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 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 nonlinear dynamic simulation model based on coordinated control of speed and flow rate for the molten salt reactor and combined cycle systems is proposed here to ensure the coordination and stability between the mol...A nonlinear dynamic simulation model based on coordinated control of speed and flow rate for the molten salt reactor and combined cycle systems is proposed here to ensure the coordination and stability between the molten salt reactor and power system.This model considers the impact of thermal properties of fluid variation on accuracy and has been validated with Simulink.This study reveals the capability of the control system to compensate for anomalous situations and maintain shaft stability in the event of perturbations occurring in high-temperature molten salt tank outlet parameters.Meanwhile,the control system’s impact on the system’s dynamic characteristics under molten salt disturbance is also analyzed.The results reveal that after the disturbance occurs,the controlled system benefits from the action of the control,and the overshoot and disturbance amplitude are positively correlated,while the system power and frequency eventually return to the initial values.This simulation model provides a basis for utilizing molten salt reactors for power generation and maintaining grid stability.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 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
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
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
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
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
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
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 molten salt leakage accident is an important issue in the nuclear safety analysis of molten salt reactors.While the molten salt leaks from the pipeline or storage tank,it will contact the insulation layer outside;...The molten salt leakage accident is an important issue in the nuclear safety analysis of molten salt reactors.While the molten salt leaks from the pipeline or storage tank,it will contact the insulation layer outside;hence,the processes of penetration and spreading play an important role in the development of leakage accidents.In this study,the penetration and diffusion of leaking molten salt(LMS)in an aluminum silicate fiber(ASF)thermal insulation layer were studied experimentally.A molten salt tank with an adjustable outlet was designed to simulate the leakage of molten salt,and the subsequent behavior in the thermal insulation layer was evaluated by measuring the penetra-tion time and penetration mass of the LMS.The results show that when the molten salt discharges from the outlet and reaches the thermal insulation layer,the LMS will penetrate and seep out from the ASF,and a higher flow rate of LMS requires less penetration time and leaked mass of LMS.As the temperature of the LMS and thickness of the ASF increased,the penetration time became longer and the leaked mass became greater at a lower LMS flow rate;when the LMS flow rate increased,the penetration time and leaked mass decreased rapidly and tended to flatten.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
The production of radionuclides(90)^Sr and(131)^I in molten salt reactors is an attractive option to address the global shortage of radionuclides.This study evaluated the production characteristics of(90)^Sr and(131)^...The production of radionuclides(90)^Sr and(131)^I in molten salt reactors is an attractive option to address the global shortage of radionuclides.This study evaluated the production characteristics of(90)^Sr and(131)^I in a modular molten salt reactor,such as equilibrium time,yield,and cooling time of isotopic impurities.The fuel burn-up of a small modular molten salt reactor was analyzed by the Triton module of the scale program,and the variation in the fission yields of the two nuclides and their precursors with burn-up time.The yield of(131)^I and~(131)Te has been increasing during the lifetime.131 I has an equilibrium time of about 40 days,a saturation activity of about 40,300 TBq,and while~(131)Te takes 250 min to reach equilibrium,the equilibrium activity was about 38,000 TBq.The yields of90 Sr and~(90)Kr decreased gradually,the equilibrium time of90 Kr was short,and(90)^Sr could not reach equilibrium.Based on the experimental data of molten salt reactor experiment,the amount of nuclide migration to the tail gas and the corresponding cooling time of the isotope impurities under different extraction methods were estimated.Using the HF-H_2 bubbling method,3.49×10^(5)TBq of(131)^I can be extracted from molten salt every year,and after13 days of cooling,the impurity content meets the medical requirements.Using the electric field method,1296 TBq of(131)^I can be extracted from the off-gas system(its cooling time is 11 days)and 109 TBq of(90)^Sr.The yields per unit power for(131)^I and(90)^Sr is approximately 1350 TBq/MW and 530 TBq/MW,respectively,which shows that molten salt reactors have a high economic value.展开更多
