Herein, we assess the129I transmutation capability of a 2250-MWt single-fluid double-zone thorium molten salt reactor(SD-TMSR) by considering two methods. One is realized by loading an appropriate amount of129I before...Herein, we assess the129I transmutation capability of a 2250-MWt single-fluid double-zone thorium molten salt reactor(SD-TMSR) by considering two methods. One is realized by loading an appropriate amount of129I before the startup of the reactor, and the amount of129I during operation is kept constant by online feeding129I.The other adopts only an initial loading of129I before startup, and no other129I is fed online during operation.The investigation first focuses on the effect of the loading of I on the Th-233U isobreeding performance. The results indicate that a233U isobreeding mode can be achieved for both scenarios for a 60-year operation when the initial molar proportion of LiI is maintained within 0.40% and 0.87%, respectively. Then, the transmutation performances for the two scenarios are compared by changing the amount of injected iodine into the core. It is found that the scenario that adopts an initial loading of129I shows a slightly better transmutation performance in comparison with the scenario that adopts online feeding of129I when the net233U productions for the two scenarios are kept equal. The initial loading of129I scenario with LiI = 0.87% molar proportion is recommended for129I transmutation in the SD-TMSR,and can transmute 1.88 t of129I in the233U isobreeding mode over 60 years.展开更多
The impregnation behavior of molten 2LiF–BeF_2(FLiBe) salt into a graphite matrix of fuel elements for a solid fuel thorium molten salt reactor(TMSR-SF) at pressures varying from 0.4 to 1.0 MPa was studied by mercury...The impregnation behavior of molten 2LiF–BeF_2(FLiBe) salt into a graphite matrix of fuel elements for a solid fuel thorium molten salt reactor(TMSR-SF) at pressures varying from 0.4 to 1.0 MPa was studied by mercury intrusion, molten salt impregnation, X-ray diffraction, and scanning electron microscopy techniques.It was found that the entrance pore diameter of the graphite matrix is less than 1.0 μm and the contact angle is about 135°. The threshold impregnation pressure was found to be around 0.6 MPa experimentally, consistent with the predicted value of 0.57 MPa by the Washburn equation. With the increase of pressure from 0.6 to 1.0 MPa, the average weight gain of the matrix increased from 3.05 to 10.48%,corresponding to an impregnation volume increase from 2.74 to 9.40%. The diffraction patterns of FLiBe are found in matrices with high impregnation pressures(0.8 MPa and1.0 MPa). The FLiBe with sizes varying from tens of nanometers to a micrometer mainly occupies the open pores in the graphite matrix. The graphite matrix could inhibit the impregnation of the molten salt in the TMSR-SF with a maximum operation pressure of less than 0.5 MPa.展开更多
A liquid fueled thorium molten salt reactor(TMSR-LF),one of the Generation IV reactors,was designed by the Shanghai Institute of Applied Physics,Chinese Academy of Sciences.This study uses the‘rt code to calculate th...A liquid fueled thorium molten salt reactor(TMSR-LF),one of the Generation IV reactors,was designed by the Shanghai Institute of Applied Physics,Chinese Academy of Sciences.This study uses the‘rt code to calculate the neutron and gamma dose rate distributions around the reactor.Multiple types of tallies and variance reduction techniques were employed to reduce calculation time and obtain convergent calculation results.Based on the calculation and analysis results,the TMSR-LF1 radiation shield with a 60-cm serpentine concrete layer and a 120-cm ordinary concrete layer is able to meet radiation requirements.The gamma dose rate outside the reactor biological shield was 16.1 mSv h-1;this is higher than the neutron dose rate of 3.71×10^(–2)mSv h^(-1).The maximum thermal neutron flux density outside the reactor biological shield was 1.899103 cm^(-2)s^(-1),which was below the 19105 cm^(-2)s^(-1)limit.展开更多
Heavy water-moderated molten salt reactors(HWMSRs)are novel molten salt reactors that adopt heavy water rather than graphite as the moderator while employing liquid fuel.Owing to the high moderating ratio of the heavy...Heavy water-moderated molten salt reactors(HWMSRs)are novel molten salt reactors that adopt heavy water rather than graphite as the moderator while employing liquid fuel.Owing to the high moderating ratio of the heavy water moderator and the utilization of liquid fuel,HWMSRs can achieve a high neutron economy.In this study,a large-scale small modular HWMSR with a thermal power of 500 MWth was proposed and studied.The criticality of the core was evaluated using an in-house critical search calculation code(CSCC),which was developed based on Standardized Computer Analyses for Licensing Evaluation,version 6.1.The preliminary fuel cycle performances(initial conversion ratio(CR),initialfissile fuel loading mass,and temperature coefficient)were investigated by varying the lattice pitch(P)and the molten salt volume fraction(VF).The results demonstrate that the temperature coefficient can be negative over the range of investigated Ps and VFs for both 233U-Th and LEU-Th fuels.A core with a P of 20 cm and a VF of 20%is recommended for 233U-Th and LEU-Th fuels to achieve a high performance of initial CR and fuel loading.Regarding TRU-Th fuel,a core with a smaller P(~5 cm)and larger VF(~24%)is recommended to obtain a negative temperature coefficient.展开更多
Axial gas-liquid separators have been adopted in fission gas removal systems for the development of thorium molten salt reactors. In our previous study, we observed an unsteady flow phenomenon in which the flow patter...Axial gas-liquid separators have been adopted in fission gas removal systems for the development of thorium molten salt reactors. In our previous study, we observed an unsteady flow phenomenon in which the flow pattern is directly dependent on the backpressure in a gas-liquid separator; however, the underlying flow mechanism is still unknown. In order to move a step further in clarifying how the flow pattern evolves with a variation in backpressure, a large eddy simulation(LES) was adopted to study the flow field evolution. In the simulation, an artificial boundary was applied at the separator outlet under the assumption that the backpressure increases linearly. The numerical results indicate that the unsteady flow feature is captured by the LES approach, and the flow transition is mainly due to the axial velocity profile redistribution induced by the backpressure variation. With the increase in backpressure,the axial velocity near the downstream orifice transits from negative to positive. This change in the axial velocity sign forces the unstable spiral vortex to become a stable rectilinear vortex.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
