FCM燃料热学性能分析

研究了利用有限元分析软件ABAQUS对全陶瓷微封装燃料(FCM燃料)芯块进行热学性能分析的方法,并对FCM燃料芯块和传统UO_2芯块的热学性能进行了对比分析。研究结果表明:FCM芯块温度分布趋势与UO_2芯块相同,但具有较大不均匀性;典型压水堆运行工况下,FCM燃料芯块的燃料温度远小于UO_2芯块的温度;在相同线功率密度下,FCM芯块温度对燃耗变化不敏感;在相同燃耗下,FCM芯块随线功率密度增加温度升高的速率相比UO_2芯块更慢。     

事故容错燃料UO2-BeO应用于压水堆燃料组件的中子学分析

本文基于中子学角度对典型压水堆中的事故容错燃料UO₂-BeO设计进行分析。选取西屋公司的2D燃料组件问题,使用组件计算程序ALPHA对不同体积分数BeO的燃料进行计算。临界及燃耗计算结果表明：在燃料中加入BeO,一方面由于中子吸收,导致反应性惩罚;另一方面由于BeO的慢化作用,导致反应性补偿,两个相反影响相互竞争共同决定UO₂-BeO燃料带来的综合效应。由反应性匹配基准可知,适量增加²³⁵U富集度对维持反应堆整个运行循环的反应性平衡十分必要,其中基准1相对于基准2和3需对燃料富集度进行较大调整才可满足寿期末得到的k_inf与参考组件一致。由反应性扰动分析结果可知,当燃料中加入BeO后,燃料温度系数随BeO体积分数的变化基本保持恒定,慢化剂温度系数降低,空泡系数增高。     

FCM燃料压水堆弥散可燃毒物中子学分析

为将全陶瓷微胶囊封装（FCM）燃料应用于小型压水堆,对FCM燃料组件开展了可燃毒物中子学设计与分析。通过寿期初引入负反应性、寿期内消耗速率和寿期末残留3个方面,对弥散在SiC基体中的弥散型可燃毒物Gd₂O₃、Er₂O₃、Sm₂O₃、Eu₂O₃、Dy₂O₃及HfO₂进行评价。FCM燃料中TRISO颗粒核芯直径达800 μm,燃料颗粒自屏效应强烈,在RMC程序中引入随机介质计算功能,对FCM燃料进行随机几何建模,保证了反应性计算精度。分析表明：Er₂O₃可作为FCM燃料堆芯的候选可燃毒物,Gd₂O₃和Eu₂O₃需结合堆芯开展进一步研究,Sm₂O₃、Dy₂O₃及HfO₂的反应性惩罚过大,不适合作为FCM燃料可燃毒物。     

FCM燃料辐照-热-力耦合性能数值研究

使用ABAQUS软件编写相应的用户自定义子程序,初步建立了全陶瓷微封装燃料(FCM)的辐照-热-力耦合性能模拟方法,采用该数值方法对TRISO燃料球的堆内性能进行了模拟,并与程序BISON计算结果进行对比,验证了该模拟方法的有效性。同时,对FCM燃料的辐照-热-力耦合性能进行了数值模拟,结果表明,FCM燃料芯块的温度分布与力学分布均表现出较强的非均匀性,裂变气体释放对于FCM燃料的性能影响较大,FCM燃料内包覆层的环向应力与FCM芯块温度变化对于功率瞬态并不敏感。     

等效天然铀燃料中子学性能分析

采用10种回收铀（RU）和贫铀（DU）成分情形,根据等效天然铀（NUE）燃料混合比计算程序ALPHA算得配成NUE燃料的混合比。以标准CANDU 6栅元结构为载体,采用WIMS程序,通过比较NUE燃料与天然铀（NU）燃料的中子学性能参数,以及NUE燃料入堆示范验证试验中实际入堆的燃料信息,对NUE燃料与NU燃料的中子学性能等效性进行了论证分析。研究表明,与NU燃料相比,各种情形下NUE燃料在无限增殖系数、卸料燃耗、冷却剂空泡反应性以及燃料温度效应等中子学性能参数上吻合较好,NUE燃料与NU燃料具备较好的中子学等效性,可应用于重水堆核电站,实现回收铀的有效利用。     

