Simulation of mixing effects in a VVER-1000 reactor

This work has been performed in the framework of the OECD/NEA thermalhydraulic benchmark V1000CT-2. This benchmark is related to fluid mixing in the reactor vessel during a MSLB accident scenario in a VVER-1000 reactor. Coolant mixing in a VVER-1000 V320 reactor was investigated in plant experiments during the commissioning of the Unit 6 of the Kozloduy nuclear power plant. Non-uniform and asymmetric loop flow mixing in the reactor vessel has been observed in the event of symmetric main coolant pump operation. For certain flow conditions, the experimental evidence of an azimuthal shift of the main loop flows with respect to the cold leg axes (swirl) was found.Such asymmetric flow distribution was analyzed with the Trio_U code. Trio_U is a CFD code developed by the CEA Grenoble, aimed to supply an efficient computational tool to simulate transient thermalhydraulic turbulent flows encountered in nuclear systems. For the presented study, a LES approach was used to simulate turbulent mixing. Therefore, a very precise tetrahedral mesh with more than 10 million control volumes has been created.The Trio_U calculation has correctly reproduced the measured rotation of the flow when the CAD data of the constructed reactor pressure vessel where used. This is also true for the comparison of cold leg to assembly mixing coefficients. Using the design data, the calculated swirl was significantly underestimated. Due to this result, it might be possible to improve with CFD calculations the lower plenum flow mixing matrices which are usually used in system codes.     

Detailed CFX-5 study of the coolant mixing within the reactor pressure vessel of a VVER-1000 reactor during a non-symmetrical heat-up test

As part of the reactor dynamics activities of FZK/IRS, the qualification of a detailed 3D CFD model of a reactor pressure vessel is a key step in safety evaluations for improving predictive capabilities and acceptability of commercial CFD tools in reactor physics. The VVER-1000 Coolant Transient Benchmark, initiated by OECD, represents an excellent opportunity for validation. In this work a CFD model for the complete VVER-1000 reactor pressure vessel is presented. Due to computational limits simplifications of the core and of some other geometrical details are introduced. The simulated scenario is the heat-up of one coolant loop in case of the isolation of a steam generator while the reactor is operating at a low power level. Two transient runs with a first and second order approximation of the spatial discretization are performed. Unexpectedly, the first order method reveals better agreement with measured data.     

Neutronic simulation of water-based nanofluids as a coolant in VVER-1000 reactor

The main goal of this study is to perform the neutronic simulation of nanofluids application to reactor core. The variation of the Bushehr VVER-1000 reactor primary neutronics parameters is investigated with using different nanofluids as coolant. In the present neutronic simulation, water-based nanofluids containing various volume fractions of Al₂O₃, Si, Zr, TiO₂, CuO, Ti, Cu and Ag nanoparticles are investigated. Optimization of type and volume fraction of nanoparticles affects the reactor neutronic characteristics. The results achieved by using WIMS and CITATION codes, show that below 0.1 percent volume fraction of Al₂O₃ is the optimum nanoparticle for normal operation and Ag/water nanofluid is suggested to use as a reactor safety enhancement tool.     

CFD study on coolant mixing in VVER-440 fuel rod bundles and fuel assembly heads

A CFD model of VVER-440 fuel assembly heads was developed based on the technical documentation of a full-scale test facility built in the Kurchatov Institute, Russia. Steady-state and transient calculations were performed to validate the model with a measurement set. Effects of the spatial resolution, turbulence models, difference schemes and different inlet boundary conditions were investigated. Inlet boundary conditions were determined with both the COBRA subchannel code and a fuel rod bundle CFD model that was built for this special purpose. The results were compared against experimental data. The sensitivity studies showed that a grid of about 8 million cells, high resolution scheme and BSL Reynolds stress model are suitable sets to provide accurate prediction for the signal of the in-core thermocouple. The best prediction was achieved with transient calculation using inlet boundary conditions generated with the CFD fuel rod bundle model. The results indicated that the coolant mixing is intensive but not perfect in the assembly head. Besides, the significant role of the outflow from the central tube was also proven. The transient runs revealed relatively large temperature fluctuations near the in-core thermocouple housing.     

