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.     

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.     

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。     

Analysis of heat source distribution in internal circulating surface heat transfer molten salt reactor

The Molten Salt Reactor (MSR) is one of the Generation IV nuclear reactor concepts that were selected by the Generation IV International Forum in 2000. The concept is based on liquid fuel instead of solid fuel assemblies. Besides the advantages, there are several aspects of operation that can hinder the realization of this reactor concept. In this paper, the authors investigate the neutronics behaviour of a new sub-concept that offers solutions for many of the technical problems. The analysis was performed using the particle transport code MCNPX 2.7. The paper focuses on the short-term and steady state heat source distribution in the fuel salt and in the graphite moderator. Accordingly, neither burn-up effects nor reactivity transients are considered. The sensitivity of the effective multiplication factor on different geometrical and material parameters was studied. The results obtained indicate that the main region of heat deposition is in the internal and external channels of the graphite moderator. Only a few percent of the total heat power is released in the graphite moderator, where the gamma and neutron related heat deposition is on the same scale. The results also prove that the heat source distribution does not change drastically upon the actuation of the control rods.     

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.     

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.     

DRAGON&DONJON程序在MSR中堆芯燃耗计算的适用性

DRAGONDONJON组件-堆芯"两步法"程序通过合理简化,理论可适用于任何堆芯与工况。使用蒙特卡罗方法 RMC(Reactor Monte Carlo code)、MCNP(Monte Carlo Neutron Particle transport code)程序验证DRADON程序是否能够承担快/热谱型熔盐堆(Molten Salt Reactor,MSR)焚烧TRU、Th U燃料燃耗计算。选出熔盐增殖堆(Molten Salt Breeder Reactor,MSBR)与熔盐锕系元素再循环和嬗变堆(Molten Salt Advanced Reactor Transmuter,MOSART)堆型进行比较,同时分别利用RMC程序验证DRAGON程序组件燃耗计算的准确性,利用MCNP程序验证DRAGON程序组件均匀化方法以及DONJON程序截面调用和程序全堆扩散的准确性。结果表明,组件燃耗计算中,TRU和Th U燃料满足燃耗计算要求;堆芯临界计算中,快/热谱堆芯计算误差均小于0.001。证明DRADON程序可以胜任快、热谱型MSR焚烧TRU、Th U燃料的物理计算任务。     

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。     

Cascade subcritical molten salt reactor (CSMSR): Main features and restrictions

The ADS-burner of minor actinides (MA) is considered of the following assumptions:

• Proton accelerator — driver has energy E_a=1 GeV and current I_a=10 mA;

• Subcriticality is Δk = 0.05 (k_eff=0.95);

• Reactor consists of two cores: core-1 of cascade amplification (CA) and core-2 of transmutation (CT);

• Molten salt NaF-ZrF₄ is used as a coolant and the nuclear fuel solvent in CT.

Technical solutions are chosen close to used in reactor technology. The main neutron physics characteristics of reactor are calculated including reactor power distribution, reactivity effects, MA burnup, thermo-hydraulics of CA fuel elements etc.Such CSMSR with power 800 MWth can incinerate 50 kg MA/year, i.e. the MA production of five thermal reactors of the same power.     

Recovery of minor actinides from spent molten salt waste and decontamination of molten salt waste

Recovery of minor actinides from spent molten salt is one of the important issues. Decontamination of spent molten salt waste is also the problem to be solved for establishment of pyrochemical reprocessing. The decontamination method of spent molten salt waste with recovery of minor actinides has been proposed. Our proposed process is based on the hydrometallurgical process. This process consists of the following processes. First, the spent molten salt waste is dissolved in aqueous solution. Next, the minor actinides are recovered by chromatographic techniques using the pyridine resin in the methanolic hydrochloric acid solution. In the last process, the spent molten salt waste is decontaminated by the cation-exchange chromatography. In the present paper, the adsorption behavior of minor actinides, rare earth elements, alkaline earth elements, and alkali metal elements on pyridine resin is reported. The demonstration experiment of the recovery of the minor actinides from simulant spent molten salt waste is also reported.     

