Uncertainty analysis of rod ejection accident in VVER-1000 reactor

Kód výsledku v IS VaVaI

<a href="https://www.isvavai.cz/riv?ss=detail&h=RIV%2F68407700%3A21220%2F19%3A00334551" target="_blank" >RIV/68407700:21220/19:00334551 - isvavai.cz</a>

Nalezeny alternativní kódy

RIV/68407700:21340/19:00334551

Výsledek na webu

<a href="https://doi.org/10.1016/j.anucene.2019.06.061" target="_blank" >https://doi.org/10.1016/j.anucene.2019.06.061</a>

DOI - Digital Object Identifier

<a href="http://dx.doi.org/10.1016/j.anucene.2019.06.061" target="_blank" >10.1016/j.anucene.2019.06.061</a>

Jazyk výsledku

angličtina

Název v původním jazyce

Uncertainty analysis of rod ejection accident in VVER-1000 reactor

Popis výsledku v původním jazyce

In present work the coupling of neutronics code PARCS and thermo-hydraulics code TRACE was performed with adoption of GRS (Gesellschaft fur Anlagen- and Reaktorsicherheit) uncertainty method for rod ejection accident in VVER-1000 reactor core. The determination of input parameters, which represent source of possible uncertainties, for both codes was based on the Phenomenon Identification and Ranking Tables (PIRT) for rod ejection accident in pressurized water reactor (PWR) established by the U.S. Nuclear Regulatory Commission (U.S. NRC). Following input parameters in PARCS were selected: delayed neutron fraction precursor groups, macroscopic cross-section data for density of moderator, temperature of fuel and temperature of moderator. For TRACE code input uncertainty of thermal conductivity of fuel, gap and cladding were defined. The results obtained from uncertainty analyses imply that macroscopic cross-section data of moderator temperature is the most influential uncertainty among the input parameters in a neutronics code. On the other hand, uncertainty of thermal conductivity of uranium fuel significantly affects the results of coupled calculation. The influence of input uncertainties were investigated on the calculated results, mainly for core thermal power, fuel centerline temperature, total reactivity, moderator density, and fuel temperature reactivity.

Název v anglickém jazyce

Uncertainty analysis of rod ejection accident in VVER-1000 reactor

Popis výsledku anglicky

In present work the coupling of neutronics code PARCS and thermo-hydraulics code TRACE was performed with adoption of GRS (Gesellschaft fur Anlagen- and Reaktorsicherheit) uncertainty method for rod ejection accident in VVER-1000 reactor core. The determination of input parameters, which represent source of possible uncertainties, for both codes was based on the Phenomenon Identification and Ranking Tables (PIRT) for rod ejection accident in pressurized water reactor (PWR) established by the U.S. Nuclear Regulatory Commission (U.S. NRC). Following input parameters in PARCS were selected: delayed neutron fraction precursor groups, macroscopic cross-section data for density of moderator, temperature of fuel and temperature of moderator. For TRACE code input uncertainty of thermal conductivity of fuel, gap and cladding were defined. The results obtained from uncertainty analyses imply that macroscopic cross-section data of moderator temperature is the most influential uncertainty among the input parameters in a neutronics code. On the other hand, uncertainty of thermal conductivity of uranium fuel significantly affects the results of coupled calculation. The influence of input uncertainties were investigated on the calculated results, mainly for core thermal power, fuel centerline temperature, total reactivity, moderator density, and fuel temperature reactivity.

Druh

J_imp - Článek v periodiku v databázi Web of Science

CEP obor

—

OECD FORD obor

10304 - Nuclear physics

Projekt

<a href="/cs/project/LM2015053" target="_blank" >LM2015053: VR-1 – Školní reaktor pro výzkumnou činnost</a><br>

Návaznosti

P - Projekt vyzkumu a vyvoje financovany z verejnych zdroju (s odkazem do CEP)

Rok uplatnění

2019

Kód důvěrnosti údajů

S - Úplné a pravdivé údaje o projektu nepodléhají ochraně podle zvláštních právních předpisů

Název periodika

Annals of Nuclear Energy

ISSN

0306-4549

e-ISSN

—

Svazek periodika

132

Číslo periodika v rámci svazku

Oct

Stát vydavatele periodika

GB - Spojené království Velké Británie a Severního Irska

Počet stran výsledku

8

Strana od-do

628-635

Kód UT WoS článku

000482247600058

EID výsledku v databázi Scopus

2-s2.0-85068213786