Investigation of the interaction of oxide melt and steel in a VVER-1000 vessel during a serious accident

The results of investigations of the interaction of oxide melt with steel, which were performed following the international OECD program Masca, are presented. The experimental conditions simulated the high-temperature stage of a serious accident in a VVER-1000 vessel. It is determined that the temperature range 2500–2600°C oxide and metallic phases which are immiscible in the fused state form as a result of the interaction of underoxidized melts C-32–C-70 (U/Zr = 1.2). Fused iron is saturated with uranium, zirconium, and oxygen. The density of the metallic phase becomes greater than the density of the oxide melt, which causes the metallic phase to move to the bottom of the pool. The compositions of the coexisting oxide and metallic phases are determined. This can serve as a basis for constructing thermodynamic models of melts in the U-Zr-O-Fe system. __________ Translated from Atomnaya énergiya, Vol. 104, No. 4, pp. 208–211, April, 2008.     

Investigation of the effect of B<Subscript>4</Subscript>C on the interaction of oxide melt and steel in a VVER-1000 vessel during a severe accident

The results of investigations of the interaction of U-Zr-B-C-O melts with steel, which are performed as part of the OECD Masca international program, are presented. It is found that, as a result of the interaction, boron and carbon become concentrated predominately in the metallic phase of the melt. As the initial mass ratio m_Fe/m_melt increases, the effect of the addition of B₄C on the melt-iron interaction decreases because the metallic phase is diluted with iron. It is concluded on the basis of a comparison of the results of the STFM-B Nos. 3, 7 experiments with the STFM-Fe Nos. 3, 7 experiments performed previously without the participation of boron carbide that the effect of boron carbide on the interaction of the oxide melt with iron decreases as the degree of oxidation of zirconium increases. Translated from Atomnaya énergiya, Vol. 105, No. 2, pp. 63–67, August, 2008.     

Investigation of ex-vessel core catcher for SBO accident in VVER-1000/V528 containment using MELCOR code

Nuclear Science and Techniques - An S-band high-gradient accelerating structure is designed for a proton therapy linear accelerator (linac) to accommodate the new development of compact,...     

Investigation of the distribution of fission products among molten fuel and reactor phases: (II). The distribution of fission products between molten iron and a molten uranium dioxide-basalt mixture

The solubility of uranium dioxide in molten basalt has been determined at 1550°C to be between 5 and 7 wt % and at 2200°C to be between 50 and 55 wt %. The distribution of some of the more important heat-producing fission products between molten iron and a molten uranium dioxide-basalt mixture has been studied. This has been done to evalutate the use of basalt as a fuel diluent and sacrificial material in a fast breeder reactor and to predict the fission product heat distribution in such a system. It was found that lanthanum, cerium and niobium were distributed to the molten uranium dioxide-basalt mixture, and molybdenum and ruthenium were distributed to the molten-iron phase.     

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.     

Investigation of the distribution of fission products among molten fuel and reactor phases: (I). The distribution of fission products between molten iron and molten uranium dioxide

The distributions of some of the more important heat-producing fission products between molten iron and molten uraniumdioxide have been studied; these experiments were carried out using an arc-melting furnace. It was found that yttrium, lanthanum, strontium, barium, zirconium, praseodymium, cerium and some niobium were distributed to the molten uranium dioxide and that molybdenum, ruthenium and some niobium were distributed to the molten iron phase.     

The retardation of the released fission products deriving from graphite structures during a hypothetical HTR-500 accident

A comprehensive computer model is presented describing the transport paths of the released fission products in the coolant gas. The transport mechanisms within the graphite are discussed in detail.An experimental assembly for the verification of the computer model is described and the measurements carried out on the retardation (retention capability) of cesium in reflector graphite are presented and compared with the calculations from the “PATRAS-CORE” program.First reliable statements under realistic conditions can be made with the example of a core heat-up accident in the HTR-500.     