The molten salt reactor(MSR), as one of the Generation Ⅳ advanced nuclear systems, has attracted a worldwide interest due to its excellent performances in safety, economics, sustainability, and proliferation resistan...The molten salt reactor(MSR), as one of the Generation Ⅳ advanced nuclear systems, has attracted a worldwide interest due to its excellent performances in safety, economics, sustainability, and proliferation resistance. The aim of this work is to provide and evaluate possible solutions to fissile 233 U production and further the fuel transition to thorium fuel cycle in a thermal MSR by using plutonium partitioned from light water reactors spent fuel. By using an in-house developed tool, a breeding and burning(B&B) scenario is first introduced and analyzed from the aspects of the evolution of main nuclides, net 233 U production, spectrum shift, and temperature feedback coefficient. It can be concluded that such a Th/Pu to Th/^(233)U transition can be accomplished by employing a relatively fast fuel reprocessing with a cycle time less than 60 days. At the equilibrium state, the reactor can achieve a conversion ratio of about 0.996 for the 60-day reprocessing period(RP) case and about 1.047 for the 10-day RP case.The results also show that it is difficult to accomplish such a fuel transition with limited reprocessing(RP is 180 days),and the reactor operates as a converter and burns the plutonium with the help of thorium. Meanwhile, a prebreeding and burning(PB&B) scenario is also analyzed briefly with respect to the net 233 U production and evolution of main nuclides. One can find that it is more efficient to produce 233 U under this scenario, resulting in a double time varying from about 1.96 years for the 10-day RP case to about 6.15 years for the 180-day RP case.展开更多
To improve the reliability and reduce energy consumption,a conceptual design of a freeze valve is proposed for the thorium-based molten salt reactor(TMSR)concept.Fins were utilized in this new design to enhance heat t...To improve the reliability and reduce energy consumption,a conceptual design of a freeze valve is proposed for the thorium-based molten salt reactor(TMSR)concept.Fins were utilized in this new design to enhance heat transfer and realize passive shut-off function,which could not be realized by the previous design.An experimental apparatus using the fluoride salt FLiNaK was constructed to conduct a series of preliminary solidification and melting experiments.In addition,the enthalpy-porosity method of ANSYS■Fluent solver was applied to simulate the solidification process of the salt at a specified operating temperature.Temperature distributions of the fluoride salt,solidification/melting time,and frozen plug effect were analyzed under natural convection heat transfer in an open space.The calculated salt temperatures exhibited good agreement with the experimental values.The results indicated that the range of effective operating temperature is 530-600℃ for the finned freeze valve.In this study,the ideal set operating temperature of the finned freeze valve was chosen as 560℃ to achieve competent performance.Moreover,560℃ is additionally the highest set operating temperature for maintaining excellent cooling performance and sustaining deep-frozen condition of the salt plug.At this set operating temperature,the simulation data indicated that the molten salt in the flat part of the finned freeze valve will completely solidify at 10.5 min.The percentage of solid salt in the flat and lower transitional parts of the valve reaches 29.60% in 30.0 min.Furthermore,the surface temperature of the proposed freeze valve is 11.10% lower compared with that of the TMSR freeze valve at a cooling gas supply of 173 m^3/h.Therefore,the new freeze valve was proven to be capable of reducing the energy consumption and realizing the passive shut-off function.展开更多
The advantages of once-through molten salt reactors include readily available fuel,low nuclear proliferation risk,and low technical difficulty.It is potentially the most easily commercialized fuel cycle mode for molte...The advantages of once-through molten salt reactors include readily available fuel,low nuclear proliferation risk,and low technical difficulty.It is potentially the most easily commercialized fuel cycle mode for molten salt reactors.However,there are some problems in the parameter selection of once-through molten salt reactors,and the relevant burnup optimization work requires further analysis.This study examined once-through graphitemoderated molten salt reactor using enriched uranium and thorium.The fuel volume fraction(VF),initial heavy nuclei concentration(HN_(0)),feeding uranium enrichment(E_(FU)),volume of the reactor core,and fuel type were changed to obtain the optimal conditions for burnup.We found an optimal region for VF and HN_(0) in each scheme,and the location and size of the optimal region changed with the degree of E_(FU),core volume,and fuel type.The recommended core schemes provide a reference for the core design of a once-through molten salt reactor.展开更多
A novel full-digital real-time neutron flux monitor(NFM) has been developed for thorium-based molten salt reactor(TMSR).The system is based on the highspeed,parallel,and pipeline processing of the field programmable g...A novel full-digital real-time neutron flux monitor(NFM) has been developed for thorium-based molten salt reactor(TMSR).The system is based on the highspeed,parallel,and pipeline processing of the field programmable gate array as well as the high-stability controller area network platform.A measurement range of 10~8 counts per second is achieved with a single fission chamber by utilizing the normalization of the count and Campbell algorithms.With the advantages of using the measurement range,system integrity,and real-time performance,digital NFM has been tested in the Xi'an pulsed reactor fission experiments and was found to exhibit superior experimental performance.展开更多
