P2M Simulation Exercise on Past Fuel Melting Irradiation Experiments

Result code in IS VaVaI

<a href="https://www.isvavai.cz/riv?ss=detail&h=RIV%2F46356088%3A_____%2F23%3AN0000018" target="_blank" >RIV/46356088:_____/23:N0000018 - isvavai.cz</a>

Result on the web

<a href="https://www.tandfonline.com/doi/full/10.1080/00295450.2023.2194270" target="_blank" >https://www.tandfonline.com/doi/full/10.1080/00295450.2023.2194270</a>

DOI - Digital Object Identifier

<a href="http://dx.doi.org/10.1080/00295450.2023.2194270" target="_blank" >10.1080/00295450.2023.2194270</a>

Result language

angličtina

Original language name

P2M Simulation Exercise on Past Fuel Melting Irradiation Experiments

Original language description

This paper presents the results of the Power To Melt and Maneuverability (P2M) Simulation Exercise on past fuel melting irradiation experiments, organized within the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Framework for IrraDiation ExperimentS (FIDES) framework by the Core Group (CEA, EDF, and SCK‧CEN) and open to all FIDES members. The exercise consisted in simulating two past power ramps where fuel melting was detected: (1) the xM3 staircase power transient [ramp terminal level (RTL) 70 kW‧m−1, average burnup 27 GWd‧tU−1], carried out in 2005 in the R2 reactor at Studsvik (Sweden), where the rodlet maintained its integrity, and (2) the HBC4 fast power transient (RTL 66 kW‧m−1, average burnup 48 GWd‧tU−1), carried out in 1987 in the BR2 reactor at SCK‧CEN (Belgium), where the cladding failed during the experiment. The exercise was joined by 13 organizations from 9 countries using 11 different fuel performance codes. In this paper, the main results of the Simulation Exercise are presented and compared to available postirradiation examinations (PIE) or on-line measurements during the power ramps (fuel and clad diameters, rod elongation, pellet-clad gap, and fission gas release). Since the focus of the Simulation Exercise is on fuel melting assessment, determination of the boundary between melted/nonmelted fuel and the consequent definition of a melting radius from PIE are first discussed. During the HBC4 ramp, fuel melting was predicted by most of the codes despite differences in the melting models. Higher discrepancies were observed for the xM3 rod that can be attributed partly to power uncertainty and partly to the limited capability of the models to describe partial melting of the fuel during this ramp. Finally, possible code developments to improve simulation results are presented.

Czech name

—

Czech description

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Type

J_imp - Article in a specialist periodical, which is included in the Web of Science database

CEP classification

—

OECD FORD branch

20305 - Nuclear related engineering; (nuclear physics to be 1.3);

Project

—

Continuities

V - Vyzkumna aktivita podporovana z jinych verejnych zdroju

Publication year

2023

Confidentiality

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

Name of the periodical

Nuclear Technology

ISSN

0029-5450

e-ISSN

1943-7471

Volume of the periodical

210

Issue of the periodical within the volume

2

Country of publishing house

US - UNITED STATES

Number of pages

7

Pages from-to

189-215

UT code for WoS article

000993607400001

EID of the result in the Scopus database

2-s2.0-85153316082