Modeling the spatial distribution of the parameters of the coolant in the reactor volume
21st Symposium of AER on VVER Reactor Physics and Reactor Safety (2011, Dresden, Germany)
Reactor dynamics and safety analysis
Abstract
In this paper the approach to the question about the spatial distribution of the parameters of the
coolant in-reactor volume. To describe the in-core space is used specially developed preprocessor.
When the work of the preprocessor in the first place, is recreated on the basis of available information
(mostly – the original drawings) with high accuracy three-dimensional description of the structures of
the reactor volume and, secondly, are prepared on this basis blocks input to the nodal system code
improved estimate ATHLET, allows to take into account the hydrodynamic interaction between the
spatial control volumes. As an example the special case of solutions of international standard problem
on the reconstruction of the transition process in the third unit of the Kalinin nuclear power plant, due
to the shutdown of one of the four Main Coolant Pumps (MCP) in operation at the rated capacity (first
download). Model-core area consists of approximately 58 000 control volumes and spatial
relationships. It shows the influence of certain structural units of the core to the distribution of the
mass floe rate of its height. It is detected a strong cross-flow coolant in the area over the baffle.
Moreover, we study the distribution of the coolant temperature at the assembly head of VVER-1000
reactor. It is shown that in the region of the top of the assembly head, where we have installation of
thermocouples, the flow coolant for internal assemblies core is formed by only from guide channel
Reactor control and protected system (CPS) Control rod (CR) flow, or a mixture of the guide channel
flow and flow from the area in front of top grid head assembly (the peripheral assemblies). It is shown
that the magnitude of the flow guide channels affects not only the position of control rods, but also the
presence of a particular type of measuring channels (Self powered neutron detector sensors (SPND) or
Temperature control (TC) sensors) in the cassette.
The given work represents generalization of the first results under the analysis of spatial
distribution of parameters of the heat-assemblies in reactor pressure volume, received with use of
specially developed spatial model 1 for inside spaces of reactor WWER-1000 for code ATHLET 2.
Object of research is the block ?3 Kalinin NPP during development work on nominal power. The
mode with switching-off of the main circulating pump (MCP) on the first loop is considered at work of
a reactor on 98,6 % of power (assemble TVSA, the first loading, 126 effective days). Discussed results
are included into a circle of the problems making the international standard problem, based on the
specified transient which detailed description is resulted in 3.
The essence of construction of spatial model is clear from the summary resulted above. Rather
well the procedure for model construction is described in works [4-6]. It is necessary to note, that in
this paper is considered only hydrodynamic problem for inside spaces of the reactor. To except of the
uncertainties, connected with calculation spatial neutron kinetic, the experimental data on spatial
power distribution, received directly from NPP during work an investigated time interval are used.
Besides these were used as boundary conditions the experimental dada for the coolant temperature at
the reactor input of corresponding loops.
Here it is necessary to specify, that we shall mean temperature of the coolant entrance data in an
installation site of gauges of temperature on an entrance stream of the coolant, i.e. at a reverser of the
coolant in the loop with switch off MCP the entrance temperature in a reactor for this loop there will
be a temperature in a hot leg , and on working loops ? temperature in a cold leg. Thus, the second
contour is excluded from consideration.
Geometrical characteristics of the main circulating loops, collectors and tubing of steam generators
are completely kept. Maintenance of volume of the coolant in the first circuit it is supervised under
actual readings of the gauge of a level in the pressurizer. Experimental dada for pressure an output of a