU)elements owing to its advantageous features of fast neutron spectrum and high TRU solubility.However,t...A molten chloride salt fast reactor(MCFR)is well suited to fuel breeding and the transmutation of transuranium(TRU)elements owing to its advantageous features of fast neutron spectrum and high TRU solubility.However,the neutron absorption cross section of 35Cl is approximately 1000 times greater than for 37Cl,which has a significant impact on the neutron physical characteristics of a MCFR.Based on an automatic online refueling and reprocessing procedure,the influences of 37Cl enrichment on neutron economy,breeding performance,and the production of harmful nuclides were analyzed.Results show that 37Cl enrichment strongly influences the neutron properties of a MCFR.With natural chlorine,233U breeding cannot be achieved and the yields of S and 36Cl are very high.Increasing the 37Cl enrichment to 97%brings a clear improvement in its neutronics property,making it almost equal to that corresponding to 100%enrichment.Moreover,when 37Cl is enriched to 99%,its neutronics parameters are almost the same as for 100%enrichment.Considering the enrichment cost and the neutron properties,a 37Cl enrichment of 97%is recommended.Achieving an optimal neutronics performance requires 99%37Cl enrichment.展开更多
Due to their unique features,such as the inherent safety,simplified fuel cycle,and continuous on-line reprocessing,molten salt reactors(MSRs)are regarded as one of the six reference reactors in the Generation IV Inter...Due to their unique features,such as the inherent safety,simplified fuel cycle,and continuous on-line reprocessing,molten salt reactors(MSRs)are regarded as one of the six reference reactors in the Generation IV International Forum(GEN-IV).Molten chloride salt fast reactors(MCFRs)are a type of MSR.Compared to molten fluoride salt reactors(MFSRs),MCFRs have a higher solubility of heavy metal atoms,a harder neutron spectrum,lower accumulation of fission products(FPs),and better breeding and transmutation performance.Thus,MCFRs have been recognized as a type of MSR with great prospects for future development.However,as the most important feature for MSRs,the effect of different reprocessing modes on MCFRs must be researched in depth.As such,this study investigated the effect of different isotopes,especially FPs,on the neutronic performance of an MCFR,such as its breeding performance.Furthermore,the characteristics of the different reprocessing modes and MCFR rates were analyzed in terms of safety,radioactivity level,neutron economy,and breeding capacity.In the end,a reprocessing method suitable for MCFRs was determined through calculation and analysis,which provides a reference for the further research of MCFRs.展开更多
Radiopharmaceuticals are used in nuclear medicine for diagnostic or therapeutic acts. The short decay half-lives of medical radioisotopes, especially those used for diagnostics, imply that they should be produced cont...Radiopharmaceuticals are used in nuclear medicine for diagnostic or therapeutic acts. The short decay half-lives of medical radioisotopes, especially those used for diagnostics, imply that they should be produced continuously and transported as quickly as possible to the medical units where they are used. Neutron-rich medical radioisotopes are generally produced in research reactors, like technetium-99m, lutetium-177, holmium-166 and iodine-131. On the other hand, proton-rich radioisotopes are produced via reactions with charged particles from accelerators like fluorine-18, gallium-67, iodine-123 and thallium-201. Beside this, innovative nuclear reactors are advocated as solutions to the issues of nuclear waste production and proliferation threats. Fast neutron, thorium-cycle and accelerator-driven subcritical (ADS) reactors are some of the most promising of them, proposed as safer fuel breeders and “waste burners”. This article examines the use of a fast thorium-cycle ADS with liquid lead-bismuth eutectic coolant for the production of molybdenum-99/technetium-99m and lutetium-177. Burnup simulation has been made with the Monte-Carlo (MC) code SERPENT. It is demonstrated that MC codes can advantageously be used to determine the optimal irradiation time for a given radioisotope in a realistic reactor core. It is also shown that fast thorium-cycle ADS is an economical option for the production of medical radioisotopes.展开更多
The objective of this study is to presume cesium corrosion process and its dominant factors in SUS316 steel, a fuel cladding material for fast breeder reactor application, based on both experimental results of cesium ...The objective of this study is to presume cesium corrosion process and its dominant factors in SUS316 steel, a fuel cladding material for fast breeder reactor application, based on both experimental results of cesium corrosion out-pile test and thermodynamic consideration. The cesium corrosion test was performed in simulated environment of high burn-up fuel pin. And main corrosion products in the specimen after the corrosion test were specified by TEM （transition electron microscopy） and SEM （scanning electron microscopy） in order to formulate a hypothesis of the cesium corrosion process. At the end of this study, it was found that the dominant factors of the corrosion process are the amount of cesium on the surface of the specimen, chromium content in the alloy, the supply rate of oxygen and temperature.展开更多
