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

CANDU 反应器容积式加热真皮熔体中自然对流换热的 CFD 分析

• 批准号: 508442-2016
• 负责人: Lightstone, Marilyn
• 金额: $ 1.91万
• 依托单位: McMaster University
• 依托单位国家: 加拿大
• 项目类别: Collaborative Research and Development Grants
• 财政年份: 2020
• 资助国家: 加拿大
• 起止时间: 2020-01-01 至 2021-12-31
• 项目状态: 已结题
• 来源: https://www.nserc-crsng.gc.ca/ase-oro/Details-Detailles_eng.asp?id=696832
• 关键词: CFD; analysis; natural; convection; heat

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.

2011年3月，福岛第一核电站遭遇地震和海啸，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，导致了放射性物质的泄漏。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，在容器内保留熔融燃料是至关重要的。在严重的事故条件下，比如在日本发生的事故，熔化的燃料和其他反应堆内部组件在反应堆容器底部形成了堆芯熔体的液体池。由于熔体内部的体积加热，自然对流流动得以发展。裂变过程产生的体积衰减热必须在容器壁处充分去除，以确保熔融材料在容器内的保留。加拿大的氘-铀（CANDU）反应堆容器（例如在达林顿核电站和布鲁斯动力核电站）被一个圆柱形屏蔽罐包围，罐内装有液态水，在发生严重事故时提供外部容器冷却。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以维持。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池到容器壁的传热，以评估反应堆容器壁外部的冷却是否足以保持其完整性。该研究将使用计算流体动力学（CFD）来获得自然对流流体流动的详细信息，以及局部（空间）和时变壁面换热率。这些将用于评估目前用于核工业设计基础以外分析的简化积分规范中基本假设的适当性。