Strategies of Corium Localization under Severe Accidents with Core Melting for New NPP Projects with VVER

The results of computational analysis of the application of in-vessel and ex-vessel melt retention strategies for VVER of different capacity is presented. The choice in favor of an in-vessel melt retention strategy for VVER-600 and an ex-vessel core catcher for VVER-1200 is proved. It is shown that the ex-vessel core catcher effectively performs its functions on severe accident management and reliably ensures the melt localization and cooling for high power reactors. The calculations of corium localization in the core catcher were carried out with the help of the HEFEST-ULR code developed at the NRC Kurchatov Institute.     

Simulation of COPO II Lo experiments on natural convection of heat generating liquid at high Rayleigh numbers

The methodology is described and the results are presented concerning numerical modeling of COPO II Lo experiments on heat transfer in liquid with internal heat generation at very high internal Rayleigh numbers typical for natural convection in a core melt that can appear during progress of severe accident at a nuclear power plant (NPP). The work is keeping in the course of development of CFD-based tool for quantitative analysis of heat transfer in a stratified molten pool of different configurations possible in severe accident scenarios with melt retention in the reactor vessel or in the VVER core catcher. Such CFD methodology would be used for testing of simplified correlation models for simulation of the core melt interaction with NPP structures in system code SOCRAT. During verification the available experimental data on the core melt thermohydraulics were analyzed, and it was concluded that they are insufficient to measures of CFD quality. The data uncertainties, along with the complexities of convective flow, uncertainties of the reactor core melt conditions, limitations of experimental possibilities and of turbulence modeling, actually constrain the multivariate CFD simulations of natural convection at very high Rayleigh numbers. RANS turbulence models only can be efficiently applied here, and they are to be checked for such purposes. In a series of numerical modeling of COPO II Lo experiments and some others, availability of a k-? realizable model with included buoyancy effects was estimated, and the optimal set of CFD options was formed for minimizing numerical artifacts. It was demonstrated that in the investigated range of Rayleigh numbers the k-? model works qualitatively correctly, but is inclined to systematical deformation of the melt boundary heat transfer distribution. This allows one to use this model for qualitative multivariate CFD estimations but requires improvement of the model or finding of its efficient and more exact equivalent.     

The procedure and results of calculations of the equilibrium isotopic composition of a demonstration subcritical molten salt reactor

A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing ²³³U from ²³²Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.     

Prediction of the thermophysical properties of molten salt fast reactor fuel from first-principles

Molten fluorides are known to show favourable thermophysical properties which make them good candidate coolants for nuclear fission reactors. Here we investigate the special case of mixtures of lithium fluoride and thorium fluoride, which act both as coolant and as fuel in the molten salt fast reactor concept. By using ab initio parameterised polarisable force fields, we show that it is possible to calculate the whole set of properties (density, thermal expansion, heat capacity, viscosity and thermal conductivity) which are necessary for assessing the heat transfer performance of the melt over the whole range of compositions and temperatures. We then deduce from our calculations several figures of merit which are important in helping the optimisation of the design of molten salt fast reactors.     

Main results of study on the interaction between the corium melt and steel in the VVER-1000 reactor vessel during a severe accident performed under the MASCA project

The interactions that take place in the corium melt in the reactor vessel in the case of a severe accident at a nuclear power plant were investigated in accordance with the MASCA international program. Results of the interaction between the oxide melt and iron (steel), partition of the main components [U, Zr, Fe (stainless steel)] between the oxide and the metal phases of the melt, partition of low-volatile simulators of fission products between the phases of the stratified core melt pool, and impact of the oxidizing atmosphere on the melt stratification are presented. The results obtained were used for prediction of thermodynamic properties of the melts belonging to the U-Zr-Fe-O system.     

Numerical simulation of turbulent natural convection of oxide heat generating melt in a core catcher at NPP with VVER-1000

The problem statement and simulation results are presented concerning turbulent natural convection in a vertical cylindrical molten pool with internal heat generation and other parameters (inner Rayleigh number Ra _i ∼ 10¹⁶–10¹⁷) corresponding to oxide core melt in a core catcher for NPP with VVER-1000. Commercial code FLUENT 6.3 was used for CFD calculations. The results on heat transfer are approximated by power law correlations for mean Nusselt numbers vs. Rayleigh number and pool height, describing the heat transfer at upper, lateral, and total boundaries of the cylinder. The influence of volumetric heat generation and material properties is studied. Spatial distribution of wall heat transfer is analyzed for different pool heights possible in the real core catcher. Along with serial calculations with isothermal boundary conditions, the cases with heat radiation conditions are considered. The results may be used for estimations of heat transfer and melt overheating in a VVER core catcher and for coefficient identification of simplified models of integrated system severe-accident codes.     