Korea imports about 97% of its energy resources as its available energy resources are extremely limited. Thus, the role of nuclear power in electricity generation is expected to become more important in future years. ...Korea imports about 97% of its energy resources as its available energy resources are extremely limited. Thus, the role of nuclear power in electricity generation is expected to become more important in future years. A fast reactor system is one of the most promising options for electricity generation with an efficient utilization of uranium resources and a reduction of radioactive wastes. Based on the experiences gained during the development of the conceptual designs for KALIMER （Korea advanced liquid metal reactor）, the KAERI （Korea Atomic Energy Research Institute） is currently developing advanced SFR （sodium cooled fast reactor） design concepts that can better meet the Gen IV （Generation IV） technology goals. The long-term advanced SFR development plan will be carried out toward the construction of an advanced SFR demonstration plant by 2028. Advanced concept design studies and the development of the advanced SFR technologies necessary for its commercialization and basic key technologies carried out by KAERI are included in this paper.展开更多
The integrity and reliability of fuel rods under both normal and accidental operating conditions are of great importance for nuclear reactors.In this study,considering various irradiation behaviors,a fuel rod performa...The integrity and reliability of fuel rods under both normal and accidental operating conditions are of great importance for nuclear reactors.In this study,considering various irradiation behaviors,a fuel rod performance analysis code,named KMC-Fueltra,was developed to evaluate the thermal–mechanical performance of oxide fuel rods under both normal and transient conditions in the LMFR.The accuracy and reliability of the KMC-Fueltra were validated by analytical solutions,as well as the results obtained from codes and experiments.The results indicated that KMC-Fueltra can predict the performance of oxide fuel rods under both normal and transient conditions in the LMFR.展开更多
This study presents a benchmark evaluation of the new ENDF/B-VIII.0 nuclear data library for the Organization for Economic Co-operation and Development/Nuclear Energy Agency Medium 1000 MWth sodium-cooled fast reactor...This study presents a benchmark evaluation of the new ENDF/B-VIII.0 nuclear data library for the Organization for Economic Co-operation and Development/Nuclear Energy Agency Medium 1000 MWth sodium-cooled fast reactor(SFR).The study presented herein covers both SFR core types,i.e.,metallic fueled(MET-1000)and oxide fueled(MOX-1000),simulated using the continuous-energy Monte Carlo Serpent2 code.The neutronics performances of the ENDF/B-VIII.0-based simulations were compared mainly to two libraries:ENDF/B-VII.1 and JENDL-4.0.The comparison includes several neutronics parameters evaluated for the beginning and end of the cycle conditions.These parameters include the effective multiplication factor keff,total effective delayed neutron fraction beff,sodium void reactivity(DqNa),Doppler constant(DqDoppler),and control rod worth(DqCR).In addition,a sensitivity study was used to reveal the major isotope/reaction pairs contributing to the discrepancy observed in the performance of the three libraries using 33 and 44-energy-group structures.展开更多
The study gives a brief introduction on development of innovated nuclear system in China,mainly focus on the materials R&D status for the sodium cooled fast reactor.With the high speed development of economy,China...The study gives a brief introduction on development of innovated nuclear system in China,mainly focus on the materials R&D status for the sodium cooled fast reactor.With the high speed development of economy,China needs a huge energy supply;at same time a more cleaning energy to reduce the carbon release is demanded.The nuclear energy is the most cleaning energy at present time,especially the innovated nuclear system which is so-called GenerationⅣpower plants has got its prior development due to its safety, economical and little fission production produced.Fast breeder reactor,as the priority development reactor type in the Gen-Ⅳnuclear system,is the key to the advanced closed fuel cycle technologies.China experimental fast reactor(CEFR ) has been completed the design,construction the synthesis system commissioning and reached its physical criticality on July 21,2010.At China Institute of Atomic Energy,the CEFR and other research facilities have been established,and extensive studies are planning to carry out in the areas of fuel and materials development.This will laid the foundation for the design and development of the future's CFR—900(China Demonstration Fast Reactor) and CCFR(China Commercial Fast Reactor). Highlights of some of materials R&D studies are discussed in this paper.展开更多
