Browse > Article
http://dx.doi.org/10.1016/j.net.2021.04.010

Assessment of three European fuel performance codes against the SUPERFACT-1 fast reactor irradiation experiment  

Luzzi, L. (Politecnico di Milano, Department of Energy, Nuclear Engineering Division)
Barani, T. (Politecnico di Milano, Department of Energy, Nuclear Engineering Division)
Boer, B. (Studiecentrum voor Kernenergie (SCK.CEN))
Cognini, L. (Politecnico di Milano, Department of Energy, Nuclear Engineering Division)
Nevo, A. Del (ENEA, FSN-ING-SIS, CR Brasimone)
Lainet, M. (Commissariat a l'Energie Atomique et aux Energies Alternatives, CEA DEC/SESC)
Lemehov, S. (Studiecentrum voor Kernenergie (SCK.CEN))
Magni, A. (Politecnico di Milano, Department of Energy, Nuclear Engineering Division)
Marelle, V. (Commissariat a l'Energie Atomique et aux Energies Alternatives, CEA DEC/SESC)
Michel, B. (Commissariat a l'Energie Atomique et aux Energies Alternatives, CEA DEC/SESC)
Pizzocri, D. (Politecnico di Milano, Department of Energy, Nuclear Engineering Division)
Schubert, A. (European Commission, Joint Research Centre, Directorate for Nuclear Safety and Security)
Uffelen, P. Van (European Commission, Joint Research Centre, Directorate for Nuclear Safety and Security)
Bertolus, M. (Commissariat a l'Energie Atomique et aux Energies Alternatives, CEA DEC/SESC)
Publication Information
Nuclear Engineering and Technology / v.53, no.10, 2021 , pp. 3367-3378 More about this Journal
Abstract
The design phase and safety assessment of Generation IV liquid metal-cooled fast reactors calls for the improvement of fuel pin performance codes, in particular the enhancement of their predictive capabilities towards uranium-plutonium mixed oxide fuels and stainless-steel cladding under irradiation in fast reactor environments. To this end, the current capabilities of fuel performance codes must be critically assessed against experimental data from available irradiation experiments. This work is devoted to the assessment of three European fuel performance codes, namely GERMINAL, MACROS and TRANSURANUS, against the irradiation of two fuel pins selected from the SUPERFACT-1 experimental campaign. The pins are characterized by a low enrichment (~ 2 wt.%) of minor actinides (neptunium and americium) in the fuel, and by plutonium content and cladding material in line with design choices envisaged for liquid metal-cooled Generation IV reactor fuels. The predictions of the codes are compared to several experimental measurements, allowing the identification of the current code capabilities in predicting fuel restructuring, cladding deformation, redistribution of actinides and volatile fission products. The integral assessment against experimental data is complemented by a code-to-code benchmark focused on the evolution of quantities of engineering interest over time. The benchmark analysis points out the differences in the code predictions of fuel central temperature, fuel-cladding gap width, cladding outer radius, pin internal pressure and fission gas release and suggests potential modelling development paths towards an improved description of the fuel pin behaviour in fast reactor irradiation conditions.
Keywords
SUPERFACT-1 irradiation experiment; Fuel performance; GERMINAL; MACROS; TRANSURANUS; MOX fuel; Assessment and benchmark;
Citations & Related Records
Times Cited By KSCI : 1  (Citation Analysis)
연도 인용수 순위
1 W. Dienst, I. Muelle-Lyda, H. Zimmermann, Swelling, densification and creep of oxide and carbide fuels under irradiation, in: Int. Conf. On Fast Breeder Reactor Performance, 5-8 March 1979, Monterey, California, USA, 1979.
2 D. Olander, Fundamental Aspects of Nuclear Reactor Fuel Elements, Technical Information Center, Office of Public Affairs, Energy Research and Development Administration, 1976.
3 T. Preusser, K. Lassmann, Current status of the transient integral fuel element performance code URANUS, in: SMiRT 7, August 1983, pp. 22-26. Chicago, USA.
