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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.

The nuclear data of Th is primary interest for the thorium-uranium (Th-U) fuel cycle, and its nuclear data accuracy is related to the conversion performance of Th-U fuel cycle. Based on the photo-neutron source for the thorium molten salt reactor (TMSR-PNS), the total cross section (TCS) of ThO2 was measured by the transmission method for the high purity thorium oxide (ThO2) samples. ThO2 samples which the thicknesses were respectively 0.9 cm and 0.608 cm were made by pressing and sintering process at 1 800 ℃. Based on the different pulse shapes of neutrons and γ-rays (n/γ) signal from the lithium iodide (LiI) scintillator detector used for the time-of-flight (TOF) calculation, γ-rays were identified by pulse shape discrimination (PSD) method to reduce the interference of the measured cross section. Combined with the trough position of TOF spectrum and its corresponding energy for the calibration targets of Ag, In and Cd, the flight distance was fitted to be 5.899 m, so as to convert TOF spectrum of neutrons into energy spectrum. Energy spectrum of 10 cm thickness boracic polythene sample was measured to deduct the influence of the scattering neutron from the background. According to the measured neutron leakage spectrum with or without ThO2 sample, experimental data of TCS for ThO2 were obtained. TCS of 16O from ENDF/B-Ⅶ.1 was deducted from TCS of ThO2 and TCS of Th was obtained. Combined with error of the sample purity and thickness, the fitting flight distance, relative standard deviation of TCS for 16O from ENDF/B-Ⅶ.1, and experimental statistical error, the uncertainty of experimental data were calculeted to be 3.25%-4.51% in the energy range from 0.02 to 0.1 eV. The results show that the measured data of TCS of Th are good agreement with the evaluated data of ENDF/B-Ⅶ.1 in the energy range from 0.02 to 0.1 eV. In the thermal energy range (E

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.

BackgroundHeavy water moderated molten salt reactor (HWMSR) adopts heavy water as moderator and heavy metal elements dissolved in fluoride salt as the fuel, which has high neutron economy. But large temperature difference of fuel salt outlet will lead to the thermal fatigue of piping components at the top of the core.PurposeThis paper aims to optimize the core design of HWMSR to minimize the fuel salt outlet temperature.MethodsBy using the developed codes of neutron-thermal hydraulic coupling calculation and core critical search calculation, the power density distribution, outlet temperature distribution, initial 233U loading and breading ratio (BR) were analyzed for the cores with different size of molten salt channels.ResultsThe calculation results demonstrate that increasing the radius of molten salt channels in the inner core zone (correspondingly decreasing the radius of molten salt channels in the outer core zone) will decrease the peak of power density and the maximum outlet temperature of molten salt while the influence on BR and 233U initial loading.ConclusionsThis study provides a valuable foundation for core optimization of HWMSR.

BackgroundThe simulation device is one-third scale of the 10 MW solid-fuel thorium molten-salt reactor (TMSR) and has the rated heat power of 370 kW. Nineteen annular channel heaters with wire coil were designed to heat the simulation device to the desired temperature.PurposeThis study aims to further investigate the effect of outer diameter of heating tube and annular width on the flow and heat transfer characteristics of FLiNaK salt in the annular channel heater.MethodsThe flow and heat transfer characteristics of annular channel heater with wire coil were simulated by using Fluent software. The effects of outer diameter of heating tube on the flow and temperature fields were investigated numerically, so did the effects of annular width on the flow and heat transfer characteristics. The overall performance of annular channel heater was evaluated by the principle of entropy production.ResultsThe simulation results show that the axial velocity, radial velocity and tangential velocity are generated and meanwhile counter-flow is formed on longitudinal section.ConclusionsThe outer diameter of heating tube has no effect on the flow and heat transfer characteristics. The larger the outer diameter of heating tube, the higher the energy efficiency. As the increase of the annular width, the pressure loss, the heat transfer coefficient decreases, and the energy efficiency gradually decreases.

