Accident Source Term Evaluation Code Research Articles

Research of aerosol removal by passive containment cooling system

Passive containment cooling system (PCCS) can effectively remove the heat from the containment and maintain the pressure and temperature in case of accident, thus ensuring the containment integrity. PCCS introduces thermophoresis and diffusiophoresis aerosol deposition and changes the distribution of fission products in the containment, thus affecting containment dose evaluation and off-site source term assessment. In this paper, a Generation III PWR nuclear power plant model with PCCS is established by using ASTEC (Accident Source Term Evaluation Code), and the removal effect of PCCS on fission product aerosol is studied under the large break loss of coolant accident. The results show that one train of PCCS can increase the aerosol removal factor for approximately 0.05/h. Taking five trains of PCCS in this study as an example, aerosol deposition effect introduced by PCCS reduces suspended aerosol in containment by 10 %–20 %, and the effect is more significant in the PCCS-locating compartment. Meanwhile, the gas flow among compartments changes the aerosol mass in the adjacency of the PCCS-locating compartment. This phenomenon weakens the removal effect in the PCCS-locating compartment and make the aerosol concentration in adjacent compartments tend to be consistent.

Validation and uncertainty analysis of ASTEC in early degradation phase against QUENCH-06 experiment

Severe Accident (SA) integral codes, such as the Accident Source Term Evaluation Code (ASTEC) developed by IRSN, are used to simulate the phenomena occurring during accident progression in Nuclear Power Plants (NPPs) up to the source term evaluation. Code validation against experimental data is fundamental to carry out deterministic safety analysis and apply these codes to NPPs. In addition, in the Best Estimate Plus Uncertainty (BEPU) framework, the quantification of the results uncertainty is needed. In the framework of the IAEA CRP I31033 “Advancing the State-of-Practice in Uncertainty and Sensitivity Methodologies for Severe Accident Analysis in Water-Cooled Reactors”, the QUENCH test-6 experiment, conducted at KIT, has been selected to develop an uncertainty analysis using the ASTEC v2.2b code. The accuracy of the best-estimate ASTEC simulation was evaluated with the Fast Fourier Transform Based Method (FFTBM) against the experimental data. Then, the uncertainty of the code results was quantified by using the probabilistic propagation of input uncertainties method, through the coupling of ASTEC with RAVEN (Risk Analysis and Virtual Environment). Beyond identifying the main sources of uncertainty affecting the simulated test, the outcomes of the work also include some general discussion on the uncertainty propagation in a SA sequence.

Application of the MOCABA algorithm and the FSTC tool for source term predictions during severe accident scenarios

This paper describes the capabilities of the Fast Source Term Calculation (FSTC) tool developed by the Karlsruhe Institute of Technology (KIT) to perform time series predictions of the radiological source term (ST) during severe accident (SA) scenarios. The prediction algorithm implemented in FSTC is based on the MOCABA data assimilation framework, developed by Framatome, and requires a training database and actual measurements. The training database is assessed by employing coupling FSTC to the European reference Accident Source Term Evaluation Code (ASTEC), developed by IRSN. The prediction capabilities of FSTC is applied to evaluate: 1) the mass of released hydrogen during the QUENCH-08 experiment; 2) the xenon release to the environment during a Medium Break Loss of Coolant Accident (MBLOCA) SA scenario in a generic KONVOI nuclear power plant (NPP). The prediction results show a very promising employment of FSTC and the MOCABA algorithm in view of supporting the emergency response team during SA scenarios.