Magnetic hysteresis properties of neutron-irradiated VVER440-type nuclear reactor pressure vessel steels

The development of non-destructive evaluation methods for irradiation embrittlement in nuclear reactor pressure vessel steels has a key role for safe and long-term operation of nuclear power plants. In this study, we have investigated the effect of neutron irradiation on base and weld metals of Russian VVER440-type reactor pressure vessel steels by measurements of magnetic minor hysteresis loops. A minor-loop coefficient, which is obtained from a scaling power-law relation of minor-loop parameters and is a sensitive indicator of internal stress, is found to change with neutron fluence for both metals. While the coefficient for base metal exhibits a local maximum at low fluence and a subsequent slow decrease, that for weld metal monotonically decreases with fluence. The observed results are explained by competing mechanisms of nanoscale defect formation and recovery, among which the latter process plays a dominant role for magnetic property changes in weld metal due to its ferritic microstructure.     

A methodology for multivariate simulation with massively parallel computing systems for NPP safety assessment: VARIA code

The paper is focused on the practical application of parallel computing techniques in uncertainty assessment in simulation of heat transfer, mechanical and some other problems related to deterministic analysis of NPP safety. A methodology is developed and implemented in VARIA computer code that performs simultaneous run of multiple simulations on a parallel computing system with further statistical analysis of the array of their results. The current version of the code allows automated preparation and execution of multivariate simulations of thermal and mechanical behavior of pressurized water reactor structures by best-estimate (BE) codes in the scope of NPP safety assessment under severe accident conditions. The number of simultaneously launched tasks is limited only by the computer cluster capacity. The VARIA code is verified on multivariate simulation with HEFEST code of thermal behavior of a core melt in the VVER-440 reactor vessel during a severe accident. The influence of the variation of input parameters (decay heat value and coefficients of the applied convective heat transfer model) on the simulation results is studied. It is concluded that the potential field of applying the program is beyond the scope of analyzing severe accidents at NPP and includes also software product quality assurance and analysis of uncertainties of obtained simulation results.     

A kinetics study of vacancy cluster evolution under VVER-type reactor condition

The vacancy clusters (VC) evolution in the neutron-irradiated VVER-type reactor pressure vessel steels is investigated, beginning at the nucleation stage and finishing in the coarsening stage. For this, characteristic VVER-type reactor conditions are considered. VC evolution in the nucleation stage is analysed on the basis of the computer simulation data. During the deterministic and the coarsening stage, elastic interaction between iron matrix and VC is accounted, that provides the stability of the peak of the size distribution function under the condition of increasing neutron fluence in correspondence to the experiments. The results are compared with the results of SANS experiments which were carried out on specimens irradiated at surveillance positions of VVER reactors. The presented approach may be used for analyses of the small inhomogeneities (about 1 nm) in the irradiated damage structure of the VVER-type steels.     

VVER Thermal Power Measured Using the Readings of Rhodium Self-powered Neutron Detector (SPND)

A procedure for determining the linear heat generation rate in a fuel assembly of VVER reactor at the location of a rhodium self-powered neutron detector (SPND) and verification of this procedure are discussed. A method of measurements of the VVER thermal power based on SPND readings is described.     

Kinetic study of copper precipitates under VVER-type reactor conditions

The copper-rich cluster evolution in the neutron-irradiated VVER steels is investigated beginning at the nucleation stage. For this, typical VVER-type reactor conditions are considered. The cluster dynamics approach is used for calculation of the density distribution of copper precipitates related to the number of Cu-atoms or radius, mean radius, volume content, number density of precipitates and the concentration of free Cu-atoms in dependence on the irradiation time. The results for time of one year are compared with the results of small angle neutron scattering experiments which were carried out on specimens irradiated at the surveillance positions of VVER reactors. It has revealed the intermediate type of the evolution kinetics between diffusion and interfacial kinetics limited regimes. The duration of the nucleation and deterministic stages is estimated. The coarsening stage does not occur.     

