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Deep burnup and high minor actinides (MA) loading are two alluring features for molten salt reactors (MSR) to incinerate the nuclear wastes. The transmutation capability of minor actinides in MSR is tightly related with the neutron spectrum, the loading of MA and the carrier salt compositions. In this work, three MSR core designs (thermal, epithermal and fast) and two types of salt compositions (Flibe and Flinak) with different solubility limits of transuranic elements are chosen for analyzing the transmutation capability of MA. With a significant mole fraction of MA loading (4% in the Flibe salt and 10% in the Flinak salt) and the continuous MA refueling, MSR acquires an excellent transmutation rate. The specific incineration rates of MA in the thermal, epithermal and fast MSR cores with the Flibe salt are about 167, 185 and 206 kg/GWth/y, respectively. With a larger loading of MA in the Flinak salt, a higher annual incineration rate of MA can be obtained, which are about 170, 206 and 247 kg/GWth/y in thermal, epithermal and fast MSRs, respectively. On the other hand, since there is a preferred neutron economy for the Flibe salt, a higher MA incineration ratio is achieved than that for the Flinak salt. When the neutron spectrum varies from the thermal to fast region, the MA incineration ratio ranges from 0.79 to 0.82 for the Flibe salt and it ranges from 0.75 to 0.81 for the Flinak salt. The transmutation capability of MA in MSR is much higher than that in solid-fueled reactors (~20 kg/GW.y in a PWR), which can provide a feasible way for reducing the current nuclear wastes.

To develop the application of 95Nb as an indicator of redox potential for fuel salt in molten salt reactor (MSR), the specific activity of 95Nb in FLiBe salt and its deposition of 95Nb on Hastelloy C276 have been studied. Experimental results indicated that the amount of 95Nb deposited on Hastelloy C276 resulted from its chemical reduction exhibited a positive correlation with the decrease of 95Nb activity in FLiBe salt and the relative deposition coefficient of 95Nb to 103Ru appeared a well correlation with 95Nb activity in FLiBe salt. Both correlations implied that the measurement of 95Nb activity deposited on Hastelloy C276 specimen might provide a quantitative approach for monitoring the redox potential of fuel salt in MSR.

The 235,238UF4 was irradiated by photo-neutrons, distribution and behavior of the fission product 95Nb from irradiated 235,238UF4 in FLiBe salt were investigated by the measurement of its activity in the salt with the γ-ray spectroscopy. The experiments indicated that a part of 95Nb deposited on the surfaces of graphite and Hastelloy, as the moderator and the structural materials of molten salt reactor (MSR), respectively, and the majority of 95Nb maintained in molten salt. Addition of lithium metal made 95Nb in salt to be reduced and settled, leading to the decrease in its activity. Degree of the decrease was found to be correlated with niobium concentration. The experimental results supported the statement proposed early by ORNL, that 95Nb might be used as a redox indicator for MSR. Finally, the problem met with on-site monitoring for redox potential in MSR was pointed, and a possible protocol to resolve the problem was proposed.

Distribution and behavior of 233Pa, essential in the thorium–uranium nuclear fuel cycle, were studied in 2LiF–BeF2 (66 : 34 mole%, FLiBe) molten salt by γ-ray spectrometry. The experiments showed that 233Pa deposited slightly on the surface of graphite crucible. The addition of Hastelloy and metallic lithium decreased the 233Pa specific activity in the salt by 1 to 2 orders of magnitude rapidly. Analysis indicated that reductive deposition of 233Pa was responsible for the rapid decrease of 233Pa specific activity in the salt. Additional experiments strongly supported the mechanism of reductive deposition of 233Pa induced by Hastelloy and metallic lithium. In view of the large deposition of 233Pa on Hastelloy, the possible influence of fissile nuclide 233U produced from 233Pa decay on the operation of thorium-based molten salt reactor was discussed.

