Based on modern theoretical knowledge and the results of representative experimental studies, the phenomenology of reflooding of fuel assemblies is considered. The parameters (test section pressure, water subcooling, peak cladding temperature at the start of the flooding, bundle power, etc.) maintained in the experiments under consideration are close to those expected when implementing measures to manage hypothetical severe accidents at pressurized water reactors. A list of processes accompanying the reflood of fuel-rod assemblies has been formulated, and specific effects have been established that can lead to a change in the local conditions of heat exchange between the cladding of fuel-rod simulators and the steam-water mixture and affect their quenching. A comparison of the results of experimental studies showed the influence of cooling water flow rate on the spread of measured values of quench time in the upper part of the fuel assembly. The view of reflood physics allowed us to analyze the results of validation of the SOCRAT code in experiments of varying phenomenological complexity (in an intact core, with an intense steam-zirconium reaction, formation of a melt). The analysis showed that the SOCRAT code correctly predicts the temperature histories of the fuel-rod simulator claddings, the quench time, and the total mass of hydrogen released during the experiment with a tendency toward slight underestimation; the modeling results do not contradict the experimental data. During validation, it was established that the thermal hydraulics model makes the greatest contribution to the assessment of the model error in calculating the quench time and the total mass of hydrogen production when modeling experiments of varying phenomenological complexity. Good predictive capabilities of the SOCRAT code confirmed the applicability of a one-dimensional approach to modeling the reflooding of fuel assemblies.

The article describes a model for calculating the containment thermal in severe accidents at nuclear power plants with a water coolant. The model is implemented in the SOCRAT integral code in the form of a CONT_TH module, which allows performing self-consistent calculation of parameters in a VVER reactor plant and a containment during a severe accident. In conjunction with other modules of the SOCRAT code, a realistic calculation of the transfer of radioactive substances to the containment is provided, which is necessary to calculate source term into the environment when justifying the safety of nuclear power plants, including probability safety assessment level 2. With a slight adaptation, the model can be used to calculate the parameters of containment in a water-cooled NPP as part of other integral codes.

The paper presents the results of numerical modeling for ballooning and burst of fuel rods claddings made of domestic and foreign alloys. The integral code SOCRAT-V1/V2 is used as a means of modeling. The uncertainty analysis of the calculation results to input uncertainties of the temperature and pressure measurements was performed. The modeling results demonstrate a good qualitative and quantitative compliance with measured times of cladding failure under partial core uncovery conditions. The results of SOCRAT-V1/V2 validation evidence on the importance of performing new experiments for domestic fuel rod cladding ballooning and burst.

The paper presents the results of qualification analysis of experimental data on high-temperature oxidation of the Russian zirconium alloys in steam, which include an assessment of the consistency of measurement results in tests, and numerical modeling of experiments with SOCRAT-V1/V2 code. Conclusions are provided on the possibility of using the SSC RIAR experimental programs for validation of severe accident codes, and the applicability of the SOCRAT-V1/V2 code for prediction of the high-temperature oxidation of the Russian zirconium claddings of fuel rods.

This paper describes an approach to evaluate the uncertainties of the results of deterministic severe accident analysis. The approach was developed in order to follow best practices of best estimate methodology and to fulfill the Russian regulatory requirements NP-001-15 in the area of atomic energy use. In accordance with these requirements, deterministic safety analyses must be accompanied with the evaluation of errors and uncertainties of the obtained results. The application of the proposed approach is demonstrated through the example of a deterministic analysis of a severe LOCA initiated by a large break at a generic nuclear power plant with VVER-1000. Within the given example a best estimate value and associated uncertainty of hydrogen mass generated at in-vessel stage is analyzed.

This paper considers the issues of safety assessment of new nuclear power plant (NPP) projects with VVER Generation III+ reactors in relation to the probability target for large release, which is subject to verification in the development of a full-scale Probabilistic Safety Assessment (PSA) Level 2. The design solutions implemented in Generation III+ reactors allow reducing the probability of a severe accident (SA) due to internal initial events to a level of 10(-7)/year. Exceeding the radioactive release criterion may thus be related mainly to the consequences of beyond-design external events. This places special demands both for the selection of SA scenarios to consider and the methods for modeling the accident progression and consequences. The paper presents a method for selecting the representative SAs in the frame of PSA of new VVER NPP projects and a practical example of radiological analysis for two bounding accidents at an arbitrary NPP using an advanced integrated computer code system.

