LiF对LiCl-KCl熔盐中电解精炼钆的影响

研究了LiF加入LiCl-KCl熔盐对钆电化学及络合行为的影响,发现LiF加入LiCl-KCl熔盐后,钆、铽的还原电位差由原来的6mV变为67mV。利用电化学方法和光谱方法研究了熔盐中钆离子和铽离子的配位结构,发现LiCl-KCl-GdCl_3(5mol%)/TbCl_3(5mol%)熔盐中存在[GdCl_6]~(3-)、[TbCl_6]~(3-)的正八面体结构;考察了LiF加入LiCl-KCl熔盐对钆、铽离子结构的影响,在LiCl-KCl-GdCl_3/TbCl_3中加入LiF后,钆离子和铽离子配位结构均为络合了3个F~-和3个Cl~-的八面体结构[GdF_3Cl_3]~(3-)和[TbF_3Cl_3]~(3-),计算得到两种八面体结构的相对累积稳定常数分别为10.98和6.38。以此为理论基础,进行了LiF对LiCl-KCl熔盐中钆电解精炼的影响研究,发现将LiF加入LiCl-KCl熔盐后进行钆电解精炼时,能以更高的去污系数分离钆。     

铝合金化技术在乏燃料干法后处理中的应用研究进展

干法后处理在未来先进核燃料循环中将发挥关键作用。由美国开发的熔盐电精炼流程是目前最具应用前景的干法后处理流程之一,但是锕系元素(An)与镧系元素(Ln)的高效分离仍然是该流程目前亟待解决的关键科学与技术问题之一。研究表明,An与Ln形成铝合金时沉积电位差较大,采用固态铝电极电解有望实现An与Ln的有效分离,从而更好地服务于分离-嬗变策略。本文针对铝合金化技术在乏燃料干法后处理中的应用研究进展进行综合阐述,重点介绍铝合金化在熔盐电精炼中的应用研究,主要包括Ln和An的铝合金化行为、An和Ln的铝合金化分离等几个方面。     

Electrolysis of Burnup-Simulated Uranium Nitride Fuels in LiCl-KCl Eutectic Melts

The electrochemical behavior of burnup-simulated uranium nitride fuels containing representative solid fission product elements, UN+Mo (Mo = 2.84 wt%), UN+Pd (Pd = 4.6 wt%) and (U, Nd)N (NdN = 8.0 wt%), was investigated in the molten LiCl-KCl eutectic salt with 0.54 wt% UCl₃ in order to clarify the effects of fission products on the dissolution of actinide nitrides and the behavior of FPs in the electrorefining of spent nitride fuel. The rest potentials of burnup-simulated UN pellets were similar to that of pure UN. The electrochemical dissolution of UN began at about _0:75V vs Ag/AgCl reference electrode in all samples as well as that of pure UN. After the electrolyses at the constant anodic potential of ?0:65––0:60V vs Ag/AgCl, most of UN was dissolved into LiCl-KCl as UCl₃ at the anode, and U was recovered in the liquid Cd cathode in all samples. Furthermore, Nd was dissolved at the anode and accumulated into LiCl-KCl as NdCl₃, while Mo and Pd were not dissolved but remained at the anode.     

A study on reactivity insertion controlled LMR cores with metallic fuel

An inherently safe core concept with metallic fuel for sodium cooled fast reactor is proposed that has a negative void reactivity at the loss of coolant events without scram as well as a small excess reactivity during the operation cycle. The relationship of sodium void reactivities and burn-up reactivities to different core configurations has been studied quantitatively to clarify the core concept for large metallic fuel reactors. It has shown that the sodium void reactivity is greatly dependent on the core shapes while the excess reactivity is on the fuel compositions. It has also indicated that the core configuration that enables to enhance the neutron streaming through the region above the active core at coolant voiding is most effective to decrease sodium void reactivity.

A 3000 MWt core composed of the flat inner core and annular outer core where the fuel volume fraction is relatively high and the sodium plenum is placed just above the active core region has been selected as a candidate core.