In this study, the behavior of fission product iodine released from the melting process of a mixture consisting of UF4 irradiated with neutrons and 2LiF–BeF2 (FLiBe) salt was studied. The experiment showed that a large amount of iodine was released immediately during melting and captured by Ni metal foils. The transient release of iodine observed in this experiment is attributed to the redox reaction between the hot atoms of the fission product iodine that cumulated due to long-time irradiation. The effect of the redox status of the molten salt on the transient release of iodine was also investigated. Based on this investigation, it was proposed that the activity ratios of 131I to salt-seeking fission products in the fuel salt, as an effective diagnostic criterion, may be used for the surveillance of the redox potential of fuel salts in a molten salt reactor.

The Heavy Water moderated Molten Salt Reactor (HW-MSR) is a novel concept of a thermal heterogeneous nuclear reactor and pursues for Thorium-Uranium (Th-U) breeding. It adopts heavy water rather than graphite as moderator while employs the same liquid fuel of molten salt reactor (MSR). Thus this new concept inherits the positive properties of MSR (feasible for online reprocessing) and heavy water reactor (high neutron moderating ratio). Furthermore it can address the problems of depleted graphite management and positive temperature coefficient due to neutron spectral shift propitious to 233U fission existed in Molten Salt Breeding Reactor (MSBR). HW-MSR consists of the primary loop system, intermediate cooling loop, energy conversion system, external cooling system and online reprocessing system. Since there is a huge temperature difference between molten salt (higher than 600 °C) and heavy water (lower than 100 °C to keep high density) in the core, effectively preventing the heat transfer between them is one main challenge. To address this problem, a thermal insulator made of Yttria Stabilized Zirconia (YSZ) is applied and analyzed. The calculation results show that both the fuel salt and moderator outlet temperature across the core can satisfy the design requirements. Similar to a traditional MSR, the HW-MSR also implements an online reprocessing system which can online extract and recycle useful fuel during operation, providing a feasible approach for Th-U fuel cycle. In addition, because of the outstanding neutron performance of heavy water moderator, a small initial 233U loading mass is required for HW-MSR, and its breeding ratio can achieve 1.073, corresponding to a doubling time of 12 years at the reprocessing cycle of 10 days. Compared with traditional light water reactor (LWR) whose transuranium (TRU) production is 69 kg/GWy, HW-MSR produces only 0.51 kg/GWy. This much lower radiotoxicity of nuclear waste in the HW-MSR would significantly enhance the sustainability of nuclear energy.

Burnup calculation is an essential topic in reactor physics analysis. And in some advanced reactors, the fuel material is continously fed into the reactor core, so that external feed must be introduced into the burnup calculations. In this work, the Laplace transform method for solving the burnup equation with external feed is presented. Based on the Laplace transform method, the Quadrature-based Rational Approximation Method (QRAM) and the Chebyshev Rational Approximation Method (CRAM) have been extended to handle the burnup equation with either constant or time-dependent external feed term. Several numerical test cases are carried out to demonstrate the correctness and high-performance of the proposed method.

Molten Salt Reactor (MSR) has the characteristics of on-line reprocessing and continuously refueling, which bring significant differences from traditional reactors in burnup calculations. To handle its specific burnup characteristics, a Molten Salt Reactor specific depletion code - MODEC has been newly developed. MODEC implements two depletion algorithms to solve the basic burnup equations: the transmutation trajectory analysis (TTA) and the Chebyshev rational approximation method (CRAM). To simulate the on-line reprocessing, the fictive decay constant method is applied. And three different methods are implemented in MODEC to solve the nonhomogeneous burnup equations in the continuously refueling problems. Moreover, it can trace in real time the evolution of the in-stockpile nuclides which is extracted by on-line reprocessing. MODEC has three calculating modes for decay, constant flux and constant power calculations. By comparing with ORIGEN-S, the validity of performance of MODEC in conventional burnup calculations, burnup calculations with on-line reprocessing and burnup calculations with continuously refueling is proved. And a comparison of the three methods of solving nonhomogeneous burnup equations is presented and discussed. Additionally, a detailed analysis of error sources in ORIGEN-S is applied and an unpublished error source is found. Finally, a specific Monte Carlo burnup procedure for actual MSR burnup calculations is developed by coupling KENO-VI with MODEC, and the burnup benchmark of Molten Salt Fast Reactor (MSFR) is calculated to validate the specific Monte Carlo burnup procedure.

