exotic nuclei structure and reaction noyaux exotiques ... - IPN - IN2P3

3D coupling of Monte-Carlo neutronics and thermalhydraulics
calculations as a simulation tool for innovative reactor concepts
IPNO Participation: N. Capellan, J.N. Wilson, S. David, O. Meplan, J. Brizi
Collaboration : LPSC Grenoble, Kurtchatov Institut
Les études relatives aux réacteurs nucléaires font appel à plusieurs disciplines dont les principales sont la
neutronique et la thermo-hydraulique. Les phénomènes physiques qui se déroulent dans le cœur d’une
centrale nucléaire comme la réaction en chaîne, le mouvement des fluides et les transferts de chaleur se
couplent de manière forte et complexe. Grâce aux avancées dans ces disciplines et la croissance massive
de la puissance des ordinateurs, cette complexité phénoménologique peut aujourd’hui être simulée en des
temps raisonnables. Le développement d’un couplage externe automatisé entre le code Monte Carlo
MCNP et le code de thermo-hydraulique/thermique COBRA-EN a été effectué. La validation du schéma
couplé a été réalisée sur un cas très complexe de cœur de réacteur et a permis de prouver la robustesse
des développements entrepris et la faisabilité d’un tel couplage.
The typical simulation methods used to study nuclear
reactor cores are neutronics codes which
model the neutron transport and chain reaction
and thermal-hydraulics codes which model the
heat transfer. Historically, these domains have often
been treated separately. However, reactor core
temperatures depend on the heat sources, and
thus on the distribution of power which evolves in
the course of time (calculated by neutronics
codes), yet these same codes require cross sections
which depend on the temperatures, and thus
of the nature of flow (solved by thermal-hydraulics
codes). In certain special energy or temperature
ranges small changes in one field can provoke
large changes in the other. For example the impact
can be large if it occurs for thermal-hydraulic properties
around enthalpy levels of phase change, and
for neutronics properties near neutron resonances.
With the rapid advances in computing power over
the last 30 years it is now possible to not only develop
the coupling of neutronics and thermalhydraulics
phenomena in a single calculation package,
but also to use precision codes for both domains,
including the use of computationally intensive
stochastic (Monte-Carlo) methods for neutronics
calculations. Such Monte-Carlo methods are
preferable to deterministic codes which analytically
solve the Boltzmann equation, because the approximations
made will not necessarily be valid for
innovative future concepts which will operate in
new domains for which there is little or no current
experimental knowledge. Monte-Carlo codes have
the ability to use nuclear data in its most complex
and complete form: continuous energy crosssection,
fine energy–angle correlations, S(α,β) tables
for thermal neutron scattering on molecular
structures.
Increasing focus on safety requirements
drives the need for coupled 3D neutronics / 3D
thermal-hydraulics tools development. Since very
good existing neutronics and thermal-hydraulics
codes have been developed independently and in
parallel, the most obvious methodology for such a
coupling can be internal (i.e. chained calculations
with exchanges of data at each step). Internal coupling
(cf. Fig. 1) requires no modifications of the
existing codes and allows a fully detailed description
of the system. However, a coupling of this type
requires large exchanges of data between the
codes. In our coupled MCNP/COBRA package, the
creation of input files and these data exchanges
are managed automatically by MURE and hidden
from the user. Furthermore, the MURE system
manages the mesh discretization and overlays:
only one geometry need to be created by the user
and it is automatically adapted and correctly formatted
for the two codes.
MURE (MCNP Utilities for Reactor Evolution) is a
code package developed at the IPN Orsay and
LPSC Grenoble written in C++. It consists of
around 25000 lines of code and represents about
15 person-years of development. The primary aim
of the MURE (ref MURE) package is to perform
nuclear reactor time-evolution using the widelyused
particle transport code MCNP5. Evolution of
chosen materials is computed by successive
MCNP calculations followed by numerical integration
of the Bateman’s equations to calculate fuel
depletion. After each MCNP run, the reactor fuel
composition is updated (cf. Fig. 2). Changes in
geometry, temperature, external feeding or fuel
extraction during the evolution can also be performed.
Fig. 2. Principle of fuel evolution in MURE
The COBRA-EN (Coolant Boiling in Rod Arrays)
version has been used for our coupling. This is a
3D sub-channel code that allows steady-state and
transient analysis of the coolant in rod arrays. The
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