4 T. Barani, D. Pizzocri, F. Cappia, L. Luzzi, G. Pastore, P. Van Uffelen, Modeling high burnup structure in oxide fuels for application to fuel performance codes. Part I: high burnup structure formation, J. Nucl. Mater. 539 (2020) 152296.   DOI
5 P. Konarski, J. Sercombe, C. Riglet-Martial, L. Noirot, I. Zacharie-Aubrun, K. Hanifi, M. Fregon ese, P. Chantrenne, 3D simulation of a power ramp including fuel thermochemistry and oxygen thermodiffusion, J. Nucl. Mater. 519 (2019) 104-120.   DOI
6 P. Chakraborty, C. Gueneau, A. Chartier, Development of a complete thermo-kinetic description of cations in the mixed oxide of uranium and plutonium, in: NuFuel-MMSNF 2019 Workshop, 4-7 November 2019, PSI, Villigen, Switzerland, 2019.
7 R. Parrish, A. Aitkaliyeva, A review of microstructural features in fast reactor mixed oxide fuels, J. Nucl. Mater. 510 (2018) 644-660.   DOI
8 F. Cappia, D. Pizzocri, A. Schubert, P. Van Uffelen, G. Paperini, D. Pellottiero, R. Macian-Juan, V.V. Rondinella, Critical assessment of the pore size distribution in the rim region of high burnup UO2 fuels, J. Nucl. Mater. 480 (2016) 138-149.   DOI
9 D. Pizzocri, F. Cappia, L. Luzzi, G. Pastore, V.V. Rondinella, P. Van Uffelen, A semi-empirical model for the formation and depletion of the high burnup structure in UO2, J. Nucl. Mater. 487 (2017).
10 M. Bober, C. Sari, G. Schumacher, Redistribution of plutonium and uranium in mixed (U, Pu) oxide fuel materials in a thermal gradient, J. Nucl. Mater. 39 (3) (1971) 265-284.   DOI
11 J. Noirot, L. Desgranges, J. Lamontagne, Detailed characterization of high burnup structures in oxide fuels, J. Nucl. Mater. 372 (2-3) (2008) 318-339.   DOI
12 V.V. Rondinella, T. Wiss, The high burn-up structure in nuclear fuel, Mater. Today 13 (12) (2010) 24-32.   DOI
13 K. Lassmann, The oxired model for redistribution of oxygen in nonstoichiometric uranium-plutonium oxides, J. Nucl. Mater. 150 (1) (1987) 10-16.   DOI
14 EERA-JPNM, INSPYRE - Investigations Supporting MOX Fuel Licensing in ESNII Prototype Reactors [Online]. Available, http://www.eera-jpnm.eu/inspyre/, 2017.
15 J.-F. Babelot, N. Chauvin, Joint CEA/JRC Synthesis Report of the Experiment SUPERFACT 1, Report JRC-ITU-TN-99/03, 1999.
16 C.T. Walker, G. Nicolaou, Transmutation of neptunium and americium in a fast neutron flux: EPMA results and KORIGEN predictions for the SUPERFACT fuels, J. Nucl. Mater. 218 (2) (1995) 129-138.   DOI
17 G. Grasso, C. Petrovich, D. Mattioli, C. Artioli, P. Sciora, D. Gugiu, G. Bandini, E. Bubelis, K. Mikityuk, The core design of ALFRED, a demonstrator for the European lead-cooled reactors, Nucl. Eng. Des. 278 (2014) 287-301.   DOI
18 A. Gallais-During, F. Delage, S. Bejaoui, S. Lemehov, J. Somers, D. Freis, W. Maschek, S. Van Til, E. D'Agata, C. Sabathier, Outcomes of the PELGRIMM project on Am-bearing fuel in pelletized and spherepac forms, J. Nucl. Mater. 512 (2018) 214-226.   DOI
19 M. Temmar, B. Michel, I. Ramiere, N. Favrie, Multi-physics modelling of the pellet-to-cladding gap closure phenomenon for SFR fuel performance codes, J. Nucl. Mater. 529 (2020), 151909.   DOI
20 C. Ronchi, C. Sari, Swelling analysis of highly rated MX-type LMFBR fuels; I. Restructuring and porosity behaviour, J. Nucl. Mater. 58 (1975) 140-152.   DOI
21 I. Mueller-Lyda, D. Freund, Referenzdaten zum thermischen und mechanischen Verhalten von hochdichtem Mischoxidbrennstoff, Primar Bericht 01/01/04 P43 B, Kernforschungszentrum Karlsruhe, Germany, 1980.
