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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 authors found that increasing the vertical equivalent damping ratio would benefit isolation effectiveness, and a 3D-CIB with a vertical equivalent damping ratio of 15~20% is appropriate and acceptable from the viewpoint of displacement control. Nuclear energy has played and will continue playing a key role in the fight against climate change due to its extremely low green-house gas emissions, and high energy density and reliability. References 1 Zhu X., Li J., Lin G., Pan R. Influence of Vertical Equivalent Damping Ratio on Seismic Isolation Effectiveness of Nuclear Reactor Building. [Extracted from the article]

• 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]

• The Laplace transform method is applied in burnup calculations with external feed. • CRAM and QRAM are extended to treat external feed. • The external feed term can have polynomial, exponentially decaying and even more complicated form of time-dependence. • The new method is equivalent to the augmented matrix method, but more flexible and easy-to-implement. 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. [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]

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. [ABSTRACT FROM AUTHOR]

• Transition to thorium fuel cycle on a HWMSR by using low enrichment uranium (LEU) is studied. • LEU with U-235 enrichment no lower than 5% can successfully achieve transition on HWMSR. • By raising LEU enrichment from 5% to 19.75%, the DT shortens from 58 years to 17.3 years at RT of 10 days. • The fuel cycle cost is comparable to that of AP1000 when the RT is longer than 30 days. • The material attractiveness of U and Pu increases and decreases respectively as raising LEU enrichment. Lacking fissile element (U-233) in nature is a key issue faced by Th-U fuel cycle, and existing available fuels should be taken as starting fuel to convert Th-232 to U-233 to realize transition to thorium fuel cycle. Low enrichment uranium (LEU, U-235 enrichment < 20%) is the most accessible fuel, and has been widely studied as starting fuel to realize transition to Th-U fuel cycle in graphite moderated molten salt reactors. Due to an inevitable introduction of large amount of fertile element U-238, however, this transition way has proven to be almost impossible. Heavy water moderated molten salt reactor (HWMSR) is an innovative molten salt reactor adopting heavy water as the moderator and heavy metal elements dissolved in fluoride salt as the fuel. A high neutron economy resulting from heavy water paves the way for the transition from LEU to thorium fuel cycle. The results obtained from this study indicated that transition can be achieved by using LEU with enrichment no lower than 5%. Raising enrichment would also enhance the breeding performance, but exert negative/positive influences on the attractiveness of uranium/plutonium contained in the fuel salt, decrease the heavy metal utilization efficiency and impose a small impact on the fuel cycle cost. [ABSTRACT FROM AUTHOR]

• Experimental tests investigate the impacts of process parameters on he-xe PCHE performance. • The effects of T in , p and m (400–600 K, 1–2 MPa, 5–45 kg/h) on PCHE are investigated. • Heat transfer efficiency reaches at least 93 %. • The nusselt number and fanning factor of he-xe correlations are proposed. Micro nuclear reactors have a wide range of applications. The printed circuit heat exchanger is considered well-suited as a heat exchanger type for reheaters and pre-coolers in the Brayton cycle of such micro nuclear reactors, due to its high compactness, high heat transfer efficiency, and ability to operate at high temperatures and pressures. This work focuses particularly on the experimental analysis of a heat exchanger for a micro-transportable gas-cooled thermal reactor. The research investigated the thermal-hydraulic performance of a helium-xenon (He-Xe) gas mixture with an 8.69 % helium mass fraction in a printed circuit heat exchanger (PCHE). Various experimental parameters, such as inlet temperatures, pressures, and mass flow rates on the hot side, are studied within the range of 400–600 K, 1–2 MPa, and 5–45 kg/h, respectively. The work evaluated the influence of mass flow rate, inlet temperature, and pressure on the thermal-hydraulic performance of the PCHE with the He-Xe gas mixture. Results demonstrated that increased mass flow rates and inlet temperatures significantly enhance the heat transfer capabilities on the He-Xe side. Besides, based on the experimental data, a method integrating genetic algorithms with the median wall temperature method is proposed to separate the thermal resistances on the hot and cold sides of the PCHE. A Nusselt number correlation is established based on the Reynolds number and the Prandtl number. Additionally, a Fanning friction factor correlation is proposed. [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]

Summary: Ex‐core transition to thorium fuel cycle is an effective way to address the issue of the lacking of the 233U in nature. It consists to extract online the thorium bred 233Pa from the core and store it outside. The obtained 233Pa stockpile will decay into the 233U (T1/2 = 27 days), and when enough 233U is cumulated, a new 233U‐based reactor can be loaded. However, this approach will inevitably provoke the heavy nuclides (HN) concentration in the core during the extraction operation, which would affect the chemical stability of fuel salt and subsequently endanger the safety of the core operation. In this study, an improved ex‐core transition with unchanged HN concentration is proposed for a 500‐MWth small modular heavy‐water moderated molten salt reactor (SM‐HWMSR). To achieve this, two operating stages are needed. In the first stage, transuranic (TRU) from the spent fuel of LWRs is used to feed the low enrichment uranium (LEU, 19.75 wt% 235U/U) started core with 300 operating days to ensure negative temperature reactivity coefficient (TRC). In the second stage, the thorium bred 233U from the first stage is mixed with TRU to refuel the core. The proposed ex‐core transition scheme aims to burn TRU, meanwhile breed 233U to start new MSRs. The fuel cycle performance over 60‐years' operation is analyzed. The results demonstrate that it is possible to realize ex‐core transition with unchanged concentration of HN in the fuel salt. As the fraction of 233U in the mixed fuel is set to be 15 mol%, it only takes 3 years to start a new reactor with the stored 233U. The radiotoxicity of TRU‐based fuel is decreased after 60 years' burning in the SM‐HWMSR core. Novelty Statement: "Ex‐core transition to thorium cycle in an SM‐HWMSR with unchanged concentration of heavy metal nuclides in the fuel salt" was investigated by Yapeng Zhang, Jianhui Wu, Jingen Chen, and Xiangzhou Cai. The results demonstrate that it is possible to realize ex‐core transition with unchanged concentration of HN in the fuel salt. As the fraction of 233U in the mixed fuel is set to be 15 mol%, it only takes 3 years to start a new reactor by the stored 233U. [ABSTRACT FROM AUTHOR]