Various sources of solid particles might exist in the coolant flow of a liquid metal cooled fast reactor(e.g.,through chemical interaction between the coolant and impurities,air,or water,through corrosion of structura...Various sources of solid particles might exist in the coolant flow of a liquid metal cooled fast reactor(e.g.,through chemical interaction between the coolant and impurities,air,or water,through corrosion of structural materials,or from damaged/molten fuel).Such particles may cause flow blockage accidents in a fuel assembly,resulting in a reduction in coolant flow,which potentially causes a local temperature rise in the fuel cladding,cladding failure,and fuel melt.To understand the blockage formation mechanism,in this study,a series of simulated experiments was conducted by releasing different solid particles from a release device into a reducer pipe using gravity.Through detailed analyses,the influence of various experimental parameters(e.g.,particle diameter,capacity,shape,and static friction coefficient,and the diameter and height of the particle release nozzle)on the blockage characteristics(i.e.,blockage probability and position)was examined.Under the current range of experimental conditions,the blockage was significantly influenced by the aforementioned parameters.The ratio between the particle diameter and outlet size of the reducer pipe might be one of the determining factors governing the occurrence of blockage.Specifically,increasing the ratio enhanced blockage(i.e.,larger probability and higher position within the reducer pipe).Increasing the particle size,particle capacity,particle static friction coefficient,and particle release nozzle diameter led to a rise in the blockage probability;however,increasing the particle release nozzle height had a downward influence on the blockage probability.Finally,blockage was more likely to occur in non-spherical particles case than that of spherical particles.This study provides a large experimental database to promote an understanding of the flow blockage mechanism and improve the validation process of fast reactor safety analysis codes.展开更多
To examine the sterilizing effect and mechanism of neutron radiation, Bacillus subtills var. niger, strain （ATCC 9372） spores were irradiated with the fast neutron from the Chinese fast burst reactor II（CFBR-II）. ...To examine the sterilizing effect and mechanism of neutron radiation, Bacillus subtills var. niger, strain （ATCC 9372） spores were irradiated with the fast neutron from the Chinese fast burst reactor II（CFBR-II）. The plate-count results indicated that the D10 value was 384.6 Gy with a neutron radiation dose rate of 7.4 Gy/min. The rudimental catalase activity of the spores declined obviously with the increase in the radiation dose. Meanwhile, under the scanning electron microscope, no visible influence of the neutron radiation on the spore configuration was detected even if the dose was increased to 4 kGy. The content and distribution of DNA double-strand breaks induced by neutron radiation at different doses were measured and quantified by pulsed- field gel electrophoresis （PFGE）. Further analysis of the DNA release percentage （PR）, the DNA breakage level （L）, and the average molecular weight, indicated that DNA fragments were obvi- ously distributed around the 5 kb regions at different radiation doses, which suggests that some points in the DNA molecule were sensitive to neutron radiation. Both PR and L varied regularly to some extent with the increase in radiation dose. Thus neutron radiation has a high sterilization power, and can induce falling enzyme activity and DNA breakage in Bacillus subtilis spores展开更多
China has decided to speed-up the nuclear power development.It is programmed that the nuclear power capacity will reach 40 GWe in 2020 and envisaged 60 GWe and 240 GWe in 2030 and 2050 respectively.The basic strategy ...China has decided to speed-up the nuclear power development.It is programmed that the nuclear power capacity will reach 40 GWe in 2020 and envisaged 60 GWe and 240 GWe in 2030 and 2050 respectively.The basic strategy of PWR-FBR matched development with Fast reactor metal fuel closed cycle for a sustainable and quick increasing nuclear energy supply is adopted.Another strategy also decided is that the partitioning and transmutation of MA will be realized using fast burner and ADS.The fast reactor engineering development will be divided into three steps:China Experimental Fast Reactor(CEFR 65 MWt/20 MWe),China Prototype/Demonstration Fast Reactor(CPFR/CDFR≥1500 MWt/600 MWe)and China Demonstration Fast Breeder Reactor(CDFBR 1000~1500 MWe).The CEFR is under installation and pre-operation testing with it's first criticality planned in 2009.The design study of CPFR is just started in 2006.Recently a discussion for the second step is under way to faster the fast reactor development by a larger than 600 MWe CPFR and as a role of CDFR.展开更多
基金the Chinese TMSR Strategic Pioneer Science and Technology Project(No.XDA02010000)the National Natural Science Foundation of China(No.91326201)。
文摘The recent development of molten salt fast reactors has generated a renewed interest in them. As compared to traditional solid fuel fast neutron systems, it has many unique advantages, e.g., lower fissile inventory,no initial criticality reserve, waste reduction, and a simplified fuel cycle. It has been recognized as an ideal reactor for achieving a closed Th–U cycle. Based on the carrier salt, molten salt fast reactors could be divided into either a molten chloride salt fast reactor(MCFR) or a molten fluoride salt fast reactor(MFFR);to compare their Th–U cycle performance, the neutronic parameters in a breeding and burning(B&B) transition scenario were studied based on similar core geometry and power. The results demonstrated that the required reprocessing rate for an MCFR to achieve self-breeding was lower than that of an MFFR.Moreover, the breeding capability of an MCFR was better than that of an MFFR;at a reprocessing rate of 40 L/day,using LEU and Pu as start-up fissile materials, the doubling time(DT) of an MFFR and MCFR were 88.0 years and 48.0 years, and 16.5 years and 16.2 years, respectively.Besides, an MCFR has lower radio-toxicity due to lower buildup of fission products(FPs) and transuranium(TRU),while an MFFR has a larger, delayed neutron fraction with smaller changes during the entire operation.
