Computational and Experimental Investigations of Accident Regimes in Standard Problems on the ISB-VVÉR Thermophysical Stand

Five standard problems of investigating accidents with loss of coolant in the first loop are studied for the first time on the basis of experiments which were performed in 1993–2001 on the ISB-VVÉR two-circuit integrated thermophysical stand, simulating the first loop of a reactor system with a VVÉR-1000 reactor. The objective of the investigations was to verify the computational thermohydraulic codes developed in this country and abroad – TECh', KORSAR, ATHLET, CATHARE, and RELAP. The results of the verification calculations on the whole agreed well with the experimental data. Most processes and phenomena which can occur in VVÉR-1000 accidents with a small and average-size coolant leak were reproduced in the experiments. Analysis of the results showed that these computational codes can be used to simulate the processes occurring during an accident.     

Some Experimental Work Performed on the BOR-60 Reactor

The experimental work performed on a BOR-60 reactor over a period of 30 years of reactor operation is briefly reviewed. The results of investigations of the neutron-physical, thermohydraulic, and dynamical characteristics and the safety parameters of the reactor are presented. The investigations performed and the analysis of transient and emergency regimes made it possible to improve the standard shutdown and cooldown systems in order to soften the temperature conditions on the reactor components.The result of a series of experiments on the safety of fast sodium reactors, among which the introduction of gas into the core, sodium boiling, blocking of the flow in the experimental fuel assembly with destruction of fuel elements, interloop leakage in the steam generators, and so on, are discussed. A complex of diagnostics systems has been developed and tested on the basis of the safety investigations.Analysis of the radiation parameters and characteristics of the reactor made it possible to develop methods and means for monitoring and improving the radiation conditions and the safety of the reactor.Experimental irradiation of various initial materials, using threshold and other reactions, enabled the serial production of radionuclides for medical purposes.     

Control of the power of a fast reactor in the self-regulation regime and reactor startup

An algorithm for controlling the power of a nuclear reactor operating in the self-regulation regime and reactor startup is proposed. The algorithm is based on investigations of a dynamical model of the reactor. It makes it possible to start up the reactor and maintain it quite accurately at a prescribed power level. However, an additional safety investigation is needed.__________Translated from Atomnaya Énergiya, Vol. 98, No. 1, pp. 18–24, January, 2005.     

IGRIK reactor operation in a two-pulse mode

The results of investigations of the operation of the IGRIK pulsed homogeneous solution research complex in the unconventional (for pulsed reactors) two-pulse mode with successive reactivity increments with generation of prompt-or delayed-neutron pulses are presented. Experimental data which give an idea of the processes occurring in the reactor core in the two-pulse regime are obtained. These data made it possible to determine whether or not the IGRIK reactor can be operated in this regime and the limitations imposed by the physical processes and special design characteristics of the system. It is shown that are pulses can be generated in several different ways. The dependences of the fission-pulse characteristics in a given regime on the initial conditions and the methods for obtaining the pulses are investigated. Methods are proposed for increasing the operational possibilities of the reactor. __________ Translated from Atomnaya énergiya, Vol. 102, No. 3, pp. 146–151, March, 2007.     

Investigations on irradiated fission chamber electrodes

Analysis of some failed in-core fission chambers in a power reactor made a series of investigations on fission chamber electrodes necessary. Two types of electrodes with different uranium coating techniques were used. One type of electrodes was first exposed to maximum neutron doses of 2 × 10²⁰ nvt at steady-state operation. Both were exposed to several reactor power pulses at the TRIGA reactor Vienna. Changes in the uranium layer on the electrode surface have been observed during steady-state operation depending on the neutron dose. Complete destruction of the uranium layer was observed after pulse irradiation with one electrode type, while the other electrode type remained in good condition.     

Investigation of the Mixing of Coolant Flows in a VVÉR Vessel

The mixing of flows in a flow-through channel of a VVÉR-1000 reactor in the case where a flow with high or low boron concentration and a flow whose temperature is different from the temperature in the reactor are fed into the same circuit is studied. To this end investigations were performed on a bench at the Special Design Office Gidropress with a model VVÉR-1000 reactor on a 1:5 scale. Regimes with the main circulation pump switched on and with natural circulation restored for an accident with a small leak are studied.Intercircuit mixing in the VVÉR-1000 reactor setup is also studied in the case where the temperature of the coolant changes in one of the circuits (regimes with a break in a steam pipe and closure of the steam channel of the turbine) or with asymmetric feeding of boron into the first loop when one of the tanks in the system for rapid introduction of boron fails. An analytical solution of the problem and the results of measurements performed on units with VVÉR-1000 reactors are presented.     

