Analysis of Th-U breeding capability for an accelerator-driven subcritical molten salt reactor

Accelerator-driven systems based on molten salt fuel have several unique advantages and features for advanced nuclear fuel utilization.The aim of this work was to study the Th-U breeding capability in such systems,known as‘‘accelerator-driven subcritical molten salt reactors’’(ADS–MSRs).Breeding capacities including conversion ratio and net~(233)U production for various subcriticalities and different minor actinides(MA)loadings were analyzed for an ADS–MSR.The results show that the subcriticality of the core has a considerable effect on the Th-U breeding.A high subcriticality is favorable to improving the conversion ratio,increasing the net~(233)U production,and reducing the doubling time.Specifically,the doubling time for k_(eff)of 0.99 is larger than 80 years,while the counterpart for k_(eff)of 0.93 is only approximately22 years.Nevertheless,in an ADS–MSR with a high initial MA loading,MA results in a non-negligible~(233)U depletion in the first two decades,while increasing the net~(233)U production compared to reactors without MA loading.During the 50 years of operation,for the subcritical reactor(k_(eff)0:97)with MA fraction increasing from 1 to 14%,the net~(233)U production increases from 3.94 to 8.24 t.     

Analysis on the neutron kinetics for a molten salt reactor

The neutron kinetics of the molten salt reactor is significantly influenced by the fuel salt flow, which leads to the analysis methods for the conventional reactors using solid fuels not being applicable for the molten salt reactors. In this study, a neutron kinetic model considering the fuel salt flow is established based on the neutron diffusion theory, which consists of two-group neutron diffusion equations for the fast and thermal neutron fluxes and six-group balance equations for delayed neutron precursors. The temperature feedback on the neutron kinetics is considered by introducing a heat transfer model in the core, in which the group constants which are dependent on the temperature are calculated by the code DRAGON. The mathematical equations are discretized and numerically calculated by developing a code, in which the fully implicit scheme is adopted for the time-dependent terms, and the power law scheme is for the convection–diffusion terms. The neutron kinetics is conducted during three transient conditions including the rods drop transient, the pump coastdown transient and the inlet temperature drop transient. The relative power changes and the distributions of the temperature, neutron fluxes and delayed neutron precursors under these three different transient conditions are obtained in the study. The results provide some valuable information for the research and design of this new generation reactor.     

Steady thermal hydraulic analysis for a molten salt reactor

The Molten Salt Reactor (MSR) can meet the demand of transmutation and breeding. In this study, theoretical calculation of steady thermal hydraulic characteristics of a graphite-moderated channel type MSR is conducted. The DRAGON code is adopted to calculate the axial and radial power factor firstly. The flow and heat transfer model in the fuel salt and graphite are developed on basis of the fundamental mass, momentum and energy equations. The results show the detailed flow distribution in the core, and the temperature profiles of the fuel salt, inner and outer wall in the nine typical elements along the axial flow direction are also obtained.     

Feasibility of a critical molten salt reactor for waste transmutation

This study assesses the feasibility of designing a Molten Salt Reactor (MSR) using the salt mixture of LiF (15 mol%), NaF (58 mol%) and BeF₂ (27 mol%) to be critical when fuelled with TRU from LWR spent fuel without exceeding the actinides solubility limit and while extracting fission products at realistic rates. The first part of the study investigated the graphite-to-MS volume ratio on the neutron balance, transmutation characteristics and graphite lifetime. It is found that a core without graphite moderator is the preferred design option; it offers the best neutron balance, most compact design and alleviated graphite lifetime problem. The second part of the study investigated sensitivity of the epithermal spectrum core to the feed composition, power density, fission products residence time and actinides loss fraction. It is found that the transmutation effectiveness improves with increasing power density and that the shorter the LWR spent fuel cooling time is, the better becomes the MSR neutron balance. The optimal MSR design offers a remarkably high transmutation capability – fissioning of as high as 99.8% of the TRU fed. The transmutation capability of the MSR is also rated in terms of final waste radiotoxicity, decay heat, spontaneous fission neutrons emission, fissile and ²³⁷Np inventory.     

