Experiments on the interactions of molten ZrO2 with water using TROI facility

Korea Atomic Energy Research Institute (KAERI) launched an intermediate scale steam explosion experiment named ‘Test for Real cOrium Interaction with water (TROI)’ using reactor material. The objective of the program is to investigate whether the corium would lead to energetic steam explosion when interacted with cold water at a low pressure. The melt/water interaction is made in a multi-dimensional test section located in a pressure vessel. The inductive skull melting, which is basically a direct inductive heating of an electrically conducting melt, is implemented for the melting and delivery of corium. In the first series of tests using several kg of ZrO₂ where the melt/water interaction is made in a heated water pool at 30–95 °C, either a quenching or a spontaneous steam explosion was observed. The spontaneous explosion observed in the present ZrO₂ melt/water experiments clearly indicates that the physical properties of the UO₂/ZrO₂ mixture have a strong effect on the energetics of steam explosion.     

Estimation of ex-vessel steam explosion pressure loads

An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment.In this article, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which is being developed for the simulation of fuel–coolant interactions. A parametric study was performed varying the location of the melt release (central, right and left side melt pour), the cavity water subcooling, the primary system overpressure at vessel failure and the triggering time for explosion calculations. The main purpose of the study was to establish the influence of the varied parameters on the fuel–coolant interaction behaviour, to determine the most challenging cases and to estimate the expected pressure loadings on the cavity walls. For the most explosive central, right side and left side melt pour scenarios a detailed analysis of the explosion simulation results was performed. The study shows that for some ex-vessel steam explosion scenarios higher pressure loads are predicted than obtained in the OECD programme SERENA phase 1.     

Change of RBMK reactivity and power during measurements of the steam coefficient of reactivity

The results of a numerical simulation of an experiment on measurement of the steam coefficient of the reactivity of a RBMK reactor are presented. The “inverted variation of the reactivity” effect is explained. The computational results show that this effect is due to the nonuniformity of the load and coolant flow rate in the channels in the core. __________ Translated from Atomnaya énergiya, Vol. 100, No. 3, pp. 171–173, March, 2006.     

Ex-vessel fuel-coolant interaction experiment in the DISCO facility in the LACOMECO project

In the frame of the LACOMECO (large scale experiments on core degradation, melt retention and containment behavior) project of the 7th European Framework Program, a test in the DISCO (dispersion of corium) facility was performed in order to analyze the phenomena which occur during an ex-vessel fuel–coolant interaction (FCI). The test is focused on the premixing phase of the FCI with no trigger used for explosion phase. The objectives of the test were to provide data concerning the dispersion of water and melt out of the pit, characterization of the debris and pressurization of the reactor compartments for scenarios, where the melt is ejected from the reactor pressure vessel (RPV) under pressure. The experiment was performed for a reactor pit geometry close to a French 900 MWe reactor configuration at a scale of 1:10. The corium melt was simulated by a melt of iron–alumina with a temperature of 2400 K. A containment pressure increase of 0.04 MPa was measured, the total pressure reached about 0.24 MPa. No spontaneous steam explosion was observed. About 16% of the initial melt (11.62 kg) remained in the RPV vessel, 60% remained in the cavity mainly as a compact crust. The fraction of the melt transported out of the pit was about 24%.     

Evaluation of Containment Failure Probability by Ex-Vessel Steam Explosion in Japanese LWR Plants