A quantitative assessment is made of the possibilities of the SOKRAT code to model the dissolution of uranium dioxide fuel by zirconium cladding melt at the initial stage of a serious accident at NPP with VVER. The methodological approach for the assessment is based on the ASME V&V 20 standard and includes an uncertainty analysis. The results of local high-temperature experiments studying the kinetics of the process are used as a technical base.

Specialized computer codes that model the behavior of aerosol particles propagating through a system of pipes or air ducts are used for assessment of aerosol particle deposition. Developed in Russia, SOCRAT/V3 is one such code. SOCRAT/V3 was used for modeling of the transport of radioactive aerosols containing the Cs-137 radionuclide through an air duct during a real emergency. The obtained results of the modeling were used to estimate the exposure dose rate (EDR) of gamma radiation near the air duct. The results of the estimation were compared with data of real measurements of the gamma-radiation EDR along the air duct. This paper proposes an approach to assessment of source term in the case of radioactive aerosol releases using (1) a thermophysical code (SOCRAT/V3), allowing modeling of physical processes that influence the formation and transport of aerosols, and (2) data of in situ measurements for the external EDR from contaminated air ducts.

The radioactive track formed on March 15, 2011 during the accident at the Fukushima-1 nuclear power plant (Japan) is reconstructed using the PROLOG–SOKRAT/V3 software system. The emission of radionuclides into the atmosphere and the radiation conditions are estimated taking account of the complex relief of the location. Comparing the computed activity, radionuclide composition and fallout density showed satisfactory agreement with actual measurements.

Some results of a numerical analysis of the serious unanticipated accident at the Fukishima-1 NPP are presented. The analysis was performed in March 2011 as part of the support given by Rosatom, including in making decisions about possible protection of the population in the Far East region of Russia. On the basis of calculations performed with the SOKRAT code, the state of the core during the first week and the possible reasons of the explosions in the first three power units were evaluated: the core of all three reactors melted and in course of its degradation from 950 to 1100 kg of hydrogen could have been formed followed by the formation of a burning mixture. The computational data on hydrogen emission due to oxidation of the melt by steam agree with experiments performed with the QUENCH and PARAMETR facilities. It can be asserted 1.5 years after the accident that post-accident studies confirm the main results of the real-time calculations.

During the serious accident at the Fukushima-1 NPP (Japan), the cores of the Nos. 2 and 3 units were cooled for a long period of time by safety systems which were not designed for operation under prolonged power loss. The serviceability of BWR-4 high-head makeup systems and the reasons for their failure are analyzed. It is shown that taking dc losses in the No. 2 unit into account the makeup system remained serviceable in the unanticipated regime. The shutdown of the makeup systems in both units could have been due to the maximum backpressure being reached at the turbine outlets or by cavitation in the turbopumps. It is shown that it is important to make a realistic assessment of the depletion time of the storage batteries in the No. 3 unit.

A brief description and the results of the verification and practical application of the Russian RATEG/SVECHA/GEFEST severe-accident code and the prospects of its development are presented.

The OECD benchmark exercise on TMI-2 plant was launched by the Working Group on the Analysis and Management of Accidents (WGAMA) of OECD/CSNI, in conjunction with the WP 5.4 “Corium and Debris Coolability – Bringing Research Results into Reactor Applications” of the EU/SARNET2 network of excellence. The main objective of this benchmark exercise is to investigate the ability of current advanced codes to predict invessel core melt progression and degraded core coolability by the analysis of different severe accident sequences, and comparing the various results from several computer codes. The planned work foresees the simulation of three representative severe accident sequences addressing core reflooding issue, starting from different degrees of core degradation, and molten core slumping into the lower plenum. Two accident sequences have been simulated in the time frame of SARNET2 activities. These first two accident sequences concern a small break loss of coolant accident (SBLOCA) and a surge line break (SLB) in station blackout (SBO) conditions. Both accident sequences were first analysed without high pressure injection until almost complete core melting, corium slumping and possible vessel failure. In a second step, core reflooding is simulated by start-up of high pressure injection at different time instants, corresponding to a pre-defined amount of core degraded materials. In this paper the code result comparison regarding the analysis of SBLOCA and SLB accident sequences and related reflooding scenarios is presented and discussed. Eleven organizations from eight countries are participating in this benchmark exercise using five different codes.