22 V. Di Marcello, V. Rondinella, A. Schubert, J. van de Laar, P. Van Uffelen, Modelling actinide redistribution in mixed oxide fuel for sodium fast reactors, Prog. Nucl. Eng. 72 (2014) 83-90.   DOI
23 M. Kato, K. Maeda, T. Ozawa, M. Kashimura, Y. Kihara, Physical properties and irradiation behavior analysis of Np- and Am-Bearing MOX Fuels, J. Nucl. Sci. Technol. 48 (4) (2011) 646-653.   DOI
24 K. Lassmann, F. Hohlefeld, The revised URGAP model to describe the gap conductance between fuel and cladding, Nucl. Eng. Des. 103 (2) (1987) 215-221.   DOI
25 C. Ronchi, C. Sari, Properties of lenticular pores in UO2, (U,Pu)O2 and PuO2, J. Nucl. Mater. 50 (1) (1974) 91-97.   DOI
26 T.C. Chawla, D.L. Graff, R.C. Borg, G.L. Bordner, D.P. Weber, D. Miller, Thermophysical properties of mixed oxide fuel and stainless steel type 316 for use in transition phase analysis, Nucl. Eng. Des. 67 (1) (1981) 57-74.   DOI
27 A. Magni, T. Barani, L. Luzzi, D. Pizzocri, A. Schubert, P. Van Uffelen, A. Del Nevo, Modelling and assessment of thermal conductivity and melting behaviour of MOX fuel for fast reactor applications, J. Nucl. Mater. 541 (2020) 152410.   DOI
28 C.F. Clement, The movement of lenticular pores in UO2 nuclear fuel elements, J. Nucl. Mater. 68 (1) (1977) 63-68.   DOI
29 K. Maeda, T. Asaga, Change of fuel-to-cladding gap width with the burn-up in FBR MOX fuel irradiated to high burn-up, J. Nucl. Mater. 327 (2004) 1-10.   DOI
30 M. Tourasse, M. Boidron, B. Pasquet, Fission product behaviour in Phenix fuel pins at high burnup, J. Nucl. Mater. 188 (1992) 49-57.   DOI
31 K. Samuelsson, J.C. Dumas, B. Sundman, M. Lainet, An improved method to evaluate the 'Joint Oxyde-Gaine' formation in (U,Pu)O2 irradiated fuels using the GERMINAL V2 code coupled to Calphad thermodynamic computations, EPJ Nucl. Sci. Technol. 6 (2020) 47.   DOI
32 V. Sobolev, S. Lemehov, N. Messaoudi, P. Van Uffelen, H.A. Abderrahim, Modelling the behaviour of oxide fuels containing minor actinides with urania, thoria and zirconia matrices in an accelerator-driven system, J. Nucl. Mater. 319 (2003) 131-141.   DOI
33 S.E. Lemehov, K. Govers, M. Verwerft, Modelling non-standard mixed oxide fuels with the mechanistic code MACROS: Neutronic and heterogeneity effects, in: IAEA-TECDOC-1416, 2003.
34 S.E. Lemehov, M. Verwerft, V. Sobolev, Thermomechanical modeling of prototypic targets containing high concentrations of minor actinides, in: Fuels and Materials for Transmutation, OECD/NEA, 2005. NEA No. 5419.
35 M. Verwerft, S. Lemehov, M. Weber, L. Vermeeren, P. Gouat, V. Kuzminov, V. Sobolev, Y. Parthoens, B. Vos, S. Van Den Berghe, H. Segura, P. Blainpain, J. Somers, G. Toury, J. McGinley, D. Staicu, A. Schubert, P. Van Uffelen, D. Haas, OMICO oxide fuels: microstructure and composition variations final report, External Report SCK.CEN-ER- 42 (2007).
36 A. Cechet, S. Altieri, T. Barani, L. Cognini, S. Lorenzi, A. Magni, D. Pizzocri, L. Luzzi, A new burn-up module for application in fuel performance calculations targeting the helium production rate in (U,Pu)O2 for fast reactors, Nucl. Eng. Technol. (2020) available online, in press.