The maximum excess reactivity of the candidate core at UTOP is about 0.4 $ and it can be reduced to approximately zero by power or inlet temperature adjustment during the operation cycle, meanwhile the sodium void reactivity is as low as -1.3 $ in negative that is enough to prevent ULOF sequences.     

Separation of Pure LiCl-KCl Eutectic Salt from a Mixture of LiCl-KCl Eutectic Salt and Rare-Earth Precipitates by Vacuum Distillation

In this study, the vacuum distillation of LiCl-KCl eutectic salt in a mixture of LiCl-KCl eutectic salt and rare-earth precipitates was carried out to evaluate the vaporization characteristics of LiCl-KCl eutectic salt. It was confirmed that the required time for salt vaporization was reduced by a reduction in the pressure. It appeared that the vaporization of LiCl-KCl eutectic salt containing rare-earth precipitates was decreased in comparison with that of pure salt because the salt adhered to the fine particles of the rare-earth precipitates. However, the distillation of the salt was almost achieved by elevating the surface area and further reducing the pressure. The distilled salt from the mixture consisted of 43.7 wt% LiCl and 56.3 wt% KCl. It is thought that the recovered salt can be reused because its composition is similar to the mixed ratio (44.2 wt% LiCl: 55.8 wt% KCl) of the salt used in an electrorefining process.     

LiCl-KCl共晶熔盐中氧离子的去除

采用NH₄Cl和HCl气体进行LiCl-KCl共晶熔盐中氧离子的去除。在使用NH₄Cl和HCl气体去除LiCl-KCl共晶熔盐中的氧离子过程中，用钇稳定氧化锆测氧电极对熔盐中的氧离子浓度变化进行测定。结果表明，HCl与熔盐中氧离子反应生成H₂O，并将反应产物水通过HCl载带出去。NH₄Cl去除氧离子的过程也是通过NH₄Cl分解的HCl与氧离子反应除去熔盐中氧离子。NH₄Cl和HCl均能有效地去除LiCl-KCl熔盐中的氧离子，使氧离子浓度降低至10^-5～10^-4 mol/kg。     

LiCl-KCl共晶熔盐中氧离子的去除

采用NH₄Cl和HCl气体进行LiCl-KCl共晶熔盐中氧离子的去除。在使用NH₄Cl和HCl气体去除LiCl-KCl共晶熔盐中的氧离子过程中，用钇稳定氧化锆测氧电极对熔盐中的氧离子浓度变化进行测定。结果表明，HCl与熔盐中氧离子反应生成H₂O，并将反应产物水通过HCl载带出去。NH₄Cl去除氧离子的过程也是通过NH₄Cl分解的HCl与氧离子反应除去熔盐中氧离子。NH₄Cl和HCl均能有效地去除LiCl-KCl熔盐中的氧离子，使氧离子浓度降低至10^-5～10^-4 mol/kg。     

Integrated Experiments of Electrometallurgical Pyroprocessing Using Plutonium Oxide

Electrometallurgical pyroprocessing is a promising technology to realize actinide fuel cycle. Integrated experiments to demonstrate electrometallurgical pyroprocessing of PuO₂ in continuous operation were carried out. In each test, 10–20 g of PuO₂ was reacted with Li reductant to form metal product. The reduction products were charged in an anode basket of the electrorefiner with LiCl-KCl-UCl₃ electrolyte. Using the anode, deposition of uranium on the solid cathode was carried out when PuCl₃/UCl₃ concentration ratio was low. After the Pu/U ratio in the salt electrolyte was increased enough, Pu and U were recovered simultaneously on a liquid cadmium cathode. By heating up the deposits for distillation of the salt and the cadmium, U metal or Pu-U alloyed metal was obtained as residues in the crucible. It was the first result to demonstrate the recovery of metal actinides in the continuous operation of pyroprocessing of oxide fuels.     

