Severe Accident Analysis Code Research Articles

Assessment of the severe accident code MIDAC based on FROMA, QUENCH-06&16 experiments

In order to meet the needs of domestic reactor severe accident analysis program, a MIDAC (Module In-vessel Degraded severe accident Analysis Code) is developed and maintained by Xi'an Jiaotong University. As the accuracy of the calculation results of the analysis program is of great significance for the formulation of severe accident mitigation measures, the article select three experiments to evaluate the updated severe accident models of MIDAC. Among them, QUENCH-06 is the international standard No.45, QUENCH-16 is a test for the analysis of air oxidation, and FROMA is an out-of-pile fuel rod melting experiment recently carried out by Xi'an Jiaotong University. The heating and melting model with lumped parameter method and the steam oxidation model with Cathcart-Pawel and Volchek-Zvonarev correlations combination in MIDAC could better meet the needs of severe accident analysis. Although the influence of nitrogen still need to be further improved, the air oxidation model with NUREG still has the ability to provide guiding significance for engineering practice.

Application of continuous Markov-chain Monte-Carlo method to multi-unit risk evaluations considering interdependence of accident progression among multiple units

ABSTRACT The accident in Fukushima Dai-ichi nuclear power plants reconfirms the necessity of the safety assessment considering multiple nuclear reactor units. However, consideration of the interdependency among safety systems or events in multiple units as well as the time dependency of accident progressions is difficult in the conventional event tree method, which is widely used in probabilistic risk assessments. Recently, the continuous Markov-chain Monte-Carlo (CMMC) method coupled with a plant safety analysis or a severe accident analysis code has been paid attention to address these issues. In the present study, the CMMC coupling method is applied to the risk assessment of multiple units to clarify the benefits and issues to be resolved of this method. Since the CMMC coupling method requires many executions of an accident analysis code, a meta-model that simplifies systems and physical phenomena in accidents is used in this study to reduce computational cost. Furthermore, the inverse transform sampling method is newly adapted. Numerical analyses of BWR accident under station blackout with loss of cooling capability are carried out considering the correlation among the availabilities of mitigation systems. The results suggested that the CMMC coupling method can quantitatively treat the interdependency and time dependency among events in multiple units.

Development of mechanistic cladding rupture model for integrated severe accident code ISAA. Part I: Module verification and application in CAP1400

Cladding rupture is an important demarcation point for severe accidents. Most severe accident analysis codes are based on simple parameter models to determine the damage of the cladding. This paper uses the thermal–mechanical model to develop a core Fuel Rod Thermal-Mechanical Behavior analysis (FRTMB) module to analyze the fuel mechanical behavior. The purpose is to judge the cladding rupture with the severe accident analysis code. The model consists of thermal, mechanical and material property parts. The correctness of the model is verified by comparing with the experimental results of related fuel rods. Finally, the FRTMB module is integrated into the severe accident analysis code to evaluate the thermal–mechanical behavior of fuel rods in different enrichment areas of CAP1400 reactor, so as to provide a basis for judging cladding rupture in severe accidents. The mechanical analysis results show that the gap will close earlier under high-power condition. The results explain the necessity for considering the fuel performance in core analysis and demonstrate applicability of the FRTMB module integrated into the severe accident analysis code.

Uncertainty studies on hydrogen source term with MAAP5 code

Abstract In the present study, a methodology is developed to quantify the uncertainties of special model parameters of the integral severe accident analysis code MAAP5. Here, the in-vessel hydrogen production during a core melt accident for Lungmen Nuclear Power Station of Taiwan Power Company, an advanced boiling water reactor, is analyzed. Sensitivity studies are performed to identify those parameters with an impact on the output parameter. For this, multiple calculations of MAAP5 are performed with input combinations generated from Latin Hypercube Sampling (LHS). The results are analyzed to determine the 95th percentile with 95% confidence level value of the amount of in-vessel hydrogen production. The calculations show that the default model options for IOXIDE and FGBYPA are recommended. The Pearson Correlation Coefficient (PCC) was used to determine the impact of model parameters on the target output parameters and showed that the three parameters TCLMAX, FCO, FOXBJ are highly influencing the in-vessel hydrogen generation. Suggestions of values of these three parameters are given.

A Newly Developed Suppression Pool Model Based on the ISAA Code

The suppression pool is an important component in a boiling water reactor nuclear power plant. Under design-basis, loss-of-coolant accident conditions, pressure in the containment increases. Gas flows from the drywell to the wetwell after the normally submerged connecting vents between the drywell and the wetwell have been purged of water through a vent clearing process so that the suppression pool may prevent pressure in the containment from exceeding the designed pressure limit. To analyze such complicated thermal-hydraulic behavior of the suppression pool under a specific accident, an advanced reasonable model should be developed. The SPARC-90 model used in MELCOR calculates the mass and energy transfer between the bubbles of the injected gas and the suppression pool, which is affected by distance efficiency and subcooling efficiency. The dedicated vent flow model used in CONTAIN can well simulate the vent clearing time, that is, the time required for the liquid level on either side of the suppression pool to drop to the level at which the vent begins to clear. However, it is necessary and better to combine them into one integrated code. This paper presents a newly developed suppression pool model based on the self-developed severe accident analysis code Integrated Severe Accident Analysis (ISAA), which combines the advantages of the dedicated vent flow model and the SPARC-90 model to analyze the suppression pool’s thermal-hydraulic behavior. The simulation results of the developed suppression pool model shows reasonableness compared with the result in the CONTAIN 2.0 code manual. The good agreement between the simulation results and the analysis results from the COLUMBIA power station final safety analysis report demonstrates the rationality and effectiveness of the developed model, although future improvement is needed.