MELCOR 2.2-ASTEC V2.2 crosswalk study reproducing SBLOCA and CSBO scenarios in a PWR1000-like reactor part II: Analysis of containment thermal-hydraulics and ex-vessel phenomena

The Accident Source Term Evaluation Code (ASTEC) and Methods of Estimation of Leakages and Consequences of Releases (MELCOR) system codes are both used by Tractebel to perform Severe Accident (SA) analyses for the Belgian Nuclear Power Plants (NPPs).This paper contains the second part of the comparative study between these two integral software focused on the SA progression in the containment after the Vessel Failure (VF).The main phenomena characterize this phase of the accident usually called ex-vessel phase are the Molten Corium Concrete Interaction (MCCI) and ex-vessel corium/debris coolability, which determine the containment Thermal-Hydraulics (TH) response and can cause its late failure. Despite their importance from a safety point of view, chemical and physical processes behind these phenomena are not yet well understood, partly because of their extreme complexity and partly because of the limitations of current experimental capabilities in reproducing them. As result, models developed and implemented in the codes to describe the MCCI and debris coolability are extremely simplified because based on limited data availability.The benchmark study consists of a comparison of the results calculated by the ASTEC and MELCOR codes during the ex-vessel phase following two different scenarios a Complete Station Blackout (CSBO) and a Small Break Loss Of Coolant Accident (SBLOCA).For each scenario, two different Severe Accident Management Strategies (SAMS) are adopted, the first one is to ensure that the corium, once fully relocated in the cavity, has top-flooding conditions throughout the transient, while the second one is to keep the reactor cavity completely dry. Furthermore, for every case reproduced the reference and best-practice models are investigated.The results show that in case of dry cavity conditions, despite the different VF timing calculated by the two codes, the final results provide very similar indications even though some dissimilarity is observed regarding the amount of fission product (FP) released through the Containment Filtering Venting System (CFVS).In contrast, in the case of wet cavity conditions, the accident progression predicted is significantly different regarding number of venting operations, TH containment conditions, and amount of concrete ablated.The discrepancies observed, should be sought in the corium and debris coolability models, which given their parametric nature and oversimplification can lead to these results.

Uncertainty and sensitivity analysis of the QUENCH-08 experiment using the FSTC tool

This paper presents results of a preliminary uncertainty and sensitivity (U&S) analysis of the QUENCH-08 experiment using the Fast Source Term Calculation (FSTC) tool, which currently is under development at KIT/INR as a part of WAME project. WAME is devoted to the development of a methodology and a tool for fast and accurate source term (ST) calculation in case of severe accident at nuclear power plant (NPP). The FSTC tool is divided into 2 parts: the U&S analysis and ST prediction (using the MOCABA approach), respectively. The description of the MOCABA approach, which is implemented in FSTC tool, will be presented – together with the results of ST evaluation – in a subsequent paper. In this paper, the FSTC tool is applied in conjunction with the Accident Source Term Evaluation Code (ASTEC) (version 2.2_b) severe accident (SA) code to the U&S analysis of the QUENCH-08 experiment to demonstrate the functionality of the tool and to check the methodology’s suitability for the application to a nuclear power plant.

Reassessment of Computational Tools for the Modeling of Heat Transfer in a Molten UO2 Pool in Natural Convection

Abstract This work, performed within the European sodium fast reactor safety measures assessment and research tools (ESFR-SMART) H2020 European project, is part of a larger framework intending to reassess the modeling of heat transfer in molten pools on SCARABEE available experimental results. This paper presents simulation results of the in-pile Bain Fondu (BF1) test, performed within the SCARABEE-N program, using accident source term evaluation code (ASTEC), Sn method implicit multifiel multiphase Eulerien recriticality (SIMMER) III, and SIMMER V simulation tools as well as comparison with its available experimental data. This program was performed in the 1980s in the frame of the Safety Assessment studies of Superphenix sodium-cooled reactor. This test was dedicated to verify the stability of a pure molten UO2 pool under decay heat conditions within natural convection and the long-term resilience of the peripheral fuel crust. The pool was generated in a stainless steel crucible by a progressive heating of a fuel pellet stack through six successive power plateaus. For the benchmark purposes, only the molten pool behavior k the last power plateau, where the pool was the largest and the fuel temperatures the highest, was investigated. Experimental data such as the axial profile of radial heat fluxes and heat transfer from the pool to the surrounding interassembly coolant as well as the axial profile of the peripheral fuel crust thickness were used for the reassessment of the simulation tools. In addition, other variables of interest not measured during the test, such as the radial and axial velocities in the pool, were also benchmarked. Finally, a critical analysis of the correlations and models used in the different simulation tools for the BF1 test modeling is also provided in the paper.