In this study, the behavior of fission product iodine released from the melting process of a mixture consisting of UF

The neutron capture cross section of 232Th 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 232Th (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 232Th (n, γ)233Th 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. The datasets are openly available at http://dx.doi.org/10.11922/sciencedb.j00113.00015.

Distribution and behavior of

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), initial fissile 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.

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 of 0.0555 m3 s−1 ensure an acceptable temperature distribution in the fertile blanket.

The production capacity of

Four I-131 production methods including irradiated TeO2 target and uranium target in the irradiation channel, batch-wise extracted iodine from the fuel salt, and online extracted solid tellurium through the by-pass loop system have been assessed in a 2 MW molten salt reactor. The latter method can produce a large annual yield of I-131 (about 155,000 Ci). The radioactivity shielding demand of the latter method is much smaller than the other I-131 production methods under the identical annual yield of I-131.

A multi-group cross section method was developed based on MOBAT code to save computing time. First, three PWR pin cell benchmark cases were computed by MOBAT, MCODE and SCALE. The results of the separate benchmark cases are consistent with each other; the maximum deviation of reactivity at a burnup depth of 100 MWd/kgU is less than 600 pcm, and the difference in nuclide concentrations is within 5%. Next, the same benchmark case was calculated by MOBAT using the multi-group cross section method. The results indicate that the method can give credible results while greatly reducing the computing time. Comparing the original MOBAT benchmark case to the results of the multi-group cross section method, the reactivity difference at a burnup depth of 100 MWd/kgU is about 100, the difference of nuclide concentrations is within 1%, and computation time was 3.5 times faster.

A Channel-type Molten Salt Fast Reactor (CMSFR) concept is studied to improve the Th-U breeding performance and to reduce the loss ratio of effective delayed neutron fraction (βeff). In the CMSFR, fuel salt is filled in the active zone of the cylindrical core for fission heat generation, while fertile salt is contained in SiC tubes distributed regularly in the core for 233U breeding and for transferring fission heat from the fuel salt to the secondary loop. To enhance the neutron capture reaction rate of 232Th, both radial and axial Th fertile blankets are adopted. Only a small part of fuel salt in the reprocessing system is adopted for fission products removing and refueling, which reduces significantly the required initial loading of fuel salt compared with a conventional Molten Salt Reactor (MSR). Based on an in-house developed analysis code for reprocessing system of MSR, the Th-U breeding capability of CMSFR with different molten salt reprocessing rates (from 10 l per day to 50 l per day) is simulated for a 50-year operation. The simulated results show that the doubling time of 233U with 233U as starting fuel shortens significantly to about 37 years compared with the Molten Salt Fast Reactor (MSFR). Furthermore, the Th-U breeding capability with Pu, TRU and Low Enriched Uranium (LEU) as starting fuels is also analyzed. The simulated results indicate that the doubling time of 233U with Pu and TRU as starting fuels is shortened to 22 years and 17 years, respectively. It indicates that both Pu and TRU can be employed for a transition approach to Th-U fuel cycle. The largest molar proportion of Pu with TRU as starting fuel during operation is about 7.1 mol% which is under the Pu solubility limit. Although the Th-233U breeding with LEU as starting fuel can also be achieved at the end of life of operation, it needs external 233U supply (about 0.6 tons) during the first 20 years operation. Both Temperature Reactivity feedback Coefficient (TRC) and βeff for all the starting scenarios are also investigated. The TRC for all the starting scenarios is always negative enough during 50-year operation. The total loss ratio of βeff with 233U, LEU and Pu/TRU as starting fuels is only 13.76%, 13.47% and 13.56%, respectively, which is much smaller than that of the MSFR. Considering the limitation of reactor period T > 15 s, the largest introduced reactivity for 233U, LEU or Pu/TRU as starting fuel shall not be greater than 110 pcm, 202 pcm and 103 pcm, respectively.

