Multiple recycling of fuel in prototype fast breeder reactor

In a thermal neutron reactor, multiple recycle of U-Pu fuel is not possible due to degradation of fissile content of Pu in just one recycle. In the FBR closed fuel cycle, possibility of multi-recycle has been recognized. In the present study, Pu-239 equivalence approach is used to demonstrate the feasibility of achieving near constant input inventory of Pu and near stable Pu isotopic composition after a few recycles of the same fuel of the prototype fast breeder reactor under construction at Kalpakkam. After about five recycles, the cycle-to-cycle variation in the above parameters is below 1%.      

Ultrasonic decontamination of prototype fast breeder reactor fuel pins

Fuel pin decontamination is the process of removing particulates of radioactive material from its exterior surface. It is an important process step in nuclear fuel fabrication. It assumes more significance with plutonium bearing fuel known to be highly radio-toxic owing to its relatively longer biological half life and shorter radiological half life. Release of even minute quantity of plutonium oxide powder in the atmosphere during its handling can cause alarming air borne activity and may pose a severe health hazard to personnel working in the vicinity. Decontamination of fuel pins post pellet loading operation is thus mandatory before they are removed from the glove box for further processing and assembly. This paper describes the setting up of ultrasonic decontamination process, installed inside a custom built fume-hood in the production line, comprising of a cleaning tank with transducers, heaters, pin handling device and water filtration system and its application in cleaning of fuel pins for prototype fast breeder reactor. The cleaning process yielded a typical decontamination efficiency of more than 99%.     

混合堆共生系统核燃料循环模式情景研究

为计算混合堆在未来燃料循环过程中起到的作用,进行了混合堆共生系统物料平衡计算。根据我国核电发展现状和中长期发展规划及中长期（2030年、2050年）发展战略研究,并充分考虑了我国经济发展速度、人口数量和人均用电量,计算得到了2100年之前,我国核电机组装机容量。假定不同堆型搭配的混合堆共生系统核燃料循环的4种情景并建立对应的物料平衡模型进行计算。计算结果表明,压水堆、混合堆和快堆共生模式能最大限度的减少天然铀的需求和节约乏燃料处置费用。     

混合堆共生系统核燃料循环模式情景研究

为计算混合堆在未来燃料循环过程中起到的作用,进行了混合堆共生系统物料平衡计算。根据我国核电发展现状和中长期发展规划及中长期(2030年、2050年)发展战略研究,并充分考虑了我国经济发展速度、人口数量和人均用电量,计算得到了2100年之前,我国核电机组装机容量。假定不同堆型搭配的混合堆共生系统核燃料循环的4种情景并建立对应的物料平衡模型进行计算。计算结果表明,压水堆、混合堆和快堆共生模式能最大限度的减少天然铀的需求和节约乏燃料处置费用。     

On feasibility of a closed nuclear power fuel cycle with minimum radioactivity

Practical implementation of a closed nuclear fuel cycle implies solution of two main tasks. The first task is creation of environmentally acceptable operating conditions of the nuclear fuel cycle considering, first of all, high radioactivity of the involved materials. The second task is creation of effective and economically appropriate conditions of involving fertile isotopes in the fuel cycle. Creation of technologies for management of the high-level radioactivity of spent fuel reliable in terms of radiological protection seems to be the hardest problem.     

Operation of CANDU power reactor in thorium self-sufficient fuel cycle

This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the same as for standard operation in uranium cycle. Two modes of operations are discussed in the paper: mode of preliminary accumulation of ²³³U and mode of operation in self-sufficient cycle. For the mode of accumulation of ²³³U it was assumed for calculations that plutonium can be used as additional fissile material to provide neutrons for ²³³U production. Plutonium was placed in fuel channels, while ²³²Th was located in target channels. Maximum content of ²³³U in target channels was estimated to be ∼13 kg/t of ThO₂. This was achieved by irradiation for six years. The start of the reactor operation in the self-sufficient mode requires ²³³U content to be not less than 12 kg/t. For the mode of operation in self-sufficient cycle, it was assumed that all channels were loaded with identical fuel assemblies containing ThO₂ and certain amount of ²³³U. It is shown that nonuniform distribution of ²³³U in fuel assembly is preferable.      

Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.     