Summary: The single‐fluid double‐zone thorium molten salt reactor (SD‐TMSR) allows for online reprocessing fission products and adding nuclear fuel, which provides a high Th‐U breeding capability for thorium utilization. Although 135Xe in the core can be online extracted by the helium bubbling system during reactor operation to improve the reactor neutron economy, it would decay to the long‐lived 135Cs with a half‐life of 2.3 × 106 years, which is undesired from the viewpoint of the development of nuclear energy. The total production of 135Cs decayed from the extracted 135Xe is about 1.83 tons for the SD‐TMSR after 60 years operation. To reduce the significant accumulation of 135Cs, we propose a cooling transmutation method, by which the produced 135Cs is reinjected into the core for transmutation. First, a separation and multigroup cooling system in the SD‐TMSR is adopted to enhance the mass fraction of decayed 135Cs in the total Cs isotopes from 29% to 75%. Then, the separated Cs isotopes with optimized mass fraction of 135Cs are online injected into the core for transmutation. The total transmuted mass of 135Cs is about 0.61 tons after 60 years operation, corresponding to the transmuted fraction of about 37.42% and the transmutation rate of 10.17 kg/(GWe y). Moreover, the net 233U production is about 2.03 tons which is much larger than the initial 233U loading inventory of the SD‐TMSR, indicating that the Th‐U breeding mode for the transmutation of 135Cs in the SD‐TMSR can be achieved over 60 years. Novelty Statement: "Transmutation of 135Cs in a single‐fluid double‐zone thorium molten salt reactor" was investigated by Kunfeng Ma, Chenggang Yu*, Jingen Chen*, and Xiangzhou Cai. A cooling transmutation method is performed to reduce the 135Cs decayed from 135Xe outside of core. It is found that a separation and multigroup cooling system improves the mass fraction of 135Cs in the total Cs isotopes from 29% to 75% and an online returning 135Cs transmutation scenario transmutes about 0.61 tons of 135Cs over 60 years. [ABSTRACT FROM AUTHOR]

He-Xe gas-cooled micro-molten salt reactor (HX-micro-MSR) is a novel reactor concept that integrates the technologies of molten salt reactor and closed Brayton cycle (CBC), which has a more compact structure compared with the other types of micro-MSRs. The reactor system probed in this paper has a thermal power of 2.5 MWth and consists of the core, fission gas collecting container, and CBC system. The He-Xe gas flows through the coolant channels in the core to transfer the nuclear heat to the CBC for electricity production. To withstand the large pressure difference between the He-Xe gas coolant (7 MPa) and molten salt (nearly atmospheric pressure), the sizes and wall thickness of coolant channels are analyzed and optimized to be 8 mm and 0.5 mm, respectively. In total, 1050 coolant channels are regularly arranged in the core to effectively remove the nuclear heat. The enrichment of U-235 in the fuel UF4 is determined to be 61% to meet the burnup goal of 5 years. To minimize the neutron leakage, the thickness of reflector is selected to be 23 cm. Six control drums with the diameter in 10.5 cm and neutron absorber layer thickness in 1 cm are symmetrically placed in the reflector to realize reactivity control under both normal and accident conditions. Radiation shield is designed and analyzed by varying the thickness of multiple shielding layers, and it is found that a combination of W layer in 14 cm and Mg(BH4)2 layer in 56 cm can satisfy the criterion of radiation dose (<1 mSv/h). The total weight of HX-micro-MSR is estimated to be 35 tonnes and falls into the acceptable range. [ABSTRACT FROM AUTHOR]

Initiating event analysis is an essential prerequisite of conducting probabilistic safety assessment for nuclear reactors, which plays an important role in improving the core design, identifying fault, and guiding operation. In order to determine the initiating event list of SIMONS (Small Innovative helium-xenon cooled Mobile Nuclear power System), preliminary researches on the initial event of SIMONS were carried out using the MLD (Main Logic Diagram) analysis method and referring to the initial event list and initial event analysis theory of other nuclear reactors such as HTGR (High Temperature Gas-cooled Reactor), MSR (Molten Salt Reactor), and PWR (Pressurized water reactor). With employing these methods, a total of 31 initial events are identified for SIMONS based on its latest conceptual design. These initial events are then divided into six groups according to the accident types, which are core heat removal increase, core heat removal decrease, abnormal reactivity and power distribution, pipeline crevasse and equipment leakage, anticipated transients without scram, and disasters (internal and external). The obtained results can provide a theoretical basis for the further safety analysis of SIMONS. [ABSTRACT FROM AUTHOR]

Microtransport with small size and a wide range of applications is very attractive for the utilization of future reactor modularization. A microreactor uses a closed Brayton cycle (CBC) to achieve high conversion efficiencies at low specific mass. The recuperator is one of the key components of CBC system which recovers heat exhausted from the turbine. A printed circuit heat exchanger (PCHE) is currently the preferred type of recuperator for the closed Brayton cycle due to its high heat transfer efficiency, high compactness, and high pressure and temperature resistance. This work intends to analyze different geometry configurations of a zigzag PCHE to increase its heat transfer efficiency while reducing its weight and size. The effect of geometric structure parameters such as channel diameter, zigzag pitch length, and zigzag angle on the Nusselt number and Fanning friction factor is investigated using a zigzag PCHE unit model. Based on the numerical simulation, the principle of least squares is employed to carry out a nonlinear fitting of the flow and heat transfer criterion correlation equations. Besides, the maximum value of the Nusselt number and minimum value of the Fanning friction factor are optimized as two conflicting objective functions using the nondominated sorting genetic algorithm-II (NSGA-II), with which a set of optimal solutions is obtained. Meanwhile, the shortest normalized distance is used to determine the compromised solution on the Pareto optimal points, and the independent variable's sensitivity analysis is performed. Finally, a multiobjective optimization analysis is conducted for the PCHE to achieve lightweight and the high heat transfer efficiency design requirements of SIMONS. [ABSTRACT FROM AUTHOR]

