AFEN method and its solutions of the hexagonal three-dimensional VVER-1000 benchmark problem

The unique features of the analytic function expansion nodal (AFEN) method in hexagonal-z geometry are described. The COREDAX code implementing the AFEN method is verified testing on the VVER-440 benchmark problem and a “simplified” VVER-1000 benchmark problem. The COREDAX code then applied to the original VVER-1000 benchmark problem exercise 2 (HZP case and HP case) provides very good results in comparison with those of other benchmark participants.     

The Analytic Function Expansion Nodal (AFEN) Method with Half-Interface Averaged Fluxes in Mixed Geometry Nodes for Analysis of Pebble Bed Modular Reactor (PBMR) Cores

The analytic function expansion nodal (AFEN) method has been successfully applied to the rectangular and hexagonal geometries in the cartesian coordinates system. In this paper, we extended the AFEN method to the cylindrical geometry in the R-Z coordinates for the analysis of pebble bed modular reactors (PBMRs). To treat the mixed geometry of rectangular and triangular nodes appearing in the lower periphery of the reactors, we used half-interface averaged fluxes as nodal unknowns. Numerical results obtained attest to their accuracy and applicability to practical problems.     

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.     

Study of thorium advanced fuel cycle utilization in light water reactor VVER-440

The first step in investigation of thorium fuel is evaluation of the results obtained from the spectral code for this type of fuel. The benchmark summarized by IAEA in 2003 was used for partial validation of the code HELIOS 1.9. The benchmark was focused on a comparison of the methods and basic nuclear data. Acceptable results of benchmark comparison allowed examining and comparing different advanced nuclear fuel cycles under light water reactor conditions, especially in VVER-440. Cycles, calculations and results for VVER-440 reactors are presented in the paper. Two of the investigated thorium based fuels include one solely plutonium–thorium based fuel, while the other one is a plutonium–thorium based fuel with a content of reprocessed uranium. The third examined fuel cycle is a cycle with an inert-matrix fuel consisting of reprocessed plutonium and minor actinides (MA) fixed in an yttria-stabilized zirconium matrix. All of them are used to carry and burn or transmute plutonium created in the classical UOX cycle. The Pu transmutation rate and cumulating of Pu with MA in the spent fuel were compared mutually and with an UOX open cycle. The fuel cycle with an inert-matrix fuel was proven to be the best cycle for minimizing the production of Pu in the VVER-440 reactors.     

Impact of cross-section generation procedures on the simulation of the VVER-1000 pump startup experiment in the OECD/DOE/CEA V1000CT benchmark by coupled 3D thermal-hydraulics/neutron kinetics models

In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat a l'Enerige Atomique (CEA), France a coupled three-dimensional (3D) thermal-hydraulics/neutron kinetics benchmark was defined. The overall objective of OECD/NEA V1000CT benchmark is to assess computer codes used in analysis of VVER-1000 reactivity transients where mixing phenomena (mass flow and temperature) in the reactor pressure vessel are complex. Original data from the Kozloduy-6 Nuclear Power Plant are available for the validation of computer codes: one experiment of pump start-up (V1000CT-1) and one experiment of steam generator isolation (V1000CT-2). Additional scenarios are defined for code-to-code comparison. As a 3D core model is necessary for a best-estimate computation of all the scenarios of the V1000CT benchmark, all participants were asked to develop their own core coupled 3D thermal-hydraulics/neutron kinetics models using the data available in the benchmark specifications and a common cross-section library. The first code-to-code comparisons based on the V1000CT-1 Exercise 2 specifications exhibited unacceptable discrepancies between two sets of results. The present paper focuses on the analysis of the observed discrepancies. The VVER-1000 3D neutron kinetics models are based on cross-section data homogenized on the assembly level. The cross-section library, provided as part of the benchmark specifications, thus consists in a set of parameterized two group cross sections representing the different assemblies and the reflectors. The origin of the observed large discrepancies was found mainly to lie in the methods used to solve the diffusion equation. The VVER reflector properties were also found to enhance discrepancies by increasing flux gradients at the core/reflector interface thus highlighting more the difficulties in some codes to handle high exponential flux gradients. This paper summarizes the different steps applied to analyze the neutronic codes and their predictions as well as the impact of cross-section generation procedures.     