The chemistry and thermodynamics of molten salt reactor fuels

A fairly detailed knowledge of the chemistry and thermodynamics of molten salt breeder reactor fuels, typically LiF-BeF₂ThF ₄-UF₄ ($\text{72-16-12- < 1}\text{mol}\text{\%}$), has been gained from measurements of heterogeneous equilibria involving various gases (e.g. HF, H₂O) and solids (usually oxides). This information combined with the thermochemical data available for the pure compounds has given formation free energies for the fual components. Their entropies can be estimated from the known entropies of the corresponding solid oxides and the charge on the cation. Their activity coefficients depend primarily on the mol % of LiF in the salt mixture. The solubility of oxides generally decreases with the square of the cation charge divided by the radius. Thus the actinide dioxides and, especially, Pa₂O₂ (or an addition compound of it) are of low solubility. Pourbaix diagrams are presented which summarize the conditions of redox potential and oxide concentration wherein the fuel may be contained without significant corrosion or precipitation reactions, or conditions wherein selective oxide precipitation may be carried out for purposes of fuel reprocessing.     

Development of a steady state analysis code for a molten salt reactor

The molten salt reactor (MSR), which is one of the ‘Generation IV’ concepts, can be used for transmutation, and production of electricity, hydrogen and fissile fuels. In this study, a single-liquid-fueled MSR is designed for conceptual research, in which no solid material is present in the core as moderator, except for the external reflector. The fuel salt flow makes the MSR neutronics different from that of conventional reactors using solid fuels, and couples the flow and heat transfer strongly. Therefore, it is necessary to study the core characteristics with due attention to the coupling among flow, heat transfer and neutronics. The standard turbulent model is adopted to establish the flow and heat transfer model, while the diffusion theory is used for the neutronics model, which consists of two-group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six groups of delayed neutron precursors. These two models which are coupled through the temperature and heat source are coded in a microcomputer program. The distributions of the velocity, temperature, neutron fluxes, and delayed neutron precursors under the rated condition are obtained. In addition, the effects of the inflow temperature, inflow velocity, and the fuel salt residence time out of the core are discussed in detail. The results provide some valuable information for the research and design of the new generation molten salt reactors.     

In-core fuel management strategy for the basket-fuel-assembly molten salt reactor

Molten salt reactor, with good economics and inherent reliability, is one of the six types of Generation IV candidate reactors. The Basket-Fuel-Assembly Molten Salt Reactor(BFAMSR) is a new concept design based on fuel assemblies composed of fuel pebbles made of TRISOcoated particles. Four refueling patterns, similar to the fuel management strategy for water reactors, are designed and analyzed for BFAMSR in terms of economy and security.The MCNPX is employed to calculate the parameters, such as the total duration time, cycle length, discharge burnup,total discharge quantity of ~(235)U, total discharge quantity of ~(239)Pu, neutron flux distribution and power distribution. The in–out loading pattern has the highest burnup and duration time, the worst neutron flux and power distribution and the lowest neutron leakage. The out–in pattern possesses the most uniform neutron flux distribution, the lowest burnup and total duration time, and the highest neutron leakage.The out–in partition alternate pattern has slightly higher burnup, longer total duration time and smaller neutron leakage than that of the out–in loading pattern at the cost of sacrificing some neutron flux distribution and power distribution. However, its alternative distribution of fuelelements cut down the refueling time. The low-leakage pattern is the second highest in burnup, and total duration time, and its neutron flux and power distributions are the second most uniform.     

氢化锆慢化熔盐堆钍铀转换性能初步分析

中子能谱对钍基燃料在熔盐堆中的利用效率及温度反馈系数等安全问题有较大影响,所以对熔盐堆新型慢化剂的研究具有重要意义。本工作基于SCALE6计算程序,对不同几何栅元结构的氢化锆栅元组件在熔盐堆的物理性能进行了研究,分别计算了中子能谱、钍铀转换比、~(233)U浓度、总温度反馈系数以及燃耗等中子物理参量。结果表明,减小六边形栅元对边距或者增加熔盐占栅元体积比可以增加钍铀转换比和改善温度反应性系数;当加入的氢化锆慢化剂体积份额为0.1时就可以将熔盐堆~(233)U初始浓度降低到2.5×10~(-2)以内;氢化锆慢化熔盐堆在超热谱条件下,其~(233)U初装载量和超铀核素产量较小,同时堆芯较为紧凑。     