Many concepts of molten salt cooled pebble bed reactor have been developed in recent decades. The packing factor (PF) of the pebble bed in the molten salt reactor should be investigated because it is of great importance for reactor design. Model experiments based on the solid fuel thorium-based molten salt reactor (TMSR-SF) were performed. Experimental results show that the PF of the pebble bed in TMSR-SF (D = 21d, H = 18d, d is pebble diameter) is about 0.57 ± 0.02. The pebble bed in liquid environment is looser than that in dry condition. The PF increases with the diameter of the reactor core and the height of the pebble bed. The geometry of the lower cone reflector may induce large variety of packing factor if the angle of reflector is smaller than the angle of repose of the pebble bed. The loading rate and flow velocity in TMSR-SF are considered to be of little influence on PF. Results from the experiments will be of reference value for the design of reactors.

Compared with the well-studied neutron poison 135 Xe, 149 Sm received much less attention in the conventional reactors because of the comparatively smaller neutron absorption cross section. But in Molten Salt Reactor (MSR), gaseous fission products such as xenon and krypton are largely removed by the gas removal system, the behavior of 149 Sm, which exists in molten salt, therefore becomes important. During MSR’s operation, the liquid fuel continuously circulates through the core (with neutron flux functioning) and outer-loop (the part of primary loop at external core, with negligible neutron flux), the nuclide evolution law is much different from that in the solid fueled reactors, thereby accurately evaluating the time behavior of 149 Sm is much needed. In this work, the control equations precisely describing the nuclide evolution of 149 Sm as well as its precursor 149 Pm entrained in the flowing salt are deduced, and a time behavior during different power cycles is then investigated for a small modular molten salt reactor based on the deduced equations. To fundamentally understand the flow effect on 149 Sm time behavior, the corresponding results are compared with those calculated by using the existent method, in which the flowing fuel is treated as the static fuel. Because the disappeared 149 Pm regardless at the core and outer-loop is all converted to 149 Sm, the flow state has a same equilibrium atomic density with that at the static state. But the time for achieving the equilibrium at the flow state is prolonged due to the time for 149 Pm decayed to 149 Sm being extended at a same 149 Pm production rate. This effect also leads to a lower “samarium peaking (overshoot)” for power up/down, as well as a lower 149 Sm concentration after scram which subsequently decreases the neutron poison (around 210 pcm) as the core start-up again. This work provides a precise description for the evolution of 149 Sm for circulating fuel, which should be included in the core design and source term analysis of safety study of a MSR.

The optimal burnup, which is essential for core design, represents the discharged fuel properties of the core at equilibrium state. The optimal burnup estimation is complex in a pebble bed reactor because the fuel elements are small, numerous and continuously recirculated through the core. In this work, a facilitative method for estimating the optimal burnup of pebble bed reactor is developed. The method employs a single pebble depletion analysis and the neutron non-leakage probability of the core to estimate the attainable burnup of the pebble fuel in the core. With the developed method, the properties of homogeneous mixture fuel of thorium and uranium were analyzed, and the relation of heavy metal utilization with C/HM (graphite-to-heavy metal atom ratio) was also studied. The result shows that, for thorium utilization with homogeneous mixture of seed uranium and thorium fuel, higher enrichment of uranium is preferred. And for a typical pebble design with 20% enriched seed uranium, the heavy metal utilization is about 0.64%, which is equivalent to traditional pressurized water reactor.