Severe accident simulation for VVER-1000 reactor using ASTEC-V2.1.1.3

Abstract The station black-out (SBO) is one of the main accident sequences to be considered in the field of severe accident research. To evaluate a nuclear power plant’s behavior in the context of this accident, the integral ASTEC-V2.1.1.3 code “Accident Source Term Evaluation Code” covers sequences of SBO accidents that may lead to a severe accident. The aim of this work is to discuss the modelling principles for the core melting and in-vessel melt relocation phenomena of the VVER-1000 reactor. The scenario of SBO is simulated by ASTEC code using its basic modules. Then, the simulation is performed again by the same code after adding and activating the modules; ISODOP, DOSE, CORIUM, and RCSMESH to simulate the ex-vessel melt. The results of the two simulations are compared. As a result of SBO, the active safety systems are not available and have not been able to perform their safety functions that maintain the safety requirements to ensure a secure operation of the nuclear power plant. As a result, the safety requirements will be violated causing the core to heat-up. Moreover potential core degradation will occur. The present study focuses on the reactor pressure vessel failure and relocation of corium into the containment. It also discusses the transfer of Fission Products (FPs) from the reactor to the containment, the time for core heat-up, hydrogen production and the amount of corium at the lower plenum reactor pressure vessel is determined.

Analysis of Sentitivity of Fission Product Iodine in Containment to Various Factors under Severe Accidents

The most serious consequence of reactor accident is that the radioactive fission products are dispersed into the environment. To study the distribution characteristics of radioactive fission product iodine in the containment under severe accident conditions, this paper assumes that the fission product iodine is released from the primary system into the containment due to the occurrence of severe accident in the nuclear power plant (NPP). Then, using the accident source term evaluation code (ASTEC), this study builds a containment construction model for the NPP, sets the boundary conditions, and calculates the chemical form, chemical properties, distribution, and change trends of different compounds, of the fission product iodine under different pH values, with or without silver (Ag) injection and under gaseous radiation conditions. The results show that the production of volatile iodine in the containment can be inhibited under alkaline conditions; the silver can promote the iodine trap in the liquid phase and reduce the iodine volatility; and the gaseous radiation environment can promote the formation of gaseous CH3I and IOx. As a result, this study can provide guidance for the removal of radioactive iodine in the containment under serious accidents.

MELCOR 2.2-ASTEC V2.2 crosswalk study reproducing SBLOCA and CSBO scenarios in a PWR1000-like reactor part I: Analysis of RCS thermal-hydraulics and in-vessel phenomena