The containment failure probability due to ex-vessel steam explosions was evaluated for Japanese BWR and PWR model plants. A stratified Monte Carlo technique (Latin Hypercube Sampling (LHS)) was applied for the evaluation of steam explosion loads, in which a steam explosion simulation code JASMINE was used as a physics model. The evaluation was made for three scenarios: a steam explosion in the pedestal area or in the suppression pool of a BWR model plant with a Mark-II containment, and in the reactor cavity of a PWR model plant. The scenario connecting the generation of steam explosion loads and the containment failure was assumed to be displacement of the reactor vessel and pipings, and failure at the penetration in the containment boundary. We evaluated the conditional containment failure probability (CCFP) based on the preconditions of failure of molten core retention within the reactor vessel, relocation of the core melt into the water pool without significant interference, and a strong triggering at the time of maximum premixed mass. The obtained mean and median values of the CCPF were 6.4x 10^?2 (mean) and 3.9x 10^?2 (median) for the BWR suppression pool case, 2.2x10^?3 (mean) and 2.8x10^?10 (median) for the BWR pedestal case, and 6.8X10^?2 (mean) and 1.4x10^?2 (median) for the PWR cavity case. The evaluation of CCFPs on the basis of core damage needs consideration of probabilities for the above-mentioned preconditions. Thus, the CCFPs per core damage should be lower than the values given above. The specific values of the probability were most dependent on the assumed range of melt flow rate and fragility curve that involved conservatism and uncertainty due to simplified scenarios and limited information.     

Contribution of γ radiation to radiation embrittlement of vessel steel

Data on the contribution of reactor γ rays to the embrittlement of nuclear reactor vessel steel are presented and discussed. It is shown qualitatively that the increase of the shift of the critical brittleness temperature in the region of the outer layers of propulsion-reactor vessels and, conversely, the decrease of this shift at the same locations in power reactors correlates with the γ radiation intensity, which changes as a result of the structural features of the reactors. In addition, it is shown that the previously published paradoxical experimental results on the irradiation of A-537 steel in the EL-3 reactor with and without cadmium foil as well as on the anomalous embrittlement of domestic vessel materials under irradiation in the core of a propulsion reactor are explained by the effect of radiation γ annealing. Translated from Atomnaya énergiya,Vol. 106, No. 1, pp. 22–28, January, 2009.     

Computational investigation of the effect of hydraulic and nondestructive test parameters on the strength reliability of the heat exchanger tubes of a steam generator

The results of calculations of the probability of a leak appearing in the tube band of steam generator in a VVéR-440 reactor system during operation are presented. The MAVR-1.1 computer code is used to calculate the probability of the formation of a leak and rupture of one of the heat exchanger tubes. The binomial distribution is used to determine the probability of the number of tubes that do not satisfy the plugging criterion. A leak in a tube bank is calculated as a sum of leaks in individual tubes. The probability of such a leak is calculated as a random sum. The calculations show that the parameters of test measures (pressure of the hydraulic tests, reliability of nondestructive testing for defects) and the sequence in which they are performed have a large effect on the failure probability of a tube bank during reactor operation. The computational results and the experience gained in operating steam generators show that the algorithm and the method developed for computing the leak probability could be helpful for estimating the strength reliability of heat exchanger tubes. __________ Translated from Atomnaya énergiya, Vol. 102, No. 4, pp. 216–221, April, 2007.     

Fuel coolant interaction experiments in TROI using a UO2/ZrO2 mixture

This paper discusses the results of steam explosion experiments using reactor material carried out under “Test for Real cOrium Interaction with water (TROI)” program. About 4–9 kg of corium melt jet is delivered into a sub-cooled water pool at atmospheric pressure. Spontaneous steam explosions are observed in four tests among six tests. The dynamic pressure, dynamic load, and morphology of debris clearly indicate the cases with steam explosion. The initial conditions and results of the experiments are discussed.     

Computational-experimental studies of the neutronics of the SM reactor with different variants of the neutron-trap arrangement

The results of experimental studies of the neutronics of the high-flux SM reactor with different arrangements of the neutron trap are presented. The MCU series of high-precision computer programs implementing the Monte Carlo method is used for computations. Experimental data on reactivity effects, the effectiveness of safety and control rods, and the coefficients of nonuniformity of energy release in the core have been obtained in experiments on a critical assembly – a physical model of the SM reactor – and directly in experiments in the reactor. The error is 4.2–10% in determining the reactivity parameters and 5–10% for the relative energy release in the fuel elements. Information on the neutron field formed in the volume of the neutron trap has been obtained for two arrangements of the beryllium and water moderators. The differential and integral energy spectra of the neutrons in the energy interval from 0.5 eV to 20 MeV are obtained for three points inside the trap (external, central series, center). The flux density of thermal, superthemal, and fast neutrons are determined.     

