Published August 2008 | Version v1
Journal article Metadata-only

MCU-FCP Code for neutron transport calculation using first collision probability method

Creators

  • 1. RNTs Kurchatovskij Inst., Moscow (Russian Federation)
  • 2. NIKIEhT im. N.A. Dollezhalya, Moscow (Russian Federation)

Description

General scheme, methods and algorithms are described, being realized in the MCU-FCP code, which permits calculating neutronic characteristics of the RBMK two-dimensional, three-dimensional cells and polycells, making allowance for fuel burnup in the course of operation using the first collision probability method. Test calculations are made and the results are compared with the ones obtained using the MCU-REA code. It is planned to use the MCU-FCP code for compiling a constant library for the Kursk NPP unit 5 calculations

Abstract (Russian)

Описаны общая схема, методы и алгоритмы, которые реализованы в программе MCU-FCP, позволяющей рассчитывать нейтронно-физические характеристики двyмepных, тpexмepных ячеек и полиячеек РБМК с учетом выгорания топлива в процессе эксплуатации методом вероятностей первых столкновений. Выполнены тестовые расчеты и проведено сравнение результатов с программой MCU-REA. Планируется использовать программу MCU-FCP для подготовки библиотеки констант при расчетах 5-го блока Курской АЭС

Additional details

Additional titles

Original title (Russian)
Программа MCU-FCP для расчета переноса нейтронов методом вероятностей первых столкновений

Publishing Information

Journal Title
Atomnaya Ehnergiya
Journal Volume
105
Journal Issue
2
Journal Page Range
p. 67-72
ISSN
0004-7163
CODEN
AENGAB

INIS

Country of Publication
Russian Federation
Country of Input or Organization
Russian Federation
INIS RN
40088576
Subject category
S22: GENERAL STUDIES OF NUCLEAR REACTORS;
Descriptors DEI
ALGORITHMS; BURNUP; COMPUTERIZED SIMULATION; FLUX DENSITY; LWGR TYPE REACTORS; M CODES; NEUTRON FLUX; NEUTRON TRANSPORT; PROBABILISTIC ESTIMATION; REACTOR CELLS; REACTOR CORES
Descriptors DEC
CALCULATION METHODS; COMPUTER CODES; GRAPHITE MODERATED REACTORS; MATHEMATICAL LOGIC; NEUTRAL-PARTICLE TRANSPORT; RADIATION FLUX; RADIATION TRANSPORT; REACTOR COMPONENTS; REACTORS; SIMULATION; WATER COOLED REACTORS

Optional Information

Notes
12 refs., 4 figs., 1 tab.