The thorium cycle for fast breeder reactors

The role that could be played by liquid metal-cooled fast breeder reactors (LMFBRs) in the utilization of India’s considerable thorium resources is reviewed in this article. Distinct advantages of thorium-based fuels over plutonium-uranium fuels in LMFBRs pertain to a more favourable coolant voiding reactivity coefficient and better fuel element irradiation stability. The poorer breeding capability of thorium-fuelled fast reactors can in principle be overcome by improved core design and development of advanced fuel concepts. The technical feasibility of such advanced thorium fuels and core designs must be established by sustained research and development. It is also necessary to efficiently close the thorium fuel cycle of fast breeder reactors by appropriate development of the fuel reprocessing and refabrication stages. The Fast Breeder Test Reactor (FBTR) at Kalpakkam is expected to be an important tool for development of thorium fuel and fuel cycle technology. A quick look at the economics of the thorium cycle for fast reactors, vis-a-vis the more conventional uranium cycle indicates only a small and acceptable cost disadvantage on account of the need for remote fabrication of recycled thorium fuel. The authors felicitate Prof. D S Kothari on his eightieth birthday and dedicate this paper to him on this occasion.     

Neutronic calculations for CANDU thorium systems using Monte Carlo techniques

In this paper,we have investigated the prospects of exploiting the rich world thorium reserves using Canada Deuterium Uranium(CANDU)reactors.The analysis is performed using the Monte Carlo MCNP code in order to understand how much time the reactor is in criticality conduction.Four different fuel compositions have been selected for analysis.We have obtained the infinite multiplication factor,k∞,under full power operation of the reactor over 8 years.The neutronic flux distribution in the full core reactor has already been investigated.     

Neutronic calculations for CANDU thorium systems using Monte Carlo techniques

In this paper, we have investigated the prospects of exploiting the rich world thorium reserves using Canada Deuterium Uranium （CANDU） reactors. The analysis is performed using the Monte Carlo MCNP code in order to understand how much time the reactor is in criticality conduction. Four different fuel compositions have been selected for analysis. We have obtained the infinite multiplication factor, k∞, under full power operation of the reactor over 8 years. The neutronic flux distribution in the full core reactor has already been investigated.     

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。     

Z箍缩驱动聚变-裂变混合堆氚燃料循环系统氚盘存量与氚自持初步分析

为研究氚自持条件,建立了Z-FFR氚分析模型,基于理论方程和氚平均滞留时间方法进行计算,得到稳态运行时排灰气处理系统、氚增殖提取系统、同位素分离系统、水去氚化系统的氚质量流分别为52.30,25.40,81.30,3.60 g/day,对应的氚盘存量为52.30,25.40,8.13,1.80 g。同时以氚质量流推导出氚自持判断条件,分析了设计参数能够满足氚自持要求,同时获得了燃烧效率、氚增殖率、提取效率与氚自持的互补关系,三者作为关键参数相互依存,于临界值、设计值、理想值之间分析了氚的自持情况。     

Improved analysis on multiple recycling of fuel in prototype fast breeder reactor in a closed fuel cycle

An FBR closed fuel cycle involves recycling of the discharge fuel, after reprocessing and refabrication, to utilize the unburnt fuel remains and the freshly bred fissile material. Our previous study in this regard for the PFBR indicated a comfortable feasibility of multiple recycling with selfsufficiency. In the present work, more refined estimations are done using the most recent nuclear data, viz. ENDF/B-VII.0, and with the most recent specification of the fuel composition. Among others, this paper brings out the importance of taking into account the energy self-shielding effects in the cross-section averages used in the study. While self-shielded averages lead to realistic predictions, unshielded averages significantly overpredict breeding in the blankets and underpredict loss in the cores.     

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.     

Investigation in thorium phosphate by NMR II-phosphorus dipolar networks

With through space and through bond experiments in two-dimensional NMR we analyze the transformation from the thorium phosphate-hydrogen phosphate hydrate (TPHPH) to the β form of the thorium phosphate diphosphate (β-TPD) in relation with the phosphorus networks. These techniques are complementary: the through space coupling gives an insight on the dipolar phosphorus networks while the through bond coupling is particularly efficient in the detection of the P₂O₇ groups. With these experiments we show that in a first step, by heating the precursor TPHPH above 250 °C, it transforms into an form of TPD. This transformation is due to the complete condensation of hydrogen phosphate groups HPO₄ into P₂O₇ entities. By heating -TPD above 950 °C it transforms into its well-known β form. The form is characterized by a hygroscopic behavior: some water molecules are present near the P₂O₇ groups that makes non-equivalent their phosphorus nuclei. PO₄ dipolar networks are always present in the form. The main effect of these PO₄ and P₂O₇ units is to give the system a channel structure and the water enters in them.     