Particulars of the choice of radiation protection for planetary-surface nuclear power plants

The results of computational studies on choosing radiation protection for planetary-surface nuclear power plants are present. Protection on the base of a 0.4–1.5 MW(t) YaEU-100 thermionic space reactor was considered for a Martian nuclear power plant and a 0.36. and 0.6 MW(t) YaEU-25 reactor was considered for a lunar reactor. The mass/size characteristics of the radiation protection were obtained for different arrangements of the nuclear power plant on the planet — directly on the surface with protection delivered or an embankment consisting of local soil and in a shaft prepared beforehand. Translated from Atomnaya énergiya, Vol. 105, No. 2, pp. 72–79, August, 2008.     

Investigation of the partitioning of fission products between the metallic and oxide phases of melt in a VVER-1000 vessel during a serious accident

The results of investigations of the interaction of U-Zr-B-C-O melt with steel, performed as part of the OECD MASCA program, are presented. It is established that the presence of Mo, Ru, Sr, Ba, Ce, and La in the melt does not qualitatively affect the interaction with structural steel and the character of the stratification of the melt in the reactor vessel. The partition factors of the fission products between the oxide and metallic phases are determined as a function of the oxidation of the melt, the ratio U/Zr, the composition of the structural steel, and the temperature. __________ Translated from Atomnaya énergiya, Vol. 105, No. 1, pp. 3–7, July, 2008.     

FCI experiments in the aluminum oxide/water system

The KROTOS facility at JRC Ispra was recently used to study experimentally melt-coolant premixing and steam explosion phenomena in Al₂O₃/water mixtures with approximately 1.5 kg melt at 2300–2400 °C. In the five tests performed the main parameter was the water subcooling, 10, 40 and 80 X, respectively. In the nearly saturated system, steam explosions could be externally triggered, which resulted in high (supercritical) explosion pressures in the test tube: KROTOS 26, 28. Without triggering, melt penetration in water and melt agglomeration on the bottom plate of the test tube could be observed, which gave rise to strong steaming during the melt cooling-down process: KROTOS 27. In the two tests KROTOS 29, 30, performed with 80 K subcooled water, self-triggered steam explosions occurred with pressures of more than 100 MPa. Post-test analysis of the debris revealed that 85% of the interacting fuel mass fragmented in particles of sizes smaller than 250 μm. An energy conversion ratio of 1.25% was estimated from vessel pressurization data taking into account the energy content in the fuel mass which fragmented to particle diameters of less than 250 μm. The test section was damaged in the test KROTOS 30.     

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.     

Effect of pump-induced cold-leg swirls on the flow field in the RPV of the EPR™: CFD investigations and comparison with experimental results

A computational fluid dynamic (CFD) model for the pressure vessel of the evolutionary pressurized reactor (EPR™) was developed and validated. The aim of this model is the simulation of transients where three-dimensional effects play a strong role, such as boron dilution and main steam line break (MSLB) scenarios. First, a full solid (CAD) model has been built, that includes all details of the reactor pressure vessel (RPV) and the internals which are important for fluid dynamic analyses. The solid model has then been used as basis for the generation of the computational mesh necessary to carry out CFD simulations. Both a hexahedral and a polyhedral mesh have been created. The CFD model has been validated against experimental results of the JULIETTE facility, a 1:5 scaled mock-up of the EPR™ reactor RPV built by AREVA and equipped with advanced instrumentation.The performances of the hexahedral and the polyhedral meshes are investigated in relation to the agreement with experimental data, convergence and CPU requirements. In addition, the effect of the cold-leg swirls on the velocity field inside the RPV is investigated. These swirls mimic the effects of the main coolant recirculation pumps on the flow field at the entrance of the RPV. It is shown that the CFD model is able to capture the shift of the maximum velocity in the downcomer annulus observed in the experimental results. Good qualitative as well as quantitative agreement with the experimental data is achieved.     

