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Fissile material dissolved in the molten salt is employed in molten salt reactor(MSR), which circulates through the primary loop. Considering the flow characteristic of fuel in MSR, the neutron poison calculation model such as Xe is different from traditional solid reactor. As an important neutron poison, 135Xe is free to migrate with the flow of molten salt in MSR. The helium bubble system of molten salt reactor can blow the fission products of krypton, xenon and other gases out of the reactor core to improve the neutron economy. The recent TENDL2021 nuclear data library estimates the thermal neutron absorption cross section of 135Xem, which is much higher than 135Xe thermal neutron absorption cross section. In order to predict the effect of xenon on core reactivity more accurately, the effect of 135Xem thermal neutron absorption cross section was added into the Xe poison model based on the lumped volume method. The migration model is made up of cover gas calculation model and xenon worth calculation model. In order to calculate the xenon worth, the loop void fraction was calculated firstly, and the loop void fraction was inserted into the xenon worth calculation model. According to the rate balance equations and mass transfer equations, the concentration of xenon in loop could be calculated to compute the xenon worth. Firstly, the model results considering the flow effect of fuel salt were verified with the experimental results of 8 MW molten salt reactor experiment (MSRE). As the results shown, with the increase of loop void fraction, the steadystate xenon worth increases first and decreases later, which is in good agreement with experimental results of MSRE. Then the influence of fuel salt types and 135Xem on steadystate xenon worth was considered in this paper. The steadystate xenon worth of the 233U system with and without 135Xem is made and compared to a 235U system, indicating that the influence of 135Xem on xenon worth cannot be ignored, which is much higher with the increase of power. Finally, the contribution of graphite, fuel salt and helium bubble to Xe worth was evaluated by the interaction between graphite and helium bubble, indicating that the interaction between helium bubble and graphite cannot be ignored.

BackgroundFLiBe is commonly used as the coolant and carrier salt in liquid molten salt reactors (MSRs). Its certain moderating properties and thermal neutron scattering attributes affect the neutronic performance of the MSR, and this in turn influences the physical design and safe operation of the reactor. Consequently, studying FLiBe's thermal neutron scattering data is essential for MSRs.PurposeThis study aims to analyze the influence of of FLiBe thermal neutron scattering on neutronic performances of a 65-MW MSR.MethodsFirst, according to the requirements, a core model of a 65-MW MSR was established by using the general Monte Carlo procedure. Then, the neutronics performance of the MSR was calculated by considering the scattering cross-section of the free gas model and FLiBe thermal neutron scattering data (e.g., the neutron spectrum, effective multiplication factor, and nuclide reactivity rate). Finally, the changes in the influence of FLiBe thermal neutron scattering effect on neutronic properties under different energy spectra were compared.ResultsThe computation results show that, by considering the thermal scattering effect of FLiBe molten salt, the neutron energy spectrum in the core of the MSR becomes harder, 235U fission rate decreases, the keff value of the reactor decreases, but the density coefficient in the temperature reaction coefficient of the fuel keeps almost unchanged, and the Doppler coefficient decreases by 0.28×10-5 K-1. With the hardening of the energy spectrum, the variation in the 235U fission rate reduction decreases, and the decrease in keff caused by thermal neutron scattering changs from 9.2×10-4 to 2×10-4.ConclusionsTherefore, it is necessary to incorporate FliBe's thermal neutron scattering data into the physical calculations for the MSR core.

BackgroundNuclear reactors have become the main energy supply for future manned Mars exploration missions due to their high energy density, high power-mass ratio and small size. Heat pipe molten salt reactor is an innovative concept combining molten salt reactor and high temperature heat pipe technology.PurposeThis study aims to carry out preliminary neutronics design of a martian surface molten salt reactor (MS)2R core with 1 MWth and a lifetime longer than 5 a.MethodsFirst of all, the core model of (MS)2R was estabished according to the requirements. It mainly included active zone (including fuel salt and heat pipe), active zone wall, reflector, reactor vessel, etc. Then, the MCNP (Monte Carlo N particle Transport Code) and MOBAT were used to optimize the core size and reactivity control of (MS)2R.ResultsThe physical design parameters of the core are obtained: the height and diameter of the core are 90.94 cm and 88.94 cm, respectively, the uranium inventory is 146.08 kg, and the core mass is 2.09×103 kg. Reactivity control is achieved by a control drum system, in which the B4C (10B enrichment is 90%) with a thickness of 1.8 cm and a wrap angle of 120° is used as the neutron absorber.ConclusionsThe control drum arrangement can meet the critical safety requirements during the lifecycle of (MS)2R under full power (1 MWth) operation for five years, and the number of heat pipes can meet the heat transfer safety limit. This study provides a basic theoretical reference for the design of planet surface molten salt reactor.