Molten Salt Oxidation of Ion-exchange Resins Doped with Toxic Metals and Radioactive Metal Surrogates

Ion-exchange resins doped with toxic metals and radioactive metal surrogates were test-burned in a bench-scale molten salt oxidation (MSO) reactor system. The purposes of this study are to confirm the destruction performance of the two-stage MSO reactor system for the organic ion-exchange resin and to obtain an understanding of the behavior of the fixed toxic metals and the sulfur in the cationic exchange resins. The destruction of the organics is very efficient in the primary reactor. The primarily destroyed products such as carbon monoxide are completely oxidized in the secondary MSO reactor. The overall collection of the sulfur and metals in the two-stage MSO reactor system appeared to be very efficient. Over 99.5% of all the fixed toxic metals (lead and cadmium) and radioactive metal surrogates (cesium, cobalt, strontium) remained in the MSO reactor bottom. Thermodynamic equilibrium calculations and the XRD patterns of the spent salt samples revealed that the collected metals existed in the form of each of their carbonates or oxides, which are non-volatile species at the MSO system operating conditions.     

Characteristics of thermochemical processes in the system C-Al-TiO2 during reprocessing of reactor graphite

A thermodynamic analysis and laboratory investigations are performed of the thermochemical processes occurring in the system C-Al-TiO₂ during the reprocessing of reactor graphite. The conditions under which hydrolytically unstable aluminum oxycarbides are formed in the final product are determined. The influence of air and moisture in the initial charge and the composition of the surrounding medium on the characteristics of the process leading to the synthesis of the final product is investigated. __________ Translated from Atomnaya énergiya, Vol. 101, No. 5, pp. 372–379, November, 2006.     

Post-reactor studies of fuel assemblies and a control-and-safety system channel of the reactor in the first in the world nuclear power plant

The results of post-reactor materials engineering studies of fuel assemblies and control-and-safety system channels, which were unsealed in the reactor of the world’s first nuclear power plant in the period from 1987 to 1998, are presented. It is established that the destruction of the fuel element, fuel-assembly tubes and a channel of the control-and-safety system was due to the formation of through transcrystallite brittle cracks and corrosion foci on the outer side of the tubes and outer cladding of the fuel elements which is in contact with the gas medium of the reactor. Chlorine was discovered on the outer surface of the outer cladding of the fuel elements in the damaged zone. The maximum chlorine concentration was observed in cracks and pits. The presence of the chlorine ion is due to corrosion fissuring and pitting corrosion.     

Computational investigations of the dynamical characteristics of the core of a periodic pulse reactor in a system with cascade neutron multiplication

The core dynamics of a fast reactor in a cascade reactor system operating in a periodic-pulse regime are examined. A model of a BN-600 fuel element is used as a computational model. Computational studies of the neutron kinetics processes in a fast rector-subcritical assembly system and the thermal dynamics of a fuel element in the core of a periodic-pulse reactor are performed. Estimates are made of the service life of a fuel element operating in a regime with repeating pulses and a number of heat loads that is admissible from the standpoint of the fatigue strength of the element.Translated from Atomnaya Ènergiya, Vol. 97, No. 4, pp. 260–269, October, 2004.     

Initial results of investigations into fission product deposition in in-pile experiments

The safe operation of a high temperature reactor (HTR) depends to a decisive extent on the behaviour of the fission products released from the fuel elements into the primary loop and on the nuclides resulting from activation. To investigate the still little-explored mechanisms of transport and deposition for fission and activation products, a comprehensive program for the study of the deposition of fission products was undertaken at the KFA in collaboration with various industrial concerns and foreign project groups, which has in the meantime led to a preliminary knowledge of these processes. It is the aim of the experimental and theoretical efforts to develop a realistic model for deposition, which permits calculating the deposition onto components of the primary loop and thus also allows an assessment of the possibility for service and repair of the individual components. Beyond this, the model should serve as the basis for the realistic consideration of the consequences of a failure in the reactor. The initial results of the Saphir-Pégase and Vampyr-AVR in-pile experiments are discussed. It was found that deposition cannot be understood in terms of pure adsorption; on the contrary, irreversible processes, such as diffusion and chemical bonding, also have to be considered. A model which includes these mechanisms is explained and its correctness discussed in terms of the experimental results. The knowledge obtained to date supports the necessity for further intensive experimental and theoretical investigations directed to the understanding of the various factors affecting the deposition process and to the determination of their magnitudes.     