From a safety point of view, it is important to study the damages and reliability of molten salt reactor structural alloy materials, which are subjected to extreme environments due to neutron irradiation, molten salt ...From a safety point of view, it is important to study the damages and reliability of molten salt reactor structural alloy materials, which are subjected to extreme environments due to neutron irradiation, molten salt corrosion, fission product attacks, thermal stress, and even combinations of these. In the past few years, synchrotron radiation-based materials characterization techniques have proven to be effective in revealing the microstructural evolution and failure mechanisms of the alloys under surrogating operation conditions. Here, we review the recent progress in the investigations of molten salt corrosion,tellurium(Te) corrosion, and alloy design. The valence states and distribution of chromium(Cr) atoms, and the diffusion and local atomic structure of Te atoms near the surface of corroded alloys have been investigated using synchrotron radiation techniques, which considerably deepen the understandings on the molten salt and Te corrosion behaviors. Furthermore, the structure and size distribution of the second phases in the alloys have been obtained, which are helpful for the future development of new alloy materials.展开更多
Delayed neutron loss is an important parameter in the safety analysis of molten salt reactors. In this study,to obtain the effective delayed neutron fraction under flow condition, a delayed neutron precursor transport...Delayed neutron loss is an important parameter in the safety analysis of molten salt reactors. In this study,to obtain the effective delayed neutron fraction under flow condition, a delayed neutron precursor transport was implemented in the Monte Carlo code MCNP. The moltensalt reactor experiment(MSRE) model was used to analyze the reliability of this method. The obtained flow losses of reactivity for 235 U and 233 U fuels in the MSRE are223 pcm and 100.8 pcm, respectively, which are in good agreement with the experimental values(212 pcm and100.5 pcm, respectively). Then, six groups of effective delayed neutron fractions in a small molten salt reactor were calculated under different mass flow rates. The flow loss of reactivity at full power operation is approximately105.6 pcm, which is significantly lower than that of the MSRE due to the longer residence time inside the active core. The sensitivity of the reactivity loss to other factors,such as the residence time inside or outside the core and flow distribution, was evaluated as well. As a conclusion,the sensitivity of the reactivity loss to the residence time inside the core is greater than to other parameters.展开更多
In this study,a numerical flow model of the fission products(FPs)in the primary loop system of a molten salt reactor(MSR)was established and solved using Mathematica 7.0.The simulation results were compared with those...In this study,a numerical flow model of the fission products(FPs)in the primary loop system of a molten salt reactor(MSR)was established and solved using Mathematica 7.0.The simulation results were compared with those of the ORIGEN-S program in the static burnup mode,and the deviation was found to be less than 10%,which indicates that the results are in good agreement.Furthermore,the FPs distribution in the primary loop system under normal operating conditions of the 2 MW MSR was quantitatively analyzed.In addition,the distribution phenomenon of the FPs under different flow rate conditions was studied.At the end of life,the FPs activity in the core region(including active region,and upper and lower plenum regions)accounted for 77.3%,and that in the hot leg #1,main pump,hot leg #2,heat exchanger,and cold leg region accounted for 1.2%,16.15%,0.99%,2.5%,and 1.9%,respectively,of the total FPs in the primary loop under normal operating conditions.The proportion of FPs in the core decreased with the increase in flow rate in the range of 2.24-22,400 cm^3 s^-1.The established analytical method and conclusions of this study can provide an important basis for radiation safety design of the primary loop,radioactive source management design,thermal-hydraulic safety analysis,and radiochemical analysis of FPs of 2 MW MSRs.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
To provide a reliable and comprehensive data reference for core geometry design of graphite-moderated and low-enriched uranium fueled molten salt reactors,the influences of geometric parameters on the temperature coef...To provide a reliable and comprehensive data reference for core geometry design of graphite-moderated and low-enriched uranium fueled molten salt reactors,the influences of geometric parameters on the temperature coefficient of reactivity(TCR)at an assembly level were characterized.A four-factor formula was introduced to explain how different reactivity coefficients behave in terms of the fuel salt volume fraction and assembly size.The results show that the fuel salt temperature coefficient(FSTC)is always negative owing to a more negative fuel salt density coefficient in the over-moderated region or a more negative Doppler coefficient in the under-moderated region.Depending on the fuel salt channel spacing,the graphite moderator temperature coefficient(MTC)can be negative or positive.Furthermore,an assembly with a smaller fuel salt channel spacing is more likely to exhibit a negative MTC.As the fuel salt volume fraction increases,the negative FSTC first weakens and then increases,owing to the fuel salt density effect gradually weakening from negative to positive feedback and then decreasing.Meanwhile,the MTC weakens as the thermal utilization coefficient caused by the graphite temperature effect deteriorates.Thus,the negative TCR first weakens and then strengthens,mainly because of the change in the fuel salt density coefficient.As the assembly size increases,the magnitude of the FSTC decreases monotonously owing to a monotonously weakened fuel salt Doppler coefficient,whereas the MTC changes from gradually weakened negative feedback to gradually enhanced positive feedback.Then,the negative TCR weakens.Therefore,to achieve a proper negative TCR,particularly a negative MTC,an assembly with a smaller fuel salt channel spacing in the under-moderated region is strongly recommended.