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
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
The neutronic properties of molten salt reactors(MSRs)differ from those of traditional solid fuel reactors owing to their nuclear fuel particularity.Based on the Monte-Carlo N particle transport code,the effects of th...The neutronic properties of molten salt reactors(MSRs)differ from those of traditional solid fuel reactors owing to their nuclear fuel particularity.Based on the Monte-Carlo N particle transport code,the effects of the size and shape of the fuel salt channel on the neutron physics of an MSR cell are investigated systematically in this study.The results show that the infinite multiplication factor(k?)first increases and then decreases with the change in the graphite cell size under certain fuel volume fraction(FVF)conditions.For the same FVF and average chord length,when the average chord length is relatively small,the k?values for different fuel salt channel shapes agree well.When the average chord length is relatively large,the k?values for different fuel salt channel shapes differ significantly.In addition,some examples of practical applications of this study are presented,including cell selection for the core and thermal expansion displacement analysis of the cell.展开更多
The commercialisation of molten salts reactors(MSRs)is hindered by the lack of structural materials capable of withstanding the corrosive environment therein.To address this problem,we herein prepared1 wt%Y_(2)O_(3)md...The commercialisation of molten salts reactors(MSRs)is hindered by the lack of structural materials capable of withstanding the corrosive environment therein.To address this problem,we herein prepared1 wt%Y_(2)O_(3)mdispersion-strengthened Ni Mo-based alloys using powder metallurgy and evaluated their potential as structural materials for MSRs based on their mechanical properties,He swelling behaviour,and molten salt corrosion resistance.In view of the strengthening provided by homogenously dispersed Y_(2)O_(3)particles,all NiMo-Y_(2)O_(3)samples exhibited ultimate tensile strengths and yield strengths exceeding those of the Hastelloy N alloy,a state-of-the-art structural material for MSRs.Moreover,the volume fraction of He bubbles in the NiMo-Y_(2)O_(3)samples(~0.3%)was lower than that in the Hastelloy N alloy(0.58%),which showed that the introduction of Y_(2)O_(3)nanoparticles effectively inhibited He swelling.All NiMo-Y_(2)O_(3)samples showed excellent resistance to molten salt corrosion(as reflected by the absence of obvious holes therein),thus holding great promise for the development of irradiation-and molten salt corrosion-resistant structural materials for high-temperature MSRs.展开更多
In this study,pure Ni was demonstrated to protect the GH3535 alloy from Te vapor corrosion because of its strong absorption capacity.Severe Te corrosion of a single GH3535 alloy sample occurred in Te vapor at 700C,whi...In this study,pure Ni was demonstrated to protect the GH3535 alloy from Te vapor corrosion because of its strong absorption capacity.Severe Te corrosion of a single GH3535 alloy sample occurred in Te vapor at 700C,which manifested as complex surface corrosion products and deep intergranular cracks.However,when pure Ni and the GH3535 alloy were put together in the vessel,the GH3535 alloy was completely protected from Te corrosion at the expense of the pure Ni.Thermodynamic calculations proved that the preferential reaction between pure Ni and Te vapor reduced the activity of Te vapor considerably,preventing the corrosion of the GH3535 alloy.Our study reveals one potential approach for protecting the alloys used in molten-salt reactors from Te corrosion.展开更多
The present study explores the potential of fission gas(Kr and Xe short life radioisotopes)released from a molten salt reactor,the separation of these noble gases using specific absorbents under well fixed conditions ...The present study explores the potential of fission gas(Kr and Xe short life radioisotopes)released from a molten salt reactor,the separation of these noble gases using specific absorbents under well fixed conditions and the utilisation of these radioisotopes for radio-diagnostics.During operation,a molten salt reactor produces noble gas radioisotopes that bubble out from the liquid fuel and that can be sampled and treated for radiopharmaceutical applications including as tols for diagnostics using radioisotopes and/or potentially in radiotherapy for specific Vviral diseases usingβ^(-)emtters.Among them^(133)Xe is currenty used fr ung diagnostics thanks to its 132.9 keγ.The use of^(85)Kr for diagnostics is also examined.Its 514 keγcpuld be used for scintigraphy,However^(133)Xe utilisation imply also itsβ^(-)(E_(mean)≈100 kev)whose mean fre pathway of 100 nm in biological tissue or in wateris much smalier than the mean pathway of the^(95)Krβ^(-).Emphasis is placed on^(133)Xe because of its potential dual ability of imag ing and as a suggested therapeutic tool of viral lung diseases.展开更多
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 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This work was supported by the Chinese TMSR Strategic Pioneer Science and Technology Project(No.XDA02010300).