Summary: A liquid‐fueled molten salt reactor (MSR) can be flexibly designed with a broad range of power density (PD) and neutron spectrum, both of which have a strong impact on the breeding performance of the fuel cycle. In this work, a thermal single fluid double‐zone Th‐based MSR (SD‐TMSR) and an improved molten salt fast reactor (IMSFR) with different PDs ranging from 40 to 300 MWth/m3 are introduced to evaluate the breeding performance. The results indicate that the breeding ratios for the pure Th‐U (BRTU) and the fissile fuel (BRfis) at equilibrium state are negatively correlated with PD. As the PD increases from 40 to 300 MWth/m3, the BRTU declines by about 9.2% and 5.6% in the SD‐TMSR and IMSFR, respectively, due to the decreases in the probability of the natural decay and extraction of 233Pa in its total vanishment and of the ratio of the 232Th capture rate to the 233U absorption rate. Meanwhile, the BRfis keeps about 6% lower than the BRTU in both the cores because of the larger absorption cross sections of 235U and 239Pu than the capture cross sections of their parent nuclides. Owing to the low capture‐to‐fission ratios of minor actinides (MAs) and the reduced absorption cross sections of fission products, the BRTU and BRfis in the IMSFR keep about 6% higher than those in the SD‐TMSR when PD varies from 40 to 300 MWth/m3. In addition, the net 233U production can achieve the highest value of 60.5 kg/year at 90 MWth/m3 in the SD‐TMSR, and it is 154.9 kg/year at 200 MWth/m3 in the IMSFR. Although the net 233U production in the IMSFR is higher than that in the SD‐TMSR, the doubling time (DT) keeps larger in the IMSFR with the increase of PD from 40 to 100 MWth/m3 due to its higher initial loading of fissile material. In addition, the highest net 233U production of 154.9 kg/year in the IMSFR with a PD of 200 MWth/m3 can be used as the igniting fissile fuel for the SD‐TMSR, in which the corresponding DT is as short as 8.3 years. In conclusion, to achieve a thorium breeding in MSRs with different neutron spectra, a small PD less than 150 MWth/m3 is preferred to achieve a positive breeding gain and a positive net 233U production in the SD‐TMSR, while a higher PD ranging from 150 to 200 MWth/m3 is more attractive to acquire a higher net 233U production of more than 146 kg/year and a shorter DT of less than 35 years in the IMSFR. [ABSTRACT FROM AUTHOR]

Summary: A thorium‐based chloride salt fast reactor (TCLFR) for transuranics (TRUs) transmutation is proposed to decrease radioactive nuclear waste. The TCLFR core uses two types of fuel: liquid nuclear fuel with TRUs and thorium being dissolved in chloride and solid transmutation rod with only TRUs fuel. The transmutation capability of TCLFR with a core power of 2500 MWth is investigated utilizing three different reactor core configurations, namely ThCore, GdCore, and TcCore, and compared to the standard Molten Chloride Fast Reactor (MCFR) while maintaining most of the geometric parameters. Important burnup parameters such as transmutation rate (TRE) and transmutation ratio (TRO) are extracted and analyzed. Besides, the neutron spectrum shift and the relevant safety parameters, including power peak factor, temperature coefficient of reactivity (TCR) and effective delayed neutron fraction (βeff) are all explored. The results demonstrate that all the TCLFR cores can transmute TRUs effectively. TRUs have a high online refueling amount of TRUs in TcCore and GdCore due to their low conversion ratio (CR). The TRE in the TcCore and GdCore is 292.77 kg/GWth/year and 258.31 kg/GWth/year, respectively, which is significantly higher than the TRE in the MCFR, which is 171.21 kg/GWth/year. As a result of the production of U‐233 compensates for the consumption of TRUs, the refueling amount of TRUs decreases, and the TRE in the ThCore is only 96.29 kg/GWth/year, which is approximately 44% lower than that in the MCFR. [ABSTRACT FROM AUTHOR]

In this work, an analytical expression based on a two-region spherical model for a High Temperature Gas cooled Reactor (HTGR), is applied to a Pebble Bed Fluoride salt cooled High temperature Reactor (PB-FHR) to calculate the average Dancoff factor for fuel kernel with various coated particle densities. The results are compared with those obtained by the Monte Carlo (MC) method. It is found that there is a good agreement between the analytical and MC, which implies that the two-region spherical model for the Dancoff factor calculation is applicable to PB-FHR. The Dancoff factors with and without FLiBe salt are both evaluated and compared to analyze the influence of FLiBe salt with different numbers of coated particles. The Dancoff factor with FLiBe salt is smaller than that without FLiBe salt from ∼ 11% to ∼ 3% when the number of coated particles in a fuel pebble increases from 5000 to 60,000. In addition, the effects of temperature and volume fraction of FLiBe salt on Dancoff factor are analyzed. Afterwards, the sensitivity of the infinite multiplication factor ( k ∞ ) to the Dancoff factor for a fuel cell of PB-FHR is further analyzed by the neutron transport calculation with a new developed routine. It is found that the infinite multiplication factor increases almost linearly with the increase of Dancoff factor, and a higher density of coated particles will lead to a sharper variation of k ∞ with Dancoff factor. [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]