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 salt reactor(MSR) is one of the six advanced reactor concepts selected by the generation Ⅳ international forum because of its advantages of inherent safety, and the promising capabilities of Th-U breeding an...A molten salt reactor(MSR) is one of the six advanced reactor concepts selected by the generation Ⅳ international forum because of its advantages of inherent safety, and the promising capabilities of Th-U breeding and transuranics transmutation. A dynamics model for the channel-type MSR is developed in this work based on a three-dimensional thermal–hydraulic model(3DTH) and a point reactor model. The 3DTH couples a three-dimensional heat conduction model and a one-dimensional single-phase flow model that can accurately consider the heat conduction between different assemblies. The 3DTH is validated by the RELAP5 code in terms of the temperature and mass flow distribution calculation. A point reactor model considering the drift of delayed neutron precursors is adopted in the dynamics model. To verify the dynamics model, three experiments from the molten salt reactor experiment are simulated. The agreement of the experimental data and simulation results was excellent.With the aid of this model, the unprotected step reactivity addition and unprotected loss of flow of the 2 MWt experimental MSR are modeled, and the reactor power and temperature evolution are analyzed.展开更多
In molten salt reactors(MSRs),the liquid fuel salt circulates through the primary loop and a part of the delayed neutron precursors(DNPs)decays outside the reactor core.To model and analyze the flow field effect of DN...In molten salt reactors(MSRs),the liquid fuel salt circulates through the primary loop and a part of the delayed neutron precursors(DNPs)decays outside the reactor core.To model and analyze the flow field effect of DNPs in channel-type liquid-fueled MSRs,a three-dimensional space-time dynamics code,named ThorCORE3D,that couples neutronics,core thermalhydraulics,and a molten salt loop system was developed and validated with the Molten Salt Reactor Experiment(MSRE)benchmarks.The effects of external loop recirculation time,fuel flow rate,and core flow field distribution on the delayed neutron fraction loss of MSRE at steadystate were modeled and simulated using the ThorCORE3D code.Then,the flow field effect of the DNPs on the system responses of the MSRE in the reactivity insertion transient under different initial conditions was analyzed systematically for the channel-type liquid-fueled MSRs.The results indicate that the flow field condition has a significant effect on the steady-state delayed neutron fractions and will further affect the transient power and temperature responses of the reactor system.The analysis results for the effect of the DNP flow field can provide important references for the design optimization and safety analysis of liquid-fueled MSRs.展开更多
The Molten Salt Reactor(MSR) is one of the six advanced reactor nuclear energy systems for further research and development selected by Generation IV International Forum(GIF),which is distinguished by its core in whic...The Molten Salt Reactor(MSR) is one of the six advanced reactor nuclear energy systems for further research and development selected by Generation IV International Forum(GIF),which is distinguished by its core in which the fuel is dissolved in molten fluoride salt.Because fuel flow in the primary loop,the depletion of MSR is different from that of solid-fuel reactors.In this paper,an MCNP5 and ORIGEN2 Coupled Burnup(MOCBurn) code for MSR is developed under the MATLAB platform.Some new methods and novel arrangements are used to make it suitable for fuel flow in the MSR.To consider the fuel convection and diffusion in the primary loop of MSR,fuel mixing calculation is carried out after each burnup time step.Modeling function for geometry with repeat structures is implicated for reactor analysis with complex structures.Calculation for a high-burnup reactor pin cell benchmark is performed using the MOCBurn code.Results of depletion study show that the MOCBurn code is suitable for the traditional solid-fuel reactors.A preliminary study of the fuel mixture effect in MSR is also carried out.展开更多
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Herein, we assess the129I transmutation capability of a 2250-MWt single-fluid double-zone thorium molten salt reactor(SD-TMSR) by considering two methods. One is realized by loading an appropriate amount of129I before the startup of the reactor, and the amount of129I during operation is kept constant by online feeding129I.The other adopts only an initial loading of129I before startup, and no other129I is fed online during operation.The investigation first focuses on the effect of the loading of I on the Th-233U isobreeding performance. The results indicate that a233U isobreeding mode can be achieved for both scenarios for a 60-year operation when the initial molar proportion of LiI is maintained within 0.40% and 0.87%, respectively. Then, the transmutation performances for the two scenarios are compared by changing the amount of injected iodine into the core. It is found that the scenario that adopts an initial loading of129I shows a slightly better transmutation performance in comparison with the scenario that adopts online feeding of129I when the net233U productions for the two scenarios are kept equal. The initial loading of129I scenario with LiI = 0.87% molar proportion is recommended for129I transmutation in the SD-TMSR,and can transmute 1.88 t of129I in the233U isobreeding mode over 60 years.