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

To improve the reliability and reduce energy consumption, a conceptual design of a freeze valve is proposed for the thorium-based molten salt reactor(TMSR)concept. Fins were utilized in this new design to enhance heat transfer and realize passive shut-off function, which could not be realized by the previous design. An experimental apparatus using the fluoride salt FLi Na K was constructed to conduct a series of preliminary solidification and melting experiments. In addition, the enthalpy-porosi...     

Radiation dose distribution of liquid fueled thorium molten salt reactor

A liquid fueled thorium molten salt reactor(TMSR-LF),one of the Generation IV reactors,was designed by the Shanghai Institute of Applied Physics,Chinese Academy...     

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

The impregnation behavior of molten 2LiF–BeF_2(FLiBe) salt into a graphite matrix of fuel elements for a solid fuel thorium molten salt reactor(TMSR-SF) at pressures varying from 0.4 to 1.0 MPa was studied by mercury intrusion, molten salt impregnation, X-ray diffraction, and scanning electron microscopy techniques.It was found that the entrance pore diameter of the graphite matrix is less than 1.0 μm and the contact angle is about 135°. The threshold impregnation pressure was found to be around 0.6 MPa experimentally, consistent with the predicted value of 0.57 MPa by the Washburn equation. With the increase of pressure from 0.6 to 1.0 MPa, the average weight gain of the matrix increased from 3.05 to 10.48%,corresponding to an impregnation volume increase from 2.74 to 9.40%. The diffraction patterns of FLiBe are found in matrices with high impregnation pressures(0.8 MPa and1.0 MPa). The FLiBe with sizes varying from tens of nanometers to a micrometer mainly occupies the open pores in the graphite matrix. The graphite matrix could inhibit the impregnation of the molten salt in the TMSR-SF with a maximum operation pressure of less than 0.5 MPa.     

Properties of phosphate glass waste forms containing fluorides from a molten salt reactor

Radioactive fluoride wastes are generated during the operation of molten salt reactors(MSRs) and reprocessing of their spent fuel.Immobilization of these wastes in borosilicate glass is not feasible because of the very low solubility of fluorides in this host.Alternative candidates are thus an active topic of research including phosphatebased glasses,crystalline ceramics,and hybrid glass-ceramic systems.In this study,mixed fluorides were employed as simulated MSRs waste and incorporated into sodium aluminophosphate glass to obtain phosphate-based waste form.These waste forms were characterized by X-ray diffraction,Raman spectroscopy,and scanning electron microscopy.Leaching tests were performed in deionized water using the product consistency test A method.This study demonstrates that up to 20 mol%of simulated radioactive waste can be introduced into the NaA1 P glass matrix,and the chemical durability is much better than that of borosilicate.The addition of Fe_2O_3 in the NaAlP glass matrix results in increases of the chemical durability at the expense of fluoride loading(to 6.4 mol%).Phosphate glass vitrification of radioactive waste containing fluorides is a potential method to treat and dispose of MSR wastes.     

小型模块化熔盐快堆燃料管理初步分析

由于燃料随熔盐流动的特性以及可以进行在线添料与处理的特点,液态燃料熔盐堆的燃耗分析与燃料管理和传统固态燃料反应堆有很大不同,需要针对液态燃料熔盐堆的特点重新开发燃耗分析与管理程序。本文针对液态燃料熔盐堆的熔盐流动特性以及在线添料与处理功能,基于MCNP5和ORIGEN2.1燃耗耦合程序,开发了适用于液态燃料熔盐堆的燃料管理程序,并应用于一种小型模块化熔盐快堆的燃料管理和分析,对比分析了5种不同运行方案以及分批在线添料情况下,运行30年期间keff的变化情况及重要核素的演化情况。计算结果表明,采用不断调整添料率的连续在线添料运行方案和固定批量添料的运行方案,都可以让小型模块化熔盐快堆维持运行在一个较小的keff波动范围之内。开发的燃料管理程序适用于液态燃料熔盐堆的研究,同时可以为液态燃料熔盐堆的设计及燃耗管理和分析提供有价值的参考。     

