کد مقاله | کد نشریه | سال انتشار | مقاله انگلیسی | نسخه تمام متن |
---|---|---|---|---|
4925297 | 1431399 | 2017 | 12 صفحه PDF | دانلود رایگان |
عنوان انگلیسی مقاله ISI
Investigation of thermo mechanical behaviour in the scaled PHWR stepped calandria vessel during severe accident
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موضوعات مرتبط
مهندسی و علوم پایه
مهندسی انرژی
مهندسی انرژی و فناوری های برق
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چکیده انگلیسی
In case of Indian Pressurized Heavy Water Reactors (PHWRs), in-calandria retention of corium is essential for mitigating core melt down accident. Benchmark calculations have shown that if corium breaches the calandria vessel and enters the calandria vault, large amount of hydrogen and other fission gases may be generated due to molten corium concrete interaction (MCCI). There is possibility of containment failure. Hence, the best option is to contain the corium inside the calandria vessel and cool it from outside by calandria vault water. PHWR calandria vessel thicknesses varies from 22Â mm to 32Â mm and have weld joints due to the steps involved. It is true that smaller calandria vessel thickness has lower temperature gradient and may promote higher heat transfer rate. However, there is an international concern about the healthiness of the thin vessel with stepped weld joints due to the thermal loading of high temperature corium melt which can lead to excessive thermal strain of vessel and weld joints and it may result in failure before the corium stabilization inside the vessel is achieved. In the present study, experiments were conducted in a scaled facility of an Indian PHWR to investigate the thermo-mechanical behaviour of simulated stepped calandria vessel with weld joints. The temperature distributions and the strain contours of vessel were recorded. Numerical analysis was carried out to validate the FEM model with the test data and also analysis was extended to predict the thermo mechanical behaviour of prototypic stepped calandria vessel.
ناشر
Database: Elsevier - ScienceDirect (ساینس دایرکت)
Journal: Nuclear Engineering and Design - Volume 322, October 2017, Pages 591-602
Journal: Nuclear Engineering and Design - Volume 322, October 2017, Pages 591-602
نویسندگان
Sumit V. Prasad, A.K. Nayak,