Improvements and validation of the transient analysis code MOREL for molten salt reactors

The liquid fuel salt used in the molten salt reactors (MSRs) serves as the fuel and coolant simultaneously. On the one hand, the delayed neutron precursors circulate in the whole primary loop and part of them decay outside the core. On the other hand, the fission heat is carried off directly by the fuel flow. These two features require new analysis method with the coupling of fluid flow, heat transfer and neutronics. In this paper, the recent update of MOREL code is presented. The update includes: (1) the improved quasi-static method for the kinetics equation with convection term is developed. (2) The multi-channel thermal hydraulic model is developed based on the geometric feature of MSR. (3) The Variational Nodal Method is used to solve the neutron diffusion equation instead of the original analytic basis functions expansion nodal method. The update brings significant improvement on the efficiency of MOREL code. And, the capability of MOREL code is extended for the real core simulation with feedback. The numerical results and experiment data gained from molten salt reactor experiment (MSRE) are used to verify and validate the updated MOREL code. The results agree well with the experimental data, which prove the new development of MOREL code is correct and effective.     

Preliminary safety analysis for heavy water-moderated molten salt reactor

The heavy water-moderated molten salt reactor(HWMSR)is a newly proposed reactor concept,in which heavy water is adopted as the moderator and molten salt dissolved with fissile and fertile elements is used as the fuel.Issues arising from graphite in traditional molten salt reactors,including the positive temperature coefficient and management of highly radio-active spent graphite waste,can be addressed using the HWMSR.Until now,research on the HWMSR has been centered on the core design and nuclear fuel cycle to explore the viability of the HWMSR and its advantages in fuel utilization.However,the core safety of the HWMSR has not been extensively studied.Therefore,we evaluate typical accidents in a small modular HWMSR,including fuel salt inlet temperature overcooling and overheating accidents,fuel salt inlet flow rate decrease,heavy water inlet temperature overcooling accidents,and heavy water inlet mass flow rate decrease accidents,based on a neutronics and thermal-hydraulics coupled code.The results demonstrated that the core maintained safety during the investigated accidents.     

Project SPHINX spent hot fuel incinerator by neutron flux (the development of a new reactor concept with liquid fuel based on molten fluorides)

The SPHINX project is dealing with a solution of some principle problems of a very promising way of nuclear waste treatment, high level wastes from spent nuclear fuel in particular, by means of transmutation of radionuclides by use of a nuclear reactor with liquid fuel based on molten fluorides, which might be a subcritical system driven by a suitable neutron source. Its superiority lies also in the fact that it makes possible to utilize actinides contained, by others, in spent nuclear fuel and so to reach a positive energy effect.

The SPHINX project has been proposed by the consortium TRANSMUTATION being established by four leading nuclear research bodies in the Czech Republic (Nuclear Research Institute Rez plc, SKODA Nuclear Machinery plc in Pilsen, Nuclear Physics Institute of Academy of Sciences in Rez and Technical University in Praha) at the end of 1996 to which Technical University in Brno (specialized for a secondary circuit problems) has associated in the year 2000. The project has been supported by the Ministry of Industry and Trade of the Czech Republic, CEZ, a.s. (Czech Electricity Generating Company) and RAWRA (Radwaste Repository Authority).

The R&D program of the SPHINX project contains an experimental part, which serves for a verification of design inputs for designing a demonstration unit of a transmuter with liquid fuel based on molten fluorides. The current status of the experimental program performance has been focused upon the irradiation of samples of molten-salt systems as well as structural materials proposed for the blanket of the SPHINX transmuter in the field of high neutron flux of research reactors.

The main aims of this program called Irradiated Probes BLANKA are the following: (1) Experimental verification of long time behavior of transmuter blanket which contains molten fluoride salts as a fuel and coolant, (2) Validation of computational code system being developed for the computation of actinides concentration in long- term operation of the transmuter, and (3) Material research on behavior of materials in neutron and gamma fields, and materials interactions on high temperature conditions.