气冷微堆燃料设计的中子学特性影响研究

为分析气冷微堆燃料设计的中子学特性影响,基于方形燃料组件模型,利用蒙特卡罗程序RMC研究了TRISO颗粒、燃料芯块在燃料设计中的主要参量对组件中子学特性的影响。研究结果表明,燃料颗粒体积占比和包覆层厚度不变时,组件寿期随燃料核芯直径的增大先显著增大,而后趋于平稳;燃料颗粒体积占比和燃料核芯直径不变时,组件寿期随包覆层厚度的增大而减小;燃料装载量不变时,芯块直径增大,组件寿期显著增大,而芯块高度影响较小;无燃料区厚度的增加对组件中子学特性有明显的负面影响,基体材料密度、基体杂质硼当量对组件中子学特性的影响较小。研究结果将为后续气冷微堆包覆颗粒弥散燃料的设计优化提供指导。     

FCM燃料压水堆弥散可燃毒物中子学分析

为将全陶瓷微胶囊封装(FCM)燃料应用于小型压水堆,对FCM燃料组件开展了可燃毒物中子学设计与分析。通过寿期初引入负反应性、寿期内消耗速率和寿期末残留3个方面,对弥散在SiC基体中的弥散型可燃毒物Gd_2O_3、Er_2O_3、Sm_2O_3、Eu_2O_3、Dy_2O_3及HfO_2进行评价。FCM燃料中TRISO颗粒核芯直径达800μm,燃料颗粒自屏效应强烈,在RMC程序中引入随机介质计算功能,对FCM燃料进行随机几何建模,保证了反应性计算精度。分析表明:Er_2O_3可作为FCM燃料堆芯的候选可燃毒物,Gd_2O_3和Eu_2O_3需结合堆芯开展进一步研究,Sm_2O_3、Dy_2O_3及HfO_2的反应性惩罚过大,不适合作为FCM燃料可燃毒物。     

混合堆增殖钍基燃料组件中子学分析

采用压水堆17×17燃料组件模型,用燃料组件参数计算程序DRAGON分别对混合堆增殖钍燃料组件和全铀组件的中子学特性进行了研究,分析组件的燃料温度系数、慢化剂温度系数及其与燃耗的关系。计算结果表明,混合堆增殖钍燃料组件和全铀组件的中子特性相似,但钍燃料组件中的乏燃料组件中的次锕系核素(MA)的含量明显减少。     

压水堆燃料组件轴向格架的中子学影响研究

基于先进压水堆燃料管理软件Bamboo-C,分别提出了轴向格架非均匀建模方法和均匀建模方法。采用轴向格架的2种不同建模方法,对福清核电厂M310堆型的燃料组件进行建模分析,通过与堆芯实测数据进行对比,检验2种不同建模方法对临界硼浓度、轴向功率分布及轴向功率偏移的影响。数值结果表明,压水堆燃料组件轴向非均匀建模方法能够显著提高堆芯关键物理参数的计算精度。      

气冷快堆燃料组件设计及中子学特性分析

气冷快堆兼具高温气冷堆的经济性和快堆的可持续性等优点,在四代堆型中具有独特的技术优势。为了适应气冷快堆高温、高中子通量的堆芯环境,本文基于耐事故燃料模型,提出了一种块状气冷快堆燃料组件设计方案,并对该组件中铀钚混合燃料中的钚含量、冷却孔道的直径及数量、栅距比、包壳及组件盒厚度等物理参数对中子学特性的影响规律开展了敏感性分析研究。分析结果表明：在研究的6个参数中,钚含量和栅距比对组件的中子学特性影响最大,冷却孔道数量主要影响组件内的功率分布,其余参数对组件中子学特性几乎无影响。最后针对块状燃料组件低冷却剂份额的特点,利用单通道模型进行组件内的温度分布计算,给出了热工限值对组件参数的要求。     

Neutronics physics analysis of a large fluoride-salt-cooled solid-fuel fast reactor with Th-based fuel

Fast reactors based on thorium fuel have enhanced inherent safety. Fluoride salt performs well as a coolant in high-temperature nuclear systems. In this paper,we present a reference core for a large fluoride-salt-cooled solid-fuel fast reactor(LSFR) using thorium–uranium fuel cycle. Neutronics physics of the LSFR reference core is investigated with 2D and 3D in-core fuel management strategy. The design parameters analyzed include the fuel volume fraction, power density level and continuous removal of fission products with 3D fuel shuffling that obtains better equilibrium core performance than 2D shuffling. A self-sustained core is achieved for all cases,and the core of 60% fuel volume fraction at 50 MW/m~3 power density is of the best breeding performance(average breeding ratio 1.134). The LSFR core based on thorium fuel is advantageous in its high discharge burn-up of 20–30% fissions per initial heavy metal atom, small reactivity swing over the whole lifetime(to simplify the reactivity control system), the negative reactivity temperature coefficient(intrinsically safe for all cases) and accepted cladding peak radiation damage. The LSFR reactor is a good alternative option for the deployment of a self-sustained thorium-based nuclear system.     

