ANS is committed to advancing, fostering, and promoting the development and application of nuclear sciences and technologies to benefit society.
Explore the many uses for nuclear science and its impact on energy, the environment, healthcare, food, and more.
Explore membership for yourself or for your organization.
Conference Spotlight
2026 Nuclear Energy Conference & Expo (NECX)
August 24–27, 2026
Dallas, TX|Hilton Anatole
Latest Magazine Issues
Aug 2026
Jan 2026
2026
Latest Journal Issues
Nuclear Science and Engineering
October 2026
Nuclear Technology
September 2026
Fusion Science and Technology
August 2026
Latest News
Porous tungsten scrubbed by glow discharge cleaning
Researchers conducted experiments in Princeton Plasma Physics Laboratory’s Lithium Tokamak Experiment-Beta (LTX-β) showing glow discharge cleaning can be used to effectively clean samples of porous tungsten—used to hold liquid lithium in fusion machine inner walls—manufactured from powder-reconstituted materials, according to a paper published in Nuclear Materials and Energy.
Tungsten is widely used for plasma-facing components in fusion machines, especially in the divertor region where materials must withstand extreme levels of power flow. According to the paper, spark plasma sintering can be used to make tungsten into spongelike samples for holding liquid lithium.
V. D’Ambrosi, J. Sercombe, S. Bejaoui, A. Chaieb, B. Baurens, R. Largenton, A. Ambard, B. Boer, G. Bonny, M. Ševeček, L. E. Herranz, F. Feria Marquez, K. Inagaki, H. Ohta, F. Boldt, J. Sappl, R. Armstrong, A. Mohamad, Y. Udagawa, C. Cozzo, J. Klouzal, M. Vitezslav, J. Corson, J. Peltonen
Nuclear Technology | Volume 210 | Number 2 | February 2024 | Pages 189-215
Research Article | doi.org/10.1080/00295450.2023.2194270
Articles are hosted by Taylor and Francis Online.
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.