  Whether for shielding applications or for criticality safety studies, solving
the neutron transport equation with good accuracy requires to take into account
the resonant structure of cross sections in part of the Unresolved Resonance
Region (URR). In this energy range even if the resonances can no longer be
resolved experimentally, neglecting them can lead to significant numerical
biases, namely in flux-based quantities. In Geant4, low energy neutrons are
transported using evaluated nuclear data libraries handled by the Neutron
High-Precision (Neutron-HP) package. In the version 11.01.p02 of the code, the
URR can only be described by average smooth cross sections that do not take
into account the statistical resonant structure of the cross sections. To
overcome this shortcoming, the treatment of the URR with the use of the
probability table method has been implemented in Geant4 and successfully
validated with the reference Monte Carlo neutron transport codes MCNP6 (version
6.2) and Tripoli-4 (version 12). These developments will be taken into account
in the next release of Geant4. All the validations of Geant4 have been
performed with probability tables generated from both the NJOY and CALENDF
pre-processing tools. Therefore Geant4 now has this unique feature to study the
relative impact of the strategies involved during the production of probability
table by the two pre-processing codes. This has been used to show that
self-shielding is important also for inelastic cross sections in the example of
238U. The tool to generate probability tables usable by Geant4 either from NJOY
or from CALENDF is made available on a dedicated GitLab repository and will be
included in Geant4.