文摘A highly accurate and precise technique for measurement of the 93 Nb(n,n’)93m Nb reaction rate was established for the material surveillance tests,etc.in fast reactors.The self-absorption effect on the measurement of the characteristic X-rays emitted by 93m Nb was decreased by the dissolution and evaporation to dryness of niobium dosimeter.A highly precise count of the number of 93 Nb atoms was obtained by measuring the niobium solution concentration using inductively coupled plasma mass spectrometry.X-rays of 93m Nb were measured accurately by means of comparing the X-ray intensity of irradiated niobium solution with that of the solution in which stable 93 Nb was added.The difference between both intensities indicates the effect of 182 Ta,which is generated from an impurity tantalum,and the intensity of X-rays from 93m Nb was evaluated.Measurement error of the 93 Nb(n,n’)93m Nb reaction rate was reduced to be less than 4%,which was equivalent to the other reaction rate errors of dosimeters used for Joyo dosimetry.In addition,an advanced technique using Resonance Ionization Mass Spectrometry was proposed for the precise measurement of 93m Nb yield,and 93m Nb will be resonance-ionized selectively by discriminating the hyperfine splitting of the atomic energy levels between 93 Nb and 93m Nb at high resolution.
文摘The feasibility of rhenium （Re） production by irradiating tungsten （W） metal in a medium size fast reactor was evaluated by using a Monte Carlo code. The fast reactor can produce about 50 kilograms of Re per every 3 years, which corresponds 10% of Japanese domestic production. The specific activity of Re can be reduced below the exemption level or even the natural Re level if W and osmium is separated after the irradiation. The use of ZrD1.7 moderator reduces the specific activity by half compared to that of ZrH1.7 case, and even the no moderator case is permissible to produce the production of Re which has lower specific reactivity than that of natural Re.
文摘The purpose of the present study is to develop a methodology to evaluate fuel discharge through the CRGT （control-rod guide tube） during CDAs （core-disruptive accidents） of SFRs （sodium-cooled fast reactors）, since fuel discharge will decrease the core reactivity and CRGTs have a potential to provide an effective discharge path. Fuel discharge contains multi-component fluid dynamics with phase changes, and, in the present study, the SFR safety analysis code SIMMER （Sn, implicit, multifield, multicomponent, Eulerian recriticality） was utilized as a technical basis. First, dominant phenomena affecting fuel discharge through the CRGT are identified based on parametric calculations by the SIMMER code. Next, validations on the code models closely relating to these phenomena were carried out based on experimental data. It was shown that the SIMMER code with some model modifications could reproduce the experimental results appropriately. Through the present study, the evaluation methodology for the molten-fuel discharge through the CRGT was successfully developed.
