In the code validation process, a standard approach to quantify the capability of an individual model to simulate a relevant process or phenomenon that occurs during an accident at NPP implies calculation of the bias, or a difference between the calculated and measured values for parameters of interest. The bias values do not always provide information about the accuracy of the mathematical model since the uncertainties that arise in numerical modeling and experiments are ignored. The paper discusses the modeling error, associated with approximations and simplifications in the boron carbide oxidation model (part of the best estimate severe accident code SOCRAT). Isothermal and transient BOX tests in the temperature range of 800-1500 degrees C are used as a reference basis for validation. The modeling error is estimated by an advanced approach based on the ASME V&V 20 standard with account for uncertainties associated with the input data and the measured values.

The SOCRAT integral code is currently the main tool for safety studies of NPPs with VVER during severe accidents. It is applied for deterministic calculations of severe accident progression in NPP licensing, support for the design of safety systems (steam and hydrogen sources into the containment to address hydrogen issues, sources of ex-vessel corium for the development of core catcher, verification of the efficiency of the system for in-vessel retention of molten corium), emergency preparedness and response, and support for the PSA-2. The containment thermal-hydraulics is simulated with the built-in CONT_TH module. The paper presents results of the CONT_TH validation against five tests performed at PSI at the middle-scale PANDA facility in the framework of EURATOM-ROSATOM ERCOSAM-SAMARA projects. The tests address containment pressurization during coolant injection, stratification build-up during light gas injection, and the effect of safety system operation (spray, cooler, heater) on gas distribution. The tests are simulated using a coarse nodalization scheme comprising 13 cells. The model capabilities are assessed based on the comparison of the predictions with the data as well as with earlier calculations of these tests performed by the project participants using CFD and LP codes. The main features of the CONT_TH models are also briefly described.

This paper is the second part of the summary presenting the results of IBRAE work over last 12 years on severe accident investigation at Unit 1 of Fukushima Daiichi with the SOCRAT code. The paper supplements the part I that was published previously, and is dedicated to the description of one-through integral simulation results of the 3 weeks of accident unfolding starting from the tsunami arrival. The integral simulation encompasses the different processes at in-vessel and ex-vessel phases of the accident including the processes in reactor building. The important issues discussed are the hypothesis on the induced leak from a steam line, the operation of isolation condenser, the possible early combustion of hydrogen in the reactor building room that houses the drywell head, as well as the time, conditions and regime of hydrogen combustion in the reactor hall, the flowrate of alternative water injection, and the depth of concrete ablation. The main criterion for checking the adequacy of assumptions and the simulation results is the agreement with factual information available including the measurement data and the results of inverse calculations of the source term. The shortcomings of current simulation with the SOCRAT code are briefly considered as a perspective for improving the realism of results in the next studies.

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.

Modelling of severe accidents (SA) at nuclear power plants (NPP) includes assessing the radioactive material release into the environment (called the << source term >>). The accuracy of the source term assessment using integral codes depends on the modeling error of individual phenomena such as the build-up of actinides and fission products (FP) in fuel prior to the accident, evolution of isotopic composition of fuel during the accident, release from fuel, transport, and deposition of actinides and FP in the reactor cooling system (RCS) and containment, their release into the environment. The application of best estimate approach with uncertainty analyses, the deterministic support of probabilistic safety assessment (PSA) level 2, and calculations within emergency response require multiple calculations of SA scenarios, therefore the integral codes have to run pretty fast. For consideration of isotopic evolution both in the irradiated fuel and in chemical compounds released from the fuel a depletion model has to be regularly called during each calculation session. To fit the requirement of smaller runtime, the integral code has to be provided with a fast-running depletion model. The fast performance of such models is payed by the predictive accuracy; therefore, it is important to assess if the accuracy is acceptable for the mentioned issues of SA analyses. This paper presents the results of study of errors in the fast-running depletion model that is intended for use in integral code during multi-variant calculations of SA scenarios. The validated parameters are concentrations of actinides and FP in light water reactor (LWR) spent fuel. Additionally, calculations of the decay heat power are provided. The basis for validation consists of the out-of-pile measurement results provided in Nuclear Energy Agency of Organization for Economic Cooperation and Development (OECD/NEA) database SFCOMPO-2.0, and the ORIGEN-2 code calculation results obtained by Japan Atomic Energy Agency (JAEA) for Units 1-3 of the Fukushima Daiichi NPP. (c) 2021 Elsevier Ltd. All rights reserved.

