Computational Methods and Tools for Neutronics Calculations for Molten-Salt Nuclear Reactors

熔盐核反应堆（MSR）是第四代国际论坛（GIF）正在研究的六种第四代先进核反应堆概念之一。加拿大是GIF框架协议的签署国，至少有一家加拿大公司正在开发MSR概念，并正在向加拿大核安全委员会申请许可证。 MSR使用铀和/或钍的熔融（液体）盐作为燃料和冷却剂。 MSR堆芯没有加压，熔盐具有高沸点，这使得放射性释放到环境中几乎是不可能的;即使在事故发生后。 作为轻水反应堆乏燃料（世界上最常见的动力反应堆燃料）中超铀元素的燃烧器，MSR很有前途，因此有可能减少此类反应堆的废物量。 它们可以提供高转换/增殖率，从而提高核燃料利用率。在MSR中，液体盐连续地循环进出堆芯，这通过向由中子学代码求解的平衡方程添加额外的“漂移”项而对此类反应堆的中子学模拟提出了独特的挑战。 因此，用于固体燃料反应堆的中子学代码不适用于液体燃料反应堆，如MSR。 虽然固体燃料反应堆的中子学计算方法和工具已经成熟，但液体燃料反应堆的计算方法和工具仍处于起步阶段，并构成了一个活跃的研究领域。到目前为止，加拿大还没有独立的MSR中子学计算工具。拟议的研究计划的重点是开发高保真，可靠的中子学计算方法和工具的MSR;工具，可以适当地考虑燃料对流的三维静态和时间相关的问题。 这种方法需要中子学和流体动力学计算的耦合。 研究将侧重于中子学方法及其与流体动力学计算的耦合，而燃料流的实际模拟将使用适用于计算流体动力学的现有软件包（例如OpenFOAM）进行。该研究将为研究人员，工程师和监管机构提供一个独立的计算工具，用于第四代MSR的设计和安全分析。 这种工具将使人们能够准确和可靠地分析现有的和新兴的MSR设计的中子特性。 该研究计划将支持五名高素质人员的培训。在更广泛的范围内，这项研究将有助于MSR的发展，MSR承诺以最小的风险和放射性废物进行无二氧化碳发电。 它还将支持参与MSR研究的加拿大核技术公司，使它们处于世界核技术发展的最前沿。*

Molten-Salt Nuclear Reactors (MSRs) are one of the six Generation-IV advanced nuclear-reactor concepts being pursued by the Generation IV International Forum (GIF). Canada is a signatory of the GIF Framework Agreement and at least one Canadian company is developing an MSR concept and is seeking licensing from the Canadian Nuclear Safety Commission. MSRs use molten (liquid) salts of uranium and/or thorium which act as both fuel and coolant. The MSR core is not pressurized and molten salts have high boiling points, which make the release of radioactivity into the environment near-impossible; even following an accident. MSRs are promising as burners of transuranic elements from spent Light-Water-Reactor fuel (the most common type of power-reactor fuel in the world) thus having the potential to reduce the amount of waste from such reactors. They can provide high conversion/breeding ratios thus increasing the nuclear-fuel utilization.****In MSRs, the liquid salt circulates continuously in and out of the core, which poses unique challenges for the neutronics simulations of such reactors by adding additional, “drift”, terms to the balance equations solved by neutronics codes. Consequently, neutronics codes in use for solid-fuel reactors are not adequate for liquid-fuel reactors, such as the MSRs. While neutronics computational methods and tools for solid-fuel reactors are mature, computational methods and tools for liquid-fuel reactors are in their infancy and constitute an active area of investigation. To date, no stand-alone neutronics computational tool for MSRs exists in Canada.****The proposed research program is centered on the development of high-fidelity, reliable neutronics computational methods and tools for MSRs; tools that can properly account for fuel convection for three-dimensional static and time-dependent problems. Such methods require the coupling of neutronics and fluid-dynamics calculations. The research will focus on the neutronics methods and their coupling with fluid-dynamics computations, while the actual simulation of the fuel flow will be performed using an existing software package applicable to Computational Fluid Dynamics (e.g. OpenFOAM).****The research will produce a stand-alone computational tool for researchers, engineers and regulators to use for the design and safety analysis of Generation-IV MSRs. Such a tool will make it possible to accurately and reliably analyze the neutronic characteristics of existing and emerging MSR designs. The training of five highly-qualified personnel will be supported by the research program. On a wider scale, the research will contribute to the development of MSRs, which promise CO2-free electricity production with minimal risk and radioactive waste. It will also support Canadian nuclear-technology companies involved in MSR research, placing them at the forefront of nuclear-technology development in the world.***