Summary: Zirconium hydride (ZrH) is considered as an alternative moderator in the molten salt reactor (MSR) due to its excellent slowing‐down ability, good thermal stability, and high radiation resistance. In the MSR, ZrH moderator is inserted in the form of rod into the fuel salt of the reactor core, and an axial and a transverse fuel salt flow occur between adjacent fuel salt channels. This inter channel flow mixing effect of fuel salt makes the molten salt reactor moderated by zirconium hydride rods (ZrH‐MSR) substantially different from other type of reactors in both neutronics and thermal hydraulics. Based on the multi‐group neutron diffusion theory and the subchannel thermal‐hydraulics model, a steady‐state analysis code for ZrH‐MSR was developed and verified in this work. Then, a 25 MWth ZrH‐MSR proposed by Transatomic Power Corporation was selected to evaluate its steady‐state characteristics due to its cost‐effective and simplified design. By applying this coupled code, the effects of control rod position, inlet fuel salt flow velocity, and residence time of fuel salt in the external loop on the distributions of the neutron fluxes, delayed neutron precursors (DNPs), and temperature for the 25 MWth ZrH‐MSR were analyzed in detail. The simulation results show that the control rod position influences the neutron fluxes greatly, which indirectly affects the distributions of DNPs and core temperature. The inlet fuel salt flow velocity and the residence time of fuel salt in the external loop have little influence on the neutron fluxes of the core but affect the distribution of DNPs significantly. These numerical results can provide valuable information for the research and design of ZrH‐MSR. [ABSTRACT FROM AUTHOR]

Summary: An increasing number of countries have taken nuclear energy as a preferred approach in response to the environment deterioration and energy supply deficit. The rapid expansion of nuclear technologies, however, would pose a great challenge to nuclear non‐proliferation, especially for the Generation IV nuclear reactor systems, which significantly differ from current nuclear fuel systems. This paper gives an overview of non‐proliferation research activities worldwide and outlines the existing problems, especially in non‐proliferation assessment. Because of numerous processes and various types of variables involved in nuclear fuel cycles (NFC), it is difficult to obtain a quantitative and objective assessment on non‐proliferation. In addition, the influences imposed by national nuclear policy on non‐proliferation have been rarely studied because of their large uncertainties, which may not precisely reflect the real non‐proliferation status in a specific country. In view of the above issues, we put forward an assessment framework by considering impact factors of national nuclear policy and by employing multi‐mathematical models to address some of the issues including subjectivity and uncertainties in the current assessment methodologies. [ABSTRACT FROM AUTHOR]

Summary: Transmutation of minor actinides (MAs) is an effective solution to reduce the long‐term radiotoxicity of nuclear waste. Online refueling and no fuel rod fabrication in thorium‐based molten salt reactors (TMSR) are two remarkable advantages of MA transmutation. However, MA solubility limit of the fuel salt and the neutron spectrum in the core are two key parameters that have a strong impact on MA transmutation capability. In this paper, three types of carrier salt with different MA solubility limits and various salt volume fractions (SVFs) are introduced in a TMSR core to evaluate the MA transmutation characteristics. The results indicate that the Flibe salt with the best neutron economy can obtain the highest MA transmutation rate (TR), while the Flinak salt with the largest MA inventory can achieve the highest specific MA transmutation consumption (STC). STC and TR for the case with the Flibe salt and 5% SVF are about 200 kg/GWth.yr and 82%, respectively, which indicates that it can consume the annual MA yields from about 3.37 typical pressurized water reactors (PWRs). Meanwhile, the STC and TR for the case with the Flinak salt and 40% SVF are about 366 kg/GWth.yr and 75%, respectively, which indicates that it can consume the annual MA yields from about 50 typical PWRs. In addition, the total radiotoxicity of MAs after 50‐year operation in TMSR can be lowered by as high as about 50%. The significant production of Pu isotopes can be reused as nuclear fuel in fast reactors. In particular, Pu‐238, with the mass fraction of more than 60% in Pu isotopes, is a useful material for many nuclear applications. In conclusion, it is feasible to transmute MAs in TMSR and thus achieve the goals of reduction of MA's long‐term radioactive hazards. [ABSTRACT FROM AUTHOR]

• A filtering model and a prediction model for equilibrium state of Molten Salt Reactors are developed. • The LightGBM model is the most favorable for filtering the burnup state and predicting the neutronic parameters. • The LightGBM model obtains the accuracy of 0.98 for classification and the relative error of 1.10% for regression. During the operation of a liquid-fueled molten salt reactor (MSR), the deep depletion to an equilibrium state can be attained with online refueling and reprocessing techniques. However, the equilibrium burnup simulation of an MSR using current depletion codes requires considerable computational cost and time, and even fails when the equilibrium conditions are unsatisfied. To efficiently evaluate the specific characteristics of the equilibrium burnup in MSRs, a fast-filtering model for equilibrium state and a fast prediction model for equilibrium neutronic properties are developed based on the machine learning (ML) technique. A database of 4860 cases including the neutronic information of equilibrium and non-equilibrium of a fuel assembly with variable geometric parameters in an MSR is first generated, and then the neutronic properties of 232Th-233U, 232Th-EU (enriched uranium) and 232Th-Pu fuel cycles are evaluated. Based on the MSR database, fifteen different ML classification algorithms are implemented to establish the fast-filtering model for equilibrium burnup state. Moreover, a model for the fast prediction of equilibrium neutronic parameters of neutron flux (Φ) and breeding ratio (BR) is developed with fifteen ML regression algorithms. Considering the various performance metrics for measuring the predicted performances of ML models in the classification and regression, the LightGBM (LGBM) model is the most favorable for filtering the burnup state and predicting the neutronic parameters in MSRs. With the test data, the LGBM model can obtain the accuracy of about 0.98 for classification and the relative error of about 1.10% for regression. [ABSTRACT FROM AUTHOR]

