The three-dimensional neutron kinetics coupled with thermal-hydraulics in RBMK accident analysis

The RBMK core is a complex ensemble of high-pressure high-temperature tubes, graphite bricks, low-pressure low-temperature control rod tubes, graphite interstitial gas passages. An about 7 MPa boiling light water crosses the around 19 m long vertical tubes (7 m active length). The lattice consisting of graphite columns and hydraulic channels is bounded by the reactor cavity whose resistant elements are the metal cylindrical tank and thick circular top and bottom plates with proper holes for the passage of tubes. Related to a typical water cooled reactor, the peculiarities of the RBMK core can be summarized as follows: (a) large dimensions – the overall core volume is by far the largest for a nuclear power plant (NPP) producing electricity; (b) use of separate moderator and coolant constituted by graphite and light boiling water, respectively – the boiling water mostly absorbs neutrons in this environment leading to the (small) positive void reactivity coefficient; (c) presence of water channels very close to each other containing coolant at different temperatures (543–557 K and 350 K for fuel channels (FC) and control and protection system (CPS) channels, respectively); (d) presence of core-wide radial, core-wide axial and local temperature gradients in the graphite bricks with temperature values in the range 330–650 K with the high-temperature values justified by the neutron moderation and gamma-heating processes. Owing to the above peculiarities, the development and the use of a three-dimensional neutron kinetics code (3D NK) coupled with a one- dimensional thermal-hydraulic (TH) code is essential in RBMK safety analyses. Two approaches have been used within the present context, i.e. use of coupled 3D NK-TH codes to support the accident analysis in the RBMK as discussed in the first of the companion papers in this journal volume: application of Korsar-Bars making use of the Unk code to derive ␭-matrices needed for Bars and of Relap5/3D-Nestle making use of the Helios code to derive the macroscopic cross-sections. Bounding transient analyses of accident scenarios including control rod withdrawal, various Loss of Coolant Accident (LOCA) and discharge of the control rod circuit, have been completed. In all of the analysed cases, starting from nominal operating conditions, modest fission power time gradients have been found, i.e. characterized by time derivative values for local and global power changes substantially smaller than current values accepted in safety analyses of light water reactors