The accuracy of system code SOCRAT/V3 to simulate the heat transfer in a vertical cylindrical pool of volumetrically heated liquid is estimated. The methodological approach for the estimation implements the ASME V&V 20 standard and includes uncertainty analysis. The study relies on simulation results of four experiments at the BAFOND facility. In the simulation of these experiments, the maximum fluid temperature in a pool is considered as major parameter. The implementation of the approach based on the ASME V&V 20 standard demonstrates that the “traditional” approach for the code accuracy estimation based on calculating the deviation of the calculated value from the measured one with subsequent statistical processing, may result in the underestimation of the modeling error interval.

The mathematical models implemented in the CABARET-SC1 code are described and the code validation results characterizing its potential in the description of key processes influencing hydrogen propagation in the reactor containment in the course of a severe accident are presented. Agreement with experimental data was achieved by refining the computational grids and adding a radiative heat-transfer model with no adjustable parameters or semi-empirical models.

The paper discusses the results of numerical assessment of LWR fuel bundle degradation phenomena during heating up in a flowing steam and top flooding with water under severe accident conditions that were investigated in the PARAMETER-SF1 test. The simulation was performed with the code SOCRAT that is widely used in Russia for safety assessment of VVER-type NPP under severe accident conditions. A set of interacting processes, including heat transfer, fuel liquefaction, oxidation of fuel rods claddings, melt relocation and oxidation, hydrogen generation is discussed. PARAMETER-SF1 was the first test in the PARAMETER program that comprised several tests with different configurations of the VVER bundle flooding (top, bottom, and combined). Investigation of the phenomenology during flooding of an overheated core at the early phase of severe accident when core keeps a mainly rod-like geometry is extremely important for understanding the possibilities and conditions for implementing the measures of severe accident management at LWR NPPs.

Experiments on overall fluid transfer and mixing in the large scale experimental facility (PANDA), where the initial helium-rich layer is eroded by a flow resulting from the interaction of a jet/plume with horizontal flow obstruction, were simulated with the CABARET scheme. Computational simulation results are provided for two mesh resolutions and are compared with the experimental data. Good agreement between calculated and measured gas temperature and the breakup of the helium layer dynamics was achieved by refining the mesh in the jet area and by adding the model of radiation heat transfer.

Experiments on atmosphere mixing and stratification in the large scale experimental facility (PANDA), where the initial helium stratification in a vessel is eroded by a vertical jet of steam, were simulated with the CABARET scheme. Good agreement between calculated and measured gas temperature and concentrations has been achieved by refining the mesh and adding the model of radiative heat transfer. The simulation results are presented and discussed.

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.

The software tools that describe various safety aspects of NPP with VVER reactor have been developed at the Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN). Functionally, the codes can be divided into two groups: the calculation codes that describe separate elements of NPP equipment and/or a group of processes and integrated software systems that allow solving the tasks of the NPP safety assessment in coupled formulation. In particular, IBRAE RAN in cooperation with the nuclear industry organizations has developed the integrated software package SOCRAT designed to analyze the behavior of NPP with VVER at various stages of beyond-design-basis accidents, including the stages of reactor core degradation and long-term melt retention in a core catcher. The general information about development, validation and applications of SOCRAT code is presented and discussed in the paper.

Analyses of the CORA-15 bundle test have been performed with the system codes ATHLET-CD (GRS) and SOCRAT (IBRAE). In the test, the behaviour of a PWR type fuel bundle, composed of 23 fuel rods and 2 absorber rods, was investigated under severe accident conditions. An important feature of the experiment different from all other CORA tests was the pressurization of fuel rods to 6.0 MPa. As a result, the fuel rods underwent ballooning and burst. In general, both codes adequately reproduced the cladding temperatures histories, the ballooning and rupture, and blockage formation due to melt relocation. The hydrogen release has been calculated within the uncertainty of the measured data.

A model of a passive autocatalytic hydrogen recombiner (RVK-500, -1000) for use in hydrodynamic calculations of hydrogen transport and recombination processes in a VVER containment shell during a severe accident is described. The model includes calculation of the efficiency of the recombiner, the hydraulic resistance to gas flow through the recombiner, and heat losses owing to radiation from the casing of the recombiner. The model was parameterized and verified on experiments performed on a stand at VTI and INPK RET for hydrogen-air compositions with hydrogen volume fraction up to 10%. Good agreement was obtained between the calculations and experiments. Its simplicity and efficacy make it possible to use the model in hydrodynamic calculations of processes with recombiners in a real containment shell.