The Accident Source Term Evaluation Code (ASTEC) and Methods of Estimation of Leakages and Consequences of Releases (MELCOR) system codes are used in Tractebel to perform source term assessments and plant Thermal Hydraulic (TH) analyses in case of Severe Accident (SA) for the Belgian Nuclear Power Plants (NPPs).These system codes were developed through intense and extensive validation activities which were mainly based on the verification of the capability of implemented physical models to reproduce the experimental results of Separate Effect Tests (SETs) and Coupled Effect Tests (CETs).However, the SETs, and in less extension the CETs, do not take into account interactions and possible synergistic or antagonistic effects which can arise among the different physical and chemical processes occurring during a SA. Thus, if a code is able to reproduce correctly most of SETs or CETs, this does not mean that it can ensure the same quality of results for the more realistic and complex Integral Effect Tests (IETs).Nevertheless, some contraindications exist also for the IETs, because of the scaling factor effect which, being difficult to quantify, could lead to development of models replicating correctly the experimental data but that are not totally reliable for reactor scale applications.In addition, waiting for the dismantling of Fukushima Dai-ichi units 1–2 and 3, which can provide new information in the field of the core degradation phenomena, in the present only data on in-vessel core melt progression for the Three Mile Island, Unit 2 (TMI-2) accident are available.Therefore, apart from simulating the TMI-2 accident scenario, there is no other way to assess models of in-vessel phenomena (H2 production, corium pool/ debris formation and relocation, etc) implemented in the codes on reactor scale.As a consequence, there are still uncertainties in the capability of these integral tools to predict the progression of postulated severe accident transients in a NPP, notably core degradation and ex-vessel phenomena.In order to investigate the differences among the physical models adopted in MELCOR 2.2 and ASTEC V2.2 codes, and evaluate the impact that they could have on the assessment of Severe Accident Management Strategies, a crosswalk study was carried-out.This paper describes the first phase of this work, which consists of a detailed comparison between these two SA tools on a well-defined plant (PWR1000-Like) with prescribed boundary and initial conditions.Furthermore, to avoid additional and unwanted sources of discrepancies between code calculations, the two reactor models were developed in parallel following the recommendations suggested by developers for core and Reactor Coolant System (RCS) nodalization.The comparative study is focused on the TH of the RCS and in-vessel phenomena, for two different accident scenarios (SBLOCA and CSBO) up to the moment of lower-head failure.Ex-vessel behaviour will be examined in the second phase of this study.The crosswalk results obtained have shown some differences and similarities on reproducing TH in the RCS and In-vessel phenomena.Especially for these latter, the discrepancies predominate mainly due to how the codes treat corium behaviour, while thanks to the harmonisation of the initial steady-state and boundary conditions, the discrepancies on the prediction of the RCS behaviour have been minimized, at least before significant core degradation takes place.

Uncertainty & sensitivity analyses for LOCA simulation of PHEBUS experiment FPT-1 using a modified SUNSET code

One task of the regulatory body during the review and assessment process is the quantification of uncertainty and sensitivity (U&S) during normal and accident conditions. Nowadays there are many codes having the capability of estimating U&S, but few of them can estimate uncertainty and sensitivity during severe accident conditions. The IRSN uncertainty propagation and sensitivity analysis tool named SUNSET is considered one of the most important tools used in U&S analysis, where it is coupled with ASTEC software package (Accident Source Term Evaluation Code) for computing U&S during severe accident condition. Unfortunately, the SUNSET tool cannot perform U&S analyses during all the transient time. The SUNSET software can only calculate U&S parameters for maximum, minimum, and last output values, therefore it was necessary to modify the SUNSET tool to calculate U&S during the whole transient time. The main objective of this study is to develop a Python script to replace the SUNSET post-processing step, extract the results, and perform U&S analysis. ASTEC code has been used for the simulation of a loss of coolant accident in the cold leg of the primary circuit of the FPT-1 test. The obtained results indicate the efficiency of the modified SUNSET tools to calculate the uncertainty during severe accidents. In general, it was found that the code is capable to model the main severe accident phenomena. SUNSET package shows good agreement with the experimental data, for the whole calculation period, until thermocouples failure moment. The obtained results show that the main safety parameters are very sensitive to the power multiplier.

Validation of Selected Cesar Friction Models of the ASTECV21 Code Based on Moby Dick Experiments