Experimental investigation of in-vessel mixing phenomena in a VVER-1000 scaled test facility during unsteady asymmetric transients

In-vessel turbulent mixing phenomena affect the time and space distribution of coolant properties (e.g., boron concentration and temperature) at the core inlet which impacts consequently the neutron kinetics response. For reactor safety evaluation purposes and to characterize these phenomena it is necessary to set and validate appropriate numerical modelling tools to improve the current conservative predictions. With such purpose, an experimental campaign was carried out by OKB Gidropress, in the framework of the European Commission Project “TACIS R2.02/02 - Development of safety analysis capabilities for VVER-1000 transients involving spatial variations of coolant properties (temperature or boron concentration) at core inlet”. The experiments were conducted on a scaled facility representing the primary system of a VVER-1000 including a detailed model of the Reactor Pressure Vessel with its internals. The simulated transients involved perturbations of coolant properties distribution providing a wide validation matrix. The main achievements of the set of experiments featuring transient asymmetric pump behaviour are presented in this paper. The potential of the obtained experimental database for the validation of thermal fluid dynamics numerical simulation tools is also discussed and the role of computational fluid dynamics in supporting the experimental data analysis is highlighted.     

Investigation of xenon transient processes in VVER-1000 at the Tianwan nuclear power plant (China)

The integral and spatial xenon transient processes in the No. 1 unit of the Tianwan nuclear power plant (China) have been studied experimentally. A measurement method which is unconventional for VVER-1000 was tested in the investigations of the integral processes: the course of the xenon process was recorded according to the variation of the critical concentration of boric acid in the reactor at the same time as the concentration was calculated in real-time. The spatial transient processes were studied for the diametric and axial free xenon oscillations of the energy release in the core. It was confirmed experimentally that axial deformations of the energy release affect the power of the reactor as well as the associated operational particularities of the automatic power regulator. Translated from Atomnaya énergiya, Vol. 105, No. 4, pp. 183–190, October, 2008.     

Numerical method for the containment of fission products at a hypothetical accident in an underground nuclear power plant

As one means to expand the siting of nuclear power plants, construction of underground plants is now under study. An underground nuclear power plant has the feature that ground surrounding the underground cavity can contain the fission products of a hypothetical accident.If it is assumed that in a hypothetical reactor accident the cooling system loses its capacity wholly or partially, and gas containing fission products is emitted into the underground cavity. As a result, temperature, gas concentration and gas pressure in the cavity increase and it can be supposed that the gas leaks up to the surface through the ground, and that ground-water contains and carries fission products. The present paper numerically simulates a course of movement as mentioned above by the finite element method and gives the underground containment effect for fission products from a hypothetical accident.     

Attack upon the cladding of oxide fuel pins by fuel and fission products

In parallel with post-irradiation examinations, a comprehensive out-of-pile experimental programme has been performed to determine the most important fission product reactions with four austenitic stainless steels at different oxygen potentials. Single as well as groups of fission products (simulated burn-up systems) have been used. Only the elements cesium, iodine and tellurium cause dangerous reactions with the cladding of an oxide fuel pin. The others are either not reactive or produced in such small quantities that their attack on the cladding is insignificant. Molybdenum is often found in the reaction zone of an irradiated oxide pin. However, according to our out-of-pile results it does not look as if molybdenum is a dangerous fission product. A decisive factor for the occurrence of reactions with the cladding is the oxygen potential in the fuel pin. As long as the O/M ratio of the fuel is markedly below 2.00, there are no dangerous reactions, neither with cesium nor with tellurium and iodine. The post-irradiation investigations (burn-up 1 to 10 at %) have shown that the cladding attack below 750 °C is most dependent on the inner wall temperature. Other factors, including fuel density, rod power and burn-up, seem to play a minor role. A noticeable reduction of the cladding attack was observed when the initial O/M ratio of the fuel was less than 1.98. A kinetic evaluation of some of the reactions observed in the out-of-pile tests has been attempted. At temperatures above 700 °C, the influence of temperature decreases markedly and the fission product concentration in the fuel becomes more important. There are indications that this also holds true for in-pile conditions.     