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。     

MCNP5在固态燃料熔盐堆功率分布计算的应用

采用蒙特卡罗输运程序MCNP5对固态燃料熔盐实验堆(TMSR-SF1)能量沉积比例及功率分布进行了计算分析。针对MCNP5不能处理缓发β及缓发γ的能量沉积问题进行了类比等效处理。对固态燃料熔盐实验堆在寿期初、寿期中、寿期末相应的能量沉积比例及功率分布进行了研究。通过计算发现,固态燃料熔盐实验堆内燃料球相比于压水堆棒状燃料元件(95%~97%左右)而言,能量沉积比例有所偏小,约为93%。同时,由于堆芯功率分布均匀,功率峰因子较小(约1.5),堆芯安全性较好。     

Determination of thorium and uranium contents in soil samples using SSNTD’s passive method

Thorium-to-uranium ratios have been determined in different soil samples using CR-39 and LR-115-II solid-state nuclear track detectors (SSNTDs). A calibration method based on determination of SSNTD registration sensitivity ratio for α-particles of thorium and uranium series has been developed. Thorium and uranium contents of the standard soil samples have been determined and compared with its known values. There is a good agreement between the results of this method and the values of standard samples. The method is simple, inexpensive, non-destructive and has a wide range of applications in environment, building materials and petroleum fields.     

新型钍基熔盐堆多物理计算模型及分析

针对熔盐堆系统特点,提出了包含堆芯及其他主回路系统在内的多物理紧密耦合计算模型,并在此基础上自主开发了多物理分析程序TANG-MSR。利用该程序进行了新型钍基熔盐堆(TMSR)的设计,并对设计方案进行了稳态及瞬态分析。相关计算结果表明,TANG-MSR所采用的多物理模型能够很好地捕捉熔盐堆的主要物理现象,提出的新型熔盐堆设计在安全性和可持续性方面表现优异。关键词：多物理模型;新型钍基熔盐堆;稳态;瞬态 Abstract: Key words:     

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.     

Structural phase transition and elastic properties of thorium pnictides at high pressure

In the present paper we have pointed out the weaknesses of the approach by Aynyas et al [1] to study the structural phase transition and elastic properties of thorium pnictides. The calculated values of phase transition pressure and other elastic properties using the realistic and actual approach are also given and compared with the experimental and previous theoretical work.      

177堆芯24个月换料可行性研究

与18个月换料相比,压水堆核电站24个月换料能减少大修次数,提高机组负荷因子,增加发电量。基于装载177组件的堆芯,通过提高新燃料组件富集度和增加批换料组件数使堆芯循环长度达到24个月换料周期要求,考虑实际24个月换料和名义24个月换料高低两种电厂可利用因子。考虑燃料组件费用、大修费用、乏燃料处理费用和发电收益等进行换料方案经济性分析评估,并和典型18个月换料经济性作比较。177堆芯平衡循环装载88组富集度为4.95%的燃料组件,能满足名义24个月换料循环长度的需要,组件平均卸料燃耗约48 GWd/tU;装载104个燃料组件的堆芯能满足实际24个换料循环长度的要求,堆芯参数满足相关安全限值要求。结果表明,177堆芯24个月换料具有可行性,其高负荷因子下的经济性与18个月换料相当。     

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.     

新型钍基熔盐堆堆芯方案及燃耗分析

采用自主开发的SONG/TANG-MSR栅格/堆芯分析程序对新型钍基熔盐堆（TMSR）进行堆芯布置与燃耗分析计算。根据前期的栅格分析相关工作,TMSR采用了无铍（BeF2）燃料熔盐、氧化铍慢化剂以及碳化硅包壳,并在组件栅格初步优化分析的基础上,通过全堆芯计算对熔盐栅格进一步优化和分析,给出了堆芯三区布置方案。该方案具有较高的增殖比,负的功率系数,以及较平的温度分布。根据该堆芯方案,在考虑熔盐在线处理情况下进行了熔盐燃耗计算分析。结果表明,堆芯具有较高的增殖比、较短的倍增时间以及长期稳定运行能力。新型的钍基熔盐设计大大提高了增殖性能,同时又确保堆芯具有足够的安全性能。     

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.     

On the role of fusion neutron source with thorium blanket in forming the nuclide composition of the nuclear fuel cycle of the Russian Federation

The possible role of available thorium resources of the Russian Federation in utilization of thorium in the closed (U–Pu)-fuel cycle of nuclear power is considered. The efficiency of application of fusion neutron sources with thorium blanket for economical use of available thorium resources is demonstrated. The objective of this study is the search for a solution of such major tasks of nuclear power as reduction of the amount of front-end operations in the nuclear fuel cycle and enhancement of its protection against uncontrolled proliferation of fissile materials with the smallest possible alterations in the fuel cycle. The earlier results are analyzed, new information on the amount of thorium resources of the Russian Federation is used, and additional estimates are made. The following basic results obtained on the basis of the assumption of involving fusion reactors with Th-blanket in future nuclear power for generation of the light uranium fraction ^232+233+234U and ²³¹Pa are formulated. (1) The fuel cycle would shift from fissile ²³⁵U to ²³³U, which is more attractive for thermal power reactors. (2) The light uranium fraction is the most “protected” in the uranium fuel component, and being mixed with regenerated uranium, it would become reduced-enrichment uranium fuel, which would relieve the problem of nonproliferation of the fissile material. (3) The addition of ²³¹Pa into the fuel would stabilize its neutron-multiplying properties, thus making it possible to implement a long fuel residence time and, as a consequence, increase the export potential of the whole nuclear power technology. (4) The available thorium resource in the vicinity of Krasnoufimsk is sufficient for operation of the large-scale nuclear power industry of the Russian Federation with an electric power of 70 GW for more than one quarter of a century. The general conclusion is that involvement of a small number of fusion reactors with Th-blanket in the future nuclear power industry of the Russian Federation would to a large extent solve its problems and increase its export potential.