In the event of a loss of integrity of the main coolant line, a large mass and energy release from the primary circuit to the containment is to be expected. The temporal evolution of such depressurization is mainly governed by the critical flow, whose correct prediction requires, in first place, a correct description of the different friction terms. Within this work, selected friction models of the CESAR module of the Accident Source Term Evaluation Code (ASTEC) V2.1 integral code are validated against data from the Moby Dick test facility. Simulations are launched using two different numerical schemes: on the one hand, the classical five equation (drift flux) approach, with one momentum conservation equation for an average fluid plus one algebraic equation on the drift between the gas and the liquid; on the other hand, the recently implemented six equation approach, where two differential equations are used to obtain the phase velocities. The main findings are listed hereafter: The use of five equations provides an adequate description of the pressure loss as long as the mass fluxes remain below 1.24 kg/cm2 s and the gas mass fractions below 5.93 × 10 − 4. Beyond those conditions, the hypotheses of the drift flux model are exceeded and the use of an additional momentum equation is required. The use of an additional momentum equation leads to a better agreement with the experimental data for a wider range of mass fluxes and gas mass fractions. However, the qualitative prediction for high gas mass fractions still shows some deviations due to the decrease of the regular friction term at the end of the test section.

Modelling issues related to molten pool behaviour in case of In-Vessel Retention strategy

The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and retain the core melt in the reactor pressure vessel. This type of Severe Accident Management (SAM) strategy has already been incorporated in the SAM guidance (SAMG) of several operating small size LWR (reactors below 500MWe, like VVER440) and is part of the SAMG strategies for some Gen III+ PWRs of higher power like the AP1000.The demonstration of IVR feasibility for high power reactors requires using less conservative models as the margins are reduced.In this paper modelling issues related to molten pool behaviour in case of IVR are highlighted based on sensitivity studies performed with the severe accident integral code ASTEC (Accident Source Term Evaluation Code) for a typical PWR reactor. In particular, the following physical processes are addressed:-Heat transfer coefficients in thin light metal layer and in heavy metal layer.-Kinetics of corium stratification and thermal exchanges between oxide and metal phases during inversion of stratification.-Relocation of steel coming from ablated vessel wall.-Characteristics of oxide crust.Analyses based on a simple configuration of corium in the lower head allow identifying potential impact of each modelling assumption on the heat flux distribution along the vessel wall, especially in transient situations. Based on this study, the most critical areas where the knowledge is too limited or where the current analytical methods should be improved are identified as being (i) the duration of situations with thin metal layer on the top of the oxide, associated to the knowledge of the kinetics of corium stratification and (ii) the possible super-heat of the metal layer at stratification inversion coupled with the risk of oxide crust failure.This work is performed in the frame of the European project IVMR (In-Vessel Melt Retention) coordinated by IRSN. This project has been launched in 2015 and gathers 27 organizations. Its main objective is the evaluation of feasibility of IVR strategy for LWR (PWR, VVER, BWR) of total power 1000 MWe or higher.

Stand-Alone Containment Analysis of the Phébus Fission Products Test 1 With the ASTEC and the MELCOR Codes

The estimation of fission products (FPs) release from the containment system of a nuclear plant to the external environment during a severe accident (SA) is a quite complex task. In the last 30–40 yr, several efforts were made to understand and to investigate the different phenomena occurring in such a kind of accidents in the primary coolant system and in the containment. These researches moved along two tracks: understanding of involved phenomenologies through the execution of different experiments and creation of numerical codes capable to simulate such phenomena. These codes are continuously developed to reflect the actual SA state of the art, but it is necessary to continuously check that modifications and improvements are able to increase the quality of the obtained results. For this purpose, also a continuous verification and validation work should be carried out. Therefore, the aim of the present work is to re-analyze the Phébus fission products test 1 (FPT-1) test employing the accident source term evaluation code (ASTEC) and MELCOR codes (respectively, ASTEC v.2.0 revision 3 patch 3 and MELCOR V2.1.6840 version). The analysis focuses on the stand-alone containment aspects of the test, and three different modelizations of the containment vessel have been developed showing that at least 15/20 control volumes (CVs) are necessary for the spatial schematization to correctly predict the test thermal hydraulics and the aerosol behavior. Furthermore, the paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and FPs behavior.