LiCl-KCl共晶熔盐的纯化

氯化锂-氯化钾共晶熔盐是电解精炼干法后处理中最常用的电解质,其含有的杂质直接影响电流效率和产物纯度。本研究分别采用高温处理、HCl气体鼓泡和恒电位电解等方法依次去除了熔盐中的易挥发物质、氧离子和金属离子等杂质,获得了较高纯度的熔盐。采用热重分析(TGA)、电化学和电感耦合等离子体原子发射光谱(ICP-AES)等方法对比了纯化前后熔盐中各杂质的含量。研究结果表明:去除易挥发杂质的最佳处理温度范围为450～650℃;去除杂质金属离子时最佳电解电位为-2.3Vvs.Ag/AgCl(摩尔分数2%),恒电位电解800s后杂质金属离子总量低于1.5×10-6 g/g(盐)。以上研究结果表明,采用高温处理、HCl气体鼓入和恒电位电解可获得纯度较高的LiCl-KCl共晶熔盐。     

Fabrication of DUPIC fuel pellets using high burn-up spent PWR fuel

Technology for the direct usage of a spent PWR fuel in CANDU reactors (DUPIC) was developed in KAERI to reduce the amount of spent fuel. DUPIC fuel pellets were fabricated using a dry processing method to re-fabricate CANDU fuel from spent PWR fuel without any intentional separation of fissile materials or fission products. The DUPIC fuel element fabrication process satisfied a quality assurance program in accordance with the Canadian standard. For the DUPIC fuels with various fuel burn-ups between 27,300 and 65,000 MWd/tU, the sintered pellet density decreased with increasing fuel burn-ups. Fission gas releases and powder properties of the spent fuel also influenced the DUPIC fuel characteristics. Measurement of cesium content released from green pellets revealed that their sintered density significantly depended on sintering temperature history. It was useful to establish a DUPIC fuel fabrication technology in which a high-burn-up fuel with 65,000 MWd/tU was treated.     

Investigation of the MSFR core physics and fuel cycle characteristics

The adoption of Th fuel in fast reactors is being reconsidered due to the potential favorable impact on actinide waste management and resource availability. A closed Th cycle leads to an actinide inventory with lower radiotoxicity and heat load for the first several thousands of years. Due to the typically low TRansUranic (TRU) Conversion Ratio (CR), Th can also be advantageous to expedite the consumption of legacy TRU. One of the main obstacles to the implementation of Th is the highly radioactive recycled fuel which requires remote handling under heavy shielding, inevitably penalizing economics and challenging conventional pin-based fuel manufacturing. From this perspective, the development of liquid-fuelled reactors, with Molten Salt Reactors regarded as the most promising, appears particularly attractive as fuel handling would be greatly simplified. The present paper investigates the fuel cycle performances of the reference GEN-IV Molten Salt Fast Reactor (MSFR) in terms of isotope evolution, radiotoxicity generation and safety-related parameters. Similarly to most MSR concepts proposed in the past, the MSFR is based on the fluoride molten salt technology, but it features the novelty of a fast neutron spectrum. Calculations are performed using state-of-the-art equilibrium-cycle methodologies, i.e., the ERANOS-based EQL3D procedure developed at the Paul Scherrer Institut and extended to the simulation of the MSFR. Selected results have been benchmarked with the Monte Carlo code PSG2/SERPENT. These results have also been used for the assessment of a diffusion module based on the COMSOL multi-physics toolkit, which is the subject of current studies aimed at efficiently simulating the peculiar MSFR transient behavior.     

