Submission Title
Coupling of 3D Space Time Neutronics with Thermal Hydraulics for Safety Analysis of Large PHWRs
Presentation Type
Invited
Start Date
18-12-2018 11:40 AM
Abstract
Present day large sized power reactors are of neutronically loosely coupled type. Deliberate flattening of the power distribution further enhances the looseness of the coupling. The flux distribution is flattened axially and radially to maximize power output in a large reactor without exceeding operating limits on maximum fuel rating. Flux flattening tends to enhance neutronic decoupling of core regions. The effect of neutronic decoupling is to accentuate reactivity effects associated with localized perturbations.
The point reactor model (PRM) results are not only inaccurate in important cases, but also generally non-conservative. Further, during power transients, the core power distribution changes quickly enough so that the delayed neutrons cannot be assumed to have the same spatial distribution as the prompt neutrons.
The reactivity transients following a Loss of Coolant Accident (LOCA) in a Pressurized Heavy Water Reactor (PHWR) are delayed supercritical. The shape of the delayed neutron distribution is similar to the power shape existing in the core prior to the event. Its effect is to retard the establishment of the asymptotic flux shapes following the initial perturbation. Also large PHWRs have two or more independent coolant loops and LOCA leads to voiding in one radial half of the reactor. The study of the neutronic transient behavior under accidental conditions in HWRs requires accurate method of solutions of multi-dimensional multi-energy group time- dependent neutron diffusion equation coupled with thermal hydraulics code.
We have developed space time kinetics code IQS3D which is based on improved quasi static method. Finite difference solution based on SLOR along with chebyshev acceleratoion or Krylov method is obtained. Two group neutron cross-section database is generated using lattice codes using 69 or 172 energy groups and supercell codes for control rod homogenisation employing few-group transport theory methods. Our physics codes are validated using observed data from different reactors and also by participating in several international benchmark exercises. We have also validated IQS3D with AECL-7236 benchmark solution which is a more realistic representation CANDU PHW reactor. The transient given for 3-D benchmark represents a non-uniform loss-of-coolant-accident (LOCA) and a subsequent asymmetric insertion of shut-off reactivity devices.
The physics code IQS3D is coupled to thermal hydraulic code ATMIKA for accident analysis as well as for transient studies. The safety analyses required to be submitted for licensing of 540 MWe and 700 MWe PHWR have been carried out using this code system. The presentation will cover the development of IQS3D and its coupling with thermal hydraulics code ATMIKA for transient and safety analyses.
Recommended Citation
Fernando, M. P.S. (2018). "Coupling of 3D Space Time Neutronics with Thermal Hydraulics for Safety Analysis of Large PHWRs," Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety.
Coupling of 3D Space Time Neutronics with Thermal Hydraulics for Safety Analysis of Large PHWRs
Present day large sized power reactors are of neutronically loosely coupled type. Deliberate flattening of the power distribution further enhances the looseness of the coupling. The flux distribution is flattened axially and radially to maximize power output in a large reactor without exceeding operating limits on maximum fuel rating. Flux flattening tends to enhance neutronic decoupling of core regions. The effect of neutronic decoupling is to accentuate reactivity effects associated with localized perturbations.
The point reactor model (PRM) results are not only inaccurate in important cases, but also generally non-conservative. Further, during power transients, the core power distribution changes quickly enough so that the delayed neutrons cannot be assumed to have the same spatial distribution as the prompt neutrons.
The reactivity transients following a Loss of Coolant Accident (LOCA) in a Pressurized Heavy Water Reactor (PHWR) are delayed supercritical. The shape of the delayed neutron distribution is similar to the power shape existing in the core prior to the event. Its effect is to retard the establishment of the asymptotic flux shapes following the initial perturbation. Also large PHWRs have two or more independent coolant loops and LOCA leads to voiding in one radial half of the reactor. The study of the neutronic transient behavior under accidental conditions in HWRs requires accurate method of solutions of multi-dimensional multi-energy group time- dependent neutron diffusion equation coupled with thermal hydraulics code.
We have developed space time kinetics code IQS3D which is based on improved quasi static method. Finite difference solution based on SLOR along with chebyshev acceleratoion or Krylov method is obtained. Two group neutron cross-section database is generated using lattice codes using 69 or 172 energy groups and supercell codes for control rod homogenisation employing few-group transport theory methods. Our physics codes are validated using observed data from different reactors and also by participating in several international benchmark exercises. We have also validated IQS3D with AECL-7236 benchmark solution which is a more realistic representation CANDU PHW reactor. The transient given for 3-D benchmark represents a non-uniform loss-of-coolant-accident (LOCA) and a subsequent asymmetric insertion of shut-off reactivity devices.
The physics code IQS3D is coupled to thermal hydraulic code ATMIKA for accident analysis as well as for transient studies. The safety analyses required to be submitted for licensing of 540 MWe and 700 MWe PHWR have been carried out using this code system. The presentation will cover the development of IQS3D and its coupling with thermal hydraulics code ATMIKA for transient and safety analyses.