基金supported by the Thorium Molten Salt Reactor Nuclear Energy System under the Strategic Priority Research Program of the Chinese Academy of Sciences(No.XDA02030000)the Frontier Science Key Program of the Chinese Academy of Sciences(No.QYZDY-SSW-JSC016)the State Key Laboratory of Particle Detection and Electronics(No.SKLPDE-KF-201811)
文摘The impregnation behavior of molten 2LiF–BeF_2(FLiBe) salt into a graphite matrix of fuel elements for a solid fuel thorium molten salt reactor(TMSR-SF) at pressures varying from 0.4 to 1.0 MPa was studied by mercury intrusion, molten salt impregnation, X-ray diffraction, and scanning electron microscopy techniques.It was found that the entrance pore diameter of the graphite matrix is less than 1.0 μm and the contact angle is about 135°. The threshold impregnation pressure was found to be around 0.6 MPa experimentally, consistent with the predicted value of 0.57 MPa by the Washburn equation. With the increase of pressure from 0.6 to 1.0 MPa, the average weight gain of the matrix increased from 3.05 to 10.48%,corresponding to an impregnation volume increase from 2.74 to 9.40%. The diffraction patterns of FLiBe are found in matrices with high impregnation pressures(0.8 MPa and1.0 MPa). The FLiBe with sizes varying from tens of nanometers to a micrometer mainly occupies the open pores in the graphite matrix. The graphite matrix could inhibit the impregnation of the molten salt in the TMSR-SF with a maximum operation pressure of less than 0.5 MPa.
基金the Chinese Academy of Sciences TMSR Strategic Pioneer Science and Technology Project(No.XDA02010000).
文摘A liquid fueled thorium molten salt reactor(TMSR-LF),one of the Generation IV reactors,was designed by the Shanghai Institute of Applied Physics,Chinese Academy of Sciences.This study uses the‘rt code to calculate the neutron and gamma dose rate distributions around the reactor.Multiple types of tallies and variance reduction techniques were employed to reduce calculation time and obtain convergent calculation results.Based on the calculation and analysis results,the TMSR-LF1 radiation shield with a 60-cm serpentine concrete layer and a 120-cm ordinary concrete layer is able to meet radiation requirements.The gamma dose rate outside the reactor biological shield was 16.1 mSv h-1;this is higher than the neutron dose rate of 3.71×10^(–2)mSv h^(-1).The maximum thermal neutron flux density outside the reactor biological shield was 1.899103 cm^(-2)s^(-1),which was below the 19105 cm^(-2)s^(-1)limit.
基金supported by the Chinese TMSR Strategic Pioneer Science and Technology Project(No.XDA02010000)the National Natural Science Foundation of China(No.11905285)+1 种基金the National Natural Science Foundation of China(No.11790321)the Frontier Science Key Program of the Chinese Academy of Sciences(No.QYZDY-SSW-JSC016)。
文摘Heavy water-moderated molten salt reactors(HWMSRs)are novel molten salt reactors that adopt heavy water rather than graphite as the moderator while employing liquid fuel.Owing to the high moderating ratio of the heavy water moderator and the utilization of liquid fuel,HWMSRs can achieve a high neutron economy.In this study,a large-scale small modular HWMSR with a thermal power of 500 MWth was proposed and studied.The criticality of the core was evaluated using an in-house critical search calculation code(CSCC),which was developed based on Standardized Computer Analyses for Licensing Evaluation,version 6.1.The preliminary fuel cycle performances(initial conversion ratio(CR),initialfissile fuel loading mass,and temperature coefficient)were investigated by varying the lattice pitch(P)and the molten salt volume fraction(VF).The results demonstrate that the temperature coefficient can be negative over the range of investigated Ps and VFs for both 233U-Th and LEU-Th fuels.A core with a P of 20 cm and a VF of 20%is recommended for 233U-Th and LEU-Th fuels to achieve a high performance of initial CR and fuel loading.Regarding TRU-Th fuel,a core with a smaller P(~5 cm)and larger VF(~24%)is recommended to obtain a negative temperature coefficient.
