A. Keresztúri

Development and validation of the three-dimensional dynamic code—KIKO3D

Abstract A three-dimensional reactor dynamics program—KIKO3D—for coupled neutron kinetics and thermohydraulics calculation of VVER type pressurized water reactor cores has been developed and benchmarked. For solution of the time dependent neutronic equations, a nodal method was used. Concerning the geometry, the symmetries of the nodes, and the concrete form of the neutronic equations to be solved inside the nodes (transport or diffusion equation), the method is general enough, only the linear anisotropy of neutron flux on the node boundaries is utilized. Generalized response matrices for the time dependent problem are introduced which can be derived also from the response matrix of the stationary problem. The so obtained time dependent matrix equations show similar structure to the non-discretized equations. Therefore, the Improved Quasi Static factorization of the time dependent matrix equations can be carried out in the usual way, leading to the point kinetic and the shape function equations. In the KIKO3D code, this general nodal method was applied for the special case of rectangular and hexagonal homogenized nodes in which the diffusion equation is to be solved. In this special case, the traditional response matrices of the stationary problem and the generalized matrices necessary for the time dependent problem can be obtained by analytical formulas. The accuracy of the introduced approximations has been validated against rectangular and hexagonal benchmark problems.

General features and validation of the recent KARATE-440 code system

In the last few years several projects aiming at the introduction of new VVER-440 fuel types and resulting in more economic fuel cycles were initiated: increased average enrichment, modification of the lattice pitch and fuel diameter, profiled enrichment, application of burnable absorber, modification of the absorber assembly coupler part. The first version of the KARATE-440 code system was elaborated in the KFKI Atomic Energy Research Institute (KFKI-AEKI) in the mid-1990s for the core design calculations of VVER-440 type reactors. The above fuel modifications and the upgraded regimes requiring more accurate calculations have necessitated the further development and validation of the code system. Owing to the new fuel types (e.g., burnable poison), greater attention has had to be paid also to the spectral calculations of the assemblies, especially to the space-dependent thermalisation in a large heterogeneous system. The fast algorithm applied to solve this problem with satisfactory accuracy is also detailed. The first part of the paper outlines the general features of KARATE, while in the next part the validation results are presented in relation to zero reactor measurements, mathematical benchmark problems and the operational data of the Paks (Hungary) Nuclear Power Plant (NPP).

First experience with a six-loop nodalization of a VVER-440 using a new coupled neutronic-thermohydraulics system KIKO3D-RETINA V1.1D

Abstract In this paper, we introduce a new, coupled neutronic-thermohydraulics system. The three-dimensional neutron kinetic code KIKO3D and the two-phase flow code RETINA V1.1D have been coupled for modeling complex transients of nuclear power plants. Using a six-loop nodalization of a VVER-440, several test calculations have been carried out. Results obtained for a trip of one main circulation pump are compared with real measurements and reference calculations provided by other neutronic-thermohydraulics systems. The ability of our coupled system is demonstrated.

Real-Time 3D Simulation of a Pressurized Water Nuclear Reactor

Fuel assemblies are very expensive parts of the nuclear reactor. Initially they were used in Hungary for 3 years, now for 4 years and soon they will stay in the core for 5 years. Each year only 1/3rd, 1/4th later 1/5th of them is replaced, therefore the change of the fuel type is a lengthy process, with mixed cores used. The authorities require that the staff should be trained to each particular core before they operate it. For this reason the simulator should be upgraded to simulate the exact behavior of each core foreseen for the next 5 years. The RETINA code (Reactor Thermo-hydraulics Interactive) is a 3D offline code, developed in our department. KIKO3D - Neutron Kinetics 3D - has been developed in our Institute, too, in the Reactor Analysis Department. Both of them should be integrated into our full-scope replica simulator, coupled, and stressed to operate parallel in real-time, using four hi-power processors. The simulation-specific details are discussed in the paper.

Qualification of the APOLLO2 lattice physics code of the NURISP platform for VVER hexagonal lattices

Abstract The experiments performed at the ZR-6 zero power critical reactor by the Temporary International Collective (TIC) and a burnup benchmark specified for depletion calculation of a VVER-440 assembly containing Gd burnable poison were used to qualify the APOLLO2.8-3.E (APOLLO2) code as a part of its ongoing validation activity. The work is part of the NURISP project, where KFKI AEKI undertook to develop and qualify some calculation schemes for hexagonal problems. Concerning the ZR-6 measurements, single cell, macro-cell and 2D calculations of selected regular and perturbed experiments are used for the validation. In the 2D cases, the radial leakage is also taken into account by the axial leakage represented by the measured axial buckling. Criticality parameter and reaction rate comparisons are presented. Although various sets of the experiments have been selected for the validation, good agreement of the measured and calculated parameters could be found by using the various options offered by APOLLO2. An additional mathematical benchmark – presented in the paper – also attests for the reliability of APOLLO2. All the test results prove the reliability of APOLLO2 for VVER core calculations.