At present, two agreements on multinational cooperation in this field have been signed: One with European Commission and one with Russian Kurchatov Institute (joint experimental programs AMPULA containing fluorides of transuranium elements like Np, Pu, Am and Cm in irradiated samples and a joint development of the ISTAR code).     

Numerical study on rotordynamic coefficients of the seal of molten salt pump

Molten salt pump is applied to pump high-temperature molten salt as the primary coolant of a molten salt reactor.The pump,generally a vertical rotor system,suffers from radial force generated by the liquid seal component,and the rotordynamic characteristics of the pump are affected considerably.In this paper,the rotordynamic coefficients of the tooth-on-stator liquid seal in molten salt pump are studied.The flow in the seal region is simulated using computational fluid dynamics technique.Parameters of the inlet loss and the pre-swirl at the inlet region of the seal are calculated.The coefficients of resistance and the wall parameters are obtained from the simulation results by data fitting method.The rotordynamic coefficients are analyzed based on the bulk-flow model of liquid seal.The rotordynamic characteristics,with and without the liquid seal,of the objective molten salt pump are inspected.The first critical speed of the rotor is found to increase.Harmonic analysis shows that the pump,being sensitive to unbalance force though,can operate safely under its design specifications.     

Implications of transmutation on the defect chemistry in crystalline waste forms

Radioactive decay within the solid state creates chemical environments which are typically incommensurate with the initial host structure. Using a combined theoretical and computational approach, we discuss this ‘transmutation problem’ in the context of the short-lived fission products Cs-137 and Sr-90. We show how a Kröger-Vink treatment is insufficient for understanding defects arising from transmutation, and present density functional theory data for chemical evolution within two prototypical hosts, CsCl and SrTiO₃. While the latter has a strong driving force for phase separation with increasing Zr content, the Cs(Ba)Cl system is surprisingly stable. The sharp difference between these two findings points to the need for better understanding of novel chemistry in nuclear waste forms.     

钍基液态燃料熔盐实验堆氙毒分析

熔盐堆作为第四代核能系统堆型之一,液态燃料形态的特点使其可以实现在线处理和在线添料。为了提高中子经济性可以利用在线处理的氦鼓泡法,将氦气通入反应堆一回路,去除堆芯内的裂变气体(如Xe、Kr)。基于钍基熔盐液态堆(Thorium Molten Salt Reactor-Liquid Fuel1,TMSR-LF1)概念设计,结合熔盐实验堆(Molten Salt Reactor Experiment,MSRE)氙毒模型,分析了鼓泡法去除氙毒中~(135)Xe扩散规律和去除效率对氙毒的影响,并给出了对应的初始有效增殖因子的变化规律。分析结果表明,虽然存在~(135)Xe会大量向石墨扩散的可能性,但是鼓泡法仍然可以有效去除TMSR-LF1堆芯内的~(135)Xe,减小堆芯毒性,提高反应性。     

The advanced high-temperature reactor: High-temperature fuel, liquid salt coolant, liquid-metal-reactor plant

The Advanced High-Temperature Reactor is a new reactor concept that combines four existing technologies in a new way: (1) coated-particle graphite-matrix nuclear fuels (traditionally used for helium-cooled reactors), (2) Brayton power cycles, (3) passive safety systems and plant designs from liquid-metal-cooled fast reactors, and (4) low-pressure liquid-salt coolants with boiling points far above the maximum coolant temperature. The new combination of technologies enables the design of a large [2400- to 4000-MW(t)] high-temperature reactor, with reactor-coolant exit temperatures between 700 and 1000°C (depending upon goals) and passive safety systems for economic production of electricity or hydrogen. The AHTR [2400-MW(t)] capital costs have been estimated to be 49 to 61% per kilowatt (electric) relative to modular gas-cooled [600-MW(t)] and modular liquid-metal-cooled reactors [1000-MW(t)], assuming a single AHTR and multiple modular units with the same total electrical output. Because of the similar fuel, core design, and power cycles, about 70% of the required research is shared with that for high-temperature gas-cooled reactors.