Computational hydrodynamics was used to perform, within the scope of the ERCOSAM–SAMARA international projects, calculations of the S1, S2 experiments on the SPOT ZO setup with a condenser-heat exchanger. The basic effects observed in the experiments are reproduced: helium pressure growth and stratification upon injection, pressure reduction and partial elimination of stratification during the operation of the condenser-heat exchanger. On the whole, the experiment agreed well with the calculations of the pressure, temperature, and composition of the gas. The small quantitative discrepancies with experiment indicate a possible impact of neglected factors, mainly, associated with inadequate knowledge of the details of the condensation/evaporation processes at and near the walls as well as with general sensitivity to variations of the initial conditions. Specifically, the deviation of the initial wall temperature or the total heat capacity of the wall appreciably influences the behavior of the pressure.

The ERCOSAM project (together with the SAMARA project) includes a set of multi-stage experiments carried out at different thermal-hydraulics facilities (TOSQAN, MISTRA, PANDA, SPOT) and their numerical simulation. The test sequences aim to investigate hydrogen concentration build-up and stratification during a postulated severe accident as well as the effect of activation of Severe Accident Management systems (SAMs) on stratification. At the initial Phases I-III of the tests stratification of the injected light gases (steam, helium) is established, pressure increases. The models of SAMs are activated at the final Phase IV.In the paper, two spray tests PE1 and PE2 performed at PANDA facility are numerically investigated aiming for code validation (FLUENT) and better understanding of physical processes. Phases I-IV of the tests were simulated. A satisfactory agreement with the experimental results on gas component mole fractions was obtained. The reasons of obtained some discrepancies in pressure and temperature were studied in multivariate calculations and discussed in the paper. The tests PE1 and PE2 were performed with different spray nozzles and initial conditions. In simulation that resulted in different flow patterns during spray operation. The sensitivity to nozzle angle and some other input parameters was investigated. Possible factors of depressurization rate caused by spray operation were also studied. The key features observed in the experiments and obtained numerically being of interest in terms of CFD applications to NPP hydrogen safety are summarized. (C) 2015 Elsevier B.V. All rights reserved.

The two-dimensional thermohydraulic model used in the thermohydraulic module of the SOKRAT-BN code is described and the computational results obtained by using it are presented. The thermohydraulic model is based on the solution of two-dimensional equations using empirical correlations for determining the intensity of the turbulent transfer of mass, energy and momentum in the radial direction. The experiments performed in the 19- and 37-pin assemblies on the Siena stand in Japan and in the 37-pin assembly in the KNS stand in Germany were calculated. The average deviations of the computed values do not exceed 10%.

Air ingress is a potential risk in some low probable situations of severe accidents in a nuclear power plant. Air is a highly oxidizing atmosphere that can lead to an enhanced Zr-based cladding oxidation and core degradation affecting the release of fission products. This is particularly true speaking about ruthenium release, due to its high radiotoxicity and its ability to form highly volatile oxides in a significant manner in presence of air. The oxygen affinity is decreasing from the Zircaloy cladding, fuel and ruthenium inclusions. It is consequently of great need to understand the phenomena governing cladding oxidation by air as a prerequisite for the source term issues in such scenarios. In the past years, many works have been done on cladding oxidation by air under severe accident conditions.This paper with in addition the paper "Part II: Synthesis of modeling results" of this journal issue aim at assessing the state of the art on this phenomenon.In this paper (Part I), the phenomenological approach is based on the analysis of experiments at different scales as separate-effect tests and integral tests like CODEX-AIT (low degree of pre-oxidation, high air flow rate), QUENCH-10 (high degree of pre-oxidation, moderate air flow rate, quench initiation at high temperature), PARAMETER-SF4 (moderate degree of pre-oxidation, low air flow rate, quench initiation at high temperature) and QUENCH-16 (moderate pre-oxidation, very low air flow rate, quench initiation at moderate temperature). The following phenomena will be described: formation of a porous nitride-oxide layer, subsequent steam penetration through this porous layer during reflood, accompanied by nitride re-oxidation and intensive cladding oxidation. (C) 2016 Elsevier Ltd. All rights reserved.

The SOCRAT computational code was developed and is used for modeling serious accidents at NPP. The present article analyzes the possibility of using the physicochemical models implemented in the SOCRAT code to calculate the processes arising in an RBMK-1000 core during serious accidents. It is shown that the modern version of the code can be used to model the rupture of RBMK fuel assemblies during accidents with complete blackout of a power-generating unit. However, individual models require additional verification under conditions characteristic for RBMK. Particular little-studied processes leading to the destruction of an RBMK core, requiring new models to be developed or extant models adapted for the SOCRAT code, are examined.