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Development and Applications of a General Coupled Thermal-hydraulic/Neutronic Model for the

Licentiatavhandling, 2008

Coupled calculations are important for the simulation of nuclear power plants when there is a strong
feedback between the neutron kinetics
and the thermal-hydraulics. A general coupled model of the Ringhals-3 Pressurized Water Reactor has been
developed for this purpose. The development is outlined in the thesis with details given in the appended papers. A PARCS model
was developed for the core calculations and a RELAP5 model for the thermal-hydraulic calculations. The RELAP5
model has 157 channels for modelling the flow in the fuel assemblies. This means that there is a one-one
correspondence radially between the neutronic and thermal-hydraulic nodalization. This detailed mapping
between the neutron kinetics and the thermal-hydraulics makes it possible to use the model for all kinds of
transient. To provide
realistic material data to the PARCS model, a cross-section interface was developed. With this interface one
can import material data from a binary CASMO-4 library file into PARCS. Due to the one-to-one mapping, any
any core loading can easily be considered. The PARCS model was benchmarked
against
measurements of the steady-state power distribution of Ringhals-3. The power shape was well reproduced by
the model. Validational work for steady-state conditions of the thermal-hydraulic was also successfully
performed. The most challenging part of the validation of a coupled model is for transients.
This is much more difficult since the dynamics of the system becomes very important.
Two transients that occurred at Ringhals-3 were chosen for the validational work. The first transient was
a Load Rejection Transient. In general the model gave good results but some problems were experienced, e.g.
the pressurizer pressure turned out to be more difficult to be correctly simulated. The second transient was
a Loss of Feed
Water transient. A malfunctioning feed water control valve closed, and therefore shut down the feed water
supply to the steam generator in one of the loops. Low level in the affected steam generator led to the SCRAM
of the reactor. Finally, the model was applied to the simulation of hypothetical Main Steam Line Break
transient. Several cases were simulated. Both Hot Full Power conditions and Hot Zero Power
conditions were used. The effects of SCRAM timing, mixing, and delayed neutrons were investigated.

reactor transients

measurements

thermal-hydraulics

core calculations

coupled calculations