Polarization Effects in the Molten Salt Electrorefining of Spent Nuclear Fuel

The effects of polarization in the electrorefining of spent nuclear fuel were analytically studied through electrotransport experiments with uranium. When the uranium concentration in molten salt was set at 0.5 wt%, polarization caused the measured cell resistance to increase from 0.16 to 0.33 Ω as the cell voltage was raised from 0.1 to 0.7 V. At 2.0 wt% uranium concentration in salt, on the other hand, the resistance was almost independent of cell voltage. A code named DEVON has been developed to estimate the effects of polarization on the electro- refiner operating condition. The Laplace equation is solved in the bulk salt region by finite element method. In the calculation, the effects of polarization are taken into account by adopting the diffusion layer model on defining the boundary conditions. The result of calculations agreed well with the measured resistance data at two sample concentrations for a diffusion layer thickness of 0.025 cm on the solid cathode. The calculations indicated that significant polarization at the cathode could be avoided by maintaining the uranium concentration in salt above 1 to 2 wt%. When it was held at 5 wt%, which is a typical level for normal operation, polarization proved to exert little influence on the cell resistance, but it was indicated to contribute appreciably toward flattering the current distribution along the cathode surface.     

Chemical diffusion coefficient calculation of U3+ in LiCl-KCl molten salt

Self-diffusion coefficient denotes the intrinsic diffusion of elements without the chemical gradient; generally, it is different from the chemical diffusion coefficient that is used in Fick's law. However, the self-diffusion coefficient can be converted to the chemical diffusion coefficient by considering the concentration-dependent activity coefficient of solute. In this study, the multiple time origins method was applied to calculate the self-diffusion coefficient of U³⁺ in LiCl-KCl eutectic salt at different temperatures and concentrations. The chemical diffusion coefficient was obtained based on the calculated results and thermodynamic theory. Calculated results show that self-diffusion coefficient decreases with concentration however chemical diffusion coefficient indicates to change little initially but appears to increase followed by decrease later within the concentration range of 0–3 at%.     

固态燃料熔盐堆自适应功率控制器设计及分析

钍基熔盐堆核能系统(Thorium-based Molten Salt Reactor,TMSR)是中国科学院首批启动实施的战略性先导科技专项,旨在研发第四代反应堆核能系统。固态燃料钍基熔盐实验堆(The Solid Fuel Thorium-based Molten Salt Experimental Reactor,TMSR-SF1)是一个10 MW热功率的氟盐冷却球床堆,目前已经完成方案设计和初步工程设计。功率控制系统是反应堆一个关键控制系统,实现反应堆正常启动、功率运行和正常停堆功能,对保证反应堆安全和稳定运行起着极其重要的作用。根据TMSR-SF1运行控制要求,结合自适应控制理论,基于Lyapunov稳定性理论设计了一种TMSR-SF1模型参考自适应功率控制器。基于TMSR仿真平台,使用MATLAB/Simulink建立了自适应功率控制系统模型,并开展了控制器特性分析。结果表明,自适应功率控制器具备良好的负荷跟随能力,抗干扰能力强、稳定性好、可靠性高,能够满足TMSR-SF1功率控制的要求,确保堆芯的输出功率与功率设定值相匹配。     

Investigation on the Neutronic Performance of a Fusion Reactor Using Flibe with Heavy Metal Fluorides

Using liquid wall between the plasma and solid first wall in a fusion reactor allows to use high neutron wall loads and could eliminate frequent replacement of the first wall structure during reactor’s lifetime. Liquid wall should have a certain effective or optimum thickness to extend solid first wall lifetime to reactor’s lifetime and supply sufficient tritium for deuterium–tritium (DT) fusion driver. This study presents the effect of thickness of flowing liquid wall containing 90 mol % Flibe+10 mol % UF₄ or ThF₄ on the neutronic performance of a magnetic fusion reactor design called APEX. Neutron transport calculations were carried out with the aid of code Scale4.3. Numerical results brought out that optimum liquid wall thickness of ∼38 cm was found for the blankets using Flibe+10% UF₄ whereas, 56 cm for that with Flibe+10% ThF₄. Significant amount of high quality fissile fuel was produced by using heavy metal salt.     

