Plutonium utilization in a small modular molten-salt reactor based on a batch fuel reprocessing scheme

A molten salt reactor(MSR)has outstanding features considering the application of thorium fuel,inherent safety,sustainability,and resistance to proliferation.However,fissile material^(233)U is significantly rare at the current stage,thus it is difficult for MSR to achieve a pure thorium-uranium fuel cycle.Therefore,using plutonium or enriched uranium as the initial fuel for MSR is more practical.In this study,we aim to verify the feasibility of a small modular MSR that utilizes plutonium as the starting fuel(SM-MSR-Pu),and highlight its advantages and disadvantages.First,the structural design and fuel management scheme of the SM-MSR-Pu were presented.Second,the neutronic characteristics,such as the graphite-irradiation lifetime,burn-up performance,and coefficient of temperature reactivity were calculated to analyze the physical characteristics of the SM-MSR-Pu.The results indicate that plutonium is a feasible and advantageous starting fuel for a SM-MSR;however,there are certain shortcomings that need to be solved.In a 250 MWth SM-MSR-Pu,approximately 288.64 kg^(233)U of plutonium with a purity of greater than 90% is produced while 978.00 kg is burned every ten years.The temperature reactivity coefficient decreases from -4.0 to -6.5 pcm K^(-1) over the 50-year operating time,which ensures a long-term safe operation.However,the amount of plutonium and accumulation of minor actinides(MAs)would increase as the burn-up time increases,and the annual production and purity of^(233)U will decrease.To achieve an optimal burn-up performance,setting the entire operation time to 30 years is advisable.Regardless,more than 3600 kg of plutonium eventually accumulate in the core.Further research is required to effectively utilize this accumulated plutonium.

Correction to:Assembly-level analysis on temperature coefficient of reactivity in a graphite-moderated fuel salt reactor fueled with low-enriched uranium

Following publication of the original article,the authors observed that both Fig.5 and Fig.4 depict the same image.Figure 5 was inaccurately referenced and displayed.The correct Fig.5 is copied below:The original article has been updated.

Effect of reprocessing on neutrons of a molten chloride salt fast reactor

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.

Assembly-level analysis on temperature coefficient of reactivity in a graphite-moderated fuel salt reactor fueled with low-enriched uranium

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.

Influences of ^7Li enrichment on Th-U fuel breeding for an Improved Molten Salt Fast Reactor(IMSFR)

The molten salt fast reactor(MSFR) shows great promise with high breeding ratio(BR),large negative temperature coefficient of reactivity,high thermal-electric conversion efficiency,inherent safety,and online reprocessing.Based on an improved MSFR optimized by adding axial fertile salt and a graphite reflector,the influences of ~7Li enrichment on Th-U breeding are investigated,aiming to provide a feasible selection for the molten salt with high fissile breeding and a relatively low technology requirement for ~7Li concentration.With the self-developed molten salt reactor reprocessing sequence based on SCALE6.1,the burn-up calculations with online reprocessing are carried out.Parameters are explored including BR,^(233)U production,double time(DT),spectrum,~6Li inventory,neutron absorption,and the tritium production.The results show that the Li enrichment of 99.95% is appropriate in the fast fission reactor.In this case,BR above 1.10 can be achieved for a long time,corresponding to the ^(233)U production of130 kg per year and DT of 36 years.After 80 years' operation,the tritium production for 99.5% is only about 7kg,and there is no obvious increase compared to that for 99.9995%.

Transition toward thorium fuel cycle in a molten salt reactor by using plutonium

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.

Core and blanket thermal-hydraulic analysis of a molten salt fast reactor based on coupling of OpenMC and OpenFOAM

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.

Study of heat transfer by using DEM–CFD method in a randomly packed pebble-bed reactor

The pebble-bed reactor is one of the most promising designs for the nuclear energy industry. In this paper,a discrete element method-computational fluid dynamics(DEM-CFD) approach that includes thermal conduction, radiation, and natural convection mechanisms was proposed to simulate the thermal-fluid phenomena after the failure of forced circulation cooling system in a pebble-bed core. The whole large-scale packed bed was created using the DEM technique, and the calculated radial porosity of the bed was validated with empirical correlations reported by researchers. To reduce computational costs, a segment of the bed was extracted, which served as a good representative of the large-scale packed bed for CFD calculation. The temperature distributions simulated with two different fluids in this DEM-CFD approach were in good agreement with SANA experimental data. The influence of the natural convection mechanism on heat transfer must be taken into account for coolants with strong convective capacity. The proposed DEM-CFD methodology offers a computationally efficient and widely applied method for understanding the heat transfer process in a pebble-bed core. The method can also be easily extended to assess the passive safety features of newly designed fluoride-salt-cooled pebble-bed reactors.

