超临界水堆堆芯新型燃料组件设计分析

超临界水堆(SCWR)作为六种第四代未来堆型中唯一的水冷反应堆,具有良好的经济性与技术延续性.本文采用最新开发的热工-物理耦合计算程序对超临界水堆方形燃料组件进行稳态热工与中子物理耦合分析,提出一种新型的超临界水堆堆芯燃料组件设计.现有单排组件设计与新型双排燃料组件设计的计算结果表明,双排组件具有功率径向分布均匀,包壳...     

超临界水冷堆双排燃料组件子通道分析

研究基于Cobra-IV程序,开发了适用于超临界水冷堆燃料组件分析的子通道程序.针对超临界水冷堆慢谱双排组件,进行了稳态计算,获取了相关组件热工水力参数.在此基础上,针对单一通道进行了瞬态计算,分析了燃料棒线功率变化和冷却剂流量变化条件下,超临界水冷堆燃料组件的流动和传热的动态响应,为超临界水冷堆组件的优化设计提供了参考.     

超临界水堆燃料组件内的排列研究

超临界水冷反应堆(SCWR)是第四代核能系统国际论坛(GIF)推荐的六种堆型中唯一的轻水堆型.SCWR和现有的轻水堆相比,具有热效率高,系统设备大大简化的优点.世界范围内的研究纷纷展开,其中燃料组件的设计优化及堆芯布置是一个重要的研究方向.本文分析比较了当前比较流行的几种燃料组件设计,在采用同一富集度燃料且不含可燃毒物的情况下,利用MCNP程序对这几种组件的当地功率峰值因子进行了计算,发现其离设计目标还有一段距离.本文分析了影响当地功率峰值因子的若干因素,发现对于正方形组件,在均匀慢化、降低当地功率峰值因子的同时也使得组件整体上慢化不足,表现为倍增因子降低,这主要与燃料棒的排列方式有关.通过对比分析发现,相对于正方形排列,改进过的六角形排列更容易解决充分慢化和均匀慢化之间的矛盾,实现组件设计的优化.     

Simulation of neutron-physical processes in the surface layer of a fuel kernel

A large change in the structure, density, and chemical and phase composition occurs in nuclear fuel with deep burnup, and an edge zone is formed. Simulation of the formation of an edge zone in a fuel kernel of thermal reactors will make it possible to suggest ways to decrease its influence on the characteristics of a fuel element. In the present work, the neutron-physical processes occurring in the peripheral layer of a fuel kernel are simulated. The distribution of nuclear reaction rates along the radius of a fuel pellet is calculated using the SCALE-4.3, MCNP-4B, and UNK computer programs. The radial dependence of the local breeding ratio is calculated. It is shown that for fresh fuel BR > 1 for fissile nuclei in a 100 μm thick layer, while the initial BR averaged over a pellet is no more than 0.5. The volume energy distribution in a 100 μm thick peripheral layer is 30% higher than the average value over a pellet. A combined pellet, where the central part possesses the standard enrichment (4–5% ²³⁵U) and the peripheral layer contains less than 0.7% ²³⁵U, is proposed to decrease the influence of the edge zone on the properties of fuel. __________ Translated from Atomnaya énergiya, Vol. 104, No. 6, pp. 353–358, June, 2008.     

新一代压水堆燃料组件技术特点及其在华专利申请

以美国西屋电气公司的Next Generation Fuel燃料组件技术特点为线索,收集了美国西屋电气公司在中国燃料组件技术方面的专利申请和专利文献,从中筛选出与NGF燃料组件技术特点符合的专利申请和专利文献,对其技术方案进行了深入剖析,从中了解西屋新一代压水堆燃料组件技术的发展趋势。     

Particulars of neutron-physical calculations of sodium-cooled fast reactors with mixed oxide fuel

The neutron-physical characteristics of reactor systems with a fast spectrum, sodium coolant, and uraniumplutonium fuel load have been analyzed on the basis of computational studies of the BFS-62-3A critical assembly and a BN-600 hybrid core with mixed oxide fuel. The large differences in the spectra in an expanded thermal range to 1 keV for the central and peripheral regions with uranium oxide and mixed oxide fuel show that spatially differentiated fission and absorption cross sections must be used for the main uranium and plutonium isotopes in neutron-physical calculations.     