文摘A nonlinear dynamic simulation model based on coordinated control of speed and flow rate for the molten salt reactor and combined cycle systems is proposed here to ensure the coordination and stability between the molten salt reactor and power system.This model considers the impact of thermal properties of fluid variation on accuracy and has been validated with Simulink.This study reveals the capability of the control system to compensate for anomalous situations and maintain shaft stability in the event of perturbations occurring in high-temperature molten salt tank outlet parameters.Meanwhile,the control system’s impact on the system’s dynamic characteristics under molten salt disturbance is also analyzed.The results reveal that after the disturbance occurs,the controlled system benefits from the action of the control,and the overshoot and disturbance amplitude are positively correlated,while the system power and frequency eventually return to the initial values.This simulation model provides a basis for utilizing molten salt reactors for power generation and maintaining grid stability.
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.
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 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 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 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 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“Strategic Priority Research Program”of the Chinese Academy of Sciences(No.XDA0201002)the Youth Innovation Promotion Association of the Chinese Academy of Sciences(No.2018301).
文摘The molten salt leakage accident is an important issue in the nuclear safety analysis of molten salt reactors.While the molten salt leaks from the pipeline or storage tank,it will contact the insulation layer outside;hence,the processes of penetration and spreading play an important role in the development of leakage accidents.In this study,the penetration and diffusion of leaking molten salt(LMS)in an aluminum silicate fiber(ASF)thermal insulation layer were studied experimentally.A molten salt tank with an adjustable outlet was designed to simulate the leakage of molten salt,and the subsequent behavior in the thermal insulation layer was evaluated by measuring the penetra-tion time and penetration mass of the LMS.The results show that when the molten salt discharges from the outlet and reaches the thermal insulation layer,the LMS will penetrate and seep out from the ASF,and a higher flow rate of LMS requires less penetration time and leaked mass of LMS.As the temperature of the LMS and thickness of the ASF increased,the penetration time became longer and the leaked mass became greater at a lower LMS flow rate;when the LMS flow rate increased,the penetration time and leaked mass decreased rapidly and tended to flatten.
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 Strategic Priority Research Program of the Chinese Academy of Sciences(No.XDA02010000)Thorium Uranium Fuel Cycle Characteristics and Key Problem Research Project(No.QYZDY-SSW-JSC016)Shanghai Natural Science Foundation(No.19ZR1468000)。
文摘The production of radionuclides(90)^Sr and(131)^I in molten salt reactors is an attractive option to address the global shortage of radionuclides.This study evaluated the production characteristics of(90)^Sr and(131)^I in a modular molten salt reactor,such as equilibrium time,yield,and cooling time of isotopic impurities.The fuel burn-up of a small modular molten salt reactor was analyzed by the Triton module of the scale program,and the variation in the fission yields of the two nuclides and their precursors with burn-up time.The yield of(131)^I and~(131)Te has been increasing during the lifetime.131 I has an equilibrium time of about 40 days,a saturation activity of about 40,300 TBq,and while~(131)Te takes 250 min to reach equilibrium,the equilibrium activity was about 38,000 TBq.The yields of90 Sr and~(90)Kr decreased gradually,the equilibrium time of90 Kr was short,and(90)^Sr could not reach equilibrium.Based on the experimental data of molten salt reactor experiment,the amount of nuclide migration to the tail gas and the corresponding cooling time of the isotope impurities under different extraction methods were estimated.Using the HF-H_2 bubbling method,3.49×10^(5)TBq of(131)^I can be extracted from molten salt every year,and after13 days of cooling,the impurity content meets the medical requirements.Using the electric field method,1296 TBq of(131)^I can be extracted from the off-gas system(its cooling time is 11 days)and 109 TBq of(90)^Sr.The yields per unit power for(131)^I and(90)^Sr is approximately 1350 TBq/MW and 530 TBq/MW,respectively,which shows that molten salt reactors have a high economic value.
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 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This work was supported by the Chinese TMSR Strategic Pioneer Science and Technology Project(No.XDA02010000)the Frontier Science Key Program of Chinese Academy of Sciences(No.QYZDYSSW-JSC016)the Shanghai Sailing Program(No.Y931021031).