Experimental and theoretical investigations on the COMET concept for ex-vessel core melt retention

In the very unlikely case of a core melt accident in a nuclear power plant, the reactor pressure vessel could fail and corium melt could be released into the reactor cavity. Subsequent processes could result in a threat of the containment integrity. As a counter-measure the implementation of a core-catcher device in nuclear power plants is envisaged. Such a core-catcher concept has been developed at the Forschungszentrum Karlsruhe (FZK, Germany) within the COMET project. It is based on water injection into the melt layer from the bottom, yielding rapid fragmentation of the corium, porosity formation and thus coolability. Detailed large scale experiments with sustained heating of melts have highlighted the sequences of flooding and cooling and have been used to optimise the COMET concept. The open porosities and large surfaces that are generated during melt solidification form a porous permeable structure that is permanently filled with the evaporating coolant water and thus allows efficient short-term and long-term removal of the decay heat. Two variants of the bottom flooding concept have been developed and seem technically mature for reactor application. Corium layers up to 0.5 m high are safely arrested and cooled by water supply with 0.2 bar overpressure.The conceptual and experimental work at FZK is accompanied by theoretical investigations at IKE, University of Stuttgart. These investigations address porosity formation as well as quenching and long-term coolability of layers with resulting porosities. The aim of the theoretical work is to get a better understanding of the underlying processes of porosity formation in order to generally support the applicability of the concept for real conditions and to allow checks and optimisation for various conditions. A model for porosity formation is presented, which assumes that this process is essentially determined by strong local pressure buildup from strong evaporation due to water injection from below and the restriction of steam removal by friction in the melt. The effect of key parameters is investigated and compared to experimental results. Agreement about the influence and importance of these parameters as well as essential quantitative effects is found.     

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.     

Estimate of the parameters of feedback on power and the stability of IBR-2 for different average power levels

Experimental and model studies of the parameters of fast feedback on power as a function of the average power of IBR-2 have been performed. Transient power processes caused by square fluctuations of reactivity have been investigated. The changes in the parameters are estimated for average power ranging from 0.5 to 1.5 MW. The results obtained are compared with data from previous experiments performed in 1984–1996. It is noted that the influence of feedback on power decreases as the reactor operating time increases. The model of a reactor with parameters of feedback on power which correspond to one series of experiments is investigated for stability by the frequency method. It is shown that at the regular average power level 1.5 MW a reactor in a self-regulating regime (i.e., without an automatic regulator) possesses an adequate margin of stability. __________ Translated from Atomnaya énergiya, Vol. 103, No. 2, pp. 89–93, August, 2007.     

Package Flow Models by Neural Network Representation for Understanding the Dynamic Behavior of Nuclear Reactor Systems

“Package Flow Model” (PFM) is a simple simulation model for intuitive understanding of various types of system dynamics. In the previous papers, the PFM was proposed and its application to the dynamic analysis of nuclear reactor systems was presented. In the present paper, the same model and same application are considered but a new representation method of the PFMs by a neural network is introduced, so that the dynamic simulation of the reactor subsystem can be performed through the calculation of corresponding neural network. Furthermore, the quasi optimum parameter values of each PFM are easily obtained by applying appropriate learning algorithm to get weight-values of the neural network.

Some case studies show that the learning process and the obtained optimum values can give us new useful information on approximate understanding of the dynamic behavior of actual processes in the system.     

Model of cladding failure estimation for a cycling nuclear unit

Nuclear reactor operating modes under multiple cyclic power changes have been promoted recently, and fuel element cladding behavior under the multiple cyclic power changes has been widely known as a key issue in terms of rod design and reliability. A model of nuclear reactor fuel rod cladding failure estimation under multiple cyclic power changes is proposed. The model is built on the basis of the following admissions of the energy version of creep theory: processes of cladding creep and destruction proceed together and affect each other, intensity of creep process is estimated by specific dispersion power W(τ), while intensity of destruction—by specific dispersion energy A(τ) accumulated during time τ. Having calculated the equivalent stress and the rate of equivalent creep strain, the condition of fuel rod cladding failure used on the basis of the energy version of the theory of creep gives us a criterion to decide if a multiple cyclic power change operating mode is permissible for a given variant of power history and coolant conditions.     

