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Assessment of Neutronic Characteristics of Accident-Tolerant Fuel and Claddings for CANDU Reactors
المؤلفون المشاركون
المصدر
Science and Technology of Nuclear Installations
العدد
المجلد 2018، العدد 2018 (31 ديسمبر/كانون الأول 2018)، ص ص. 1-17، 17ص.
الناشر
Hindawi Publishing Corporation
تاريخ النشر
2018-03-01
دولة النشر
مصر
عدد الصفحات
17
الملخص EN
The objective of this study was to evaluate accident-tolerant fuel (ATF) concepts being considered for CANDU reactors.
Several concepts, including uranium dioxide/silicon carbide (UO2-SiC) composite fuel, dense fuels, microencapsulated fuels, and ATF cladding, were modelled in Serpent 2 to obtain reactor physics parameters, including important feedback parameters such as coolant void reactivity and fuel temperature coefficient.
In addition, fuel heat transfer was modelled, and a simple accident model was tested on several ATF cases to compare with UO2.
Overall, several concepts would require enrichment of uranium to avoid significant burnup penalties, particularly uranium-molybdenum (U-Mo) and fully ceramic microencapsulated (FCM) fuels.
In addition, none of the fuel types have a significant advantage over UO2 in terms of overall accident response or coping time, though U-9Mo fuel melts significantly sooner due to its low melting point.
Instead, the different ATF concepts appear to have more modest advantages, such as reduced fission product release upon cladding failure, or reduced hydrogen generation, though a proper risk assessment would be required to determine the magnitude of these advantages to weigh against economic disadvantages.
The use of uranium nitride (UN) enriched in N15 would increase exit burnup for natural uranium, providing a possible economic advantage depending on fuel manufacturing costs.
نمط استشهاد جمعية علماء النفس الأمريكية (APA)
Younan, Simon& Novog, David R.. 2018. Assessment of Neutronic Characteristics of Accident-Tolerant Fuel and Claddings for CANDU Reactors. Science and Technology of Nuclear Installations،Vol. 2018, no. 2018, pp.1-17.
https://search.emarefa.net/detail/BIM-1214961
نمط استشهاد الجمعية الأمريكية للغات الحديثة (MLA)
Younan, Simon& Novog, David R.. Assessment of Neutronic Characteristics of Accident-Tolerant Fuel and Claddings for CANDU Reactors. Science and Technology of Nuclear Installations No. 2018 (2018), pp.1-17.
https://search.emarefa.net/detail/BIM-1214961
نمط استشهاد الجمعية الطبية الأمريكية (AMA)
Younan, Simon& Novog, David R.. Assessment of Neutronic Characteristics of Accident-Tolerant Fuel and Claddings for CANDU Reactors. Science and Technology of Nuclear Installations. 2018. Vol. 2018, no. 2018, pp.1-17.
https://search.emarefa.net/detail/BIM-1214961
نوع البيانات
مقالات
لغة النص
الإنجليزية
الملاحظات
Includes bibliographical references
رقم السجل
BIM-1214961
قاعدة معامل التأثير والاستشهادات المرجعية العربي "ارسيف Arcif"
أضخم قاعدة بيانات عربية للاستشهادات المرجعية للمجلات العلمية المحكمة الصادرة في العالم العربي
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