Transition toward thorium fuel cycle in a molten salt reactor by using plutonium

The molten salt reactor(MSR), as one of the Generation Ⅳ advanced nuclear systems, has attracted a worldwide interest due to its excellent performances in safety, economics, sustainability, and proliferation resistance. The aim of this work is to provide and evaluate possible solutions to fissile 233 U production and further the fuel transition to thorium fuel cycle in a thermal MSR by using plutonium partitioned from light water reactors spent fuel. By using an in-house developed tool, a breeding and burning(BB) scenario is first introduced and analyzed from the aspects of the evolution of main nuclides, net 233 U production, spectrum shift, and temperature feedback coefficient. It can be concluded that such a Th/Pu to Th/~(233)U transition can be accomplished by employing a relatively fast fuel reprocessing with a cycle time less than 60 days. At the equilibrium state, the reactor can achieve a conversion ratio of about 0.996 for the 60-day reprocessing period(RP) case and about 1.047 for the 10-day RP case.The results also show that it is difficult to accomplish such a fuel transition with limited reprocessing(RP is 180 days),and the reactor operates as a converter and burns the plutonium with the help of thorium. Meanwhile, a prebreeding and burning(PBB) scenario is also analyzed briefly with respect to the net 233 U production and evolution of main nuclides. One can find that it is more efficient to produce 233 U under this scenario, resulting in a double time varying from about 1.96 years for the 10-day RP case to about 6.15 years for the 180-day RP case.     

Development of a MCNP5 and ORIGEN2 based burnup code for molten salt reactor

The Molten Salt Reactor(MSR) is one of the six advanced reactor nuclear energy systems for further research and development selected by Generation IV International Forum(GIF),which is distinguished by its core in which the fuel is dissolved in molten fluoride salt.Because fuel flow in the primary loop,the depletion of MSR is different from that of solid-fuel reactors.In this paper,an MCNP5 and ORIGEN2 Coupled Burnup(MOCBurn) code for MSR is developed under the MATLAB platform.Some new methods and novel arrangements are used to make it suitable for fuel flow in the MSR.To consider the fuel convection and diffusion in the primary loop of MSR,fuel mixing calculation is carried out after each burnup time step.Modeling function for geometry with repeat structures is implicated for reactor analysis with complex structures.Calculation for a high-burnup reactor pin cell benchmark is performed using the MOCBurn code.Results of depletion study show that the MOCBurn code is suitable for the traditional solid-fuel reactors.A preliminary study of the fuel mixture effect in MSR is also carried out.     

Process optimization of a closed-chamber distillation system for the recovery of FLiNaK molten salt

Low-pressure distillation has been proposed as a suitable technique for the recovery of carrier salt from molten salt reactor spent fuel. A closed-chamber disti...     

Numerical study of the dynamic characteristics of a single-layer graphite core in a thorium molten salt reactor

A reactor core in a thorium molten salt reactor uses graphite as a moderator and reflector. The graphite core is a multi-layered arrangement of graphite bricks that are loosely connected to each other using a system of keys and dowels. Consequently, the graphite core is a type of discrete stack structure with highly nonlinear dynamic behavior. Hence, it is important to investigate the dynamic characteristics of the graphite core. In this study, a threedimensional single-layer graphite core model, which is a part of the thorium molten salt reactor side reflector structure, was analyzed using the explicit method in ABAQUS 2016 to study the core dynamic behavior when subjected to different excitations. The design parameters,such as the diameter of the dowel, the gap between key and keyway and the bypass flow gap between two adjacent bricks, were also considered in this model. To reduce excessive demands on available computational resources considering the effect of molten salt, the spring–dashpot model was applied to model the interaction forces between the molten salt and graphite bricks. Numerical simulation results show that the effect of molten salt is a reduction inthe peak maximal principal stress, and a larger gap between two bricks is beneficial to maintain the integrity of the graphite core under earthquake loading. The results obtained by the simulation can be used as a reference for future designs of a molten salt graphite core.