Neutronics physics analysis of a large fluoride-salt-cooled solid-fuel fast reactor with Th-based fuel

A synchrotron radiation source called TURKAY is proposed as a sub-project of the Turkish Accelerator Center Project. The storage ring of TURKAY is a low emittance synchrotron and the radiation ranges between 0.01 and 60 keV can be generated from the insertion devices and bending magnets placed on it. The injector system of the facility will mainly consist of a 150 MeV linac and full energy booster. In this study, we present design considerations and beam dynamics studies of the pre-injector linac and booster ring for TURKAY.     

Dynamic analysis of a pressurized water reactor

An apparent malfunction in a pressurized water reactor system has been investigated using fluctuation analysis. Both frequency-domain and time-domain analyses have been used and the results obtained by the two methods have been compared. The recording was performed by a relatively simple, cheap system giving high recording precision and the analyses were performed on an IBM 370 digital computer. It is shown that, while considerable information can be derived from frequency-domain analyses, a misinterpretation can occur in some cases. Time-domain correlation, as normally performed, was not very informative. However, time-domain correlation on bandwidth-limited time-series proved to be very valuable and could remove the misinterpretation of the frequency-domain analyses. The bandwidth limitation was performed by digital filters.     

Reactivity control technique for a pressurized water reactor with an inventive TRISO fuel particle composition

In this research paper a reactivity control technique has been suggested for the conceptual design of a compact sized pressurized water reactor (PWR) core with an inventive tristructural-isotropic (TRISO) fuel particle composition. This conceptual design is a light water cooled and moderated reactor which utilizes TRISO fuel particles in PWR technology. The use of TRISO fuel in PWR technology improves integrity of the design due to its fission fragments retention ability. The fuel provides first retention barrier within fuel itself against the release of fission fragments that makes this design concept safer and environment friendly. The suggested TRISO fuel particle composition has a small amount of Pu-240 with 2.0 w/o in the place of U-238 which acts as reactivity suppressor. Reactor codes WIMS-D/4 and CITATION have been used for simulation and core design modeling. Results reveals that the amount of excess reactivity can be reduced significantly by using a small amount of Pu-240 in TRISO fuel which in turns reduces the number of gadolinia rods in the core required for excess reactivity control and completely eliminates the requirement of soluble boron system. Therefore the effective and optimal use of reactivity suppressor and burnable poison suppresses and flattens the core excess reactivity throughout the core life and hence number of control rods can be reduced without compromising on the shutdown margin.     

Theoretical analysis of long-lived radioactive waste in pressurized water reactor

Background The accelerator-driven subcritical transmuta-tion system(ADS)is an advanced technology for the harmless disposal of nuclear waste.A theoretical analy...     

A class-based search for the in-core fuel management optimization of a pressurized water reactor

The In-Core Fuel Management Optimization (ICFMO) is a prominent problem in nuclear engineering, with high complexity and studied for more than 40 years. Besides manual optimization and knowledge-based methods, optimization metaheuristics such as Genetic Algorithms, Ant Colony Optimization and Particle Swarm Optimization have yielded outstanding results for the ICFMO. In the present article, the Class-Based Search (CBS) is presented for application to the ICFMO. It is a novel metaheuristic approach that performs the search based on the main nuclear characteristics of the fuel assemblies, such as reactivity. The CBS is then compared to the one of the state-of-art algorithms applied to the ICFMO, the Particle Swarm Optimization. Experiments were performed for the optimization of Angra 1 Nuclear Power Plant, located at the Southeast of Brazil. The CBS presented noticeable performance, providing Loading Patterns that yield a higher average of Effective Full Power Days in the simulation of Angra 1 NPP operation, according to our methodology.     