基金supported by the National Natural Science Foundation of China(Grant Nos.52175224,52205262,52235005)。
文摘Accurate determination of inner wall temperature fluctuations is critical for thermal fatigue assessment in sodiumcooled fast reactors(SFR)piping systems,but remains inaccessible for direct measurement due to extreme operational conditions involving high temperature and chemical activity of liquid sodium.To overcome this challenge,this study proposes a self-adaptive Gaussian process regression(GPR)approach.The large eddy simulations(LES)of hot and cold liquid sodium mixing in T-junction pipes are conducted to quantify intense thermal-fluid interactions,revealing that inner wall temperature fluctuations are significantly higher than those at the outer walls.Building on these insights,we develop a self-adaptive GPR approach that integrates tree-structured composite kernel optimization with gradient-based hyperparameter tuning.The resulting approach accurately predicts inner wall temperature fluctuations using only outer wall measurements and corresponding operational parameters,achieving a predictive performance of determination coefficient R^(2)>0.95,and retaining robustness(R^(2)>0.75)even when trained on limited datasets.The proposed self-adaptive GPR approach offers non-intrusive,real-time thermal diagnostics for SFR piping systems,utilizing composite kernels that afford clear physical interpretability.Moreover,it provides a promising tool for safety monitoring in reactor cores,heat exchangers,and other nuclear components requiring high-fidelity thermal transient analysis.
基金supported by the National Natural Science Foundation of China(Nos.12105170,12135008)Science and Technology on Reactor System Design Technology Laboratory.
文摘The accurate modeling of depletion,intricately tied to the solution of the neutron transport equation,is crucial for the design,analysis,and licensing of nuclear reactors and their fuel cycles.This paper introduces a novel multi-group Monte-Carlo depletion calculation approach.Multi-group cross-sections(MGXS)are derived from both 3D whole-core model and 2D fuel subassembly model using the continuous-energy Monte-Carlo method.Core calculations employ the multi-group Monte-Carlo method,accommodating both homogeneous and specific local heterogeneous geometries.The proposed method has been validated against the MET-1000 metal-fueled fast reactors,using both the OECD/NEA benchmark and a new refueling benchmark introduced in this paper.Our findings suggest that microscopic MGXS,produced via the Monte-Carlo method,are viable for fast reactor depletion analyses.Furthermore,the locally heterogeneous model with angular-dependent MGXS offers robust predictions for core reactivity,control rod value,sodium void value,Doppler constants,power distribution,and concentration levels.
基金supported by Science and Technology on Reactor System Design Technology Laboratory,Chengdu,China(LRSDT2020106)
文摘In this study,a multi-physics and multi-scale coupling program,Fluent/KMC-sub/NDK,was developed based on the user-defined functions(UDF)of Fluent,in which the KMC-sub-code is a sub-channel thermal-hydraulic code and the NDK code is a neutron diffusion code.The coupling program framework adopts the"master-slave"mode,in which Fluent is the master program while NDK and KMC-sub are coupled internally and compiled into the dynamic link library(DLL)as slave codes.The domain decomposition method was adopted,in which the reactor core was simulated by NDK and KMC-sub,while the rest of the primary loop was simulated using Fluent.A simulation of the reactor shutdown process of M2LFR-1000 was carried out using the coupling program,and the code-to-code verification was performed with ATHLET,demonstrating a good agreement,with absolute deviation was smaller than 0.2%.The results show an obvious thermal stratification phenomenon during the shutdown process,which occurs 10 s after shutdown,and the change in thermal stratification phenomena is also captured by the coupling program.At the same time,the change in the neutron flux density distribution of the reactor was also obtained.
文摘Owing to the inherent instability of the natural circulation system,flow instability can easily occur during the operation of a natural circulation lead-cooled fast reactor,especially during the startup phase.A comprehensive startup scheme for SNCLFR-100,including primary and secondary circuits,is proposed in this paper.It references existing more mature startup schemes in various reactor types.It additionally considers the restriction conditions on the power increase in other schemes and the characteristics of lead-based coolant.On this basis,the multi-scale coupling code ATHLET-OpenFOAM was used to study the flow instability in the startup phase under different power-step amplitudes and power duration times.The results showed that obvious flow instability phenomena were found in the different startup schemes,such as the short-term backflow phenomenon of the core at the initial time of the startup.Moreover,an obvious increase in the flow rate and temperature to the peak value at the later stage of a continuous power rise was observed,as well as continuous oscillations before reaching a steady state.It was determined that the scheme with smaller power-step amplitude and a longer power duration time requires more time to start the reactor.Nevertheless,it will be more conducive to the safe and stable startup of the reactor.
