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VTT TECHNICAL RESEARCH CENTRE OF FINLAND LTD
PALAMA
SAFIR2014 Final seminar
19th – 20th March 2015
Ville Tulkki
Polttoaineen laaja-alainen mallinnus
PALAMA
 Project focused on analysis on
nuclear fuel behaviour
 4 years, 12.1 person-years, 14
researchers, 1.45 M€
 Deliverables include 36
technical reports, 15
conference papers, 7 journal
papers, 2 Master’s theses and
1 Licentiate thesis
17/04/2015
 Asko Arkoma, Santtu
Huotilainen, Silja Häkkinen,
Anitta Hämäläinen, Timo
Ikonen, Seppo Kelppe,
Joonas Kättö, Henri
Loukusa, Anna Nieminen,
Jan-Olof Stengård, Elina
Syrjälahti, Ville Tulkki, Ville
Valtavirta, Tuomas Viitanen
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PALAMA Highlights
 Completion of methodology for statistical assesment of the number of
failed rods during large break LOCA
 Methodology development initiated in SAFIR2010 POKEVA
 Development of FINIX fuel module
 Improves fuel description in wide variety of host codes
 New methodology to describe cladding response to stresses
 To replace widely used strain hardening rule
17/04/2015
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Statistical evaluation of
failed rods during large
break LOCA
Statistical methodology
 Holistic approach that provides a fraction of failed rods
 95% confidence that in 95% of LB LOCAs the actual number will be the
same or less
 59 global APROS LB LOCA simulations, 1000 local fuel performance
simulations per global case
 Also power histories from reactor physics codes required
 First application: EPR analysis performed for STUK
 STUK agreed to allow the publication of the results
 A. Arkoma et al., Nuclear Engineering and Design Volume 285, 15 April
2015, Pages 1–14
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Calculation system
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Result
 At 95/95 confidence and probability, at most 1.2% of all the rods in the
reactor fail
 10% maximum allowed for this kind of scenario
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FINIX fuel module
FINIX – fuel behavior model and interface for
multiphysix applications
Reactor
physics
clad temperature
Fuel
behavior
17/04/2015
coolant conditions
Thermal
hydraulics
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FINIX – fuel behavior model and interface for
multiphysix applications
Reactor
physics
clad temperature
FINIX
17/04/2015
coolant conditions
Thermal
hydraulics
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FINIX models
 Coupled thermomechanical
models
 Solves the time-dependent radial
heat equation, in several
independent axial nodes
 Rigid pellet (no elastic
deformations)
 Cladding modeled as a thinwalled tube
 Thermal and mechanical solutions
coupled through gap conductance
and gap pressure
 Burnup effects can be taken into
account in the initial state if given
by, e.g., FRAPCON
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FINIX coupling
Fast transient with Serpent 2
 Currently coupled to
 Serpent 2
 TRAB-1D/3D and Hextran
 Provides realistic description of
fuel behaviour
TMI-1 power peak with TRAB-1D
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FINIX
 More information on FINIX in PALAMA special
article in the SAFIR2014 proceedings
 Also: ”Module for thermomechanical modeling
of LWR fuel in multiphysics simulations”, T.
Ikonen et al. Annals of Nuclear Energy (2014),
accepted manuscript, in press.
 FINIX currently licensed to Aalto, KIT
(Germany), Necsa (S. Africa) and CV Rez
(Czech Republic)
 FINIX development continues in SAFIR2018
PANCHO project
17/04/2015
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Cladding viscoelastic
response
Cladding creep response to changing
conditions
 Cladding protects the fuel pellets
from the environment and contains
radioactive nuclides
 Creep deformation at high
temperature and pressure
 Commonly assumed that changes in
conditions can be accounted by
strain hardening rule
 Accumulated strain invariant during
changes
 Experiments show this is not true
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Viscoelastic model
 Metal anelastic behaviour
 Slow reversible deformation
 Usually assumed insignificant
 Irradiation and fuel behaviour
make it relevant in cladding
 Can be illustrated with mechanical
analogs
 Springs and dashpots parallel
and in series
 Anelastic + permanent strain =
viscoelasticity
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Cladding creep
response
 Viscoelastic model enables
reproduction of cladding response
to transient stresses
V. Tulkki, T. Ikonen, Journal of Nuclear Materials,
445 (2014) pp. 98-103.
17/04/2015
V. Tulkki, T. Ikonen Journal of Nuclear Materials,
457 (2015) , pp. 324-329
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Stress relaxation behaviour
 Viscoelastic model also
capable of simulating stress
relaxation behaviour
 Figure displays model
performance in stress
relaxation with coefficients
fitted to creep experiment
 Doctoral thesis ”Modelling fuel
behaviour and cladding
viscoelastic response” due this
summer
V. Tulkki, T. Ikonen, “Modelling anelastic contribution to
nuclear fuel cladding creep and stress relaxation”,
submitted to Journal of Nuclear Materials (2015)
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Summary
 PALAMA was 4 year project focused on fuel
performance issues
 Wide variety of case studies, model
development, code development and analysis
 Several highlights presented here
 Completion of methodology for statistical
assesment of the number of failed rods during
large break LOCA
 Development of FINIX fuel module
 New methodology to describe cladding response
to stresses
 More can be found from the SAFIR2014 Final
seminar proceedings
17/04/2015
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