Control rod drop hydrodynamic analysis for a pressurized water reactor

During operation of nuclear power reactors, reactivity initiated accidents can take place such as a control rod drop. If this occurs, the reactivity increases significantly and leads to an enhancement in power, fuel temperature and damage of reactor eventually. Exact assessment of these accidents depends on the hydrodynamic information. In this research, it is tried to simulate the unsteady flow field around the control rod for a pressurized water reactor power plant. In order to simulate the flow field around the control rod inside the guide tube, averaged Navier–Stokes equations accompanied by the layering dynamic mesh strategy have been used. The information exchange between the two computational stationary and moving grids, the computational grid around the control rod and the grid next to the guide tube, has been taken place through the interface. It was concluded that the time duration of control rod to reach the bottom of the core depends on the leakage. It was also observed that the velocity and acceleration of the control rod would be reduced by decreasing leakage flow rate and in certain leakages, the acceleration of the control rod approaches zero due to equilibrium conditions. During this research, a correlation based on the achieved data was proposed which would provide useful information on the relation between the leakage and the time for control rod to reach the bottom of the core.     

Thermal hydraulic analysis for grid supported pressurized water reactor cores

This paper presents the methodology and results for thermal hydraulic analysis of grid supported pressurized water reactor cores using U(45% wt)-ZrH_1.6 hydride fuel in square arrays. The same methodology is applied to the design of UO₂ oxide fueled cores to provide a fair comparison of the achievable power between the two fuel types. Steady-state and transient design limits are considered. Steady-state limits include: fuel bundle pressure drop, departure from nucleate boiling ratio, fuel temperature (average for UO₂ and centerline/peak for U-ZrH_1.6), and fuel rod vibrations and wear. Transient limits are derived from consideration of the loss of flow and loss of coolant accidents, and an overpower transient.In general, the thermal hydraulic performance of U-ZrH_1.6 and UO₂ fuels is very similar. Slight power differences exist between the two fuel types for designs limited by rod vibrations and wear, because these limits are fuel dependent. Large power increases are achievable for both fuels when compared to the reference core power output of 3800 MW_th. In general, these higher power designs have smaller rod diameters and larger pitch-to-diameter ratios than the reference core geometry. If the pressure drop across new core designs is limited to the pressure drop across the reference core, power increases of ∼400 MW_th may be realized. If the primary coolant pumps and core internals could be designed to accommodate a core pressure drop equal to twice the reference core pressure drop, power increases of ∼1000 MW_th may be feasible.     

Impacts and fluid-structure interaction in pressurized water reactor safety analysis

A numerical method is described for the analysis of coupled three-dimensional fluid-structure motion with impacts between structural parts at rigid or flexible supports with small clearances. The method is used for the analysis of the blowdown loadings and the response of internal structures in the vessel of a pressurized water reactor (PWR) in the hypothetical event of a sudden break of a coolant inlet pipe. The method is a generalization of the existing code FLUX which simulates the three-dimensional fluid-structure motion by means of an implicit time integration scheme. The additional supports with clearances are taken into account by applying support forces to the freely moving fluid-structure system. The forces are determined such that the kinematic constraints are enforced at each time step. Numerically, these forces are determined efficiently using a precomputed influence matrix which defines the dynamic displacement per time step at each support due to a unit force at each other support. According to the actually “active” supports the relevant influence matrix in constructed. Energy is conserved for rigid supports and for supports which are so flexible that the impact time is large in comparison to the time steps. Treatment of plastic supports is possible.An application of the new method is demonstrated by analysis of the core barrel motion in a PWR with and without impacts at the lower core barrel edge and at the upper flange. The results show the large effects of such impacts in changing the global motions. Large local impact forces and accelerations appear. The interaction with the fluid reduces these loads. By proper design of the supports the resultant stresses can be minimized. Thus the method can be used to demonstrate and enlarge nuclear reactor safety.     

Neutronics assessment of the use of thorium fuels in current pressurized water reactors

The use of thorium fuel in current PWRs in a once-through fuel cycle is an attractive option due to potential advantages such as high conversion ratio and low minor actinide generation. The current neutronics assessments indicate that the thorium fuel cycle could supplement the current uranium–plutonium fuel cycle to improve operational performance and spent fuel consideration in current PWRs without core and subassembly modifications. Neutronics safety parameters in the PWR cores with the thorium fuels are within the range of current PWRs.The PWR cores with thorium fuels have significantly higher conversion ratios which could enable efficient fuel utilization. Further, it is shown that the use of thorium as a fertile material can reduce minor actinide generation and the radio-toxicity of spent fuels. In considerations related to proliferation resistance, the results of the current analyses show no significant difference between the studied thorium fuels and the standard oxide fuel for the assumed characteristics and burnup levels.