Prediction of the thermophysical properties of molten salt fast reactor fuel from first-principles

Molten fluorides are known to show favourable thermophysical properties which make them good candidate coolants for nuclear fission reactors. Here we investigate the special case of mixtures of lithium fluoride and thorium fluoride, which act both as coolant and as fuel in the molten salt fast reactor concept. By using ab initio parameterised polarisable force fields, we show that it is possible to calculate the whole set of properties (density, thermal expansion, heat capacity, viscosity and thermal conductivity) which are necessary for assessing the heat transfer performance of the melt over the whole range of compositions and temperatures. We then deduce from our calculations several figures of merit which are important in helping the optimisation of the design of molten salt fast reactors.     

Possibility of utilizing the D-3He fuel cycle with 3He production in a spherical tokamak reactor

Possible operating regimes of a spherical tokamak reactor based on the D-³He fuel cycle with ³He production are considered. The parameters of the plasma and magnetic system are calculated for several versions corresponding to the high power efficiency (with a power gain factor in plasma of Q = 20) in a reactor with an aspect ratio of A = 1.5. According to calculations, for an axial magnetic field in vacuum of B ₀ = 2 T, a plasma radius of a = 3 m, an average 〈β〉 value of 0.53, and a plasma temperature of 〈T〉 = 48 keV, the reactor power can reach P _fus = 500 MW. In order to achieve a power of P _fus = 1500 MW in a reactor with a = 2 m, 〈β〉 = 0.36, and 〈T〉 = 40 keV, the magnetic field should be increased to B ₀ = 5 T.     

铅冷快堆M2LFR-1000堆芯燃料管理方案设计

堆芯燃料管理是反应堆设计中极为重要而且复杂的工作,直接影响着堆芯的经济性。目前国内外对于压水堆等传统热堆已有了较为丰富和成熟的燃料管理计算方法,但对于快堆,由于其中子能谱硬,与传统热堆相比有着不同的控制方式和功率分布,快堆的堆芯燃料管理缺乏系统研究。针对中国科学技术大学自主研发的强迫循环冷却的铅基快堆M2LFR-1000,应用SRAC/COREBN软件包进行堆芯燃耗计算,根据燃耗深度提取核素核子密度,计算伪平衡循环参数进行燃料管理预估,然后进行首循环装料、过渡循环和平衡循环燃料管理方案设计。结果表明:对M2LFR-1000堆芯外区燃料换料组件Pu的富集度进行优化,可以延长换料周期到540 d,提高平均卸料燃耗深度;伪平衡循环结果与平衡循环基本一致,伪平衡循环可以用于燃料管理预估。     

Analysis of features of hydrodynamics and heat transfer in the fuel assembly of prospective sodium reactor with a high rate of reproduction in the uranium-plutonium fuel cycle

The fast sodium reactor fuel assembly (FA) with U–Pu–Zr metallic fuel is described. In comparison with a “classical” fast reactor, this FA contains thin fuel rods and a wider fuel rod grid. Studies of the fluid dynamics and the heat transfer were carried out for such a new FA design. The verification of the ANSYS CFX code was provided for determination of the velocity, pressure, and temperature fields in the different channels. The calculations in the cells and in the FA were carried out using the model of shear stress transport (SST) selected at the stage of verification. The results of the hydrodynamics and heat transfer calculations have been analyzed.     

Modelling of a closed two-dimensional reactor bounded by a wavy wall

In this paper, we propose a model for a two-dimensional closed reactor bounded by a wavy wall. The left, right and top walls of the reactor are assumed to be flat surfaces while the bottom wall is a wavy surface. In order to formulate a model for such a reactor, we introduce a coordinate transformation into the dimensionless equations of a rectangular closed domain. Then the resulting equations illustrate the phenomena for a closed reactor bounded by a wavy wall. We solve these equations using the finite difference method. The astonishing results are that the intensity of streamlines and the maximum temperature within the reactor significantly increase with an increase of the number of waves in the bottom wall, the amplitude of waves and the Frank-Kamenetskii number. Converse characteristics are observed for higher values of the enlargement of a wave. Moreover, larger Rayleigh number induces stronger vortices in the flow field and reduces the maximum temperature. The Nusselt number at the bottom wavy wall is found to increase for higher values of the Frank-Kamenetskii number and the amplitude of a wave. A transition from the steady-state to the oscillatory convection is identified for a certain value of the Frank-Kamenetskii number. However, for a low value of the Rayleigh number, there occurs a transition from the steady-state to an explosion for increasing value of the Frank-Kamenetskii number. Results also demonstrate that the critical value of the Frank-Kamenetskii number, for which a transition from the steady-state to the oscillatory convection occurs, is higher for increasing values of the number of waves, the enlargement of a wave and the amplitude of a wave.     