基金supported by the National Natural Science Foundation of China(Nos.11535009 and 51406114)
文摘Axial gas-liquid separators have been adopted in fission gas removal systems for the development of thorium molten salt reactors. In our previous study, we observed an unsteady flow phenomenon in which the flow pattern is directly dependent on the backpressure in a gas-liquid separator; however, the underlying flow mechanism is still unknown. In order to move a step further in clarifying how the flow pattern evolves with a variation in backpressure, a large eddy simulation(LES) was adopted to study the flow field evolution. In the simulation, an artificial boundary was applied at the separator outlet under the assumption that the backpressure increases linearly. The numerical results indicate that the unsteady flow feature is captured by the LES approach, and the flow transition is mainly due to the axial velocity profile redistribution induced by the backpressure variation. With the increase in backpressure,the axial velocity near the downstream orifice transits from negative to positive. This change in the axial velocity sign forces the unstable spiral vortex to become a stable rectilinear vortex.
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.
基金supported by the Chinese TMSR Strategic Pioneer Science and Technology Project(No.XDA02010000)the National Natural Science Foundation of China(No.91326201)the Frontier Science Key Program of the Chinese Academy of Sciences(No.QYZDY-SSW-JSC016)
文摘The molten salt reactor(MSR), as one of the Generation Ⅳ advanced nuclear systems, has attracted a worldwide interest due to its excellent performances in safety, economics, sustainability, and proliferation resistance. The aim of this work is to provide and evaluate possible solutions to fissile 233 U production and further the fuel transition to thorium fuel cycle in a thermal MSR by using plutonium partitioned from light water reactors spent fuel. By using an in-house developed tool, a breeding and burning(B&B) scenario is first introduced and analyzed from the aspects of the evolution of main nuclides, net 233 U production, spectrum shift, and temperature feedback coefficient. It can be concluded that such a Th/Pu to Th/^(233)U transition can be accomplished by employing a relatively fast fuel reprocessing with a cycle time less than 60 days. At the equilibrium state, the reactor can achieve a conversion ratio of about 0.996 for the 60-day reprocessing period(RP) case and about 1.047 for the 10-day RP case.The results also show that it is difficult to accomplish such a fuel transition with limited reprocessing(RP is 180 days),and the reactor operates as a converter and burns the plutonium with the help of thorium. Meanwhile, a prebreeding and burning(PB&B) scenario is also analyzed briefly with respect to the net 233 U production and evolution of main nuclides. One can find that it is more efficient to produce 233 U under this scenario, resulting in a double time varying from about 1.96 years for the 10-day RP case to about 6.15 years for the 180-day RP case.
基金supported by the National Natural Science Foundation of China(No.91326201)the Strategic Pilot Technology Chinese Academy of Sciences(No.XDA0201002)
文摘To improve the reliability and reduce energy consumption,a conceptual design of a freeze valve is proposed for the thorium-based molten salt reactor(TMSR)concept.Fins were utilized in this new design to enhance heat transfer and realize passive shut-off function,which could not be realized by the previous design.An experimental apparatus using the fluoride salt FLiNaK was constructed to conduct a series of preliminary solidification and melting experiments.In addition,the enthalpy-porosity method of ANSYS■Fluent solver was applied to simulate the solidification process of the salt at a specified operating temperature.Temperature distributions of the fluoride salt,solidification/melting time,and frozen plug effect were analyzed under natural convection heat transfer in an open space.The calculated salt temperatures exhibited good agreement with the experimental values.The results indicated that the range of effective operating temperature is 530-600℃ for the finned freeze valve.In this study,the ideal set operating temperature of the finned freeze valve was chosen as 560℃ to achieve competent performance.Moreover,560℃ is additionally the highest set operating temperature for maintaining excellent cooling performance and sustaining deep-frozen condition of the salt plug.At this set operating temperature,the simulation data indicated that the molten salt in the flat part of the finned freeze valve will completely solidify at 10.5 min.The percentage of solid salt in the flat and lower transitional parts of the valve reaches 29.60% in 30.0 min.Furthermore,the surface temperature of the proposed freeze valve is 11.10% lower compared with that of the TMSR freeze valve at a cooling gas supply of 173 m^3/h.Therefore,the new freeze valve was proven to be capable of reducing the energy consumption and realizing the passive shut-off function.
基金supported by the Shanghai Sailing Program(No.19YF1457900)Chinese TMSR Strategic Pioneer Science and Technology Project(No.XDA02010000)+1 种基金National Natural Science Foundation of China(No.12005290)Youth Innovation Promotion Association of the Chinese Academy of Sciences(No.2020261)。
文摘The advantages of once-through molten salt reactors include readily available fuel,low nuclear proliferation risk,and low technical difficulty.It is potentially the most easily commercialized fuel cycle mode for molten salt reactors.However,there are some problems in the parameter selection of once-through molten salt reactors,and the relevant burnup optimization work requires further analysis.This study examined once-through graphitemoderated molten salt reactor using enriched uranium and thorium.The fuel volume fraction(VF),initial heavy nuclei concentration(HN_(0)),feeding uranium enrichment(E_(FU)),volume of the reactor core,and fuel type were changed to obtain the optimal conditions for burnup.We found an optimal region for VF and HN_(0) in each scheme,and the location and size of the optimal region changed with the degree of E_(FU),core volume,and fuel type.The recommended core schemes provide a reference for the core design of a once-through molten salt reactor.