Validation of a new continuous Monte Carlo burnup code using a Mox fuel assembly

The reactivity of nuclear fuel decreases with irradiation (or burnup) due to the transformation of heavy nuclides and the formation of fission products. Burnup credit studies aim at accounting for fuel irradiation in criticality studies of the nuclear fuel cycle (transport, storage, etc…). The principal objective of this study is to evaluate the potential capabilities of a newly developed burnup code called “BUCAL1”. BUCAL1 differs in comparison with other burnup codes as it does not use the calculated neutron flux as input to other computer codes to generate the nuclide inventory for the next time step. Instead, BUCAL1 directly uses the neutron reaction tally information generated by MCNP for each nuclide of interest to determine the new nuclides inventory. This allows the full capabilities of MCNP to be incorporated into the calculation and a more accurate and robust analysis to be performed. Validation of BUCAL1 was processed by code-to-code comparisons using predictions of several codes from the NEA/OCED. Infinite multiplication factors (k_∞) and important fission product and actinide concentrations were compared for a MOX core benchmark exercise. Results of calculations are analysed and discussed.     

Computational and experimental investigations of the neutron-physical characteristics of the GT-MHR core

The main results of computational and experimental investigations of the neutron-physical characteristics of the GT-MHR high-temperature gas-cooled reactor are presented: analysis of the possible reasons for the uncertainties in the calculations of the neutron-physical characteristics of a core with plutonium fuel using different computer programs, computational and experimental investigations taking account of the experiments on the Astra critical assembly simulating the ring-shaped GT-MHR core, and the effect of using different nuclear data on the temperature coefficient of reactivity computed with high-accuracy computational codes such as MONTEBURNS-MCNP 5-ORIGEN 2.1. __________ Translated from Atomnaya énergiya, Vol. 102, No. 1, pp. 63–68, January, 2007.     

CAP1400燃料组件用新锆合金研究

在Zr-Sn-Nb系合金的基础上添加微量合金元素Ge和Si等,采用真空电弧熔炼,制备了多种新锆合金。使用透射电子显微镜(Transmission electron microscope,TEM)对合金基体进行显微组织分析,分别通过堆外高压釜腐蚀试验、定氢分析仪和万能材料试验机对合金的腐蚀、吸氢和拉伸性能进行评估。结果表明,常规工艺处理后,SZA-4和SZA-6合金均发生了完全再结晶,第二相细小、均匀弥散分布在晶粒内和晶界上;SZA-4和SZA-6合金在三种水化学条件下均具有优良的耐腐蚀性能,SZA-6合金的耐腐蚀性能优于参考合金,SZA-4合金的耐腐蚀性能略优于SZA-6合金;SZA-6合金的吸氢性能略优于SZA-4合金;两种合金的拉伸性能满足设计要求。基于SZA-4和SZA-6合金优良的耐腐蚀、吸氢和力学性能,未来将有望用于CAP1400自主化燃料组件。     

A new method to predict Grid-To-Rod Fretting in a PWR fuel assembly inlet region

Grid-To-Rod Fretting (GTRF) is one of the main causes of leaking fuel in a Pressurized Water Reactor (PWR). GTRF is caused by grid-to-rod gap, secondary flow, and axial/lateral turbulence caused pressure fluctuations within the fuel assembly, which produces rod vibration and wear. The cross flow and vortex shedding phenomenon produce low frequency vibration forces on fuel rods. In some plants, leaking fuel has been detected at the fuel inlet region of fuel assembly designs that do not have Protective Grid (P-grid) which, in addition to providing debris protection, also provides lateral stability against vibration. In order to understand the root cause of the fuel leaks, a thorough investigation of the flow field at the fuel inlet region is required. Leaking fuel has also been detected in the fuel inlet region in transition cores. In the transitional core arrangement, there are different fuel assembly designs next to each other. Due to the structure difference, there will be cross flow between fuel assemblies, which may be the initiating factor for fuel leaks.A method based on Computational Fluid Dynamics (CFD) has been developed in Westinghouse to predict the GTRF in the fuel inlet region. The fuel inlet region consists of the lower core plate, the bottom nozzle, the fuel rods, the thimble rods, the P-grid, and the bottom grid. This study employed CFD to investigate the unsteady forces on the fuel rods under typical reactor in-core conditions. Two fuel assembly (FA) inlet regions with and without the P-grid were simulated. The time history of the unsteady force components on fuel rods was recorded. Fast Fourier Transform (FFT) analyses were carried out for the force history. Compared to the data from operating plants, the new method predicted synchronized excitation forces on the rods that leaked in real operation. The CFD results also demonstrated the advantage of using the P-grid. GTRF at the fuel inlet region can be significantly reduced when the P-grid is used in Westinghouse fuel assembly designs.     