基金supported by the Chinese TMSR Strategic Pioneer Science and Technology Project(No.XDA02010000)the Frontier Science Key Program of the Chinese Academy of Sciences(No.QYZDY-SSW-JSC016)。
文摘In the core of a molten salt fast reactor(MSFR),heavy metal fuel and fission products can be dissolved in a molten fluoride salt to form a eutectic mixture that acts as both fuel and coolant.Fission energy is released from the fuel salt and transferred to the second loop by fuel salt circulation.Therefore,the MSFR is characterized by strong interaction between the neutronics and the thermal hydraulics.Moreover,recirculation flow occurs,and nuclear heat is accumulated near the fertile blanket,which significantly affects both the flow and the temperature fields in the core.In this work,to further optimize the conceptual geometric design of the MSFR,three geometries of the core and fertile blanket are proposed,and the thermal-hydraulic characteristics,including the three-dimensional flow and temperature fields of the fuel and fertile salts,are simulated and analyzed using a coupling scheme between the open source codes OpenMC and OpenFOAM.The numerical results indicate that a flatter core temperature distribution can be obtained and the hot spot and flow stagnation zones that appear in the upper and lower parts of the core center near the reflector can be eliminated by curving both the top and bottom walls of the core.Moreover,eight cooling loops with a total flow rate of0.0555 m3 s-1 ensur an acceptable temperature distribusure an acceptable temperature distribution in the fertile blanket.
基金supported by the Chinese TMSR Strategic Pioneer Science and Technology Project(No.XDA02010000)the Frontier Science Key Program of the Chinese Academy of Sciences(No.QYZDY-SSW-JSC016)the Nuclear Energy Development Project(No.20154602098)
文摘A molten chloride salt fast reactor(MCFR)is well suited to fuel breeding and the transmutation of transuranium(TRU)elements owing to its advantageous features of fast neutron spectrum and high TRU solubility.However,the neutron absorption cross section of 35Cl is approximately 1000 times greater than for 37Cl,which has a significant impact on the neutron physical characteristics of a MCFR.Based on an automatic online refueling and reprocessing procedure,the influences of 37Cl enrichment on neutron economy,breeding performance,and the production of harmful nuclides were analyzed.Results show that 37Cl enrichment strongly influences the neutron properties of a MCFR.With natural chlorine,233U breeding cannot be achieved and the yields of S and 36Cl are very high.Increasing the 37Cl enrichment to 97%brings a clear improvement in its neutronics property,making it almost equal to that corresponding to 100%enrichment.Moreover,when 37Cl is enriched to 99%,its neutronics parameters are almost the same as for 100%enrichment.Considering the enrichment cost and the neutron properties,a 37Cl enrichment of 97%is recommended.Achieving an optimal neutronics performance requires 99%37Cl enrichment.
基金supported by the Chinese TMSR Strategic Pioneer Science and Technology Project (No.XDA02010000)the Frontier Science Key Program of Chinese Academy of Sciences (No.QYZDY-SSW-JSC016)the Shanghai Sailing Program (No.20YF1457600).
文摘Due to their unique features,such as the inherent safety,simplified fuel cycle,and continuous on-line reprocessing,molten salt reactors(MSRs)are regarded as one of the six reference reactors in the Generation IV International Forum(GEN-IV).Molten chloride salt fast reactors(MCFRs)are a type of MSR.Compared to molten fluoride salt reactors(MFSRs),MCFRs have a higher solubility of heavy metal atoms,a harder neutron spectrum,lower accumulation of fission products(FPs),and better breeding and transmutation performance.Thus,MCFRs have been recognized as a type of MSR with great prospects for future development.However,as the most important feature for MSRs,the effect of different reprocessing modes on MCFRs must be researched in depth.As such,this study investigated the effect of different isotopes,especially FPs,on the neutronic performance of an MCFR,such as its breeding performance.Furthermore,the characteristics of the different reprocessing modes and MCFR rates were analyzed in terms of safety,radioactivity level,neutron economy,and breeding capacity.In the end,a reprocessing method suitable for MCFRs was determined through calculation and analysis,which provides a reference for the further research of MCFRs.
文摘Radiopharmaceuticals are used in nuclear medicine for diagnostic or therapeutic acts. The short decay half-lives of medical radioisotopes, especially those used for diagnostics, imply that they should be produced continuously and transported as quickly as possible to the medical units where they are used. Neutron-rich medical radioisotopes are generally produced in research reactors, like technetium-99m, lutetium-177, holmium-166 and iodine-131. On the other hand, proton-rich radioisotopes are produced via reactions with charged particles from accelerators like fluorine-18, gallium-67, iodine-123 and thallium-201. Beside this, innovative nuclear reactors are advocated as solutions to the issues of nuclear waste production and proliferation threats. Fast neutron, thorium-cycle and accelerator-driven subcritical (ADS) reactors are some of the most promising of them, proposed as safer fuel breeders and “waste burners”. This article examines the use of a fast thorium-cycle ADS with liquid lead-bismuth eutectic coolant for the production of molybdenum-99/technetium-99m and lutetium-177. Burnup simulation has been made with the Monte-Carlo (MC) code SERPENT. It is demonstrated that MC codes can advantageously be used to determine the optimal irradiation time for a given radioisotope in a realistic reactor core. It is also shown that fast thorium-cycle ADS is an economical option for the production of medical radioisotopes.