Process resulting in the destruction of the No. 4 unit reactor at the Chernobyl nuclear power plant

The results of an investigation of structural fragments of the core of the destroyed reactor in the No. 4 unit of the Chernobyl nuclear power plant are analyzed. It is shown on the basis of investigations of the fission product distribution over the cross-section of the graphite blocks and the determination of the physical properties of graphite that the temperature of the graphite blocks, including the reflector, at the moment they were ejected from the core exceeded 1000°C. The heat content of the fuel was estimated on the basis of an analysis of fragments of the dispersed uranium dioxide particles and an analysis of possible graphite dispersal mechanisms at the moment of the explosion. It is shown that energy sufficient for dispersing and partial vaporization of the fuel and for dispersing the graphite could have been introduced into the fuel during the accident process. Analysis confirms the possibility of a core destruction scenario with ejection from the shaft and ejection of part of the fuel in the form of vapor and dispersed particles into the atmosphere. __________ Translated from Atomnaya énergiya, Vol. 104, No. 6, pp. 319–328, June, 2008.     

Radioactivity Confinement Studies Within the SEAL Program

In the framework of the European SEAL program, investigations have been performed with the aim of optimizing the second confinement function and plant layout with respect to normal operation as well as abnormal operation, including accident conditions. This has been done for two conceptual fusion reactor designs: one using water as the coolant and the other using helium. The starting point of these investigations was the SEAFP project design. For the water-cooled reactor design the studies were focused on design options such as pressure suppression spray system, pressure suppression pool with closed containment or with venting to gravel bed filter and stack, and separate expansion volume optionally operated with a vacuum and equipped with spray system. Similar analyses were performed for the helium-cooled reactor design. The analyses were focused on design options comprising a single, large confinement volume or a vent duct connected to an expansion volume operated at vacuum in comparison with the SEAFP Model 1. The thermal-hydraulic analyses performed with the MELCOR code provide an integrated assessment of the cooling loop and confinement system dynamics.     

Multipulse operation of the IGRIK reactor

The operation of the IGRIK pulsed homogeneous solution research complex in a multipulse regime followed by a reactivity increase is examined. The experimental data give an idea of the processes occurring in the reactor core when the multipulse regime is implemented. These data showed that the IGRIK reactor can be operated in this regime, and they made it possible to determine the range of the pulse parameters and the limitations which the physical processes and the special design features of the system impose. It is shown that there are several possible variants for the generation of a series of successive pulses. The dependences of the fission-pulse characteristics in this regime on the initial conditions and the pulse implementation methods are investigated. __________ Translated from Atomnaya énergiya, Vol. 102, No. 2, pp. 98–104, February, 2007     

PARIS project: Radiolytic oxidation of molecular iodine in containment during a nuclear reactor severe accident: Part 1. Formation and destruction of air radiolysis products—Experimental results and modelling

In case of a hypothetical severe accident in a nuclear LWR (light water reactor), the high radiation fields reached in the reactor containment building due to the release of fission products from the reactor core could induce air radiolysis. The air radiolysis products could, in turn, oxidise gaseous molecular iodine into aerosol-borne iodine-oxygen-nitrogen compounds. Thereby, this reaction involves a change of iodine speciation and a decrease of iodine volatility in the reactor containment atmosphere. Kinetic data were produced within the PARIS project on the air radiolysis products formation and destruction, and on their reaction with molecular iodine, with the objective of developing and validating existing kinetic models.The current paper includes the non-iodine tests of the PARIS project whose objective was to determine the rates of formation and destruction of air radiolysis products in the presence of both structural containment surfaces (decontamination coating (“paint”) and stainless steel), aerosol particles such as silver rich particles (issued from the control rods) in boundary conditions representative for LWR or PHEBUS facility containments.It is found that the air radiolysis products concentration increases with dose and tend to approach saturation levels at doses higher than about 1 kGy. This behaviour is more evident in oxygen/steam atmospheres, producing ozone, than in air/30% (v/v) steam atmospheres, the latter favouring the model-predicted on-going production of nitrogen dioxide even at very high doses. No significant effect of temperature, dose rate and hydrogen addition (4%, v/v) was observed. Furthermore, the inserted surfaces do not exhibit significant effects on the air radiolysis concentrations. However, these “non-noticeable influence” could be due to a masking of small effects by the appreciable scattering of the experimental air radiolysis product concentrations.The PARIS results are then analysed using two different kinetic models, an empirical and a mechanistic one. The kinetic constants within an empirical model including formation and destruction of air radiolysis products, derived from PARIS results, are in reasonable agreement with those determined previously for lower steam fractions.From the mechanistic model IODAIR-IRSN, it is concluded that ozone is the predominant air radiolysis product at low doses in air/steam atmospheres. At doses higher than 1 kGy, nitrogen dioxide becomes increasingly important, both due to an increase in its concentration and due to a simultaneous decrease in ozone concentration.