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

2011年3月，福岛第一核电站发生地震和海啸，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壁，导致放射性物质失去安全壳。对环境进行了重大的辐射释放，超过10万人从该地区疏散。福岛第一核电站的灾难突显出，在船上保存熔融燃料是至关重要的。在严重的事故条件下，比如在日本发生的情况，熔化的燃料和反应堆内部的其他部件会在反应堆容器的底部形成一个熔化的液态铯熔池。自然对流是由于熔体内部的体积加热而形成的。从裂变过程中产生的体积衰变热必须在容器壁上充分去除，以确保熔融材料在容器内保持不变。加拿大重铀(CANDU)反应堆容器(例如达灵顿核电站和布鲁斯电力核电站)被一个装有液态水的圆柱形屏蔽罐包围，在发生严重事故时，该屏蔽罐提供外部容器冷却。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来确保保持反应堆容器的完整性。在核工业中，积分或集中参数程序主要使用经验换热关联式来预测从熔化的真皮熔池到容器壁的热传递，以便评估反应堆容器壁外部的冷却是否足以维持其完整性。拟议的研究将使用计算流体动力学(CFD)来获得关于自然对流流体流动的详细信息，以及局部(空间)和时变的壁面换热速率。这些将被用来评估简化的积分代码中的基本假设的适当性，这些代码目前用于核工业中的超设计基础分析。

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.