BackgroundHelium-xenon gas cooled nuclear reactor adopts helium-xenon gas as the coolant and is able to apply the integral Brayton cycle. It has advantages in miniaturization and lightweight design, which has attracted more and more attention worldwide. But rare severe accidents studies have been carried out for the small helium-xenon gas cooled reactor. Probabilistic safety assessment (PSA) is an important method to evaluate the safety of a reactor system. The results obtained by PSA can provide a valuable basis for improving the core design, identifying fault, guiding operation, etc. whilst the occurrence frequency of initial events is required for PSA analysis.PurposeThis paper aims to evaluate the accident occurrence frequency for the small helium-xenon gas cooled reactor to provide an input for PSA analysis.MethodsBased on the main technical characteristics of a small innovation helium-xenon cooled mobile nuclear power system (SIMONS), the occurrence frequency of typical accidents was analyzed by referring to the operational experiences and data of high temperature gas cooled reactor (HTGR) and light water reactor (LWR). Intensity analyses were performed on the frequency of accidents such as increase or decrease of core heat removal, abnormal reactivity and power distribution, pipe break and abnormal equipment leakage, anticipated transient without scram (ATWS), and loss of offsite power (LOOP).ResultsThe calculation results show that the accidents of heat removal increase and decrease, abnormal reactivity and power distribution, pipe break and abnormal equipment leakage, ATWS, and LOOP have an occurrence frequency of 3.90×10-2 RY-1, 2.36×10-1 RY-1, 2.69×10-2 RY-1, 6.50×10-2 RY-1, 2.69×10-2 RY-1 and 4.60×10-2 RY-1, respectively.ConclusionsThe calculated results can be taken as the inputs for the PSA study of the SIMONS, providing basic reference value for further PSA analysis of mobile nuclear power system.

BackgroundBoron neutron capture therapy (BNCT) is a promising treatment for cancer. In recent years, D-T neutron source has shown great potential as a suitable neutron source for BNCT. However, the neutron energy produced by a D-T neutron source is 14.1 MeV, which cannot be straightforwardly used in a BNCT, so beam shaping is necessary.PurposeThis study aims to design beam shaping assembly (BSA) to make the neutron beam of D-T neutron source suitable for BNCT.MethodsThe Monte Carlo simulation program MCNP5 was employed to design the corresponding BSA for D-T neutron source. The advantages of using natural uranium as neutron multiplicative layer was firstly verified by simulation. Then, the moderating layer was simulated together the supplementary moderating layer, reflective layer, shielding layer and thermal neutron absorption layer using different materials and thickness to obtain optimal configuration of these materials.ResultsSimulation results show that an ideal therapeutic neutron beam is obtained by using a 14 cm radius natural uranium as neutron multiplicative, 50 cm thick BiF3 and 10 cm thick TiF3 as a double-moderating layer, a 17 cm thick AlF3 as a supplementary moderator, a 0.2 mm thick Cd as a thermal neutron absorbing layer, a 3.5 cm thick Pb as a γ shielding layer and a 10 cm thick Pb as a reflecting layer. Combination of these materials is suggested to optimize and design as the neutron BSA.ConclusionsThe air outlet parameters of output neutron beam using BSA designed in this study can well meet the recommended values of the International Atomic Energy Agency (IAEA).

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.

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]

The evaporation behaviours of various components were investigated by using a low pressure distillation method in a 2LiF-BeF 2 -ZrF 4 mixture containing irradiated ThF 4 and UF 4 . The experiment showed that BeF 2 and ZrF 4 were found to mainly condensate at the outer cover, the coolest zone, and their relative volatilities vs. LiF were 9.8 and 32.2, respectively, while for ThF 4 and UF 4, at four different temperature zones the values were almost constant, at 0.1 and 0.3. The radioactivity of various nuclides was further detected using gamma spectrometer analysis. 137 Cs was hardly observed due to long half-time decay. 233 Pa was found to co-evaporate with the carrier salt, while 239 Np mainly remained in the residual salt as 237 U. In different temperature zones, the decontamination factors of rare earth in receiver salts ranged from 10 to 10 3 . On the basis of the investigation, it was proposed that the distribution of various nuclides after distillation, may be helpful to design the feasible condensate system to recover the carried salt in a molten salt reactor., Competing Interests: There are no conflicts to declare., (This journal is © The Royal Society of Chemistry.)

The production capacity of 89 Sr and 90 Sr in the 2 MW MSR are evaluated. The gaseous 89 Kr and 90 Kr are extracted from the core through the helium bubbling system, and then decay to 89 Sr and 90 Sr, respectively. In order to improve purity of 89 Sr product, two cooling devices are adopted in the 89 Sr and 90 Sr production system. The annual yields of 89 Sr and 90 Sr are about 9000 Ci and 32 Ci, respectively, and the impurity of 89 Sr product is less than 2 ppm which can meet the medical requirement., (Copyright © 2021 Elsevier Ltd. All rights reserved.)