37 G. Pastore, L. Luzzi, V. Di Marcello, P. Van Uffelen, Physics-based modelling of fission gas swelling and release in UO2 applied to integral fuel rod analysis, Nucl. Eng. Des. 256 (2013) 75-86.   DOI
38 P. Garcia, A. Miard, Availability of Setup for Mechanical Measurements on UO2 at CEA/DEC, INSPYRE Milestone MS8, 2019.
39 P. Martin, G. Jouan, L. Medyk, E. Nilly, Availability of the Micro and Nano Indentation Devices at CEA/DMRC, INSPYRE Milestone MS10, 2019.
40 D. Pizzocri, T. Barani, L. Luzzi, SCIANTIX: a new open source multi-scale code for fission gas behaviour modelling designed for nuclear fuel performance codes, J. Nucl. Mater. 532 (2020) 152042.   DOI
41 T.R. Pavlov, F. Kremer, R. Dubourg, A. Schubert, P. Van Uffelen, Towards a More Detailed Mesoscale Fission Product Analysis in Fuel Performance Codes: a Coupling of the TRANSURANUS and MFPR-F Codes, TopFuel2018 - Reactor Fuel Performance, September 30 - October 4 2018, Prague, Czech Republic.
42 EERA-JPNM, EERA-JPNM website [Online]. Available, http://www.eera-jpnm.eu/, 2017.
43 D. Pizzocri, G. Pastore, T. Barani, A. Magni, L. Luzzi, P. Van Uffelen, S.A. Pitts, A. Alfonsi, J.D. Hales, A model describing intra-granular fission gas behaviour in oxide fuel for advanced engineering tools, J. Nucl. Mater. 502 (2018) 323-330.   DOI
44 S. Lemehov, F. Jutier, Y. Parthoens, B. Vos, S. Van Den Berghe, M. Verwerft, N. Nakae, MACROS benchmark calculations and analysis of fission gas release in MOX with high content of plutonium, Prog. Nucl. Energy 57 (2012) 117-124.   DOI
45 G. Wallez, P.E. Raison, A.L. Smith, N. Clavier, N. Dacheux, High-temperature behavior of dicesium molybdate Cs2MoO4: implications for fast neutron reactors, J. Solid State Chem. 215 (2014) 225-230.   DOI
46 EERA-JPNM, GEMMA European H2020 Project [Online]. Available, http://www.eera-jpnm.eu/gemma/, 2017.
47 NEA, Primary Radiation Damage in Materials, vol. 9, Report NEA/NSC/DOC, 2015.
48 IAEA, Fuel Materials for Fast Reactors [Online]. Available, https://www.iaea.org/projects/crp/t12031, 2019.
49 K. Govers, S. Lemehov, M. Verwerft, M. Hou, Interatomic potentials for atomicscale simulations of UO2, in: Enlarged Halden Programme Group Meeting - 737 Proceedings of the Fuels & Materials Sessions, vol. 2, Halden, Norway, 2007.
50 K. Govers, S. Lemehov, M. Hou, M. Verwerft, Comparison of interatomic potentials for UO2. Part I: static calculations, J. Nucl. Mater. 366 (1-2) (2007) 161-177.   DOI
51 S. Lemehov, V. Sobolev, R. Thetford, Prognosis of thermomechanical behaviour of cercer and cermet fuels in EFIT-400 transmuter, in: Int. Workshop on Technology and Components of Accelerator Driven Systems, March 2010, pp. 15-17. Karlsruhe, Germany.
52 H.A. Abderrahim, D. De Bruyn, M. Dierckx, R. Fernandez, L. Popescu, M. Schyns, A. Stankovskiy, G. Van Den Eynde, D. Vandeplassche, MYRRHA accelerator driven system programme: recent progress and perspectives, Nucl. Power Eng. 2 (2019) 29-41.
53 GIF (Generation IV International Forum), GIF R&D Outlook for Generation IV Nuclear Energy Systems - 2018 Update, 2018.