Based on the advanced high temperature reactor (AHTR), which is an advanced concept combining attractive attributes by adopting liquid salt and coated particle fuel, more and more reactor designs that use fuel pebbles are developed. The bed structure in liquid salt cooled pebble bed reactor is important for reactor design. To investigate the bed structure in the reactor, the Pebble Recirculation Experiment Device (PRED) and the Pebble Bed Dense Experiment facility (PBDE), which were mock-ups for TMSR-SF (Thorium-based Molten Salt Reactor with Solid Fuel) were developed. Packing experiments were performed under different conditions. The results show that, the packing factor of TMSR-SF is supposed to be about 0.574, which is much less than 0.60 in gaseous environment; the loading rate of pebbles and the velocity of inlet flow has great effect on the bottom shape of the pebble bed; the structure of the pebble bed remains stable during normal operation and changes under accident conditions; loss of the coolant in the core may cause rearrangement of pebble bed; increase in packing factor that caused by strong vibration may induce reactivity in reactor core. The pebble bed behavior under different conditions that investigated in this paper provides the basis phenomenological analysis for reactor design. Moreover, the running of PRED demonstrates the feasibility of the methods for pebble loading and defueling in liquid salted cooled pebble bed reactors.

The Back-n facility at the China Spallation Neutron Source (CSNS Back-n facility) is a newly-built time-of-flight (TOF) spectrometer for neutron nuclear data measurements. The Energy Resolution Function (ERF) of a neutron spectrometer is essential for the analysis of the data measured on the facility, especially in the resolved energy region, as the ERF represents the intrinsic broadening effect of the measured resonance peaks. For the constructing of the ERF, the most crucial component need to be investigated is the neutron production and transport in the spallation target, which could only be inferred from detailed Monte-Carlo simulations. In the present work, the simulation is performed with the Monte-Carlo toolkit Geant4 on account of its flexible features of particle tracking and parameter recording. The production and transport of neutrons in the target–moderator–reflector (TMR) assembly of the Back-n facility are simulated, and the simulation data are analyzed to deduce the distributions of the time spent of neutrons’ transport in the TMR assembly. Based on the distributions of neutrons with different energies from eV to MeV, the ERF of the Back-n facility is constructed by adapting the RPI resolution function-form which is extensively used in neutron nuclear data evaluation. The energy dependence of the ERF is also parameterized to ensure its applicability for the nuclear data evaluation in the resolved energy region. Furthermore, the standard neutron capture cross section of 197Au is taken to verify the accuracy of the ERF by comparing the measured data and the theoretical ones from the ENDF/B-VIII.0 library.

Transuranic (TRU) elements can be used as an alluring fuel in a graphite-moderated Molten Salt Reactor (MSR) with liquid fuel and online reprocessing. However, the positive temperature feedback coefficient (TFC) stemming from the resolvable fission resonances of Pu-239 and Pu-241 at thermal neutron energies should be addressed. The aim of this study is to propose an effective solution to ameliorate the positive TFC, by adding Er-167 into the fuel salt or adjusting the minor actinides (MA) ratio in TRU. Er-167, with a strong capture resonance cross section of about 10,000 barns at 0.5 eV, is one of the most promising poisons. After adding Er-167 into the fuel salt with hectogram scale, the fuel TFC is decreased significantly to below −10 pcm/K and the total TFC to −2 pcm/K correspondingly. Similar to Er-167, some MA isotopes, especially Np-237 (at 0.5 eV) and Am-241 (at 0.3 eV, 0.6 eV and 1.1 eV) reveal significant capture resonance cross sections (>1000 barns). In this way, increasing the MA ratio in TRU is another available approach to obtain a negative TFC. The simulated results show that adjusting the MA ratio up to 50% can ameliorate the TFC efficiently. Combing the negative TFC and the salt solubility limit, only a narrow region of the fuel volume fraction lower than 25% can meet the two requirements, and the eligible region becomes smaller for a lower MA ratio.