Experimental determination of the leakage of solid fission products at a temperature simulating an accident at a nuclear power plant

Institute of Nuclear Reactors, Russian Scientific Center "Kurchatovskii Institut." Translated from Atomnaya Énergiya, Vol. 75, No. 5, pp. 363-367, November, 1993.     

Crack initiation and arrest in a pressurized thermal shock test for a model pressure vessel made of VVER-440 reactor pressure vessel steel

A joint pressure vessel integrity research programme involving three partners is being carried out during 1990–1995. The partners are the Central Research Institute of Structural Materials “Prometey” from Russia, IVO International Ltd (IVO) from Finland, and the Technical Research Centre of Finland (VTT). The main objective of the research programme is to increase the reliability of the VVER-440 reactor pressure vessel safety analysis. This is achieved by providing material property data for the VVER-440 pressure vessel steel, and by producing experimental understanding of the crack behaviour in pressurized thermal shock loading for the validation of different fracture assessment methods. The programme is divided into four parts: pressure vessel tests, material characterization, computational fracture analyses, and evaluation of the analysis methods. The testing programme comprises tests on two model pressure vessels with artificial axial outer surface flaws. The first model vessel had circumferential weld seam at the mid-length of the vessel. A special embrittling heat treatment is applied to the vessels before tests to simulate the fracture toughness at the end-of-life condition of a real reactor pressure vessel. The sixth test on the first model led to crack initiation followed by arrest. After the testing phase, material characterization was performed. Comparison of calculated and experimental data generally led to a good correlation, although the work is being continued to resolve the discrepancies between the measured initiation and arrest properties of the material.     

Release of fission products from the reactor fuel of the bilibinsk nuclear power plant under conditions which simulate a severe accident

Power Physics Institute (FÉI). Translated from Atomnaya Énergiya, Vol. 74, No. 5, pp. 416–421, May, 1993.     

Modelling of heat and mass transfer processes during core melt discharge from a reactor pressure vessel

The objective of this paper is to study the heat and mass trasnfer processes related to core melt discharge from a reactor vessel in a light water reactor severe accident. The phenomenology modelled includes the convection in, and heat transfer from, the melt pool in contact with the vessel lower head wall, the fluid dynamics and heat transfer of the melt flow in the growing discharge hole and multi-dimensional heat conduction in the ablating lower head wall. A research programme is underway at the Royal Institute of Technology (Kungliga Tekniska Högskolan, KTH) to (1) identify the dominant heat and mass transfer processes determining the characteristics of the lower head ablation process: (2) develop and validate efficient analytical/computational models for these processes; (3) apply models to assess the character of the melt discharge process in a reactor-scale situation; (4) determine the sensitivity of the melt discharge to structural differences and variations in the in-vessel melt progression scenarios. The paper also presents a comparison with recent results of vessel hole ablation experiments conducted at KTH with a melt simulant.     

Estimate of the Effect of the Time Interval between Technical Inspections on the Strength Reliability of Certain First-Loop Components of a VVER-1000 Reactor

The results of a computational investigation of the possibility of and the safety conditions for switching from a 4- to an 8-yr time interval between technical inspections of the main circulation pipeline and the pressure compensator vessel in VVER-1000 reactors are presented. To this end, calculations of the critical and admissable sizes of defects in the main metal and weld seams in the operating regime, in accident situations, and during earthquakes have been performed. Calculations of the time for a through defect to reach a critical size for different operating periods have been performed. The influence of the hydrotesting pressure and the time interval between such tests on the operational safety and the effect of the time between technical inspections on the reliability of the indicated first-loop components of a VVER-1000 reactor with respect to the criterion of fracture resistance, taking account of the probabilistic nature of the initial data, is analyzed.The calculations were performed using normative and certified procedures geared toward the typical characteristics of steel and structures and the conditions of fabrication, assembly, and operation.__________Translated from Atomnaya Energiya, Vol. 98, No. 4, pp. 267–273, April, 2005.