KnE Engineering

ISSN: 2518-6841

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THE VERIFICATION OF THE COMPLEX PROGRAMS SERPENT 2 AND SCALE (SAS2) FOR ANALYSING THE SAFETY CHARACTERISTICS OF FA REACTOR VVER-1000 AT ALL THE OPERATION STAGES

Published date:Feb 21 2018

Journal Title: KnE Engineering

Issue title: XIII International Youth Scientific and Practical Conference "FUTURE OF ATOMIC ENERGY - AtomFuture 2017"

Pages:201-218

DOI: 10.18502/keg.v3i3.1620

Authors:
Abstract:

Currently, to estimate the parameters of the Spent Nuclear Fuel (SNF) and their evolution with time many of the international software are used. This work is dedicated to the evaluation, analysis and comparison the safety characteristics of the fuel Assembly which used in reactor types VVER-1000 obtained in the present work and by other authors [1] using different software packages. The result of calculation for the characteristic safety nuclear fuel at many stages of the Nuclear Power Planet (NPP) was also calculated in this work.

References:

[1] Emmett M. B. Calculational Benchmark Problems for VVER-1000 Mixed Oxide Fuel Cycle (2000).


[2] Bomboni E, Cerullo N, Fridman E, et al (2010) Comparison among MCNP-based depletion codes applied to burnup calculations of pebble-bed HTR lattices. Nucl Eng Des 240:918–924. doi: 10.1016/j.nucengdes.2009.12.006


[3] Chersola D (2016) Application of new neutronic and burnup Monte Carlo based codes to the study of nuclear fuel cycles for GFR and VVER systems. University of Genova, Italy.


[4] Chersola D, Lomonaco G, Marotta R, Mazzini G (2014) No Title. Nucl Eng Des 273:542– 554.


[5] Deen R., Woodruff W., Costescu C., Leopando L. (2000) WIMS-ANL User Manual, Rev. 4.


[6] Fowler T., Vondy D. (1969) NUCLEAR REACTOR CORE ANALYSIS CODE: CITATION.


[7] Kalugin M., Oleynik D., Shkarovsky D. (2015) Overview of the MCU Monte Carlo software package. Ann Nucl Energy 82:54–62.


[8] RSICC Computer Code Collection MCNP4 Oak Ridge National Laboratory, CCC-700.


[9] SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluations, ORNL/TM-2005/39, Version 5, Vols. I–III, April 2005. Available from Radiation Safety Information Computational Center at Oak Ridge National Laboratory as.


[10] Авдеев EФ, Чусов ИА, Карпенко АА (2010) Верификация Струйной Методики Расчета Гидродинамики Активной Зоный Реакторов При Блокировке Сечения ТВС. Ядерная энергетика 2:115–124.


[11] Бурцев С. (2017) Анализ влияния числа Прандтля на значение коэффициента восстановления температуры. Наука и Образование МГТУ им НЭ Баумана 3:78– 96. doi: 10.7463/0317.0001115.


[12] Головко ЮЕ, Кощеев ВН, Ломаков ГБ, et al (2014) Версии Констант БНАБ и Программы подготовки Критичности. Ядерная энергетика 2:99–108.


[13] Онегин M., Рыжов ИВ (2011) Верификация Пргораммы MURE Для Расчета Остаточного Топлива Ялерных Реакторов. Вопросы Атомной Науки и Техники.


[14] Хайлов АМ, Иванников АИ, Орленко СП, et al (2015) Расчёт поглощённых доз фотонного и нейтронного излучения в эмали и дентине зубов человека методом Монте - Карло Введение. Радиация и риск 2:93–106.


[15] Leppänen J (2015) Serpent – a Continuous-energy Monte Carlo Reactor Physics Burnup Calculation Code.


[16] Gauld I. C. Hermann O. W. SAS2: A Coupled One-Dimensional Deplition And Shielding Analysis Module. OAK RIDGE National Laboratory Oak Ridge, Tennessee 37831-6170. ORNL/TM-2005/39 Version 5 Vol. I, Book 3, Sect. S2.


[17] Schlenker M (2014) Multi-physical Developments for Safety Related Investigations of Low Moderated Boiling Water Reactors. Karlsruher Institut für Technologie (KIT) genehmigte.

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