Summary: Using low enrichment uranium as driver fuel under once‐through mode in molten salt reactor (MSR) attracts more and more attention because of its fuel availability, no new technology, and nuclear nonproliferation. It is regarded as a wise research and development road to shorten deployment time of MSRs and to prepare techniques and experiences for thorium‐uranium breeding of MSRs in the future. However, this fuel management is still faced with some different technical routes, such as the selection of carrier salts, the enrichment of uranium, with or without thorium, and the recycle necessary of spent nuclear fuel. Therefore, various fuel cycle schemes were compared and analyzed using an in‐house developed fuel management code MOBAT. Different graphite assemblies were optimized by changing the salt volume fraction in core and dimension to find a region for best fuel utilization and negative temperature reactivity coefficient. Prismatic block with 10% volume fraction of molten salt is considered as a good assembly type because of its significant space shielding effect of U‐238. For carrier salts, LiF‐BeF2 with 99.995% enriched Li‐7 displays higher fuel utilization and lower cost of fuel cycle than NaF‐BeF2, while the tritium production at the beginning of life will be two orders of magnitude higher. For fuel enrichment, 20% enriched uranium is recommended because the background of neutron absorption from carrier salt and graphite will be more significant in lower enrichment condition. Importantly, it shows that thorium is a good breed and burned fuel in situ and could improve the fuel utilization by 20%. Also, offline reprocessing to recover the uranium is a commendable scheme when the cost of offline reprocessing is lower than 400 $/kgHN. [ABSTRACT FROM AUTHOR]

Summary: The thorium‐uranium (Th‐U) fuel cycle is considered as a potential approach to ensure a long‐term supply of nuclear fuel. Small modular molten salt reactor (SMMSR) is regarded as one of the candidate reactors for Th utilization, since it inherits the merits of both MSR and small modular reactor. The Th utilization in a 220‐MWe SMMSR with the once‐through fuel cycle mode is investigated first. Then, the SMMSR with batch and online fuel processing modes is investigated second for comparison, considering the progressive development of fuel reprocessing technology. To keep a negative temperature reactivity feedback coefficient (TRC), a configuration for fuel salt volume fraction (SVF) equal to 15%, with a mixed fuel of low enriched uranium (LEU) and thorium at an operation time of 5 years is recommended for the once‐through mode, corresponding to the Th energy contribution (ThEC) of 37.6% and natural U and Th utilization efficiency (UE) of 0.51%. Considering the solubility limit of heavy nuclide (HN) proportion (below 18.0 mol%) in the fuel salt, the total operation time of the SMMSR shall be less than 50 years for the batch reprocessing mode with a 5‐year reprocessing interval time. In this case, the ThEC and UE can be improved to about 47.4% and 0.99%, respectively. Finally, the Th utilization and fuel sustainability are analyzed at a lifetime of 50 years for the online reprocessing fuel cycle mode, including both the only online fission products (FPs) removing scheme and the fuel transition scheme from LEU to 233U. For the former scheme, the ThEC and UE can be further improved to 58.6% and 1.52%, respectively. For the latter scheme, 233Pa is extracted continuously from the core to breed and store 233U. If a total reactor lifetime of 50 years is assumed, the operation time using LEU as starting and feeding fuel for 6 years is required, and the bred 233U during this 6‐year operation can start and maintain the reactor criticality for the remaining 44 years. In this case, the ThEC is improved significantly to 89.1% corresponding to a UE of 2.74%. [ABSTRACT FROM AUTHOR]

Highlights • The characteristics of the thorium fuel cycle transition with TRUs as starting fuel for a smTMSR are analyzed. • Five different fuel fractions in the core are optimized with regarding to the U-233 production in the thermal spectrum region. • These analyses can provide solutions for 233U shortage of the thorium fuel cycle deployment in smTMSRs. Abstract Small modular reactors (SMRs) can provide an energy option with low carbon emission, enhanced safety conviction, convenient construction and operation. Meanwhile, molten salt reactors (MSRs) have been recognized as one of the reference reactors of the Generation IV International Forum (GIF) with the unique potential (inherent safety, no fuel fabrication, online fuel reprocessing, etc). Combining advantages of SMRs and MSRs, the small modular Th-based molten salt reactor (smTMSR) has excellent performances such as inherent safety, economics, sustainability, and proliferation resistance. Moreover, thorium is still an attractive fuel in a smTMSR with characteristics of online refueling and reprocessing. In this paper, we attempt to study the capability of the transition to thorium fuel cycle in a thermal smTMSR by analyzing the neutronic characteristics using TRUs as starting fuel. Optimizations are conducted to produce the required amount of U-233 for starting a new smTMSR based on various fuel fractions in thermal energy region. It is concluded that the U-233 production is superior with higher fuel fraction in the core, which indicates that a thermal smTMSR with a proper fuel fraction can achieve the transition to thorium fuel cycle. The HM and Pu mole proportions which may have a negative influence on the molten salt stability are analyzed. Moreover, the temperature feedback coefficient as a key safety parameter is also discussed. With a suitable fuel fraction in thermal region, the smTMSR can well satisfy the restrictions of HM and Pu mole proportions and safety demands. [ABSTRACT FROM AUTHOR]

Highlights • Th-U breeding capability in the CMSFR with different starting fuels is analyzed. • A 233U breeder mode can be achieved with a doubling time of 37 years. • Both Pu and TRU can be employed for a transition approach to Th-U fuel cycle. • The total loss ratio of β eff with different starting scenarios is about 13.5%. Abstract 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 (β eff). 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 β eff 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 β eff 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. [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]

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. [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]

