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Neutronics and Safety Studies on a Research Reactor Concept for an Advanced Neutron Source

机译:先进中子源研究堆概念的中子学和安全性研究

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This paper presents preliminary neutronics and thermal hydraulics safety analysis results for a low-enriched uranium (LEU) fueled research reactor concept being studied at the National Institute of Standards and Technology (NIST). The main goal of this research reactor is to provide advanced sources for neutron scattering experiments with a particular emphasis given to high intensity cold neutron sources (CNSs). A tank-in-pool type reactor with an innovative horizontally split compact core was developed in order to maximize the yield of the thermal flux trap in the reflector area. The reactor concept considered a 20 MW thermal power and a 30-day operating cycle. For non-proliferation purposes, a LEU fuel (U_(3)Si_(2)-Al) with 19.75 wt% enrichment was used. The core performance characteristics of an equilibrium cycle with several representative burnup states—including startup and end of cycle—were obtained using the Monte Carlo–based code MCNP6. The estimated maximum perturbed thermal flux of the core is ~5.0 × 10~(14) n/cm~(2)-s. The calculated brightness of the CNS demonstrates an average gain factor of ~4 compared to the current source operated at the existing NIST reactor. Sufficient reactivity control worth and shutdown margins were provided by hafnium control elements. Reactivity coefficients were evaluated to ensure negative feedback. Thermal hydraulics safety studies of the reactor were performed using the multi-channel safety analysis code PARET. Steady-state analysis shows that the peak cladding temperature and minimum critical heat flux ratio are less than design limits with sufficient safety margins. Detailed transient analyses for a couple of hypothetical design-basis accidents show that no fuel damage or cladding failure would occur with the protection of reactor scrams. All these study results suggest this new research reactor concept offers a demonstrable potential to greatly expand the cold neutron capability with a 20 MW power and certified LEU fuels.
机译:本文介绍了美国国家标准技术研究院(NIST)正在研究的低浓铀燃料研究堆概念的初步中子学和热力学安全性分析结果。该研究堆的主要目的是为中子散射实验提供先进的来源,特别着重于高强度冷中子源(CNS)。为了使反射器区域的热通量阱的产量最大化,开发了具有创新的水平分裂紧凑型堆芯的池中式反应器。反应堆概念考虑了20兆瓦的火力和30天的运行周期。为了不扩散,使用了浓缩量为19.75 wt%的LEU燃料(U_(3)Si_(2)-Al)。使用基于蒙特卡洛的代码MCNP6获得了具有几个代表性燃耗状态(包括循环的开始和结束)的平衡循环的核心性能特征。磁芯的估计最大摄动热通量为〜5.0×10〜(14)n / cm〜(2)-s。与在现有NIST反应堆上运行的电流源相比,CNS的计算亮度显示出〜4的平均增益系数。 ha控制元素提供了足够的反应性控制价值和关闭裕度。评价反应性系数以确保负反馈。使用多通道安全分析代码PARET对反应堆进行了热力学安全性研究。稳态分析表明,包层的峰值温度和最小临界热通量比小于设计极限,并具有足够的安全裕度。对一些假设的设计基准事故的详细瞬态分析表明,在保护反应堆后,不会发生燃料损坏或覆层故障。所有这些研究结果表明,这种新的研究堆概念具有可证明的潜力,可通过20兆瓦的功率和经过认证的LEU燃料大大扩展冷中子的能力。

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