A VVER-1000 LEU and MOX assembly computational benchmark analysis using the lattice burnup code EXCEL

Utilization of Mixed Uranium–Plutonium Oxide (MOX) fuel in VVER-1000 reactors envisages the core physics analysis using computational methods and validation of the related computer codes. Towards this objective, an international experts group has been established at OECD/NEA. The experts group facilitates sharing of existing information on physics parameters and fuel behaviour. Several benchmark exercises have been proposed by them with intent to investigate the core physics behaviour of a VVER-1000 reactor loaded with 2/3rd of low enriched uranium (LEU) fuel assemblies (FA) and 1/3rd of weapons grade mixed oxide (MOX) FA. In the present study an attempt is made to analyse ‘AVVER-1000LEUandMOXAssemblyComputationalBenchmark’ and predict the neutronics behaviour at the lattice level. The lattice burnup code EXCEL, developed at Light Water Reactor Physics Section, BARC is employed for this task. The EXCEL code uses the 172 energy group ‘JEFF31GX’ cross-section library in WIMS-D format. Assembly level fuel depletion calculations are performed up to a burnup of 40 MWD/kg of heavy metal (HM). Studies are made for the parametric variations of fuel and moderator temperatures, coolant density and boron content in the coolant. Both operational and off-normal states are analysed to determine the corresponding infinite neutron multiplication factor (k_∞). Pin wise isotopic compositions are computed as a function of burnup. Isotopic compositions in different annular regions of Uranium–Gadolinium (UGD) pin, fission rate distributions in UGD, UO₂ and MOX pin cells are also computed. The predicted results are compared with the benchmark mean results.     

An optimization method based on combination of cellular automata and simulated annealing for VVER-1000 NPP loading pattern

This paper introduces a design methodology in the context of finding new and innovative design principles by means of optimization techniques. In this method cellular automata (CA) and simulated annealing (SA) were combined and used for solving the optimization problem. This method contains two principles that are neighboring concept from CA and accepting each displacement basis on decreasing of objective function and Boltzman distribution from SA that plays role of transition rule. Proposed method was used for solving fuel management optimization problem in VVER-1000 Russian reactor. Since the fuel management problem contains a huge amount of calculation for finding the best configuration for fuel assemblies in reactor core this method has been introduced for reducing the volume of calculation. In this study reducing of power peaking factor inside the reactor core of Bushehr NPP is considered as the objective function. The proposed optimization method is compared with Hopfield neural network procedure that was used for solving this problem and has been shown that the result, velocity and qualification of new method are comparable with that. Besides, the result is the optimum configuration, which is in agreement with the pattern proposed by the designer.     

六角形格林函数节块法

本文基于保角变换思想将格林函数节块法应用于六角形几何,该模型采用保角变换将六角形节块变换为矩形节块,对变换后的矩形节块扩散方程横向积分并应用第二类边界条件的格林函数法进行求解。基于此模型编制了堆芯三维多群稳态程序NACK。利用NACK程序计算了不带反射层二维VVER-1000、三维两群VVER-440和带不连续因子的二维基准题。计算结果表明,有效增殖因数k_eff的误差均小于50 pcm,组件功率分布最大相对误差小于2%,验证了程序的正确性。     

Experience of irradiation embrittlement in the Loviisa reactor pressure vessels

The water gap between the wall and the core of the RPV (Reactor Pressure Vessel) in a VVER-440 plant is small compared with typical Western type LWR₅. The neutron fluence on the RPV wall is, consequently, much higher in a VVER-440 plant. In older VVER-440 plants the material of the RPV, especially the horizontal core weld, contains so much impurities (P- and Cu-content) that the irradiation embrittlement has become a problem. On bases of fracture mechanics analyses in Loviisa, IVO has been forced to make several measures to ensure safe operation of the plants. According to IVO's current understanding, both plants may be in operation for the design life without annealing of the RPVs.     