Reprocessing separation efficiency (RSE) is one important fuel cycle indicator which determines the loss of actinides left in the nuclear waste and the recovered actinides for reuse in the core, affecting the nuclear waste management, fuel utilization and other fuel cycle characteristics. By changing the reprocessing cycle time (RCT) from 10 days to 360 days, the impacts of RSE varying from 99.9% to 99.999% on the core actinides inventory evolution, breeding ratio (BR) and nuclear waste radiotoxicity were investigated for a Heavy Water moderated Molten Salt Reactor (HWMSR), which is a newly proposed molten salt reactor (MSR) that adopts heavy water rather than graphite as the moderator while employs the usual liquid fuel as in a traditional MSR. The obtained results demonstrate that the inventories of transuranic (TRU) and uranium except U-233 at equilibrium and their transition time to equilibrium both decreases as RSE declines due to their increased mass loss in reprocessing, which in turn brings down the parasitic neutron absorption in the core. Consequently, the U-233 inventory required for maintaining critical operation for a lower RSE drops. While the Th-232 inventory rises, since it is online refueled to maintain the total heavy metal (HM) content in the core constant to ensure the electrochemical stability of the fuel salt. As a result, BR is relatively improved as RSE decreases. But the increased U-233 production resulting from the improved BR could not compensate for the U-233 loss during reprocessing and leads to a longer doubling time for a lower RSE. The radiotoxicity of nuclear waste increases as RSE decreases since more actinides are lost to the nuclear waste. The above effects caused by RSE are mitigated as the RCT prolongs since the frequency for reprocessing the fuel salt of primary loop over a given period decreases, leading to a decrease of HM loss. For an RSE level of 99.9%, the shortest doubling time appears at the RCT of 30 days rather than 10 days because the mitigation of U-233 loss overwhelms the efficiency decrease of FPs removal and Pa-233 extraction. By balancing the Th-U breeding and radiotoxicity, an RCT of 60 days is recommended for HWMSR for the RSE of 99.9%.

The total neutron cross sections of natural beryllium in the neutron energy region of 0.007 to 0.1 eV were measured by using a time-of-flight (TOF) technique at the Shanghai Institute of Applied Physics (SINAP). The low energy neutrons were obtained by moderating the high energy neutrons from a pulsed photo-neutron source generated from a 16 MeV electron linac. The time dependent neutron background component was determined by employing the 12.8 cm boron-loaded polyethylene (PEB) (5% w.t.) to block neutron TOF path and using the Monte Carlo simulation methods. The present data was compared with the fold Harvey data with the response function of the photo-neutron source (PNS, phase-1). The present measurement of total cross section of natBe for thermal neutrons based on PNS has been developed for the acquisition of nuclear data needed for the Thorium Molten Salt Reactor (TMSR).

Breeder reactors are considered as a unique tool for fully exploiting natural resources. Fast breeder reactors based on thorium fuel can enhance inherent safety. Fluoride salt has good performance as a coolant in high-temperature nuclear systems. However, there is some doubt about the fuel breeding ability using fluoride salt coolant for fast spectrum due to its moderating ability. The aim of this study was to choose a proper fluoride salt mixture for Liquid-salt-cooled Solid-fuel Fast Reactor (LSFR) based on thorium-uranium fuel and give parametric studies to provide a design window for flexible self-sustaining core design. Infinite assembly model was used to analyze the salt selection from five candidate fluorides for fast spectrum as coolant. Combining neutron balance analysis with linear least squares fitting method based on 0D model, parametric studies at the neutron balance equation unique solution with burn-up for several parameters such as fuel volume fraction, removing fission gases process, total neutron losses and power density were presented in this paper. It was found that BeF 2 -NaF was a promising coolant in the five candidate fluoride salt mixture. This study proved that the design of a self-sustaining core for fluoride-salt-cooled fast breeder based on thorium fuel is achievable. A design window was found in the definition of a self-sustaining core for various fuel volume fractions and neutron loss fractions. Core design and fuel management strategy will be given in future.