基金supported by the National Natural Science Foundation of China(Nos.11375195 and 11575184)the National Magnetic Confinement Fusion Energy Development Research(No.2013GB104003)
文摘A novel full-digital real-time neutron flux monitor(NFM) has been developed for thorium-based molten salt reactor(TMSR).The system is based on the highspeed,parallel,and pipeline processing of the field programmable gate array as well as the high-stability controller area network platform.A measurement range of 10~8 counts per second is achieved with a single fission chamber by utilizing the normalization of the count and Campbell algorithms.With the advantages of using the measurement range,system integrity,and real-time performance,digital NFM has been tested in the Xi'an pulsed reactor fission experiments and was found to exhibit superior experimental performance.
基金supported by the National key research and development program of China(Nos.2016YFB0700401 and 2016YFB0700404)Natural Science Foundation of Shanghai(Nos.19ZR1468200 and 18ZR1448000)+2 种基金National Natural Science Foundation of China(Nos.51671154,51601213 and 51671122)Strategic Priority Research Program of the Chinese Academy of Sciences(No.XDA02004210)Youth Innovation Promotion Association,Chinese Academy of Science(No.2019264)
文摘From a safety point of view, it is important to study the damages and reliability of molten salt reactor structural alloy materials, which are subjected to extreme environments due to neutron irradiation, molten salt corrosion, fission product attacks, thermal stress, and even combinations of these. In the past few years, synchrotron radiation-based materials characterization techniques have proven to be effective in revealing the microstructural evolution and failure mechanisms of the alloys under surrogating operation conditions. Here, we review the recent progress in the investigations of molten salt corrosion,tellurium(Te) corrosion, and alloy design. The valence states and distribution of chromium(Cr) atoms, and the diffusion and local atomic structure of Te atoms near the surface of corroded alloys have been investigated using synchrotron radiation techniques, which considerably deepen the understandings on the molten salt and Te corrosion behaviors. Furthermore, the structure and size distribution of the second phases in the alloys have been obtained, which are helpful for the future development of new alloy materials.
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Delayed neutron loss is an important parameter in the safety analysis of molten salt reactors. In this study,to obtain the effective delayed neutron fraction under flow condition, a delayed neutron precursor transport was implemented in the Monte Carlo code MCNP. The moltensalt reactor experiment(MSRE) model was used to analyze the reliability of this method. The obtained flow losses of reactivity for 235 U and 233 U fuels in the MSRE are223 pcm and 100.8 pcm, respectively, which are in good agreement with the experimental values(212 pcm and100.5 pcm, respectively). Then, six groups of effective delayed neutron fractions in a small molten salt reactor were calculated under different mass flow rates. The flow loss of reactivity at full power operation is approximately105.6 pcm, which is significantly lower than that of the MSRE due to the longer residence time inside the active core. The sensitivity of the reactivity loss to other factors,such as the residence time inside or outside the core and flow distribution, was evaluated as well. As a conclusion,the sensitivity of the reactivity loss to the residence time inside the core is greater than to other parameters.
基金supported by the Chinese Academy of Sciences TMSR Strategic Pioneer Science and Technology Project(No.XDA02010000)The Frontier Science Key Program of Chinese Academy of Sciences(No.QYZDY-SSW-JSC016)
文摘In this study,a numerical flow model of the fission products(FPs)in the primary loop system of a molten salt reactor(MSR)was established and solved using Mathematica 7.0.The simulation results were compared with those of the ORIGEN-S program in the static burnup mode,and the deviation was found to be less than 10%,which indicates that the results are in good agreement.Furthermore,the FPs distribution in the primary loop system under normal operating conditions of the 2 MW MSR was quantitatively analyzed.In addition,the distribution phenomenon of the FPs under different flow rate conditions was studied.At the end of life,the FPs activity in the core region(including active region,and upper and lower plenum regions)accounted for 77.3%,and that in the hot leg #1,main pump,hot leg #2,heat exchanger,and cold leg region accounted for 1.2%,16.15%,0.99%,2.5%,and 1.9%,respectively,of the total FPs in the primary loop under normal operating conditions.The proportion of FPs in the core decreased with the increase in flow rate in the range of 2.24-22,400 cm^3 s^-1.The established analytical method and conclusions of this study can provide an important basis for radiation safety design of the primary loop,radioactive source management design,thermal-hydraulic safety analysis,and radiochemical analysis of FPs of 2 MW MSRs.