54 T. Beck, V. Blanc, J.M. Escleine, D. Haubensack, M. Pelletier, M. Phelip, B. Perrin, C. Venard, Conceptual design of ASTRID fuel sub-assemblies, Nucl. Eng. Des. 315 (2017) 51-60.   DOI
55 ESFR-SMART, ESFR-SMART European H2020 Project [Online]. Available, http://esfr-smart.eu/, 2017.
56 L. Luzzi, A. Cammi, V. Di Marcello, S. Lorenzi, D. Pizzocri, P. Van Uffelen, Application of the TRANSURANUS code for the fuel pin design process of the ALFRED reactor, Nucl. Eng. Des. 277 (2014) 173-187.   DOI
57 S. Lemehov, V. Sobolev, P. Van Uffelen, Modelling thermal conductivity and self-irradiation effects in mixed oxide fuels, J. Nucl. Mater. 320 (1-2) (2003) 66-76.   DOI
58 F. Cappia, B.D. Miller, J.A. Aguiar, L. He, D.J. Murray, B.J. Frickey, J.D. Stanek, J.M. Harp, Electron microscopy characterization of fast reactor MOX Joint Oxyde-Gaine (JOG), J. Nucl. Mater. 531 (2020), 151964.   DOI
59 European Commission, TRANSURANUS Handbook, Joint Research Centre, Karlsruhe, Germany, 2020.
60 C. Prunier, F. Boussard, L. Koch, M. Coquerelle, Some specific aspects of homogeneous Am and Np based fuels transmutation through the outcomes of the SUPERFACT experiment in Phenix fast reactor, in: Global'93 Int. Conf, September 1993, pp. 12-17. Seattle, Washington, USA.
61 H. Tobbe, Das Brennstabrechenprogramm IAMBUS zur Auslegung von Schellbruter Brennst aben, Interatom - Technischer Bericht 75 (1975), 65.
62 B. Michel, I. Ramiere, I. Viallard, C. Introini, M. Lainet, N. Chauvin, V. Marelle, A. Boulore, T. Helfer, R. Masson, J. Sercombe, J.C. Dumas, L. Noirot, S. Bernaud, Two fuel performance codes of the PLEIADES platform: ALCYONE and GERMINAL, in: J. Wang, X. Li, C. Allison, J. Hohorst (Eds.), Nuclear Power Plant Design and Analysis Codes - Development, Validation and Application, Woodhead Publishing Series in Energy, Elsevier, 2021, pp. 207-233. Chap. 9.
63 A. Magni, A. Del Nevo, L. Luzzi, D. Rozzia, M. Adorni, A. Schubert, P. Van Uffelen, The TRANSURANUS fuel performance code, in: J. Wang, X. Li, C. Allison, J. Hohorst (Eds.), Nuclear Power Plant Design and Analysis Codes - Development, Validation and Application, Woodhead Publishing Series in Energy, Elsevier, 2021, pp. 161-205. Chap. 8.
64 H.A. Abderrahim, P. Baeten, A. Sneyers, M. Schyns, P. Schuurmans, A. Kochetkov, G. Van Den Eynde, J.-L. Biarrotte, Partitioning and transmutation contribution of MYRRHA to an EU strategy for HLW management and main achievements of MYRRHA related FP7 and H2020 projects: MYRTE, MARISA, MAXSIMA, SEARCH, MAX, FREYA, ARCAS, Nucl. Sci. Technol. 33 (2020) 1-8.
65 K. Lassmann, Uranus - a computer programme for the thermal and mechanical analysis of the fuel rods in a nuclear reactor, Nucl. Eng. Des. 45 (1978) 325-342.   DOI
66 S. Lemehov, V. Sobolev, M. Verwerft, H.A. Abderrahim, Comparative studies of different target designs for minor actinides transmutation, in: GLOBAL 2005 Int. Conf., October 2005, pp. 9-13. Tsukuba, Ibaraki, Japan.
67 K. Govers, D. Terentyev, M. Hou, S. Lemehov, "Molecular dynamics study of mixed oxide fuels : issues and perspectives", in: 43rd Plenary Meeting of the European Working Group - Hot Laboratories and Remote Handling, 2005. Petten, The Netherlands.