Minor actinide (MA) accumulated in spent fuel is the primary contributor to the long-term radiological hazards of high-level nuclear waste. Due to its outstanding features such as fast neutron spectrum, no fuel assembly fabrication, on-line refueling and reprocessing, Molten Salt Fast Reactor (MSFR) is regarded as one of the candidate reactors for MA incineration. In the present work, we evaluate the MA incineration capability for a 500 MWth FLiNaK MSFR by considering on-line MA feeding into the fuel salt and/or 233 U extraction from the fertile salt, where the initial FLiNaK fuel salt is composed of 70 mol% (LiF, NaF and KF) and 30 mol% (ThF 4 , 233 UF 4 and MAF 3 ). The simulated results show that, during a 50-year operation, the MA incineration efficiency with on-line MA feeding and 233 U extraction scenario is significantly higher compared with the case without either MA feeding or 233 U extraction. Based on the scenario of on-line MA feeding and 233 U extraction, the initial MA loading is further adjusted to keep a small excess reactivity during the whole operation. And the MA incineration amount, the depletion ratio between MA and total heavy nuclides and the MA incineration efficiency at 50-year operating time are 7301 kg, 77.4% and 33.3 kg/TWh, respectively, with a doubling time of 39-year for 233 U breeding. We also analyze and discuss the evolutions of the fuel temperature feedback coefficient, the effective delayed neutron fraction and the intensity of spontaneous fission neutron during the whole operating time. [ABSTRACT FROM AUTHOR]

As one of the generation IV reactors, molten salt reactors (MSRs) have attracted much attention in recent years. Generally, molten salt acting as both fuel and coolant circulates through graphite channels and external primary loop in a liquid-fueled MSR with a channel-type core. To accurately simulate the neutronics of channel-type MSRs, a GPU paralleled high-fidelity neutron transport code ThorMOC has been improved by implementing a 3D rasterized coarse mesh finite difference (RCMFD) method and a quasi-2D delayed neutron precursor (DNP) drift model. A Molten Salt Reactor Experiment (MSRE) model and a small Thorium-based MSR (sTMSR) model with stationary and flowing fuel are used for the verification of the improved ThorMOC. The reference results of the two stationary reactor models are obtained from Monte Carlo code OpenMC, which indicates that ThorMOC performs a high accuracy for stationary MSR simulations. The reactivity loss caused by fuel flowing in the steady state MSRE is calculated by ThorMOC and agrees well with the measured value from the MSRE. Moreover, the influences of geometry simplifications of fuel channels and downcomer on the reactivity loss are evaluated by comparing their results with those of the high-fidelity model. The reactivity loss in the sTMSR core is computed, and the effects of fuel velocity in core channels and residence time outside the core on the reactivity loss are also analyzed. The whole-core transport calculations of the 3D MSRE and sTMSR core models with 8 energy groups cost several hundred seconds. • A high-fidelity neutron transport code has been developed for liquid-fueled MSRs. • A quasi-2D DNP drift model was implemented and verified by the MSRE. • Geometric approximations commonly used in nodal codes are eliminated. • CPU and GPU parallel computing are supported. [ABSTRACT FROM AUTHOR]

An integral benchmark experiment for neutron leakage spectra from ThO 2 samples has been performed with a 14.8 MeV D-T neutron source at China Institute of Atomic Energy (CIAE) to validate the evaluated nuclear data related to the design of Thorium Molten Salt Reactor (TMSR). The leakage spectra from ThO 2 samples were measured at 60° and 120° emission angles with the Time of Flight (TOF) method. The measured spectra are well reproduced as a whole by the Monte Carlo neutron transport code simulations with the ENDF/B-VII.1, JENDL-4.0 and CENDL-3.1 evaluated nuclear data, while some discrepancies in the major reaction channels including (n, el), (n, inl), (n, 2n) and (n, 3n), still exist. Detailed information on the neutron product distribution with the evaluated data from the above three libraries is presented to explain the discrepancies. From the comparisons, the data of the above three libraries need to be improved as required in different energy ranges. [ABSTRACT FROM AUTHOR]

• The recovery of uranium from FLiBe-ZrF 4 molten-salt containing irradiated UF 4 and ThF 4 by the fluoride volatility process (FVP) was carried out. The radioactivity detection method was applied for in-situ monitoring of the FVP. • The FVP was monitored by detecting the change of the γ activities of key nuclides in the molten salt and the NaF adsorber. • The overall process γ decontamination factors (DFs) of 95Zr, 103Ru, 132Te, 140La,140Ba, 141Ce, 143Ce and 233Pa were greater than 105∼106. • 239Np could be co-separated and recovered with uranium by fluorination from the fuel salt. • The separated 233Pa from the molten salt in the fluorination-sorption step could be trapped in the NaF trap, resulting in a high γ DF in the uranium product. The recovery of uranium from FLiBe-ZrF 4 (LiF-BeF 2 -ZrF 4 : 65–30–5 mol%) molten-salt containing irradiated UF 4 and ThF 4 by the fluoride volatility process (FVP) was carried out. Gamma radiography was applied for in-situ monitoring of the fluorination and desorption steps successfully. The results confirmed that it was feasible to monitor the FVP by detecting the changes of the γ activities of key nuclides in the molten salt and the NaF adsorber. The behavior and distribution of actinides and fission products in each unit of the FVP were investigated by the determination of their activities with the γ-ray spectroscopy. The results indicated the conversion and volatilization rate of UF 4 was greater than 99.8%. And the recovery yield of uranium was approximately 80% in two cold traps. A small amount of uranium remained in the molten salt, others was found in the NaF trap and the NaF adsorber. In addition, small amounts of uranium were also found in the off-gas, in the desorption step. As for 95Zr, 140La, 140Ba, 141Ce, 143Ce and 233Pa, existing as nonvolatile fluorides, remained in the fluorinated salt with the overall process gamma (γ) decontamination factors (DFs) in the UF 6 product greater than 105∼106. Meanwhile, 99Mo, 131I and 239Np existing as volatile fluorides, were found in the UF 6 product, NaF absorber and off-gas with the overall process γ DFs in the UF 6 product less than 10. In the case of 103Ru and 132Te also existing as volatile fluorides, the overall process γ DFs in the UF 6 product still reached approximately 105∼106 due to their separation from the UF 6 product by the NaF adsorber. [Display omitted] [ABSTRACT FROM AUTHOR]