文摘The objective of this study is to presume cesium corrosion process and its dominant factors in SUS316 steel, a fuel cladding material for fast breeder reactor application, based on both experimental results of cesium corrosion out-pile test and thermodynamic consideration. The cesium corrosion test was performed in simulated environment of high burn-up fuel pin. And main corrosion products in the specimen after the corrosion test were specified by TEM （transition electron microscopy） and SEM （scanning electron microscopy） in order to formulate a hypothesis of the cesium corrosion process. At the end of this study, it was found that the dominant factors of the corrosion process are the amount of cesium on the surface of the specimen, chromium content in the alloy, the supply rate of oxygen and temperature.
文摘Korea imports about 97% of its energy resources as its available energy resources are extremely limited. Thus, the role of nuclear power in electricity generation is expected to become more important in future years. A fast reactor system is one of the most promising options for electricity generation with an efficient utilization of uranium resources and a reduction of radioactive wastes. Based on the experiences gained during the development of the conceptual designs for KALIMER （Korea advanced liquid metal reactor）, the KAERI （Korea Atomic Energy Research Institute） is currently developing advanced SFR （sodium cooled fast reactor） design concepts that can better meet the Gen IV （Generation IV） technology goals. The long-term advanced SFR development plan will be carried out toward the construction of an advanced SFR demonstration plant by 2028. Advanced concept design studies and the development of the advanced SFR technologies necessary for its commercialization and basic key technologies carried out by KAERI are included in this paper.
文摘The integrity and reliability of fuel rods under both normal and accidental operating conditions are of great importance for nuclear reactors.In this study,considering various irradiation behaviors,a fuel rod performance analysis code,named KMC-Fueltra,was developed to evaluate the thermal–mechanical performance of oxide fuel rods under both normal and transient conditions in the LMFR.The accuracy and reliability of the KMC-Fueltra were validated by analytical solutions,as well as the results obtained from codes and experiments.The results indicated that KMC-Fueltra can predict the performance of oxide fuel rods under both normal and transient conditions in the LMFR.
基金the Research Institute of Science and Engineering at the University of Sharjah(No.1802040790-P).
文摘This study presents a benchmark evaluation of the new ENDF/B-VIII.0 nuclear data library for the Organization for Economic Co-operation and Development/Nuclear Energy Agency Medium 1000 MWth sodium-cooled fast reactor(SFR).The study presented herein covers both SFR core types,i.e.,metallic fueled(MET-1000)and oxide fueled(MOX-1000),simulated using the continuous-energy Monte Carlo Serpent2 code.The neutronics performances of the ENDF/B-VIII.0-based simulations were compared mainly to two libraries:ENDF/B-VII.1 and JENDL-4.0.The comparison includes several neutronics parameters evaluated for the beginning and end of the cycle conditions.These parameters include the effective multiplication factor keff,total effective delayed neutron fraction beff,sodium void reactivity(DqNa),Doppler constant(DqDoppler),and control rod worth(DqCR).In addition,a sensitivity study was used to reveal the major isotope/reaction pairs contributing to the discrepancy observed in the performance of the three libraries using 33 and 44-energy-group structures.
文摘The study gives a brief introduction on development of innovated nuclear system in China,mainly focus on the materials R&D status for the sodium cooled fast reactor.With the high speed development of economy,China needs a huge energy supply;at same time a more cleaning energy to reduce the carbon release is demanded.The nuclear energy is the most cleaning energy at present time,especially the innovated nuclear system which is so-called GenerationⅣpower plants has got its prior development due to its safety, economical and little fission production produced.Fast breeder reactor,as the priority development reactor type in the Gen-Ⅳnuclear system,is the key to the advanced closed fuel cycle technologies.China experimental fast reactor(CEFR ) has been completed the design,construction the synthesis system commissioning and reached its physical criticality on July 21,2010.At China Institute of Atomic Energy,the CEFR and other research facilities have been established,and extensive studies are planning to carry out in the areas of fuel and materials development.This will laid the foundation for the design and development of the future's CFR—900(China Demonstration Fast Reactor) and CCFR(China Commercial Fast Reactor). Highlights of some of materials R&D studies are discussed in this paper.