Extension of ASTEC-Na capabilities for simulating reactivity effects in Sodium Cooled Fast Reactor

The EU-JASMIN project (7th FP of EURATOM) has been centred on the development and validation of the new severe accident analysis code ASTEC-Na (Accident Source Term Evaluation Code) for Sodium-cooled Fast Reactors (SFR). The development of such computational tool being able to assist safety analysis of innovative reactor concepts is of crucial importance. One of the challenging issues when modelling SFRs is the neutronic reactivity feedbacks. This paper presents the model implemented in ASTEC-Na for representing the reactivity effects in SFR as well as the benchmarking results of a ULOF transient against SAS-SFR code results. It has been verified that the models are correctly implemented and that ASTEC-Na is now able to calculate reactivity feedbacks not only in the sodium single phase, but also after boiling onset and fuel in-pin relocation.

A numerical methodology for estimation of volatile fission products release from nuclear fuel

In this paper we have developed a numerical methodology which is capable of estimating volatile fission product release from nuclear fuel under changing irradiation conditions with incorporation of all physical phenomena’s. The present study bridges the gap between previous analytical studies by providing estimation of unstable fission product release under accidental condition with and without incorporation of precursor’s history effect. Based on the present methodology a computer model “FIPRAP-Fission Product Release Analysis Program” has been developed and validated with standard fission product release problems and benchmarked against internationally accepted CORSOR-BOOTH model as well as Accident Source Term Evaluation Code (ASTEC). The result of the proposed model lies within 15% of internationally accepted models.

Core Melt Composition at Fukushima Daiichi: Results of Transient Simulations with ASTEC

The Accident Source Term Evaluation Code (ASTEC) is used to perform numerical simulations of the accidents at the Fukushima Daiichi nuclear power station in the frame of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) project. At present, simulations are available for Units 1, 2, and 3 of Fukushima Daiichi and for 6 days from the earthquake. A clear lesson from phase 1 of the project was that the uncertainties in the functioning of the safety systems and in accident progression are still large and there are many explanations for the measured thermohydraulic behavior. Rather than focusing on the thermohydraulic key parameters for which comparisons with measurements are available, this paper will address melt composition computation results that may provide insights relevant for the decommissioning process.When molten corium relocates from the core down to the vessel lower head, the melt jets interact with water and may be totally or partially fragmented depending on the level of water. A U-Zr-O-Fe molten pool may form in the lower head, and because of chemical reactions, separation between nonmiscible metallic and oxide phases may occur. The models implemented in ASTEC enable the simulation of these phenomena. Up to five different axisymmetric corium layers in the vessel bottom head can be formed, which are, from bottom to top, a debris layer, a heavy metallic layer, an oxide layer, a light metallic layer, and another debris layer. An important process is the UO2 fuel reduction to metallic uranium by nonoxidized zirconium, which results in uranium transport to the dense metallic layer as demonstrated in the MAterial SCAling (MASCA) program.Because of the large consensus on the accident progression of Fukushima Daiichi Unit 1, in this paper we present complex melt compositions before vessel failure for the current best-estimate cases for Unit 1. We do not present similar work performed for Units 2 and 3.It should be underlined that in the case of vessel bottom failure, a part of this complex melt will relocate to the pedestal and molten core–concrete interaction will take place enhancing other complex physical phenomena with possible large consequences on the melt chemical composition and behavior.

Low doses exposure to uranium from in vivo/in vitro experiments: Deleterious or adaptive responses induced?

This paper presents simulation results of the CABRI BI1 test using the code ASTEC-Na, currently under development, as well as a comparison of the results with available experimental data. The EU-JASMIN project (7th FP of EURATOM) centres on the development and validation of the new severe accident analysis code ASTEC-Na (Accident Source Term Evaluation Code) for sodium-cooled fast reactors whose owner and developer is IRSN. A series of experiments performed in the past (CABRI/SCARABEE experiments) and new experiments to be conducted in the new experimental sodium facility KASOLA have been chosen to validate the developed ASTEC-Na code.One of the in-pile experiments considered for the validation of ASTEC-Na thermal–hydraulic models is the CABRI BI1 test, a pure loss-of-flow transient using a low burnup MOX fuel pin. The experiment resulted in a channel voiding as a result of the flow coast-down leading to clad melting. Only some fuel melting took place.Results from the analysis of this test using SIMMER and SAS-SFR codes are also presented in this work to check their suitability for further code benchmarking purposes.