文摘The neutronic properties of molten salt reactors(MSRs)differ from those of traditional solid fuel reactors owing to their nuclear fuel particularity.Based on the Monte-Carlo N particle transport code,the effects of the size and shape of the fuel salt channel on the neutron physics of an MSR cell are investigated systematically in this study.The results show that the infinite multiplication factor(k?)first increases and then decreases with the change in the graphite cell size under certain fuel volume fraction(FVF)conditions.For the same FVF and average chord length,when the average chord length is relatively small,the k?values for different fuel salt channel shapes agree well.When the average chord length is relatively large,the k?values for different fuel salt channel shapes differ significantly.In addition,some examples of practical applications of this study are presented,including cell selection for the core and thermal expansion displacement analysis of the cell.
基金supported by the National Natural Science Foundation of China(Grant Nos.11975304,12022515,11805261,11805256)Youth Innovation Promotion Association,Chinese Academy of Sciences(Grant No.Y202063)。
文摘The commercialisation of molten salts reactors(MSRs)is hindered by the lack of structural materials capable of withstanding the corrosive environment therein.To address this problem,we herein prepared1 wt%Y_(2)O_(3)mdispersion-strengthened Ni Mo-based alloys using powder metallurgy and evaluated their potential as structural materials for MSRs based on their mechanical properties,He swelling behaviour,and molten salt corrosion resistance.In view of the strengthening provided by homogenously dispersed Y_(2)O_(3)particles,all NiMo-Y_(2)O_(3)samples exhibited ultimate tensile strengths and yield strengths exceeding those of the Hastelloy N alloy,a state-of-the-art structural material for MSRs.Moreover,the volume fraction of He bubbles in the NiMo-Y_(2)O_(3)samples(~0.3%)was lower than that in the Hastelloy N alloy(0.58%),which showed that the introduction of Y_(2)O_(3)nanoparticles effectively inhibited He swelling.All NiMo-Y_(2)O_(3)samples showed excellent resistance to molten salt corrosion(as reflected by the absence of obvious holes therein),thus holding great promise for the development of irradiation-and molten salt corrosion-resistant structural materials for high-temperature MSRs.
基金the National Natural Science Foundation of China(Nos.U2032205,52171023,51971238,and 52005492)Natural Science Foundation of Shanghai(Nos.20ZR1468600 and 19ZR1468200)+1 种基金Shanghai Sailing Program(No.19YF1458300)the Youth Innovation Promotion Association,Chinese Academy of Science(No.2019264)。
文摘In this study,pure Ni was demonstrated to protect the GH3535 alloy from Te vapor corrosion because of its strong absorption capacity.Severe Te corrosion of a single GH3535 alloy sample occurred in Te vapor at 700C,which manifested as complex surface corrosion products and deep intergranular cracks.However,when pure Ni and the GH3535 alloy were put together in the vessel,the GH3535 alloy was completely protected from Te corrosion at the expense of the pure Ni.Thermodynamic calculations proved that the preferential reaction between pure Ni and Te vapor reduced the activity of Te vapor considerably,preventing the corrosion of the GH3535 alloy.Our study reveals one potential approach for protecting the alloys used in molten-salt reactors from Te corrosion.
文摘The present study explores the potential of fission gas(Kr and Xe short life radioisotopes)released from a molten salt reactor,the separation of these noble gases using specific absorbents under well fixed conditions and the utilisation of these radioisotopes for radio-diagnostics.During operation,a molten salt reactor produces noble gas radioisotopes that bubble out from the liquid fuel and that can be sampled and treated for radiopharmaceutical applications including as tols for diagnostics using radioisotopes and/or potentially in radiotherapy for specific Vviral diseases usingβ^(-)emtters.Among them^(133)Xe is currenty used fr ung diagnostics thanks to its 132.9 keγ.The use of^(85)Kr for diagnostics is also examined.Its 514 keγcpuld be used for scintigraphy,However^(133)Xe utilisation imply also itsβ^(-)(E_(mean)≈100 kev)whose mean fre pathway of 100 nm in biological tissue or in wateris much smalier than the mean pathway of the^(95)Krβ^(-).Emphasis is placed on^(133)Xe because of its potential dual ability of imag ing and as a suggested therapeutic tool of viral lung diseases.