High burn-up operation and MOX burning in LWR; Effects of burn-up and extended cooling period of spent fuel on vitrification and disposal

Looking ahead to final disposal of high-level radioactive waste arising from further utilization of nuclear energy, the effects of high burn-up of light-water reactors (LWR) with UO₂ and MOX fuel and extended cooling period of spent fuel on waste management and disposal were discussed. It was assumed that the waste loading of waste glass is restricted by three factors: heat generation rate, MoO₃ content, and platinum group metal content. As a result of evaluation for effects of extended cooling period, the waste loading of waste glass from both UO₂ and MOX spent fuel could be increased in the current vitrification technology. For the storage of waste glass from MOX spent fuel with higher waste loading, however, those waste glass require long storage period prior to geological disposal because decay heat of ²⁴¹Am contributes significantly. Therefore, the evaluation of effects of Am separation on the storage period was performed. Furthermore, heat transfer calculation was carried out in order to evaluate the temperature of buffer material in a geological repository. The results showed, 70 to 90% of Am separation is sufficiently effective in terms of thermal feasibility of a repository.     

Change of fuel-to-cladding gap width with the burn-up in FBR MOX fuel irradiated to high burn-up

In order to study the dependence of the gap width change on the burn-up, the fuel-to-cladding gap widths were investigated by ceramography in a large number of FBR MOX fuel pins irradiated to high burn-up. The dependence of gap widths on the burn-up was closely connected with the formations of JOG (joint oxyde-gaine) and rim structure. The gap widths decreased gradually due to the fuel swelling until ∼30 GWd/t, but beyond this burn-up the dependence showed two different tendencies. With the increase of burn-up, the gap widths decreased due to the increase of fuel swelling in the low fuel temperature region where the rim structure was observed, but they increased in the high fuel temperature region where the JOG rich in Cs and Mo formed in the gap.     

(JAERI) Fourth International Symposium on Advanced Nuclear Energy Research

To increase the uranium recovery rate of molten salt electrorefining step in pyrometallurgical reprocessing of metallic fast reactor fuels, tests were carried out using electrorefiners equipped with mechanisms for scraping cathode deposits. After the modifications in the design of the anode basket and scraper mechanism, no stalling of the anode and scraper rotation due to interference by cathode deposits occurred. Under the condition that codissolution of zirconium and uranium was allowed in order to obtain maximum throughput, a current of 400–450A was maintained until 82% of the initially loaded uranium was recovered. The uranium recovery rate for the same duration reached 789 g U/h (32.9 g U/h_L per electrode volume). On the assumption that an electrorefiner operates for 20 h/d and 200 d/y in an actual pyrometallugical reprocessing facility, this result corresponds to a uranium recovery rate of 3.16 t U/y using only one electrode assembly of about 30 cm diameter, which should be a sufficiently high performance for practical use. From these results, the engineering feasibility of uranium recovery using an electrorefiner with cathode deposit scraper mechanism has been demonstrated.     

Recovery of Neptunium by Electrolysis of NpN in LiCl-KCl Eutectic Melts

The electrochemical behavior of neptunium nitride, NpN, in the LiCl-KCl eutectic melt containing NpCl₃ at 450, 500 and 550°C was investigated from the viewpoint of the application of electrochemical refining in a fused salt to nitride fuel cycle. The electrochemical dissolution of NpN began nearly at the potential theoretically evaluated, though this reaction was irreversible owing to small partial pressure of N₂ in the salt and the reaction rate was slow. Under the electrolysis in the NpCl₃-LiCl-KCl eutectic melt, NpN was dissolved into the salt as Np³⁺ at the anode, and Np metal was deposited at the cathode. About 0.5 g of Np metal was obtained by heating the deposit containing the salt at 800°C for 3.6 ks.     

高燃耗对乏燃料贮存和运输的影响研究

本文对高燃耗对乏燃料包壳结构完整性的影响进行了分析。探讨了影响包壳结构完整性的重要温度限值,即燃料包壳温度限值、包壳溶解温度以及韧脆转变温度(DBTT)。给出了分析包壳结构完整性的方法,对拟在干式贮存设施内贮存超过20年的容器性能及贮存后运输时乏燃料组件的结构完整性进行了分析,并给出了相关建议。