不同燃料循环方案的可持续性与经济性分析

基于核燃料循环政策技术的成熟度,选取了一次通过循环方案(OTC)、单次复用循环方案(TTC)、快堆闭式循环方案(FRC)及混合堆闭式循环方案(HRC)四种典型的核燃料循环方案进行分析。采用平衡物质流模型对不同燃料循环方案的可持续性进行研究,基于平准化电力成本计算方法对不同方案的燃料成本和乏燃料处置成本进行分析。研究结果表明：闭式燃料循环可极大减少核废料产生; 燃料可自持的FRC方案及HRC方案可使用贫铀做燃料而不消耗天然铀; 仅考虑燃料成本和乏燃料处置成本时,HRC方案的经济性最高而TTC方案的经济性最差。     

超临界水堆燃料组件选型论证研究

开展SCWR燃料组件性能分析及设计论证,分析研究SCWR组件设计目标,对比分析各类组件设计理念的物理热工性能与结构可实现性,论证选取综合性能优化的SCWR燃料组件方案。组件选型论证研究表明,大水棒方形组件方案采用结构设计较为简单的单水棒、组合式方形燃料组件,能够为燃料元件提供充分、均匀的慢化,而且慢化剂和冷却剂分流简单,具备良好的物理热工性能与结构可实现性,满足SCWR组件设计目标。     

超临界水堆燃料组件选型论证研究

开展SCWR燃料组件性能分析及设计论证,分析研究SCWR组件设计目标,对比分析各类组件设计理念的物理热工性能与结构可实现性,论证选取综合性能优化的SCWR燃料组件方案。组件选型论证研究表明,大水棒方形组件方案采用结构设计较为简单的单水棒、组合式方形燃料组件,能够为燃料元件提供充分、均匀的慢化,而且慢化剂和冷却剂分流简单,具备良好的物理热工性能与结构可实现性,满足SCWR组件设计目标。     

Optical spectrum of a closed cycle MHD generator plasma

The spectra of argon-cesium plasmas, used as working fluids of closed-cycle MHD generators, were investigated in the visible and near-i.r. regions by using a Hilger-Watts medium quartz spectrograph. A number of argon and cesium lines was observed, as well as some lines originating from the impurities present in the cesium capsule. The molecular bands of aluminum monoxide have also been detected. The population temperatures estimated from relative intensities of cesium lines lie below those obtained by using either the line-reversal method or recombination-radiation measurements. The intensities of some neutral cesium lines follow the Boltzmann population relation at the population temperature. The Zeeman effect and reversed profiles of cesium resonance lines were not observed because of low resolution of the spectrograph used and because of light scattering on the film.     

Frequency dependence of acoustic resonances on blockage Position in a fast reactor subassembly wrapper

Monitoring changes of the acoustic resonance frequencies in the subassemblies of sodium cooled fast reactors is a potential method for blockage detection. The dependence of the resonance frequencies on blockage position along the subassembly has been investigated experimentally and the results compared with model predictions.     

Assessment of sensitivity of neutron-physical parameters of fast neutron reactor to purification of reprocessed fuel from minor actinides

The work is devoted to computational investigation of the dependence of basic physical parameters of fast neutron reactors on the degree of purification of plutonium from minor actinides obtained as a result of pyroelectrochemical reprocessing of spent nuclear fuel and used for manufacturing MOX fuel to be reloaded into the reactors mentioned. The investigations have shown that, in order to preserve such important parameters of a BN-800 type reactor as the criticality, the sodium void reactivity effect, the Doppler effect, and the efficiency of safety rods, it is possible to use the reprocessed fuel without separation of minor actinides for refueling (recharging) the core.     

ITER驱动混合堆次临界包层燃料区结构设计与分析

基于国际热核聚变实验反应堆磁约束聚变-裂变混合能源系统次临界包层的物理-热工设计结果,提出了聚变-裂变混合堆次临界包层燃料区结构设计方案,包括纵骨支撑结构、燃料区结构和锆包壳结构。运用Pro/e建模软件建立了次临界包层燃料区结构模型,并利用ANSYS-workbench mechanical有限元分析软件对纵骨式支撑结构开展了初步力学分析,得到了燃料部件和纵骨式多层支撑结构的最大Tresca应力值、应力分布云图和总变形量,其中最大应力为87.04 MPa,最大变形量为0.17mm。按照第3强度理论校核,计算结果表明纵骨式次临界包层结构各部件能够满足强度要求。