• A GPU-based 3D neutron transport code is newly developed. • Calculating the modified total source on-the-fly to significantly reduce the GPU memory requirement. • An azimuthal angle decomposition algorithm is proposed to reduce GPU memory consumption further. • Geometric and physical information are written into a VTK file for visualization. A GPU-based three-dimensional (3D) neutron transport code, named ThorMOC, is developed in this work. A GPU-parallelized transport sweeping algorithm based on a planar method of characteristics (MOC) is implemented in ThorMOC and a two-level coarse mesh finite difference (CMFD) method is used to accelerate the convergence of the MOC transport solution. To reduce the memory burden of 3D calculation, the transport sweeping algorithm is improved by azimuthal angle decomposition. This developed ThorMOC code is verified by a Takeda benchmark and a C5G7 3D extension benchmark, and the performance of the azimuthal angle decomposition algorithm is evaluated by the whole core calculation of a C5G7 hexagonal benchmark. The numerical results obtained from ThorMOC show a good agreement with those from the Monte Carlo code OpenMC. Besides, the azimuthal angle decomposition algorithm can reduce the GPU storage requirement by more than 40% for the C5G7 hexagonal benchmark. [ABSTRACT FROM AUTHOR]

As one of the six candidate reactors chosen by the Generation IV International Forum (GIF), Molten Salt Fast Reactor (MSFR) has many outstanding advantages and features for advanced nuclear fuel utilization. Effective transmutation of minor actinides (MA) could be attained in this kind of fast reactor, which is of importance in the future closed nuclear fuel cycle scenario. In this work, we attempt to study the MA transmutation capability in a MSFR with power of 500 MWth by analyzing the neutronics characteristics for different MA loadings. The calculated results show that MA loading plays an important role in the reactivity evolution of the MSFR. A larger MA loading is favorable to improving the MA transmutation performance and simultaneously to reducing the fissile consumption. When MA = 18.17 mol%, the transmutation fraction can achieve to about 95% on iso-breeding. We also find that although the fuel temperature coefficient (FTC) decreases with the increasing MA loading, it is still negative enough to keep the safety of the MSFR during the whole operation time. The MA contribution to the effective delayed neutron fraction (EDNF) and the intensity of spontaneous fission neutron (ISFN) are also analyzed. Also MA loading can affect the EDNF during the operation and the ISFN of the MSFR is dominated by 244 Cm. Finally, we analyze the effect of the core power on MA transmutation capability. The result shows that for all the operating powers the depletion ratio of MA to HN increases with time and reaches a maximum value. And additional MA should be fed into the fuel salt before the MA depletion ratio reaches the peak value to improve its transmutation capability. The net mass of the transmuted MA during the 50 years operation for 500 MWth is 5620 kg which is very close to that of 1000 MWth. [ABSTRACT FROM AUTHOR]

The in situ breed-and-burn (B&B) strategy in fast reactor is capable of offering very high fuel utilization efficiency. CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy) or TWR (Traveling Wave Reactor) is a possible scheme to implement the B&B strategy. The approach in realizing the equilibrium state of CANDLE is based on neutron source supply from ignition zone with enriched uranium or plutonium. And burnable poison is usually employed to compensate the extremely high excess reactivity during the startup. In the present work, the optimization of the starter fuel with only plutonium and depleted uranium is performed to reduce the initial positive excess reactivity swing and to flatten the power distribution. The optimization is done with refer to a nominal steady state CANDLE core. In this regard the swing of the reactivity is limited during the startup. And the initial power shape is flattened significantly compared with that without optimization. It is proved that constructing the ignition zone with refer to the equilibrium state of a CANDLE is effective and feasible to improve the operation performance during the startup. [ABSTRACT FROM AUTHOR]

Monte Carlo transport code Geant4 has been successfully utilised to study of neutron-induced fission reaction for Th in the transport neutrons generated from H(d,n)He neutron source. The purpose of this work is to examine the applicability of Monte Carlo simulations for the computation of fission reaction process. For this, Monte Carlo simulates and calculates the characteristics of fission reaction process of Th(n,f), such as the fission yields distribution, kinetic energy distribution, fission neutron spectrum and decay γ-ray spectrum. This is the first time to simulate the process of neutron-induced fission reaction using Geant4 code. Typical computational results of neutron-induced fission reaction of Th(n,f) reaction are presented. The computational results are compared with the previous experimental data and evaluated nuclear data to confirm the certain physical process model in Geant4 of scientific rationality. [ABSTRACT FROM AUTHOR]

Heavy water moderated molten salt reactor (HWMSR) is a newly proposed molten salt reactor. It adopts heavy water as the moderator and molten salt dissolved with actinide elements as the fuel and coolant as well. These unique features endow HWMSR a high neutron economy as performed in CANDU (CANada Deuterium Uranium) pressure tube heavy water reactor and a capability to implement the processes of online refueling and reprocessing as done in the conventional MSRs, making natural uranium (NU) utilization and together with U-233 production become feasible. The obtained results in this study demonstrate that the fuel utilization efficiency of HWMSR with NU fuel cycle can achieve 1%, slightly better than that of the CANDU reactor at the same physical size and power level. Meanwhile, around 2662 kg U-233, which can start 3 new Th–U fuel cycle HWMSR cores, can be produced over 20 years' operation by online refueling thorium and extracting Pa-233. • Natural uranium (NU) utilization on a heavy water moderated molten salt reactor (HWMSR) is studied. • The fuel utilization efficiency of NU on HWMSR can achieve 1%, higher than that of CANDU reactor and LWR. • Around 2662 kg U-233 can be produced over 20 years' operation by online refueling thorium and extracting Pa-233. [ABSTRACT FROM AUTHOR]