Experimental spectrum of reactor antineutrinos and spectra of main fissile isotopes

Within the period between the years 1988 and 1990, the spectrum of positrons from the inverse-beta-decay reaction on a proton was measured at the Rovno atomic power plant in the course of experiments conducted there. The measured spectrum has the vastest statistics in relation to other neutrino experiments at nuclear reactors and the lowest threshold for positron detection. An experimental reactor-antineutrino spectrum was obtained on the basis of this positron spectrum and was recommended as a reference spectrum. The spectra of individual fissile isotopes were singled out from the measured antineutrino spectrum. These spectra can be used to analyze neutrino experiments performed at nuclear reactors for various compositions of the fuel in the reactor core.     

不同类型核燃料对热管冷却反应堆燃耗性能的影响

热管冷却反应堆采用固态反应堆设计理念,具有功率密度高、结构紧凑、固有安全性高等特点,在深空探索、深海勘探、偏远地区等场景中具有广阔的应用前景。核燃料作为热管冷却反应堆的重要组成部分,不同类型核燃料在堆芯燃耗分析时会呈现不同的中子学性能。基于美国爱达荷国家实验室(INL)提出的热管冷却反应堆INL Design A,利用清华大学蒙特卡罗中子输运程序RMC (Reactor Monte Carlo code)建立堆芯物理模型,选取UO₂,(U_0.9Pu_0.1)O₂,U-10Zr,U-8Pu-10Zr,UN,UC这6种核燃料开展燃耗计算,分析了不同核燃料、不同功率水平对热管冷却反应堆堆芯燃耗性能的影响。计算结果表明:在堆芯燃耗深度相同情况下(20.8 GW·d·t?1),装载U-8Pu-10Zr燃料的堆芯所需235U富集度最低(9.8%),具有较好的U-Pu增殖性能。堆芯功率处于5 MW的热管冷却反应堆,燃料中241Pu的存在不仅没起到增大堆芯燃耗深度的作用,反而导致堆芯剩余反应性和堆芯寿期末次锕系核素(MAs)的产量增大,影响反应堆的安全性与经济性。因此,对于装载含有Pu燃料的小功率长寿期热管冷却反应堆,需重点关注241Pu对堆芯燃耗性能的影响。     

Dynamic characteristics of a VK-50 reactor operating under conditions of the loss of a normal feedwater flow

The paper describes the conditions of the ATWS type with virtually complete cessation of the feed-water flow at the operating power level of a reactor of the VK-50 type. Under these conditions, the role of spatial kinetics in the system of feedback between thermohydraulic and nuclear processes with bulk boiling of the coolant in the reactor core is clearly seen. This feature determines the specific character of experimental data obtained and the suitability of their use for verification of the associated codes used for calculating water-water reactors.     

Determination of the relative power density distribution in a heterogeneous reactor from the results of measurements of the reactivity effects and the neutron importance function

A method for experimental determination of the relative power density distribution in a heterogeneous reactor based on measurements of fuel reactivity effects and importance of neutrons from a californium source is proposed. The method was perfected on two critical assembly configurations at the NARCISS facility of the Kurchatov Institute, which simulated a small-size heterogeneous nuclear reactor. The neutron importance measurements were performed on subcritical and critical assemblies. It is shown that, along with traditionally used activation methods, the developed method can be applied to experimental studies of special features of the power density distribution in critical assemblies and reactors.     

Reactor physics experimental data for VVER type reactors

An international research program was carried out between 1972 and 1990 into the physics of VVER type reactors. The present paper gives a summary on the second item, i.e. the reactor physics experimental data. The neutron economy is discussed in practical neutron chain reactions.     

Operation of a tokamak reactor in the radiative improved mode

The operation of a nuclear fusion reactor has been simulated within a model based on experimental results obtained at the TEXTOR-94 tokamak and other facilities in which quasistationary regimes were achieved with long confinement times, high densities, and absence of the edge-localized mode. The radiative improved mode of confinement studied in detail at the TEXTOR-94 tokamak is the most interesting such regime. One of the most important problems of modern tokamaks is the problem of a very high thermal load on a divertor (or a limiter). This problem is quite easily solved in the radiative improved mode. Since a significant fraction of the thermal energy is reemitted by an impurity, the thermal loading is significantly reduced. As the energy confinement time τ_E at high densities in the indicated mode is significantly larger than the time predicted by the scaling of ITERH-98P(y, 2), ignition can be achieved in a facility much smaller than the ITER facility at plasma temperatures below 20 keV. The revealed decrease in the degradation of the confinement time τ_E with an increase in the introduced power has been analyzed.