Influence of cross-section uncertainties on the calculated results from a thermo-hydraulics code

Kód výsledku v IS VaVaI

<a href="https://www.isvavai.cz/riv?ss=detail&h=RIV%2F68407700%3A21340%2F20%3A00342530" target="_blank" >RIV/68407700:21340/20:00342530 - isvavai.cz</a>

Výsledek na webu

<a href="https://doi.org/10.1016/j.anucene.2020.107598" target="_blank" >https://doi.org/10.1016/j.anucene.2020.107598</a>

DOI - Digital Object Identifier

<a href="http://dx.doi.org/10.1016/j.anucene.2020.107598" target="_blank" >10.1016/j.anucene.2020.107598</a>

Jazyk výsledku

angličtina

Název v původním jazyce

Influence of cross-section uncertainties on the calculated results from a thermo-hydraulics code

Popis výsledku v původním jazyce

This study deals with the influence of uncertainties in selected cross-section data in a neutronics code while their impact on the results from thermo-hydraulics code is investigated. Particular cross-section data were defined according to the Phenomenon Identification and Ranking Tables (PIRT) for a pressurized water reactor. Uncertainties were identified for the following data: b – delayed neutron fraction precursors and transport, absorption, m-fission, j-fission, down scattering for moderator temperature, moderator density and fuel temperature. As a neutronics code, PARCS in conjunction with thermohydraulics code TRACE was applied. Uncertainty analyses were carried out by the code DAKOTA where coupled PARCS/TRACE calculations were run 146 times in order to meet a two-sided 95/95 confidence interval. In the thermo-hydraulic code thermal power of the reactor core and inner temperature of the fuel rod were studied. This research shows that the most influential parameters associated with crosssection parametrization are moderator temperature and fuel temperature.

Název v anglickém jazyce

Influence of cross-section uncertainties on the calculated results from a thermo-hydraulics code

Popis výsledku anglicky

This study deals with the influence of uncertainties in selected cross-section data in a neutronics code while their impact on the results from thermo-hydraulics code is investigated. Particular cross-section data were defined according to the Phenomenon Identification and Ranking Tables (PIRT) for a pressurized water reactor. Uncertainties were identified for the following data: b – delayed neutron fraction precursors and transport, absorption, m-fission, j-fission, down scattering for moderator temperature, moderator density and fuel temperature. As a neutronics code, PARCS in conjunction with thermohydraulics code TRACE was applied. Uncertainty analyses were carried out by the code DAKOTA where coupled PARCS/TRACE calculations were run 146 times in order to meet a two-sided 95/95 confidence interval. In the thermo-hydraulic code thermal power of the reactor core and inner temperature of the fuel rod were studied. This research shows that the most influential parameters associated with crosssection parametrization are moderator temperature and fuel temperature.

Druh

J_imp - Článek v periodiku v databázi Web of Science

CEP obor

—

OECD FORD obor

10304 - Nuclear physics

Projekt

<a href="/cs/project/EF16_013%2F0001790" target="_blank" >EF16_013/0001790: Posílení a rozvoj výzkumu na ČVUT v Praze s využitím výzkumné infrastruktury 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

2020

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

145

Číslo periodika v rámci svazku

May

Stát vydavatele periodika

GB - Spojené království Velké Británie a Severního Irska

Počet stran výsledku

8

Strana od-do

—

Kód UT WoS článku

000540700300031

EID výsledku v databázi Scopus

2-s2.0-85085269483