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 Youth Innovation Promotion Association CAS (No.2022258)the National Natural Science Foundation of China (No.12175300)+1 种基金the Chinese TMSR Strategic Pioneer Science and Technology Project (No.XDA02010000)the Young Potential Program of Shanghai Institute of Applied Physics,Chinese Academy of Sciences (No.E1550510)。
文摘To provide a reliable and comprehensive data reference for core geometry design of graphite-moderated and low-enriched uranium fueled molten salt reactors,the influences of geometric parameters on the temperature coefficient of reactivity(TCR)at an assembly level were characterized.A four-factor formula was introduced to explain how different reactivity coefficients behave in terms of the fuel salt volume fraction and assembly size.The results show that the fuel salt temperature coefficient(FSTC)is always negative owing to a more negative fuel salt density coefficient in the over-moderated region or a more negative Doppler coefficient in the under-moderated region.Depending on the fuel salt channel spacing,the graphite moderator temperature coefficient(MTC)can be negative or positive.Furthermore,an assembly with a smaller fuel salt channel spacing is more likely to exhibit a negative MTC.As the fuel salt volume fraction increases,the negative FSTC first weakens and then increases,owing to the fuel salt density effect gradually weakening from negative to positive feedback and then decreasing.Meanwhile,the MTC weakens as the thermal utilization coefficient caused by the graphite temperature effect deteriorates.Thus,the negative TCR first weakens and then strengthens,mainly because of the change in the fuel salt density coefficient.As the assembly size increases,the magnitude of the FSTC decreases monotonously owing to a monotonously weakened fuel salt Doppler coefficient,whereas the MTC changes from gradually weakened negative feedback to gradually enhanced positive feedback.Then,the negative TCR weakens.Therefore,to achieve a proper negative TCR,particularly a negative MTC,an assembly with a smaller fuel salt channel spacing in the under-moderated region is strongly recommended.
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 National Natural Science Foundation of China Key Program(No.91326201)
文摘A molten salt reactor(MSR) is one of the six advanced reactor concepts selected by the generation Ⅳ international forum because of its advantages of inherent safety, and the promising capabilities of Th-U breeding and transuranics transmutation. A dynamics model for the channel-type MSR is developed in this work based on a three-dimensional thermal–hydraulic model(3DTH) and a point reactor model. The 3DTH couples a three-dimensional heat conduction model and a one-dimensional single-phase flow model that can accurately consider the heat conduction between different assemblies. The 3DTH is validated by the RELAP5 code in terms of the temperature and mass flow distribution calculation. A point reactor model considering the drift of delayed neutron precursors is adopted in the dynamics model. To verify the dynamics model, three experiments from the molten salt reactor experiment are simulated. The agreement of the experimental data and simulation results was excellent.With the aid of this model, the unprotected step reactivity addition and unprotected loss of flow of the 2 MWt experimental MSR are modeled, and the reactor power and temperature evolution are analyzed.
基金supported by Strategic Pilot Science and Technology Project of Chinese Academy of Sciences (No. XD02001005)
文摘In molten salt reactors(MSRs),the liquid fuel salt circulates through the primary loop and a part of the delayed neutron precursors(DNPs)decays outside the reactor core.To model and analyze the flow field effect of DNPs in channel-type liquid-fueled MSRs,a three-dimensional space-time dynamics code,named ThorCORE3D,that couples neutronics,core thermalhydraulics,and a molten salt loop system was developed and validated with the Molten Salt Reactor Experiment(MSRE)benchmarks.The effects of external loop recirculation time,fuel flow rate,and core flow field distribution on the delayed neutron fraction loss of MSRE at steadystate were modeled and simulated using the ThorCORE3D code.Then,the flow field effect of the DNPs on the system responses of the MSRE in the reactivity insertion transient under different initial conditions was analyzed systematically for the channel-type liquid-fueled MSRs.The results indicate that the flow field condition has a significant effect on the steady-state delayed neutron fractions and will further affect the transient power and temperature responses of the reactor system.The analysis results for the effect of the DNP flow field can provide important references for the design optimization and safety analysis of liquid-fueled MSRs.
基金supported by the‘‘Strategic Priority Research Program’’of the Chinese Academy of Science(No.XDA02010000)
文摘The Molten Salt Reactor(MSR) is one of the six advanced reactor nuclear energy systems for further research and development selected by Generation IV International Forum(GIF),which is distinguished by its core in which the fuel is dissolved in molten fluoride salt.Because fuel flow in the primary loop,the depletion of MSR is different from that of solid-fuel reactors.In this paper,an MCNP5 and ORIGEN2 Coupled Burnup(MOCBurn) code for MSR is developed under the MATLAB platform.Some new methods and novel arrangements are used to make it suitable for fuel flow in the MSR.To consider the fuel convection and diffusion in the primary loop of MSR,fuel mixing calculation is carried out after each burnup time step.Modeling function for geometry with repeat structures is implicated for reactor analysis with complex structures.Calculation for a high-burnup reactor pin cell benchmark is performed using the MOCBurn code.Results of depletion study show that the MOCBurn code is suitable for the traditional solid-fuel reactors.A preliminary study of the fuel mixture effect in MSR is also carried out.
