کد مقاله | کد نشریه | سال انتشار | مقاله انگلیسی | نسخه تمام متن |
---|---|---|---|---|
4925862 | 1431412 | 2016 | 15 صفحه PDF | دانلود رایگان |
عنوان انگلیسی مقاله ISI
Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event
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کلمات کلیدی
OECDMDNBRVERA-CSdNBRPWRMain steam line breakEOCCASLCOBRA-TFCTFINLChFDNBHFPBEPULocaRCCAFOMMSLBHZPHPCConsortium for Advanced Simulation of Light Water ReactorsK. Thermal Hydraulics - K. حرارتی هیدرولیکIdaho National Laboratory - آزمایشگاه ملی آیداهوBest estimate plus uncertainty - بهترین برآورد همراه با عدم قطعیتSensitivity analysis - تحلیل حساسیتHigh flow - جریان بالاLow flow - جریان کمLoss-of-Coolant Accident - حوادث ناشی از خنک کنندهDeparture from nucleate boiling - خروج از جوش هسته ایCFD - دینامیک سیالاتComputational fluid dynamics - دینامیک سیالات محاسباتیPressurized Water Reactor - راکتور آب تحت فشارOrganisation for Economic Co-operation and Development - سازمان همکاری اقتصادی و توسعهCritical heat flux - شار حرارتی بحرانیCorrelation coefficient - ضریب همبستگیHigh performance computing - محاسبات با کارایی بالاFigure of merit - معیار شایستگیEnd Of Cycle - پایان چرخهUncertainty quantification - کمی سازی عدم قطعیت
موضوعات مرتبط
مهندسی و علوم پایه
مهندسی انرژی
مهندسی انرژی و فناوری های برق
پیش نمایش صفحه اول مقاله
چکیده انگلیسی
VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was applied to simulate core behavior of a typical Westinghouse-designed 4-loop pressurized water reactor (PWR) with 17Â ÃÂ 17 fuel assemblies in response to two main steam line break (MSLB) accident scenarios initiated at hot zero power (HZP) at the end of the first fuel cycle with the most reactive rod cluster control assembly stuck out of the core. The reactor core boundary conditions at the most DNB limiting time step were determined by a system analysis code. The core inlet flow and temperature distributions were obtained from computational fluid dynamics (CFD) simulations. The two MSLB scenarios consisted of the high and low flow situations, where reactor coolant pumps either continue to operate with offsite power or do not continue to operate since offsite power is unavailable. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks' nonparametric statistical approach. In this demonstration of BEPU application, 59 full core simulations were performed for each accident scenario to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. A parametric goodness-of-fit approach was also applied to the results to obtain the MDNBR value at the 95/95 tolerance limit. Initial sensitivity analysis was performed with the 59 cases per accident scenario by use of Pearson correlation coefficients. The results show that this typical PWR core retains design margin with respect to the MDNBR safety limit for both of the MSLB accident scenarios. The scenario with available offsite power was more restrictive in terms of MDNBR than the scenario without offsite power.
ناشر
Database: Elsevier - ScienceDirect (ساینس دایرکت)
Journal: Nuclear Engineering and Design - Volume 309, 1 December 2016, Pages 8-22
Journal: Nuclear Engineering and Design - Volume 309, 1 December 2016, Pages 8-22
نویسندگان
C.S. Brown, H. Zhang, V. Kucukboyaci, Y. Sung,