Hot channel calculation methodologies in case of VVER-1000/1200 reactors

Abstract In VVER-1000/1200 type reactors much higher and geometrically more complex assemblies are applied compared to the VVER-440 design. Additionally, assembly shrouds are not used in the improved reactor types. This implies that the existing sub-channel wise models and hot channel calculation methodologies used in the safety assessment of VVER-440 reactors have to be improved. For this purpose a full core thermal hydraulic model was developed. In the first part of the paper the new model and its connections with the reactor physics of KARATE-1200 code system are demonstrated with results regarding the sub-channel outlet temperature and DNBR. In the second part of the paper the hot channel calculation methodologies to be used in the transient safety analyses of VVER-1200 type reactors are discussed. The appropriate frame parameter concept concerning the limitation of the heating up of the coolant and the single closed sub-channel approach vs. the multi-channel approach are studied, as well.

Application of discontinuity factors and group constants generated by SERPENT in the KIKO3 DMG code

Abstract Discontinuity factors and node-wise group constants were generated by the SERPENT Monte Carlo code and applied in the KIKO3 DMG nodal code. The methodology was tested by calculating a typical VVER-440 calculation benchmark. A reference solution of the benchmark was calculated by using also the SERPENT code and the accuracy of the different approaches was checked against this latter solution.

A methodology for the estimation of the radiological consequences of a Loss of Coolant Accident

Abstract For calculation of the radiological consequences of Large Break Loss of Coolant (LBLOCA) events, a set of various computer codes modeling the corresponding physical processes, disciplines and their appropriate subsequent data exchange are necessary. For demonstrating the methodology applied in MTA EK, a LBLOCA event at shut down reactor state – when only limited configuration of the Emergency Core Cooling System (ECCS) is available – was selected. In this special case, fission gas release from a number of fuel pins is obtained from the analyses. This paper describes the initiating event and the corresponding thermal hydraulic calculations and the further physical processes, the necessary models and computer codes and their connections. Additionally the applied conservative assumptions and the Best Estimate Plus Uncertainty (B+U) evaluation applied for characterizing the pin power and burnup distribution in the core are presented. Also, the fuel behavior processes. Finally, the newly developed methodology to predict whether the fuel pins are getting in-hermetic or not is described and the the results of the activity transport and dose calculations are shown.

Uncertainties of the KIKO3D-ATHLET calculations using the Kalinin-3 benchmark (Phase II) data

Abstract The best estimate simulation of three-dimensional phenomena in nuclear reactor cores requires the use of coupled neutron physics and thermal-hydraulics calculations. However these analyses should be supplemented by the survey of the corresponding uncertainties. In this paper the uncertainties of the coupled KIKO3D-ATHLET calculations are presented for a VVER-1000 type core using the OECD NEA Kalinin-3 (Phase II) benchmark data, although only the neutronic uncertainties are considered and further simplifications are applied and discussed. Additionally, this study has been performed in the conjunction with the OECD NEA UAM benchmark, as well. In the first part of the paper, the uncertainties of the effective multiplication factor, the assembly-wise radial power distribution, the axial power distribution, the rod worth, etc. are presented at steady-state. After that some uncertainties of the transient calculations are discussed for the considered switch-off of one Main Circulation Pump (MCP) type transient.

Investigation of the hot-channel calculation methodology in case of shroud-less assemblies

Abstract The fulfillment of the safety analysis acceptance criteria and the normal operation safety related limitations are usually evaluated by using separate hot-channel or/and hot-assembly thermal hydraulic calculations, especially for the closed VVER-440 assemblies. However, the shroud-less assemblies of the expected new NPP units in Hungary or even those of some recently foreseen VVER-440 are raising several questions; like what is the role of the coolant mixing between the neighboring assemblies and can the closed assembly approximation be regarded as conservative for DBA analyses. Focusing an anticipated transient without scram (ATWS) event, the paper evaluates the thermal hydraulic calculation methodology of the shroud-less assemblies by supplementing the computational domain step by step with the closest pins and subchannels of the neighboring assemblies. We concluded that assembly-wise coolant cross-flows and significant changes in the hot-channel DNBR can appear in case of shroud-less assemblies.