Burn-Up Dependent Modeling of Fuel-to-Clad Gap Conductance and Temperature Predictions for Mixed-Oxide Fuel in the ESFR-SMART Core

被引：2

作者：

Lavarenne, Jean ^{[1
]}

Bubelis, Evaldas ^{[2
]}

Davies, Una ^{[3
]}

Gianfelici, Simone ^{[2
]}

Gicquel, Solene ^{[4
]}

Krepel, Jiri ^{[5
]}

Lainet, Marc ^{[6
]}

Lindley, Ben ^{[7
]}

Mikityuk, Konstantin ^{[5
]}

Murphy, Christophe ^{[1
]}

Perrin, Benoit ^{[8
]}

Pfrang, Werner ^{[2
]}

Ponomarev, Alexander ^{[5
]}

Schubert, Arndt ^{[9
]}

Shwageraus, Eugene ^{[10
]}

Van Uffelen, Paul ^{[11
]}

机构：

[1] Jacobs, 5 Kings Point House,Queen Mother Sq, Dorchester DT1 3BW, England

[2] Karlsruhe Inst Technol KIT, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany

[3] Univ Cambridge, Dept Engn, Trumpington St, Cambridge CB2 1PZ, England

[4] EDF, EDF R&D, Ave Renardieres, F-77818 Ecuelles, Moret Sur Loing, France

[5] Paul Scherrer Inst PSI, Forsch Str 111, CH-5232 Villigen, Switzerland

[6] French Alternat Energies & Atom Energy Comm CEA, Cadarache, F-13115 St Paul Les Durance, France

[7] Univ Wisconsin, Engn Hall,1415 Engn Dr, Madison, WI 53706 USA

[8] Framatome France, 1 Pl Jean Miller, F-92400 Courbevoie, France

[9] European Commiss, Joint Res Ctr JRC, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshalen, Germany

[10] Univ Cambridge, Engn Dept, Trumpington SI, Cambridge CB2 1PZ, England

[11] European Commis, Joint Res Ctr JRC, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshalen, Germany

来源：

JOURNAL OF NUCLEAR ENGINEERING AND RADIATION SCIENCE | 2022年 / 8卷 / 01期

关键词：

CODE; BEHAVIOR; VALIDATION;

D O I：

10.1115/1.4050479

中图分类号：

TL [原子能技术]; O571 [原子核物理学];

学科分类号：

0827 ; 082701 ;

摘要：

Accurate coupled neutronic-thermal-hydraulic analysis of sodium fast reactor (SFRs), including European sodium fast reactor-safety measures assessment and research tools (ESFR-SMART), requires an accurate calculation of the fuel-to-clad gap conductance. This is typically derived using a fuel performance code and is dependent on fuel rating and burn-up history, among other parameters. In this paper, the gap conductance of the mixed oxide fuel (MOX) pins in the ESFR-SMART project is investigated through predictive modeling in seven independent fuel performance codes, to provide confidence in results and to aid the understanding of the uncertainties associated with gap conductance predictions. A parameterization is then defined. A single pin model for both the inner and outer fuel was built. The fuel was burned for 2100 effective full power days, with the axial power distribution varying over time. This paper presents a comparison of the gap conductance and predicted fuel temperature distribution between codes. The values produced from the codes are then combined to produce a best estimate prediction of the gap conductance expressed as a function of nodal fuel rating and burn-up, for both inner and outer fuel for all seven codes. A 2D fit was applied to the data thus obtained. The spread between results is such that, to 95% confidence, gap conductance predictions may vary from the correlation by up to a factor of similar to 4. The gap conductance results obtained show a general increase of conductance with fuel rating and burn-up, from 0.22 at 0 burn-up and 10 kW/m to 0.45 W/cm/K at 0 burn-up and 50 kW/m and to 1.00 W/cm/K at 150GWd/t and 50 kW/m. Some spread between codes has been noted and appears to be consistent with the spread published earlier by several code developers. There is good agreement between codes at low burn-up for both the central and surface fuel temperature predictions. The spread between codes increases with burn-up due to multiple phenomena, including gap re-opening, joint oxyde gain (JOG) formation, and clad swelling.

引用

页数：11

共 2 条

[1] STRUCTURE DEVELOPMENT IN MIXED-OXIDE FUEL AT LOW BURN-UP
BAKER, RB
LEGGETT, RD
CHASTAIN, SA
AMERICAN CERAMIC SOCIETY BULLETIN, 1976, 55 (04): : 442 - 442
[2] Microstructural modeling of thermal conductivity of high burn-up mixed oxide fuel
Teague, Melissa
Tonks, Michael
Novascone, Stephen
Hayes, Steven
JOURNAL OF NUCLEAR MATERIALS, 2014, 444 (1-3) : 161 - 169

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