68 K. Govers, S. Lemehov, M. Hou, M. Verwerft, Molecular dynamics simulation of helium and oxygen diffusion in UO2±x, J. Nucl. Mater. 395 (1-3) (2009) 131-139.   DOI
69 S.E. Lemehov, V.P. Sobolev, M. Verwerft, Predicting thermo-mechanical behaviour of high minor actinide content composite oxide fuel in a dedicated transmutation facility, J. Nucl. Mater. 416 (1-2) (2011) 179-191.   DOI
70 Y. Philipponneau, Thermal conductivity of (U, Pu)O2-x mixed oxide fuel, J. Nucl. Mater. 188 (C) (1992) 194-197.   DOI
71 L. Luzzi, T. Barani, A. Magni, D. Pizzocri, A. Schubert, P. Van Uffelen, M. Bertolus, V. Marelle, B. Michel, B. Boer, S. Lemehov, A. Del Nevo, Internal report describing the irradiation experiments selected for the assessment of fuel performance codes, INSPYRE Report R7.1 (2019).
72 P. Van Uffelen, G. Pastore, V. Di Marcello, L. Luzzi, Multiscale modelling for the fission gas behaviour in the TRANSURANUS Code, Nucl. Eng. Technol. 43 (6) (2011) 477-488.   DOI
73 GIF (Generation IV International Forum), Annual Report 2019, 2019.
74 A.L. Smith, G. Kauric, L. van Eijck, K. Goubitz, G. Wallez, J.C. Griveau, E. Colineau, N. Clavier, R.J.M. Konings, Structural and thermodynamic study of dicesium molybdate Cs2Mo2O7: implications for fast neutron reactors, J. Solid State Chem. 253 (2017) 89-102. May.   DOI
75 M. Lainet, B. Michel, J.C. Dumas, M. Pelletier, I. Ramiere, GERMINAL, a fuel performance code of the PLEIADES platform to simulate the in-pile behaviour of mixed oxide fuel pins for sodium-cooled fast reactors, J. Nucl. Mater. 516 (2019) 30-53.   DOI
76 H. Matzke, Gas release mechanisms in UO2 - a critical review, Radiat. Eff. 53 (1980) 219-242.   DOI
77 K. Lassmann, H. Benk, Numerical algorithms for intragranular fission gas release, J. Nucl. Mater. 280 (2) (2000) 127-135.   DOI
78 T. Barani, E. Bruschi, D. Pizzocri, G. Pastore, P. Van Uffelen, R.L. Williamson, L. Luzzi, Analysis of transient fission gas behaviour in oxide fuel using BISON and TRANSURANUS, J. Nucl. Mater. 486 (2017) 96-110.   DOI
79 M. Charles, M. Bruet, Gap Conductance in a Fuel Rod: Modelling of the FURET and CONTACT Results, CEA, Centre d'etudes nucl eaires de Grenoble, Grenoble, France, 1984.
80 K. Govers, S. Lemehov, M. Verwerft, In-pile Xe diffusion coefficient in UO2 determined from the modeling of intragranular bubble growth and destruction under irradiation, J. Nucl. Mater. 374 (3) (2008) 461-472.   DOI
81 R. Delville, R&D programme for the fuel qualification of the research fast reactor MYRRHA, in: IAEA-TECDOC-CD-1689, 2011.
82 S. Lemehov, Modeling Fuel Fragmentation and Particle Size Distribution under Normal Operation and Accidental Conditions, Mol, Belgium, 2016.
83 S. Lemehov, MYRRHA MOX Thermo-Physical Properties: Cracked Fuel Mechanics and Axial Thermal Expansion, Mol, Belgium, 2018.
84 V. Di Marcello, A. Schubert, J. Van De Laar, P. Van Uffelen, Extension of the TRANSURANUS plutonium redistribution model for fast reactor performance analysis, Nucl. Eng. Des. 248 (2012) 149-155.   DOI
85 C.F. Clement, M.W. Finnis, Plutonium redistribution in mixed oxide (U, Pu)O2 nuclear fuel elements, J. Nucl. Mater. 75 (1) (1978) 193-200.   DOI
86 Euratom, ESNII+ - Preparing ESNII for HORIZON 2020 [Online]. Available, https://cordis.europa.eu/project/id/605172/, 2013.