用于燃耗计算的三维MCCOOR程序系统

介绍了由标准程序MCNP、COUPLE、ORIGEN-S组成的耦合程序系统MCCOOR的结构和功能,用VVER等轻水堆栅元和燃料组件的多个Benchmark模型进行了检验.本文列举了在VVER-1000带可燃毒物Gd的燃料组件Benchmark模型上,分别用UO2和MOX燃料的检验结果.所有检验结果表明:MCCOOR的反应性和核素成分的计算结果与Benchmark的结果在误差范围内一致.     

Simulation of the VVER-1000 pump start-up experiment in the OECD/DOE/CEA V1000CT benchmark by the FLICA4/CRONOS2 coupled code system

In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat à l'Energie Atomique (CEA), France, a coupled 3-D thermal–hydraulics/neutron kinetics benchmark was defined. The overall objective of OECD/NEA V1000CT benchmark [Ivanov, B., Ivanov, K., Groudev, P., Pavlova, M., Hadjiev, V., 2002. VVER-1000 Coolant Transient Benchmark (V1000-CT). Phase 1 – Final Specifications, NEA/NSC/DOC] is to assess computer codes used in the analysis of VVER-1000 reactivity transients where mixing phenomena (mass flow and temperature) in the reactor pressure vessel are complex. Original data from the Kozloduy-6 Nuclear Power Plant are available for the validation of computer codes: one experiment of pump start-up (V1000CT-1) and one experiment of steam generator isolation (V1000CT-2). The CEA presented results for the V1000CT-1 Exercise 2 using a coupling of FLICA4 [Toumi, I., Gallo, D., Bergeron, A., Royer, E., Caruge, D., 2000. FLICA4: a three dimensional two-phase flow computer code with advanced numerical methods for nuclear applications. Nuclear Engineering and Design 200, 139–155] and CRONOS2 [Akherraz, B., Baudron, A.M., Lautard, J.J., Magnaud, C., Moreau, F., Schneider, D., Gonzales, M., 2004. Manuel de Référence CRONOS 2.6. Technical Report SERMA/LENR/RT/04-3433/A, CEA] via the coupling tool ISAS [Toumi, I., et al., 1995. Specifications of the general software architecture for code integration in ISAS. Euratom Fusion Technology, ITER task S81TT-01/1]. The FLICA4/CRONOS2 VVER-1000 model is based on the data available in the benchmark specifications. This paper summarizes the FLICA4/CRONOS2 model build-up with the associated sensitivity studies and presents the CEA results for V1000CT-1 Exercise 2 as well as a comparison with experimental results at hot power steady state (HP SS).     

Simulation of the VVER-1000 pump start-up experiment of the OECD V1000CT benchmark with CATHARE and TRAC-PF1

In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat a l'Energie Atomique (CEA), France a coupled three-dimensional (3D) thermal-hydraulics/neutron kinetics benchmark for VVER-1000 was defined. The benchmark consists of calculation of a pump start-up experiment labelled V1000CT-1 (Phase 1), as well as a vessel mixing experiment and main steam line break (MSLB) transient labelled V1000CT-2 (Phase 2), respectively. The reference nuclear plant is Kozloduy-6 in Bulgaria. The overall objective is to assess computer codes used in the analysis of VVER-1000 reactivity transients. A specific objective is to assess the vessel mixing models used in system codes. Plant data are available for code validation consisting of one experiment of pump start-up (V1000CT-1) and one experiment of steam generator isolation (V1000CT-2). The validated codes can be used to calculate asymmetric MSLB transients involving similar mixing patterns. This paper summarizes a comparison of CATHARE and TRAC-PF1 system code results for V1000CT-1, Exercise 1, which is a full plant point kinetics simulation of a reactor coolant system (RCS) pump start-up experiment. The reference plant data include integral and sector average parameters. The comparison is made from the point of view of vessel mixing and full system simulation. CATHARE used a six-sector multiple 1D vessel thermal-hydraulic model with cross flows and TRAC used a six-sector, 18-channel coarse-mesh 3D vessel model. Good agreement in terms of integral parameters and inter-loop mixing is observed.     