Molten salt reactors (MSRs) are one type of GEN-IV advanced reactors that adopt melt mixtures of heavy metal elements and molten salt as both fuel and coolant. The liquid fuel allows MSRs to perform online refueling, reprocessing, and helium bubbling. The fuel utilization, safety, and economics can be enhanced, while some new physical mechanisms and phenomena emerge simultaneously, which would significantly complicate the numerical simulation of MSRs. The dual roles of molten fuel salt in the core lead to a tighter coupling of physical mechanisms since the released fission energy will be absorbed immediately by the molten salt itself and then transferred to the primary heat exchanger. The modeling of multi-physics coupling is regarded as one important aspect of MSR study, attracting growing attention worldwide. Up to now, great efforts have been made in the development of MSR multi-physics coupling models over the past 60 years, especially after 2000, when MSR was selected for one of the GEN-IV advanced reactors. In this paper, the development status of the MSR multi-physics coupling model is extensively reviewed in the light of coupling models of N-TH (neutronics and thermal hydraulics), N-TH-BN (neutronics, thermal hydraulics, and burnup) and N-TH-BN-G (neutronics, thermal hydraulics, burnup, and graphite deformation). The problems, challenges, and development trends are outlined to provide a basis for the future development of MSR multi-physics coupling models. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

• A novel concept of molten salt reactor moderated by heavy water (HW-MSR) is proposed. • HW-MSR inherits the positive properties of MSR and heavy water reactor. • HW-MSR can address the problems of depleted graphite management and positive temperature coefficient existed in MSR. • BR and doubling time of HW-MSR are respective 1.073 and 12 years at the reprocessing cycle of 10 days. 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. [ABSTRACT FROM AUTHOR]

Highlights • The MODEC code has been developed for MSR depletion system. • TTA-based and CRAM-based methods of treating external feed are implemented and compared. • MODEC is able to trace in real time the evolution of the in-stockpile nuclides which is extracted by on-line reprocessing. • Code-to-code comparisons prove the accuracy of MODEC. • A detailed error analysis of ORIGEN-S has been conducted. Abstract 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. [ABSTRACT FROM AUTHOR]

To obtain an outstanding neutron economy and Th-U (Thorium-Uranium) breeding performance, a quite high 7Li enrichment up to 99.995% is generally required for the designs of molten salt reactors with FLiBe carrier salt. While such high 7Li enrichment would undoubtedly bring great technical challenges and high costs to the molten salt preparation. Meanwhile, the 6Li neutron absorption cross section differs greatly under different neutron spectra, which would probably mean the choice of 7Li enrichment can be different for the molten salt reactors with different neutron spectra. With the aim to provide an appropriate reference for the choice of 7Li enrichment in the engineering design of molten salt reactors, key parameters (including 233U inventory, neutron spectrum, temperature coefficient, 7Li enrichment, 7Li and 6Li neutron absorption rate, breeding ratio, relative 233U production and doubling time) and their time evolution are systematically explored based on the in-house developed code MSR-RS. The results show that 99.99% and 99.95% in 7Li enrichment introduce limited influence on Th-U breeding performance for a thermal and fast reactor respectively, and their temperature coefficient are negative. By trading off the cost and technologies required for enriching 7Li, the above two 7Li enrichment are therefore recommended. • The influence of 7Li enrichment on Th-U fuel breeding performance for molten salt reactor under different spectra is investigated. • Parameters including 233U inventory, spectrum, temperature coefficient, 7Li enrichment, neutron absorption rate, breeding ratio, relative 233U production and doubling time are explored. • The traditional definition of BR and its relation with 233U net production are discussed. • By taking the cost and technology of enriching 7Li, 99.99% and 99.95% are recommended for TMSBR-T and TMSBR-F respectively. [ABSTRACT FROM AUTHOR]

Recent studies related to the thermal neutron scattering (TNS) data in reactors were generally focused on fuels and moderators, while the TNS data for structural materials has received little attention. The TNS effect of the GH3535 alloy, a nickel-based alloy widely used in molten salt reactor (MSR), is studied in this work. The partial phonon densities of states (PDOS) for the main metals (Ni, Mo, Cr and Fe) in the alloy, which are required for the TNS data generation, are calculated using the density functional perturbation theory (DFPT) based on a special quasi-random structure (SQS) of the alloy. The TNS cross sections for the main metals in the alloy are generated using the modified NJOY code. To preliminarily validate the TNS data, the calculated scattering cross section, combined with the evaluated absorption cross section of the alloy, is compared with the experimental total cross section, which shows a good agreement. Moreover, an integral neutronics experiment of neutron leakage spectrum is designed to further validate the TNS data. The neutron leakage spectrum is simulated by the Monte Carlo method, which can be compared with the future experimental result. Based on the experiment geometry modeling, the TNS effects of the GH3535 alloy are analyzed at multiple incident neutron energies, sample thicknesses and temperatures, respectively. It is found that the TNS effect of the GH3535 alloy on the neutron leakage spectrum cannot be ignored, and the scattering cross section of the alloy with high precision should be included in designs of MSR. • The partial phonon densities of states for main metals in the GH3535 alloy are calculated. • The thermal neutron scattering (TNS) data for the alloy is generated. • The impacts of the alloy's TNS effects on neutron leakage spectra are analyzed. [ABSTRACT FROM AUTHOR]