• A rasterized CMFD method is developed to deal with generalized boundaries. • A LFMSR benchmark is proposed and used to validate the rasterized CMFD method. • More than 100 times speedup is obtained for the LFMSR benchmark. The method of characteristics (MOC) is widely employed in neutron transport solvers to obtain accurate solution of a direct transport calculation for a whole reactor core model. Coarse mesh finite difference (CMFD) method is often used to speedup the convergence of MOC solution especially for a large geometry. In this work, a rasterized CMFD method is proposed to improve the convergence rate of the high-order neutron transport calculation for reactor cores with generalized boundaries. The rasterized CMFD method is implemented in a 2D neutron transport MOC code ThorMOC which has been developed to perform the high-fidelity criticality calculation for channel type molten salt reactors (MSRs). The rasterized CMFD method is validated by two benchmarks, a C5G7 hexagonal benchmark and a LFMSR (Liquid-Fueled Molten Salt Reactor) benchmark. The computational result obtained from ThorMOC accelerated by the rasterized CMFD method shows a good agreement with that from the Monte Carlo code OpenMC for the two benchmarks. The rasterized CMFD method can reduce the MOC stand-alone solution time of the LFMSR benchmark by more than 100 times. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

• Analysis of ameliorating the positive TFC for a TRU-fueled MSR is analyzed. • Two effective solutions are proposed to ameliorate the positive TFC. • With a hectogram of 167Er or MA ratio over 50%, the TFC turns to be negative. 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. [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]

• Two possible 238Pu production scenarios are proposed and assessed. • 238Pu production can meet the demand of RTG fuel. • Producing 238Pu has a small influence on the original design objective of the MSFR. 238Pu can be used as the heat source in radioisotope thermoelectric generators for deep space exploration, while it is currently in shortage because of the suspension of 238Pu production in the USA and Russia for many years. Two possible scenarios for the production of 238Pu as a byproduct in a molten salt fast reactor (MSFR) are assessed in this work. For the first scenario, the MSFR is started with low-enriched uranium (LEU), and neptunium is extracted from the fuel salt and then fed into the fertile salt of thorium for high-purity 238Pu production. For the second scenario, the ignition material for the MSFR is 233U, and highly purified 238Pu is produced by directly extracting plutonium from the fuel salt. The simulated results indicate that on average 15.48 kg 238Pu with the quality (ratio of 238Pu mass to total mass of Pu isotopes) larger than 98.6% can be produced per year during a 60-year operation for scenario I. In parallel, 8.31 kg 238Pu with the quality larger than 91.9% can be produced annually for scenario II. The 238Pu production has a small impact on the neutronic performance of the core for the two scenarios except that bred 238U is reduced by 922 kg for scenario I and by 261 kg for scenario II during the 60-year lifetime. The 238Pu production has a very tiny effect on the radiotoxicity of discharged nuclear fuel for scenario I, while it can reduce the total radiotoxicity of discharged nuclear fuel by three times for scenario II. [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]

• The properties of a sm-TMSR using TRUs and Th as fuels are analyzed. • The U233 demands with thorium fuel cycle in various thermal sm-TMSRs are evaluated. • These analyses can provide solutions for transition to thorium fuel cycle in sm-TMSRs. There is currently a renewed interest in small modular reactor (SMR) and molten salt reactor (MSR) concepts, since the former is recognized as the most accessible form of nuclear energy with a lower cost compared to conventional reactors while the latter is considered as one of the reference reactors of the Generation IV International Forum (GIF) suited for thorium fuel cycle. The TRUs from the spent fuel of LWRs containing more than 50% of fissile fuels may be used to start up a MSR due to its outstanding features of no fuel assembly and online/ batch reprocessing. This study aims to analyze the characteristics of a small modular thorium-based molten salt reactor (sm-TMSR) using TRUs and Th as fissile and fertile fuels towards thorium fuel cycle with a batch reprocessing mode. The neutronic properties using TRUs as an igniting fuel in the core configurations with the fuel salt fraction (SF) of 18.4%/27.4% and the reprocessing periods (RPs) of 5/10 years are first analyzed and compared. To obtain a high production of U233, the configuration with SF = 27.4% and RP = 5 years is recommended. The recovered uranium production and the purity of U233 can achieve ∼69 kg/y and over 75%, respectively. Furthermore, the recovered uranium demand with thorium fuel cycle in the 27.4% core and other thermal sm-TMSRs are evaluated. In the 27.4% core, the recovered uranium demand during 40-year operation is ∼38 kg/y. For the new sm-TMSRs with the SF ranging from 2% to 18.4%, the recovered uranium demand varies from 148 to 48 kg/y. Finally, transition scenarios to thorium fuel cycle are proposed. The transition time is about 22 years in the 27.4% core while it is about 28 years in the SF = 18.4% core. The simulated results demonstrate that it is feasible to achieve the transition from TRUs to thorium fuel cycle in sm-TMSRs if a batch reprocessing mode can be adopted. [ABSTRACT FROM AUTHOR]

• MA characteristics from core start-up to equilibrium of three TMSRs are analyzed. • Mass equilibrium time strongly depends on the vanishing coefficients. • Fast MSRs produce less MA inventory and radiotoxicity for a short operation time. • Fast MSRs would loss its less-radiotoxic advantage for a long-term operation. A liquid-fueled molten salt reactor (MSR) can be flexibly designed as a thermal, epithermal or fast system. Meanwhile, available online fuel reprocessing can inherently interconnect the closed fuel cycle with the reactor, which would have a substantial impact on the MA production as well as nuclear waste management of a MSR. In this work, three thorium-based MSRs (thermal, epithermal and fast spectra) are chosen for analysis. The results indicate that the mass equilibrium time of a nuclide strongly depends on the vanishing coefficient of its dominant precursor nuclides and of itself. Owing to the low generation rate of MA caused by small vanishing coefficient, the fast TMSR shows a lower radiotoxicity over the thermal and epithermal TMSRs for a short operation time. Nevertheless, this advantage gradually losses as the operation time further progresses and so that the MA inventory in the fast reactor can also achieve equilibrium state. [ABSTRACT FROM AUTHOR]