Hydrodynamics of coolant flow in the circulational loop of atomic power plants with an RBMK-1000 reactor

All-Union Scientific-Research Institute of Atomic Power Stations. Translated from Atomnaya Énergiya, Vol. 71, No. 4, pp. 287–293, October, 1991.     

Application of global optimization to VVER-1000 reactor diagnostics

Problems of reactor equipment diagnostics are formulated as inverse eigenvalue problems. Numerical methods of solving the inverse problems are presented. Incompleteness of spectral data results in the error function being non-convex. As the function has numerous local minima, it is necessary to use global optimization methods. Two different strategies are discussed: the modified TRUST algorithm and the algorithm that reduces the original problem to a one-dimensional form. The outcome algorithm that combines the strategies is proposed. Results of computational experiments are presented to illustrate the efficiency of the approach.     

HEXTRAN–SMABRE calculation of the VVER-1000 coolant transient-1 benchmark

The VVER-1000 coolant transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three-dimensional neutron kinetic core models. It concerns switching on a main coolant pump when the other three main coolant pumps are in operation. The problem is based on an experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, an extreme scenario concerning a control rod ejection after switching on a main coolant pump was calculated. At VTT the three-dimensional advanced nodal code HEXTRAN is used for the core dynamics, and the system code SMABRE as a thermal hydraulic model for the primary and secondary loop. The parallelly coupled HEXTRAN–SMABRE code has been in production use since early 1990s, and it has been extensively used for analyses of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used at VTT. The whole core calculation is performed with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation were specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Parametric studies have been performed for selected parameters.     

Application of coulped code method for the analysis of VVER-1000 coolant transient benchmark

In the light of the sustained development in computer technology, the possibilities of code capabilities have been enlarged substantially. Consequently, advanced safety evaluations and design optimizations, which were not possible few years ago, can now be performed. Nowadays, it becomes possible to switch to new generation of computational tools, namely, coupled code (CC) technique. The application of such method is mandatory for the analysis of transient events where strong coupling between the core neutronics and the primary circuit thermal-hydraulics exits, and more especially when asymmetrical processes take place in the core leading to local space-dependent power generation. Through the current study, a demonstration of the maturity level achieved in the calculation of 3-D core performance during complex accident scenarios is emphasized. The study is followed by a typical application through which the main features and limitations of this technique are discussed.     

Comparison of microstructural features of radiation embrittlement of VVER-440 and VVER-1000 reactor pressure vessel steels

Comparative microstructural studies of both surveillance specimens and reactor pressure vessel (RPV) materials of VVER-440 and VVER-1000 light water reactor systems have been carried out, following irradiation to different fast neutron fluences and of the heat treatment for extended periods at the operating temperatures. It is shown that there are several microstructural features in the radiation embrittlement of VVER-1000 steels compared to VVER-440 RPV steels that can cause changes in the contributions of different radiation embrittlement mechanisms for VVER-1000 steel.     

Neutronic study of nanofluids application to VVER-1000

The change in neutronic parameters of the VVER-1000 nuclear reactor core attributable to the use of nanoparticle/water (nanofluid) as coolant is presented in this paper. Optimization of type and volume fraction of nanoparticles in water that affect the safety enhancement of core primary parameters is intended in this study. Reactivity change, radial and axial local peaking factors (LPPF), and the consequence of nanoparticle deposition on fuel clad are investigated.     

Xenon transient simulation of the VVER-1000 nuclear reactor using adiabatic approximation

The neutronic behavior of very slow transients like fuel burn up and xenon studies can be performed with the sequence of instantaneous criticality calculations. Such a scheme is known as the adiabatic approximation. ¹³⁵Xe as a fission product has an enormous thermal absorption cross section, on the order of a million barns, therefore the study of xenon poisoning and its effect on flux and power distribution is very important in thermal reactors. In this work xenon transient analysis of the VVER-1000 nuclear reactor and its effect on the flux and power distribution from reactor start up to xenon saturation and the change of power from nominal to 25% of nominal is carried out using WIMS and CITATION codes. We used the WIMS cell calculation code and found some relations between xenon concentration and group constants of different FA (Fuel Assemblies); in this way we bypassed the WIMS running at each time step. Also, the CITATION code is used as a core calculation code to find the effective multiplication factor as well as flux and power distributions. In order to link WIMS and CITATION codes and facilitate numerous executions, a VISUAL BASIC program has been developed. The results have a good agreement with the safety analysis report of the reference plant such that the relative differences in most cases are less than 10%.     