Molten Salt Reactor (MSR) is a liquid fueled reactor, in which the sparingly soluble fission gases including xenon continuously circulate with the fuel salt in the primary loop, and meanwhile are removed by helium bubbling. This unique operation feature introduces multiple mechanisms and in turn complicates the behavior of xenon, necessitating being accurately modeled. In this study, a migration model of 135Xe and 135mXe in the primary loop for MSR is established based on the equations that describe the nuclide concentration balance and mass transfer among the helium bubbles, fuel salt and graphite. The developed model is validated with the experimental data of Molten Salt Reactor Experiment (MSRE), and presents a better agreement compared with the previous xenon migration model. Sensitivity analysis of some key parameters including injected void fraction, bubble size, mass transfer coefficient, temperature and pressure, on the xenon poison under steady state are then analyzed. The results demonstrate that the injected void fraction, mass transfer coefficient and core operation pressure would impose a significant influence on the xenon poison, which should be paid particular attention in the core design and safety analysis. [ABSTRACT FROM AUTHOR]

• Different fuel scenarios in a molten salt reactor are evaluated for the Cf-252 conversion capability. • An alluring quality factor of Cf-252 with its mass fraction in Cf over 80% can be obtained for the selected targets. • The target with a larger atomic mass is more apt to produce Cf-252, and the Cf-252 production can be further improved with the increasing mole fraction of the target. Californium-252 (Cf-252) is a powerful neutron source for various applications including neutron radiography, reactor startup, cancer therapy, etc. However, the demand for Cf-252 exceeds its supply at present, and therefore becomes one of the most important restrictions on its applications. Molten salt reactor (MSR) has unique advantages of no fuel rod fabrication, online reprocessing and very deep burnup, which permits flexible core design, substantial target loading into the fuel salt and sufficient depletion time, and is therefore expected to provide an effective solution for Cf-252 production. In this study, four different targets (DU, Pu, Am and Cm) and various mole fractions of the Am target in an MSR are proposed to evaluate the Cf-252 conversion capability. The results indicate that an alluring quality factor of Cf-252 with its mass fraction of over 80% in Cf can be obtained for the selected targets. The highest inventory and conversion rate for Cf-252 with the Cm target can achieve about 700 g and 0.14%, respectively, during the first 10-year operation. The Cf-252 production can be further improved with the increased mole fraction of target but with a deteriorated conversion rate of Cf-252 due to the hardened neutron spectrum and the weakened neutron captures of most isotopes in the reaction chains of Cf-252 formation. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

• Influences of reprocessing separation efficiency (RSE) on HWMSR fuel cycle performances are analyzed. • Lowering RSE would improve the breeding ratio but prolong the double time. • Radiotoxicity of lost heavy metal decreases with RSE and reprocessing cycle time (RCT). • An RCT of 60 days is recommended for HWMSR with the RSE of 99.9%. 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%. [ABSTRACT FROM AUTHOR]

The production capacity of 89Sr and 90Sr in the 2 MW MSR are evaluated. The gaseous 89Kr and 90Kr are extracted from the core through the helium bubbling system, and then decay to 89Sr and 90Sr, respectively. In order to improve purity of 89Sr product, two cooling devices are adopted in the 89Sr and 90Sr production system. The annual yields of 89Sr and 90Sr are about 9000 Ci and 32 Ci, respectively, and the impurity of 89Sr product is less than 2 ppm which can meet the medical requirement. • The production capacity of 89Sr and 90Sr in the 2 MW Molten Salt Reactor are evaluated and the annual yield of 89Sr and 90Sr are about 9000 Ci and 32 Ci, respectively. • The production of 89Sr by the decay chain of 89Kr->89Rb->89Sr through the helium bubbling system has been assessed. • The cooling time of Kr in cooling device 1 is optimized to be 6 min to improve the quality of 89Sr product. [ABSTRACT FROM AUTHOR]

Four I-131 production methods including irradiated TeO 2 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. • The I-131 production capability in a 2 MW Molten Salt Reactor with different methods is evaluated. • The annual production activity of I-131 for online extracted solid tellurium through the BPLS method is about 155,000 Ci. • The radioactivity shielding demand for the latter method is much smaller than the other I-131 production methods. [ABSTRACT FROM AUTHOR]

The 99Mo production in a 2 MW molten salt reactor using liquid low-enriched uranium (LEU) fuel has been evaluated. The batch-wise extraction period of 99Mo is optimized to be one day corresponding to 9415 6-day Ci/week of the 99Mo production rate. The required amount of uranium is only 4.77 kg annually. The required chemically reprocessed amount of FPs is about 58.4 g annually, accounting for only 4.9% of the solid LEU target method under the identical production capacity of 99Mo. • The 99Mo production capability in a 2 MW Molten Salt Reactor using LEU fuel is evaluated. • The extraction period of one day is selected corresponding to 9415 6-day Ci/week of the 99Mo production rate. • The annually required amount of LEU is only 4.77 kg and no Pu is required to be reprocessed. • The annually reprocessed amount of FPs is about 58.4 g accounting for only 4.9% of the solid LEU target method. [ABSTRACT FROM AUTHOR]