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Validation of the COTENP Code: A Steady-State Thermal-Hydraulic Analysis Code for Nuclear Reactors with Plate Type Fuel Assemblies
المؤلفون المشاركون
Maiorino, J. R.
Castellanos-Gonzalez, Duvan A.
Moreira, João Manoel Losada
Carajilescov, Pedro
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
2018-12-02
دولة النشر
مصر
عدد الصفحات
17
الملخص EN
This article presents the validation of the Code for Thermal-hydraulic Evaluation of Nuclear Reactors with Plate Type Fuels (COTENP), a subchannel code which performs steady-state thermal-hydraulic analysis of nuclear reactors with plate type fuel assemblies operating with the coolant at low pressure levels.
The code is suitable for design analysis of research, test, and multipurpose reactors.
To solve the conservation equations for mass, momentum, and energy, we adopt the subchannel and control volume methods based on fuel assembly geometric data and thermal-hydraulic conditions.
We consider the chain or cascade method in two steps to facilitate the analysis of whole core.
In the first step, we divide the core into channels with dimensions equivalent to that of the fuel assembly and identify the assembly with largest enthalpy rise as the hot assembly.
In the second step, we divide the hot fuel assembly into subchannels with size equivalent to one actual coolant channel and similarly identify the hot subchannel.
The code utilizes the homogenous equilibrium model for two-phase flow treatment and the balanced drop pressure approach for the flow rate determination.
The code results include detailed information such as core pressure drop, mass flow rate distribution, coolant, cladding and centerline fuel temperatures, coolant quality, local heat flux, and results regarding onset of nucleate boiling and departure of nucleate boiling.
To validate the COTENP code, we considered experimental data from the Brazilian IEA-R1 research reactor and calculated data from the Chinese CARR multipurpose reactor.
The mean relative discrepancies for the coolant distribution were below 5%, for the coolant velocity were 1.5%, and for the pressure drop were below 10.7%.
The latter discrepancy can be partially justified due to lack of information to adequately model the IEA-R1 experiment and CARR reactor.
The results show that the COTENP code is sufficiently accurate to perform steady-state thermal-hydraulic design analyses for reactors with plate type fuel assemblies.
نمط استشهاد جمعية علماء النفس الأمريكية (APA)
Castellanos-Gonzalez, Duvan A.& Moreira, João Manoel Losada& Maiorino, J. R.& Carajilescov, Pedro. 2018. Validation of the COTENP Code: A Steady-State Thermal-Hydraulic Analysis Code for Nuclear Reactors with Plate Type Fuel Assemblies. Science and Technology of Nuclear Installations،Vol. 2018, no. 2018, pp.1-17.
https://search.emarefa.net/detail/BIM-1215010
نمط استشهاد الجمعية الأمريكية للغات الحديثة (MLA)
Castellanos-Gonzalez, Duvan A.…[et al.]. Validation of the COTENP Code: A Steady-State Thermal-Hydraulic Analysis Code for Nuclear Reactors with Plate Type Fuel Assemblies. Science and Technology of Nuclear Installations No. 2018 (2018), pp.1-17.
https://search.emarefa.net/detail/BIM-1215010
نمط استشهاد الجمعية الطبية الأمريكية (AMA)
Castellanos-Gonzalez, Duvan A.& Moreira, João Manoel Losada& Maiorino, J. R.& Carajilescov, Pedro. Validation of the COTENP Code: A Steady-State Thermal-Hydraulic Analysis Code for Nuclear Reactors with Plate Type Fuel Assemblies. Science and Technology of Nuclear Installations. 2018. Vol. 2018, no. 2018, pp.1-17.
https://search.emarefa.net/detail/BIM-1215010
نوع البيانات
مقالات
لغة النص
الإنجليزية
الملاحظات
Includes bibliographical references
رقم السجل
BIM-1215010
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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