CFD analysis of natural convection heat transfer in volumetrically heated corium melt for CANDU reactors

2011年3月，福岛第一核电站遭受地震和海啸，导致1号、2号和3号机组发生严重事故。 在这次事故中，熔化的燃料冲破了反应堆压力容器的壁，导致放射性物质的泄漏。 有大量的放射性物质释放到环境中，超过10万人从该地区撤离。 福岛第一核电站的灾难突出表明，容器内熔融燃料的保留至关重要。在严重的事故条件下，例如在日本发生的事故，熔化的燃料和其他内部反应堆部件在反应堆容器底部形成熔化的堆芯熔化物的液体池。由于熔体内的体积加热而产生自然对流。 来自裂变过程的体积衰变热必须在容器壁处充分地移除，以确保熔融材料的容器内保留。 加拿大氘-铀（CANDU）反应堆容器（例如在达林顿核发电站和布鲁斯电力核发电站）被包含液态水的圆柱形屏蔽罐包围，液态水用于在严重事故的情况下提供外部容器冷却。 数值分析用于评估严重事故的后果，并确定是否有足够的安全裕度，以确保反应堆容器的完整性得到保持。 在核工业中，积分或集总参数代码主要使用经验传热相关性来预测从熔融堆芯熔化物池到容器壁的传热，以便评估反应堆容器壁的外部的冷却是否足以保持其完整性。 拟议的研究将使用计算流体动力学（CFD）来获得有关自然对流流体流动以及局部（空间）和随时间变化的壁传热率的详细信息。 这些将用于评估简化积分代码中基本假设的适当性，该代码目前用于核工业中的超设计基准分析。

In March 2011, the Fukushima Daiichi nuclear power plant was subjected to an earthquake and tsunami resulting in a severe accident to Units 1, 2, and 3. During this accident molten fuel breached the reactor pressure vessel wall which resulted in a loss of containment of radioactive material. There was significant radiological release to the environment and over 100,000 people evacuated from the area. The Fukushima Daiichi disaster highlighted that in-vessel retention of molten fuel is of paramount importance. Under severe accident conditions, such as what occurred in Japan, the molten fuel and other internal reactor components form a liquid pool of corium melt at the bottom of the reactor vessel. Natural convection flows develop due to the volumetric heating within the melt. The volumetric decay heat from the fission process must be adequately removed at the vessel wall in order to ensure in-vessel retention of the molten material. Canada deuterium-uranium (CANDU) reactor vessels (for example at Darlington Nuclear Generating Station and Bruce Power Nuclear Generating Station) are surrounded by a cylindrical shield tank containing liquid water which acts to provide external vessel cooling in the event of a severe accident. Numerical analysis is used to assess the consequences of severe accidents and to determine if an adequate safety margin is in place to ensure that the reactor vessel integrity is maintained. In the nuclear industry, integral or lumped parameter codes primarily use empirical heat transfer correlations to predict the heat transfer from the molten corium pool to the vessel wall in order to assess if cooling of the exterior of the reactor vessel wall is sufficient to maintain its integrity. The proposed research will use computational fluid dynamics (CFD) to obtain detailed information on the natural convection fluid flow, and local (spatial) and time-varying wall heat transfer rates. These will be used to assess the appropriateness of underlying assumptions in the simplified integral codes which are currently used for beyond design basis analysis in the nuclear industry.