Analysis of Th-U breeding capability for an accelerator-driven subcritical molten salt reactor

Accelerator-driven systems based on molten salt fuel have several unique advantages and features for advanced nuclear fuel utilization.The aim of this work was to study the Th-U breeding capability in such systems,known as‘‘accelerator-driven subcritical molten salt reactors’’(ADS–MSRs).Breeding capacities including conversion ratio and net^(233)U production for various subcriticalities and different minor actinides(MA)loadings were analyzed for an ADS–MSR.The results show that the subcriticality of the core has a considerable effect on the Th-U breeding.A high subcriticality is favorable to improving the conversion ratio,increasing the net^(233)U production,and reducing the doubling time.Specifically,the doubling time for k_(eff)of 0.99 is larger than 80 years,while the counterpart for k_(eff)of 0.93 is only approximately22 years.Nevertheless,in an ADS–MSR with a high initial MA loading,MA results in a non-negligible^(233)U depletion in the first two decades,while increasing the net^(233)U production compared to reactors without MA loading.During the 50 years of operation,for the subcritical reactor(k_(eff)0:97)with MA fraction increasing from 1 to 14%,the net^(233)U production increases from 3.94 to 8.24 t.

Application of material-mesh algebraic collapsing acceleration technique in method of characteristics——based neutron transport code

The algebraic collapsing acceleration(ACA)technique maximizes the use of geometric flexibility of the method of characteristics(MOC).The spatial grids for loworder ACA are the same as the high-order transport,which makes the numerical solution of ACA equations costly,especially for large-size problems.To speed-up the MOC transport iterations effectively for general geometry,a coarse-mesh ACA method that involves selectively merging fine-mesh cells with identical materials,called material-mesh ACA(MMACA),is presented.The energy group batching(EGB)strategy in the tracing process is proposed to increase the parallel efficiency for microscopic crosssection problems.Microscopic and macroscopic crosssection benchmark problems are used to validate and analyse the accuracy and efficiency of the MMACA method.The maximum errors in the multiplication factor and pin power distributions are from the VERA-4 B-2 D case with silver-indium-cadmium(AIC)control rods inserted and are 104 pcm and 1.97%,respectively.Compared with the single-thread ACA solution,the maximum speed-up ratio reached 25 on 12 CPU cores for microscopic cross-section VERA-4-2 D problem.For the C5 G7-2 D and LRA-2 D benchmarks,the MMACA method can reduce the computation time by approximately one half.The present work proposes the MMACA method and demonstrates its ability to effectively accelerate MOC transport iterations.

Preliminary analysis of fuel cycle performance for a small modular heavy water-moderated thorium molten salt reactor

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.

Effect of 37Cl enrichment on neutrons in a molten chloride salt fast reactor

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.

Measurement of ^(232)Th(n,γ)cross section at the CSNS Back-n facility in the unresolved resonance region from 4 keV to 100 keV

The neutron capture cross section of ^(232)Th was measured at the neutron time-of-flight facility Back-n of China Spallation Neutron Source(CSNS)for the first time.The measurement was performed with 4 hydrogen-free deuterated benzene C6D6 liquid scintillation detectors,in the ES#2 experiment station on the beam line,at a distance of about 76 m from the neutron-production assembly.The total energy detection principle in combination with the pulse height weighting technique(PHWT)was applied to analyze the measured data.Results of the ^(232)Th(n,γ)reaction cross section in the unresolved resonance region from 4 keV to 100 keV were obtained,which shows a good agreement with the existing experimental data from EXFOR,as well as with the evaluated data from the ENDF/B-VIII.0 and CENDL-3.1.In addition,the excitation function of ^(232)Th(n,γ)^(233)Th reaction in the unresolved resonance region was theoretically calculated by using the code TALYS-1.95.By fitting the experimental cross section and theoretical data,the average parameters in the unresolved resonance region were extracted.

Transmutation of 129I in a single-fluid double-zone thorium molten salt reactor

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.

Influence of 235U enrichment on the moderator temperature coefficient of reactivity in a graphite-moderated molten salt reactor

To optimize the temperature coefficient of reactivity(TCR)for a graphite-moderated and liquid-fueled molten salt reactor,the effects of fuel salt composition on the fuel salt temperature coefficient of reactivity(FSTC)were investigated in our earlier work.In this study,we aim to provide a more comprehensive analysis of the TCR by considering the effects of the graphite-moderator temperature coefficient of reactivity(MTC).The effects of^235U enrichment and heavy metal(HM)proportion in the salt mixture on the MTC are investigated from the perspective of the six-factor formula based on a full-core model.For the MTC(labeled“αTM”),the temperature coefficient of the fast fission factors(αTM(ε))is positive,while those of the resonance escape probability(αTM(p)),the thermal reproduction factor(αTM(η)),the thermal utilization factor(αTM(f)),and the total non-leakage probability(αTM(A))are negative.The results reveal that the magnitudes ofαTM(ε)andαTM(p)for the MTC are similar.Thus,variations in the MTC with^235U enrichment for different HM proportions are mainly dependent onαTM(η),αTM(A),andαTM(f),but especially on the former two.To obtain a more negative MTC,a lower HM proportion and/or a lower 235U enrichment is recommended.Together with our previous studies on the FSTC,a relatively soft neutron spectrum could strengthen the TCR with a sufficiently negative MTC.