Molten Salt Reactor (MSR) employs fissile material dissolved in the fluoride salt as fuel which continuously circulates through the primary loop with the flow cycle time being a few tens of seconds. The nuclei evolution law is quite different from that in a solid fuel reactor. In this paper, we analytically deduce the nuclei evolution law of 135 Xe and 135 I which are entrained in the flowing salt, evaluate its concentration changing with the burnup time, and validate the result with the SCALE6. The circulation of fuel salt could decrease the concentration of 135 Xe and 135 I, and the reduction can achieve to around 40% and 50% for 135 Xe and 135 I respectively at a small power level (e.g., 2 MW) when the core has the same fuel salt volume as that of the outer-loop. Furthermore, it can be found that the reduction is inversely proportional to the core to outer-loop volume ratio, but uncorrelated with the mass flow rate under normal operating condition of a MSR. At low core power scale, the flow effect on 135 Xe concentration reduction is apparent, but it is mitigated as the core power scale increases because of the rise of 135 I concentration, which raises its decay to 135 Xe and compensates the loss of 135 Xe due to decay at the outer-loop. The decreased 135 Xe concentration results in a core reactivity increase varying from around 150 pcm to 1000 pcm depending on the core power and core to outer-loop volume ratio.

Herein, we assess the 129I 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 of 129I before the startup of the reactor, and the amount of 129I during operation is kept constant by online feeding 129I. The other adopts only an initial loading of 129I before startup, and no other 129I 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 a 233U 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 of 129I shows a slightly better transmutation performance in comparison with the scenario that adopts online feeding of 129I when the net 233U productions for the two scenarios are kept equal. The initial loading of 129I scenario with LiI = 0.87% molar proportion is recommended for 129I transmutation in the SD-TMSR, and can transmute 1.88 t of 129I in the 233U isobreeding mode over 60 years.

Thermal neutron scattering data has an important influence on the calculation and design of reactor with a thermal spectrum. However, as the only liquid fuel in the Gen-IV reactor candidates, the research on the thermal neutron scattering effect of coolant and somewhat moderator FLiBe has not been carried out sufficiently either experimentally or theoretically. The effect of FLiBe thermal neutron scattering on reactivity of TMSR-LF (thorium molten salt reactor - liquid fuel), TMSR-SF (thorium molten salt reactor - solid fuel) and MSRE (molten salt reactor experiment) were investigated and compared. Results show that the effect of FLiBe thermal neutron scattering on reactivity depends to some extent on the fuel-graphite volume ratio of core. Calculations indicate that FLiBe thermal neutron scattering of MSRE (with the hardest spectrum) has the minimum effect of 41 pcm on reactivity, and FLiBe thermal neutron scattering of TMSR-SF (with the softest spectrum) has the maximum effect of -94 pcm on reactivity, and FLiBe thermal neutron scattering of TMSR-LF has an effect of -61 pcm on reactivity at 900 K.

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 $$^{235}$$U 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 “$$\alpha _{{\mathrm{TM}}}$$”), the temperature coefficient of the fast fission factors ($$\alpha _{{\mathrm{TM}}}(\varepsilon )$$) is positive, while those of the resonance escape probability ($$\alpha _{{\mathrm{TM}}}(p )$$), the thermal reproduction factor ($$\alpha _{{\mathrm{TM}}}(\eta )$$), the thermal utilization factor ($$\alpha _{{\mathrm{TM}}}(f )$$), and the total non-leakage probability ($$\alpha _{{\mathrm{TM}}}(\varLambda )$$) are negative. The results reveal that the magnitudes of $$\alpha _{{\mathrm{TM}}}(\varepsilon )$$ and $$\alpha _{{\mathrm{TM}}}(p )$$ for the MTC are similar. Thus, variations in the MTC with $$^{235}$$U enrichment for different HM proportions are mainly dependent on $$\alpha _{{\mathrm{TM}}}(\eta )$$, $$\alpha _{{\mathrm{TM}}}(\varLambda )$$, and $$\alpha _{{\mathrm{TM}}}(f )$$, but especially on the former two. To obtain a more negative MTC, a lower HM proportion and/or a lower $$^{235}$$U 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.

The