Thermal-hydraulic performance analysis of a subchannel with square and triangle fuel rod arrangements using the entropy generation approach

The present paper discusses entropy generation in fully developed turbulent flows through a subchannel,arranged in square and triangle arrays. Entropy generation is due to contribution of both heat transfer and pressure drop. Our main objective is to study the effect of key parameters such as spacer grid, fuel rod power distribution,Reynolds number Re, dimensionless heat power ω, lengthto-fuel-diameter ratio λ, and pitch-to-diameter ratio ξ on subchannel entropy generation. The analysis explicitly shows the contribution of heat transfer and pressure drop to the total entropy generation. An analytical formulation is introduced to total entropy generation for situations with uniform and sinusoidal rod power distribution. It is concluded that power distribution affects entropy generation.A smoother power profile leads to less entropy generation.The entropy generation of square rod array bundles is more efficient than that of triangular rod arrays, and spacer grids generate more entropy.     

超临界水冷混合堆快谱区组件物理——热工分析

提出了超临界水冷混合堆快谱区多层燃料组件设计方案.应用MCNP程序为该组件建立计算模型,并进行了相应的物理计算;同时运用子通道分析程序STAFAS对多层燃料组件子通道进行了初步的稳态热工分析.计算结果表明:超临界水冷混合堆快谱区多层燃料组件燃料转换比超过1.0,并且获得负的冷却剂空泡反应性系数;燃料包壳表面最高温度约为595℃,低于设计准则规定的上限值,同时组件各子通道出口冷却剂温度均匀性较好.通过对燃料棒径敏感性分析可知,较大棒径组件燃料转换比较大,但也会导致热通道包壳表面温度峰值升高.     

采用哑棒修复燃料组件的核特性分析方法及其工程应用

在核电厂正常运行过程中,由于一回路杂物的存在或燃料操作失误,出现了少量燃料棒损伤的情况,通过采用哑棒替换损伤燃料棒可修复损伤燃料组件并回堆使用,可避免降低核电厂运行经济性。本文通过模拟采用不锈钢和锆合金哑棒替换破损燃料棒对燃料组件进行修复,分析修复后燃料组件中子学特性及修复燃料组件对堆芯运行核特性参数的影响机理,评估采用哑棒修复燃料组件并回堆使用对堆芯运行安全的影响,对采用哑棒修复燃料组件建立了完整的核设计分析方法和流程。该方法对采用哑棒修复燃料组件的核设计分析具有广泛的适用性,对采用修复燃料组件的堆芯换料设计具有实际的指导意义。该分析方法和流程的建立在国内反应堆物理分析领域尚属首次,目前该技术已应用于恰希玛一期核电厂堆芯换料设计的工程实践。     

Thermal-hydraulic and safety aspects of UO2 fuel for low-power reactors

The low enriched uranium UO₂ (about 19.75% U²³⁵) fuel is proposed to be used in low-power research reactors. The thermal-hydraulic and dynamic characteristics are examined in this paper. The fuel behaves similarly to the actual highly enriched uranium fuel in the normal daily operation for both Miniature Neutron Source Reactors and SLOWPOKEs, the cladding temperature reaching about 60 °C. During the simulation of a design basis accident the reactor power peak and temperatures are found to be higher than in the case of the highly enriched uranium fuel for MNSRs, the power peak touching 135 kW, and the cladding temperature reaching over 110 °C in this case. Nevertheless the fuel can be safely used in these reactors.