Influence of α-Al_2O_3 and AlF_3 on pyrohydrolysis of Li_3AlF_6

In this study, Li_3AlF_6 was employed to simulate the molten salt LiF–BeF_2 to explore its pyrohydrolysis behavior and that of its components, i.e., LiF and AlF_3,respectively. The influence of the accelerators α-Al_2O_3 and AlF_3 on the pyrohydrolysis of LiF and Li_3AlF_6 was investigated. Finally, the solid pyrohydrolytic products were characterized by means of X-ray diffraction, and the corresponding reaction mechanisms were proposed. These experimental results indicated that AlF_3 was completely hydrolyzed to the corresponding oxide α-Al_2O_3 at 650 ℃ in 1 h, whereas the complete hydrolysis of LiF and Li_3AlF_6 required the assistance of either α-Al_2O_3 or AlF_3 under the same conditions. The influence of the accelerator α-Al_2O_3 and AlF_3 on the pyrohydrolytic behavior of Li_3AlF_6 provides references for future research studies on the pyrohydrolysis of LiF–BeF_2 and multi-component molten salts.

Analysis of burnup performance and temperature coefficient for a small modular molten‑salt reactor started with plutonium

In a thorium-based molten salt reactor(TMSR),it is difficult to achieve the pure 232Th–^(233)U fuel cycle without sufficient^(233)U fuel supply.Therefore,the original molten salt reactor was designed to use enriched uranium or plutonium as the starting fuel.By exploiting plutonium as the starting fuel and thorium as the fertile fuel,the high-purity^(233)U produced can be separated from the spent fuel by fluorination volatilization.Therefore,the molten salt reactor started with plutonium can be designed as a^(233)U breeder with the burning plutonium extracted from a pressurized water reactor(PWR).Combining these advantages,the study of the physical properties of plutonium-activated salt reactors is attractive.This study mainly focused on the burnup performance and temperature reactivity coefficient of a small modular molten-salt reactor started with plutonium(SM-MSR-Pu).The neutron spectra,^(233)U production,plutonium incineration,minor actinide(MA)residues,and temperature reactivity coefficients for different fuel salt volume fractions(VF)and hexagon pitch(P)sizes were calculated to analyze the burnup behavior in the SM-SMR-Pu.Based on the comparative analysis results of the burn-up calculation,a lower VF and larger P size are more beneficial for improving the burnup performance.However,from a passive safety perspective,a higher fuel volume fraction and smaller hexagon pitch size are necessary to achieve a deep negative feedback coefficient.Therefore,an excellent burnup performance and a deep negative temperature feedback coefficient are incompatible,and the optimal design range is relatively narrow in the optimized design of an SM-MSR-Pu.In a comprehensive consideration,P=20 cm and VF=20%are considered to be relatively balanced design parameters.Based on the fuel off-line batching scheme,a 250 MWth SM-MSR-Pu can produce approximately 29.83 kg of ^(233)U,incinerate 98.29 kg of plutonium,and accumulate 14.70 kg of MAs per year,and the temperature reactivity coefficient can always be lower than−4.0pcm/K.

Effects of fuel salt composition on fuel salt temperature coefficient(FSTC)for an under-moderated molten salt reactor(MSR)

With respect to a liquid-fueled molten salt reactor(MSR),the temperature coefficient of reactivity mainly includes the moderator temperature coefficient(MTC)and the fuel salt temperature coefficient(FSTC).The FSTC is typically divided into the Doppler coefficient and the density coefficient.In order to compensate for the potentially positive MTC,the FSTC should be sufficiently negative,and this is mostly optimized in terms of the geometry aspect in pioneering studies.However,the properties of fuel salt also directly influence the FSTC.Thus,the effects of different fuel salt compositions including the^(235)U enrichment,heavy metal proportion in salt phase(HM proportion),and the^7Li enrichment on FSTC are investigated from the viewpoint of the essential six-factor formula.The analysis is based on an undermoderated MSR.With respect to the Doppler coefficient,the temperature coefficient of the fast fission factors(a_T(ξ))is positive and those of the resonance escape probability(a_T(p)),thermal reproduction factor(a_T(g)),thermal utilization factor(a_T(f)),and total non-leakage probability(a_T(λ))are negative.With respect to the density coefficient,a_T(p)and a_T(g)are positive,while the others are negative.The results indicate that the effects of the^(235)U enrichment and HM on FSTC are mainly reflected in a_T(e)and a_T(p),which are the dominant factors when the neutron spectrum is relatively hard.Furthermore,the^7Li enrichment influences FSTC by a_T(f)and a_T(λ),which are the key factors in a relative soft spectrum.In order to obtain a more negative FSTC for an under-moderated MSR,the possible positive density coefficient,especially its a_T(p),should be suppressed.Thus,a lower^(235)U enrichment(albeit higher than a certain value,5 wt%in this article)along with a lower HM proportion and/or a higher^7Li enrichment are recommended.The analyses provide an approach to achieve a more suitable fuel salt composition with a sufficiently negative FSTC.

Development of a dynamics model for graphite-moderated channel-type molten salt reactor

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.

Th–U cycle performance analysis based on molten chloride salt and molten fluoride salt fast reactors

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.