Calculation of neutron fluence in the region of the pressure vessel for the history of different reactors by using the Monte-Carlo-method

Embrittlement of pressure vessel material caused by neutron irradiation is a very important problem for VVER-440 reactors. For the estimation of the fracture risk highly reliable neutron fluence values are necessary. For this reason a special theoretical determination of space dependent neutron fluences has been performed mainly on the basis of Monte-Carlo calculations. The described method allows the accurate calculation of neutron fluences near the pressure vessel in the height of the core region for all reactor histories and loading cycles in an efficient manner. To illustrate the accuracy of the suggested method a comparison with experimental results was done. The calculated neutron fluence values can be used for planning the loading schemes of each reactor according to the safety requirements against brittle fracture.     

Method for detecting unsealed fuel elements in VVER and RBMK fuel assemblies and test results

A method is described for detecting unsealed fuel elements in VVER and RBMK fuel assemblies in a cooling pond. The method is based on detecting water which has seeped under the cladding of an unsealed fuel element. The results of testing the method on unsealed VVER-440, -1000, and RBMK-1000 fuel assemblies are presented. Translated from Atomnaya énergiya, Vol. 106, No. 2, pp. 84–88, February, 2009.     

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.     

Analysis of natural circulation phenomena in VVER-1000

In all light water reactors (LWR), natural circulation is an important passive heat removal mechanism. In the present paper, the natural circulation phenomena are studied with reference to step-wise coolant inventory reduction and a small break loss-of-coolant-accident (SBLOCA) in the cold leg of VVER-1000. The natural circulation flow map (NCFM) approach is considered to evaluate the natural circulation performance of the VVER-1000 NPP also comparing VVER-1000 and PWR systems. Three different elevations between heat source (core) and heat sink (steam generators) zones have been considered in order to characterize the buoyancy force in a VVER-1000. The influence of power and the cold legs loop seal upon the natural circulation performance is also evaluated. In the second part, a series of SBLOCA simulations with break area ranging from 0.5 to 11.7% of the cold leg cross sectional area are performed starting with the VVER-1000 system in nominal conditions. The effect of Emergency Core Cooling System (ECCS) including passive and active parts of ECCS are evaluated. The simulations were performed by the help of the system code RELAP5. Within the framework of the qualification of the adopted computational tools, the results are compared with experimental data from Kozloduy NPP unit 6 test and PSB-VVER integral test facility available from the literature. Namely, the qualification of the adopted nodalisation in steady state conditions is achieved by using experimental data. The accuracy of selected results have been estimated in quantitative terms by applying the fast Fourier transform based method (FFTBM). Finally, the relevance and the potential for the occurrence of the reflux condensation mode, i.e., one of the Natural Circulation regimes, for cooling of reactor core in VVER-1000 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.     

VVER-1000 neutronics calculation with ENDF/B-VII data

The initial release of the ENDF/B-VII nuclear data library is verified for VVER-1000 reactors. For neutronics calculation, the MCNP code based on the Monte-Carlo method is applied. Continuous-energy cross-sections for use with MCNP are calculated with the NJOY code. Isotopics for burned fuel is calculated with the WIMSD code. Calculated criticality, pin-to-pin power distribution, time-dependent critical concentration of soluble boron, worth of the control rods, average fuel assembly powers and time-dependent axial power distribution are compared to the corresponding experimental values for both zero-power VVER-1000 model, created at the LR-0 experimental facility, and the first fuel cycle of a real VVER-1000 reactor. For all of these parameters, neutronics calculation with ENDF/B-VII is in good agreement with the measurement. Moreover, for VVER-1000 neutronics calculation, ENDF/B-VII provides better results than ENDF/B-VI.