Analysis of ASTEC-Na capabilities for simulating a loss of flow CABRI experiment

This paper presents simulation results of the CABRI BI1 test using the code ASTEC-Na, currently under development, as well as a comparison of the results with available experimental data. The EU-JASMIN project (7th FP of EURATOM) centres on the development and validation of the new severe accident analysis code ASTEC-Na (Accident Source Term Evaluation Code) for sodium-cooled fast reactors whose owner and developer is IRSN. A series of experiments performed in the past (CABRI/SCARABEE experiments) and new experiments to be conducted in the new experimental sodium facility KASOLA have been chosen to validate the developed ASTEC-Na code.One of the in-pile experiments considered for the validation of ASTEC-Na thermal–hydraulic models is the CABRI BI1 test, a pure loss-of-flow transient using a low burnup MOX fuel pin. The experiment resulted in a channel voiding as a result of the flow coast-down leading to clad melting. Only some fuel melting took place.Results from the analysis of this test using SIMMER and SAS-SFR codes are also presented in this work to check their suitability for further code benchmarking purposes.

Evaluation of corium behavior in the lower plenum of the reactor vessel during a severe accident

Corium behavior in the lower plenum of the reactor vessel during a severe accident is very important, because this affects the failure mechanism of the lower spherical reactor vessel and the thermal load to the outer reactor vessel under the IVR-ERVC (in-vessel corium retention through external reactor vessel cooling). A detailed analysis of the in-vessel melt progression during a SBLOCA (small break loss of coolant accident) without the SI (safety injection) of the APR (advanced power reactor) 1400 was conducted to determine the corium mass in the lower plenum upon a reactor vessel failure using the SCDAP/RELAP5 computer code. GEMINI code analyses were conducted to determine the final thermodynamic states of the melt compositions using the SCDAP/RELAP5 results. The heights of the metallic and oxide layers are 1.46m and 0.46m, respectively. Finally, the corium behavior in the lower plenum of the reactor vessel under the IVR-ERVC was analyzed using the ASTEC (accident source term evaluation code) computer code. A two layer corium formation of the lower oxide and the overlying metallic layers was analyzed. The corium temperature, the lower head vessel temperature, and the final reactor vessel geometry in the lower plenum under the IVR-ERVC were estimated for the APR1400.

Thermodynamic properties of gaseous ruthenium species.

The review of thermodynamic data of ruthenium oxides reveals large uncertainties in some of the standard enthalpies of formation, motivating the use of high-level relativistic correlated quantum chemical methods to reduce the level of discrepancies. The reaction energies leading to the formation of ruthenium oxides RuO, RuO2, RuO3, and RuO4 have been calculated for a series of reactions. The combination of different quantum chemical methods has been investigated [DFT, CASSCF, MRCI, CASPT2, CCSD(T)] in order to predict the geometrical parameters, the energetics including electronic correlation and spin-orbit coupling. The most suitable method for ruthenium compounds is the use of TPSSh-5%HF for geometry optimization, followed by CCSD(T) with complete basis set (CBS) extrapolations for the calculation of the total electronic energies. SO-CASSCF seems to be accurate enough to estimate spin-orbit coupling contributions to the ground-state electronic energies. This methodology yields very accurate standard enthalpies of formations of all species, which are either in excellent agreement with the most reliable experimental data or provide an improved estimate for the others. These new data will be implemented in the thermodynamical databases that are used by the ASTEC code (accident source term evaluation code) to build models of ruthenium chemistry behavior in severe nuclear accident conditions. The paper also discusses the nature of the chemical bonds both from molecular orbital and topological view points.