基金supported by the Basic and Applied Basic Research Foundation of Guangdong Province(Nos.2021A1515010343,2022A1515011582)the Science and Technology Program of Guangdong Province(Nos.2021A0505030026,2022A0505050029).
文摘Various sources of solid particles might exist in the coolant flow of a liquid metal cooled fast reactor(e.g.,through chemical interaction between the coolant and impurities,air,or water,through corrosion of structural materials,or from damaged/molten fuel).Such particles may cause flow blockage accidents in a fuel assembly,resulting in a reduction in coolant flow,which potentially causes a local temperature rise in the fuel cladding,cladding failure,and fuel melt.To understand the blockage formation mechanism,in this study,a series of simulated experiments was conducted by releasing different solid particles from a release device into a reducer pipe using gravity.Through detailed analyses,the influence of various experimental parameters(e.g.,particle diameter,capacity,shape,and static friction coefficient,and the diameter and height of the particle release nozzle)on the blockage characteristics(i.e.,blockage probability and position)was examined.Under the current range of experimental conditions,the blockage was significantly influenced by the aforementioned parameters.The ratio between the particle diameter and outlet size of the reducer pipe might be one of the determining factors governing the occurrence of blockage.Specifically,increasing the ratio enhanced blockage(i.e.,larger probability and higher position within the reducer pipe).Increasing the particle size,particle capacity,particle static friction coefficient,and particle release nozzle diameter led to a rise in the blockage probability;however,increasing the particle release nozzle height had a downward influence on the blockage probability.Finally,blockage was more likely to occur in non-spherical particles case than that of spherical particles.This study provides a large experimental database to promote an understanding of the flow blockage mechanism and improve the validation process of fast reactor safety analysis codes.
基金supported by Defense Key Laboratory of Nuclear Wastes and Environmental Safety Scientific Research Fund of the Southwest University of Science and Technology of China (No.07JGZB07)the China Academy of Engineering Physics Developing Fund
文摘To examine the sterilizing effect and mechanism of neutron radiation, Bacillus subtills var. niger, strain （ATCC 9372） spores were irradiated with the fast neutron from the Chinese fast burst reactor II（CFBR-II）. The plate-count results indicated that the D10 value was 384.6 Gy with a neutron radiation dose rate of 7.4 Gy/min. The rudimental catalase activity of the spores declined obviously with the increase in the radiation dose. Meanwhile, under the scanning electron microscope, no visible influence of the neutron radiation on the spore configuration was detected even if the dose was increased to 4 kGy. The content and distribution of DNA double-strand breaks induced by neutron radiation at different doses were measured and quantified by pulsed- field gel electrophoresis （PFGE）. Further analysis of the DNA release percentage （PR）, the DNA breakage level （L）, and the average molecular weight, indicated that DNA fragments were obvi- ously distributed around the 5 kb regions at different radiation doses, which suggests that some points in the DNA molecule were sensitive to neutron radiation. Both PR and L varied regularly to some extent with the increase in radiation dose. Thus neutron radiation has a high sterilization power, and can induce falling enzyme activity and DNA breakage in Bacillus subtilis spores
文摘China has decided to speed-up the nuclear power development.It is programmed that the nuclear power capacity will reach 40 GWe in 2020 and envisaged 60 GWe and 240 GWe in 2030 and 2050 respectively.The basic strategy of PWR-FBR matched development with Fast reactor metal fuel closed cycle for a sustainable and quick increasing nuclear energy supply is adopted.Another strategy also decided is that the partitioning and transmutation of MA will be realized using fast burner and ADS.The fast reactor engineering development will be divided into three steps:China Experimental Fast Reactor(CEFR 65 MWt/20 MWe),China Prototype/Demonstration Fast Reactor(CPFR/CDFR≥1500 MWt/600 MWe)and China Demonstration Fast Breeder Reactor(CDFBR 1000~1500 MWe).The CEFR is under installation and pre-operation testing with it's first criticality planned in 2009.The design study of CPFR is just started in 2006.Recently a discussion for the second step is under way to faster the fast reactor development by a larger than 600 MWe CPFR and as a role of CDFR.
