Published June 2005 | Version v1
Journal article Metadata-only

Analysis of uncertainties of calculations of the loss-of-coolant accidents at the first power unit of the Kursk NPP

Creators

  • 1. NIKIEhT im. N.A. Dollezhalya, Moscow (Russian Federation)

Description

A technique for statistical analysis of uncertainties of thermophysical calculations of the loss-of-coolant accidents is developed. The technique permits to determine with a given reliability the confidence interval for an estimated parameter and to estimate the extent of proximity of the variant calculations to the base (nondefected) one. This technique is tested within frameworks of a profound basis of the safety of the Kursk NPP first power unit during analysis by means of the RELAP5/MOD3.2 code for two types of accidents with rupture of a group collector

Abstract (Russian)

Разработана методика статистического анализа неопределенности теплофизических расчетов, позволяющая построить с заданной надежностью доверительный интервал для оцениваемого параметра и оценить степень близости вариантных расчетов к базовому (неотклоненному). Методика апробирована в рамках углубленного обоснования безопасности 1-го энергоблока Курской АЭС для анализа с помощью кода RELAP5/MOD3.2 двух видов аварий с разрывом раздаточного группового коллектора

Additional details

Additional titles

Original title (Russian)
Анализ неопределенности расчетов аварий с потерей теплоносителя для 1-го энергоблока Курской АЭС

Publishing Information

Journal Title
Atomnaya Ehnergiya
Journal Volume
98
Journal Issue
6
Journal Page Range
p. 422-428
ISSN
0004-7163
CODEN
AENGAB

INIS

Country of Publication
Russian Federation
Country of Input or Organization
Russian Federation
INIS RN
37070181
Subject category
S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
Descriptors DEI
COMPARATIVE EVALUATIONS; COMPUTER CALCULATIONS; KURSK-1 REACTOR; LOSS OF COOLANT; PARAMETRIC ANALYSIS; PIPELINES; R CODES; REACTOR SAFETY; RELIABILITY; RUPTURES; STATISTICAL MODELS; TESTING
Descriptors DEC
ACCIDENTS; COMPUTER CODES; ENRICHED URANIUM REACTORS; EVALUATION; FAILURES; GRAPHITE MODERATED REACTORS; LWGR TYPE REACTORS; MATHEMATICAL MODELS; POWER REACTORS; REACTOR ACCIDENTS; REACTORS; SAFETY; THERMAL REACTORS; WATER COOLED REACTORS

Optional Information

Notes
7 refs., 2 tabs.