AP1000反应堆主泵屏蔽套制造工艺浅析

简要地从材料、成形、焊接、热处理几个方面对我国引进的第三代核电站AP1000反应堆主泵屏蔽套的制造工艺进行了浅析,阐明了在屏蔽套制造过程中应该注意的问题,对于实现我国反应堆主泵的国产化具有一定的积极意义。     

Analyses for VVER-1000/320 reactor for spectrum of break sizes along with SBO

Severe accident analysis of a reactor is an important aspect for evaluation of source term. This in turn helps in emergency planning and severe accident management (SAM). Analyses have been carried out for VVER-1000 (V320) reactor following LOCA along with station blackout (SBO) to generate information on these aspects. Availability and unavailability of hydro-accumulators (HAs) are also considered for this study. Integral code ASTEC V1.3 (jointly developed by IRSN, France, and GRS, Germany) is used for analysing the transients. The predictions of different severe accident parameters like vessel rupture time, hydrogen and corium production and radioactivity release to containment have been compared for a spectrum of break sizes to provide information for probabilistic safety analysis (PSA) level-2 and severe accident management (SAM) guidelines.     

Comparison of RELAP5 calculations of VVER-1000 coolant transient benchmark phase 1 at different power

This paper provides comparisons between experimental data of “MCP switching on when the other three MCPs are in operation” and RELAP5 calculations with different initial levels of the reactor power 29.45% and 27.47% from the nominal.

The reference power plant for this analysis is Unit 6 at the Kozloduy nuclear power plant (NPP) site. RELAP5/MOD3.2 computer code has been used to simulate the investigated transient. Operational data from Kozloduy NPP have been used for the purpose of assessing how the RELAP5 model compares against plant data. During the plant-commissioning phase at Kozloduy NPP Unit 6 a number of experiments have been performed. One of them is switching on MCP when the other three MCPs are in operation.

This investigation was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory, under the International Nuclear Safety Program (INSP) of the United States Department of Energy.     

Effects of various spacer grid modeling on the neutronic parameters of the VVER-1000 reactor

The main objective of this paper is to study the effects of various spacer grid models on the neutronic parameters of a VVER-1000 reactor. Specifically, the data of the nuclear power plant at the Bushehr site, which is of a VVER-1000 type, will be studied. Three models, representing the spacer grids along the fuel assemblies are presented. These three models are the homogeneous and the heterogeneous local spacer grid models and the shroud spacer grid model. In the homogeneous and the heterogeneous models, the spacer grids are considered at their actual locations in the axial direction. The only difference between the two models is that in the homogeneous model, the spacer grids are homogenized with the coolant while in the heterogeneous model, the spacer grids are modeled around the fuel cells at their exact axial positions. In the shroud model, the spacer grids are modeled in the shroud region containing the coolant and are not necessarily placed at their appropriate axial positions.     

Incorporation of the procedure for monitoring coolant boiling in a VVER-1000 core into the in-reactor noise diagnostics system

A method for performing diagnostics of coolant boiling in a VVER-1000 core is described. This method was developed at the Russian Science Center Kurchatov Institute and is based on monitoring the increase of the sensitivity of neutron flux noise to the fluctuations of the coolant parameters. This procedure is intended to be used as part of the in-reactor noise diagnostics system. It is now in commercial operation of the No. 3 unit of the Kalinin and two units of the Tianwan (China) nuclear power plants. Translated from Atomnaya énergiya, Vol. 105, No. 2, pp. 79–82, August, 2008.