Dynamic simulation analysis of molten salt reactor-coupled air-steam combined cycle power generation system

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.

Preliminary safety analysis for heavy water-moderated molten salt reactor

The heavy water-moderated molten salt reactor(HWMSR)is a newly proposed reactor concept,in which heavy water is adopted as the moderator and molten salt dissolved with fissile and fertile elements is used as the fuel.Issues arising from graphite in traditional molten salt reactors,including the positive temperature coefficient and management of highly radio-active spent graphite waste,can be addressed using the HWMSR.Until now,research on the HWMSR has been centered on the core design and nuclear fuel cycle to explore the viability of the HWMSR and its advantages in fuel utilization.However,the core safety of the HWMSR has not been extensively studied.Therefore,we evaluate typical accidents in a small modular HWMSR,including fuel salt inlet temperature overcooling and overheating accidents,fuel salt inlet flow rate decrease,heavy water inlet temperature overcooling accidents,and heavy water inlet mass flow rate decrease accidents,based on a neutronics and thermal-hydraulics coupled code.The results demonstrated that the core maintained safety during the investigated accidents.

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.

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.

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.

Characterization of molten 2LiF–BeF_2 salt impregnated into graphite matrix of fuel elements for thorium molten salt reactor

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.

Application of Monte Carlo method to calculate the effective delayed neutron fraction in molten salt reactor

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.

Synchrotron radiation-based materials characterization techniques shed light on molten salt reactor alloys

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.

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.

Study on dynamic characteristics of fission products in 2 MW molten salt reactor

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.

Numerical and experimental investigation of a new conceptual fluoride salt freeze valve for thorium-based molten salt reactor

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.

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.

Radiation dose distribution of liquid fueled thorium molten salt reactor

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.

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.

Research on the effect of the heavy nuclei amount on the temperature reactivity coefficient in a small modular molten salt reactor

Small modular thorium-based graphite-moderated molten salt reactors (smTMSRs), which combine the advantages of small modular reactors and molten salt reactors, are regarded as a wise development path to speed deployment time. In a smTMSR, low enriched uranium and thorium fuels are used in once-through mode, which makes a marked difference in their neutronic properties compared with the case when a conventional molten salt breeder reactor is used. This study investigated the temperature reactivity coefficient (TRC) in a smTMSR, which is mainly affected by the molten salt volume fraction (VF) and the heavy nuclei concentration in the fuel salt (HN). The fourfactor formula method and the reaction rate method were used to indicate the reasons for the TRC change, including the fuel density effect, the fuel Doppler effect, and the graphite thermal scattering effect. The results indicate that only the fuel density has a positive effect on the TRC in the undermoderated region. Thermal scattering from both salt and graphite has a significant negative influence on the TRC in the overmoderated region. The maximal effective multiplication factor, which shows the highest fuel utilization, is located at 10% VF and 12 mol% HN and is still located in the negative TRC region. In addition, on increasing the heavy nuclei amount from 2 mol% HN to 12 mol% HN (VF = 10%), the total TRC undergoes an obvious change from - 11 to - 3 pcm/K, which implies that the change in the HN caused by the fuel feed online should be small to avoid potential trouble in the reactivity control scheme.

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.

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.

Flow field effect of delayed neutron precursors in liquid-fueled molten salt reactors

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.

Study on the production characteristics of(131)^I and(90)^Sr isotopes in a molten salt reactor

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.

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.