Effect of melt water interaction configuration on the process of steam explosion

Steam explosion experiments are performed at various modes of melt water interaction configuration using prototypic corium melt. The tests are performed to simulate both melt water interaction in a partially flooded cavity and melt water interaction in a cavity with submerged reactor. The tests are performed using zirconia and corium melts. The behavior of melt jet fragmentation during the flight in the air and fragmentation and mixing of melt jet in water is investigated by a high-speed video visualization and by comparison of debris size distribution and morphology of debris. Strength of steam explosion is estimated by measuring dynamic pressure and dynamic force.     

A use of prototypic material for the investigation of severe accident progression

An experimental research platform using corium melts is established for the understanding of safety related important phenomena during a severe accident progression. The research platform includes TROI facility for corium water interaction experiments and VESTA facility for corium-structural material interaction experiments. A cold crucible technology is adapted and improved for a generation of 5–100 kg of corium melts at various compositions. TROI facility is used for experiments to investigate premixing and explosion behaviors during a fuel coolant interaction process. More than 70 experiments using corium at various compositions were performed to simulate steam explosion phenomena in a reactor situation. The results indicate that the conversion efficiency of steam explosion for corium is less than 1%. VESTA facility is used to investigate molten corium-structural material interaction phenomena. VESTA facility consists of two cold crucibles. One crucible is used for the melting of charged material and pouring of corium melt. The other crucible is used for the corium-structural material interaction while providing an induction heating to simulate the decay heat. The results of an experiment on the interaction between corium melt and a specimen made of Inconel performed in the VESTA facility is reported.     

Load carrying capacity of a reactor vessel head under a corium slug impact from a postulated in-vessel steam explosion

The load carrying capacity of the pressure vessel head to withstand an in-vessel steam explosion is investigated. Firstly, as a key problem, the impact of molten core material against the vessel head is studied by model experiments scaled down 1:10. Structural details are considered carefully. The results are converted to reactor dimensions using similarity theory. This approach was checked by simplified liquid-structure impact experiments in different scale. Secondly, the upward acceleration of molten core material is studied by computational models. As results the mechanical energies which the vessel head can withstand are presented.     

PGV-1500 steam generator. New problems and solutions

The basic questions concerning the development of a steam generator for a nuclear power plant with a VVé R-1500 reactor are presented. The basic design requirements which follow for steam generators from experience in operating analogs at nuclear power plants and taking account of the requirements for a reactor system are presented. The special features inherent to horizontal-type steam generators, which have been mastered and are used in nuclear power plants in our country, are noted. The domestic and world operating experience is taken into account in the development of the design. It is concluded that the design of the PGV-1500 steam generator satisfies the requirements for the concept of a VVéR reactor facility for a 1500 MW(e) unit of a nuclear power plant and is competitive on the world market for power-generating equipment for nuclear power plants. __________ Translated from Atomnaya énergiya, Vol. 99, No. 6, pp. 416–425, December, 2005. An erratum to this article is availabel at .     

Flow analyses using RELAP5/MOD3 code for OPR1000 under the external reactor vessel cooling

External reactor vessel cooling (ERVC) is considered as one of the most promising severe accident management strategies for an in-vessel corium retention (IVR). Heat removal capacity and water availability at the vessel outer surface can be key factors determining the success of ERVC measures. In this study, for the investigations on the effect of water availability in case of ERVC, flow analyses using the RELAP5/MOD3 code were performed. The analyses were focused to examine the flow behavior inside the reactor pressure vessel (RPV) insulator of the OPR1000 (Optimized Power Reactor 1000 MWe) under a cavity flooding. The current flow analyses results show that for the accident scenarios of station black out (SBO) and 9.6 in. large break loss of coolant accident (LBLOCA) under the ERVC, steam could not ventilate through the insulator and the pressure inside the RPV insulator increased abruptly. This induced a water sweep out and steam domination in the flow path inside the insulator. These flow analyses results indicate that sufficient water ingression and steam venting through the insulator can be a key factor determining the success of the ERVC in the operating nuclear power plant, OPR1000. According to the results of the sensitivity studies for the venting area, in terms of an effective flow circulation inside the insulator, an optimal venting is to assign four holes having a diameter of 0.3 m at the upper exit (hot leg level) of the insulator. And the effect of the inlet flow area at the insulator bottom is rather minor when compared to that of the outlet flow area of a steam venting.