Browsing by Author "Lapins, Janis"

Now showing 1 - 1 of 1

Open Access
An advanced three-dimensional simulation system for safety analysis of gas cooled reactors
(2016) Lapins, Janis; Laurien, Eckart (Prof. Dr.-Ing. habil.)
The neutron transport programme TORT-TD that solves the neutron transport equation in discrete ordinates for stationary as well as transient problems is used for neutronics calculations. The transient solution of the neutron transport equation is performed by making use of a time-dependent neutron source, xenon/iodine dynamics are implemented as well. The programme ATTICA3D applies the porous medium approach for flow in packed beds according to Ergun. This approach uses a quasi-steady state formulation for the momentum equation while time dependent formulations are employed for mass conservation, and energy conservation for both, the solid and gaseous phase. For spatial discretisation of the conservation equations, the finite volume method is used. For material properties, gas densities, heat transfer etc. a set of constitutive equations completes the set of differential equations. Time integration in ATTICA3D is realised applying a modified Newton-Raphson method which linearizes and subsequently solves the set of equations. It can automatically adapt the time step width within user specified limits. Within this work, the mass and energy conservation equations are modified so that chemical reactions as consequence of water or air ingress can be simulated, i.e. mass sources for CO, CO2, H2 and sinks for H2O and O2 or heat sources and enthalpy transport. The heat generated by chemical reaction is either added to the solid or the gaseous phase. The corrosion rates were implemented according to experimental findings for fuel and reflector graphite. Steam or air might enter the primary circuit through a break in the steam generator or an opening of the primary circuit. Steam entering the core region will moderate neutrons, reduce the leakage and thereby increase power. The corrosion was validated for the NACOK experiment performed within the RAPHAEL project where temperature evolution under corrosion and total burn-off had to be simulated simultaneously. TORT-TD and ATTICA3D exchange data (power distributions or fuel and moderator temperature distribution, possibly hydrogen distribution) by means of a common interface that interpolates values that are exchanged on mutual computational grids by volumetric averaging. As verification for the proper operation of the interface, the steady state of the transient PBMR-400 benchmark was used. After obtaining a coupled steady state, the transient exercises are performed to test the proper working of the interface in time dependent cases. Here, the cold helium ingress, the total control rod withdrawal case and the total control rod ejection case were simulated and compared to results of other partakers of the benchmark. Also, the coupled system was validated for a full power temperature distribution experiment in the Chinese experimental reactor HTR-10 where good agreement could be reached with the measurements. The coupled HTR simulation system TORT-TD/ATTICA3D was then applied for single control rod ejection cases for both the PBMR-400 and the HTR-PM. These cases require a 180° model of the reactor. As preparatory works, the control rod cross sections were adjusted to yield the same reactivity increase as the grey curtain model for the PBMR and with MCNP5 for the HTR-PM. Since there are strong shielding effects by neighbouring rods, the power increase was moderate due to strong Doppler and moderator feedbacks. For the HTR-PM, coupled calculations for water ingress cases are simulated. This also tested the whole computational sequence, i.e. steam transport into the core by ATTICA3D, then transfer of hydrogen densities (from hydrogen or from steam) to TORT-TD via the interface, interpolation of the macroscopic cross sections which changes the power density, and the feedback to ATTICA3D. Additionally, an anticipated transient without scram is simulated where shutdown of the reactor is achieved by the temperature feedback effects. For both, the design basis accident and the anticipated transient without scram, the power increases - lacking experimental results - were compared to published results produced with the TINTE code and are, again, in good agreement.