Improvement of integrative severe accident analysis code, SPECTRA

Improvement of integrative severe accident analysis code, SPECTRA

Development of the sodium fire module in integrated severe accident analysis code, SPECTRA

Development of the sodium fire module in integrated severe accident analysis code, SPECTRA

Numerical simulation of CS28-1 experiment by using CANDU severe accident analysis code, CAISER

Due to the demands for an accurate and realistic severe accident code in the PHWR safety analysis, a new severe accident code called CAISER is being developed at Korea Atomic Energy Research Institute (KAERI). The CAISER code simulates core degradation phenomena occurring in a calandria tank, and it consists of two main modules: A fuel rod degradation module and a fuel channel degradation module. The fuel rod degradation module simulates the severe accident phenomena in a fuel channel, which includes a core uncovery, fuel rods heatup, hydrogen generation due to steam-Zr oxidation, fuel rods slumping, fuel rods melting and relocation, and thermal interaction of relocated molten mass with a pressure tube or a calandria tube. The fuel channel degradation module simulates the overall severe accident phenomena in a calandria tank, including the sagging of a fuel channel, debris bed formation caused by a fuel channel failure, the molten pool formation, and the calandria tank failure. Each module is tracking mass and energy changes for the main components, which are fuel, cladding, pressure tube, calandria tube in a fuel rod degradation module, and debris bed, metallic and oxidic corium pools in a fuel channel degradation module. These two modules are closely interconnected to simulate phenomena both in a fuel channel and in a calandria tank at the same time. In this paper, the numerical modelling in the CAISER code is described, including the node system and the governing equations for each main component. For the purpose of CAISER code validation, numerical simulations were conducted for the CS28-1 experiment. The simulation results show that the temperature distributions calculated by CAISER are in line with the experimental data.

Development of a semi-empirical model-based COrium COolability analysis tool (COCOA) validated against a large scale corium experiment, FARO

As molten corium is released from the reactor pressure vessel (RPV) under severe accidents, the molten corium is going to experience various complicated phenomena. If there is no steam explosion, the corium is going to be ejected through the reactor vessel, fragmented, and cooled through the interaction between the corium and water existing in the cavity during its falling stage.Our research group has recently developed a COrium COolability Analysis tool named COCOA. COCOA is the transient calculation tool for the evaluation of the phenomena from the molten corium release to debris formation through the corium falling stage of FCI. COCOA was corporately developed to calculate the transient results of the pressure, the released energy, the average water temperature, fragmented particle size distribution, particulate debris fraction etc., during the corium falling stage of the FCI phase.COCOA is validated against FARO, which is a large-scale corium experiment. Comparing to COMETA, which is a 2D severe accident analysis code, COCOA showed better accuracies on the prediction of the pressure, the released energy, and the particulate debris fraction during the corium falling stage of the FCI phase. COCOA gives 16.9% and 26.5% of NRMSD for the pressure and the released energy, while COMETA does 42.2% and 36.0% of NRMSD. Also, COCOA and COMETA predict the particulate debris fraction with 27.22% and 39.25% of RMSE, respectively. Since COCOA is a semi-empirical model-based simple 1-D tool, only small computational effort is required for the calculation.

Three-dimensional numerical study on the effect of sidewall crust thermal resistance on transient MCCI by improved MPS method

Molten Corium and Concrete Interaction (MCCI) is the phenomenon that could occur in the severe accident of nuclear power plant (NPP) after the failure of reactor pressure vessel (RPV). There are considerable uncertainties regarding the MCCI initial conditions, including melt pouring, which is harder to be considered in many severe accident analysis codes. In this study, the improved MPS method with explicit pressure method was used to simulate the pouring stage of HECLA-3 transient MCCI experiment. The anisotropic ablation was obtained at a lower pouring rate, which was in good agreement with the experiment. It is found that the initial sidewall crust formation is the main reason for the ablation difference of MCCI caused by low melt pouring rate. And the mechanism of adiabatic effect of the crust layer on MCCI was revealed by the comparison of simulation results with three different crust heat transfer models or assumptions.

Chemical forms of uranium evaluated by thermodynamic calculation associated with distribution of core materials in the damaged reactor pressure vessel

ABSTRACT To suggest efficient process of the fuel debris treatment after the retrieval from the Fukushima Daiichi Nuclear Power Plant (1F), thorough investigation is indispensable on potential source of U in the fuel debris. The present study aims to estimate chemical forms of U in the in-vessel fuel debris, especially in the minor phases such as metallic phases, by performing the thermodynamic calculation under various conditions considering material relocation and changing environment during the accident progression in the 1F Unit 2. Input conditions for the thermodynamic calculation such as composition, temperature, and oxygen amount were set mainly based on the transient change in the core material distribution which had been calculated with severe accident analysis codes such as MAAP and MELCOR. The result showed that chemical form of U varied depending on the local amount of Fe and O. In regions of low steel content, the U-containing metallic phase was dominated by α-(Zr,U)(O), while regions of high steel content were dominated by Fe2(Zr,U) (Laves phase). A few percent of U transferred to the metallic phases were highly expected under reducing conditions. Therefore, those metallic phases should be one of the potential sources of U.

Development of a reduced order model for severe accident analysis codes by singular value decomposition aiming probabilistic safety margin analysis

ABSTRACT Severe accident codes (e.g. MAAP, RELAP, and MELCORE) model various physical phenomena during severe accidents. Many analyses using these codes for safety margin evaluation are impractical due to large computational costs. Surrogate models have an advantage of quickly reproducing multiple results with a low computational cost. In this study, we apply the singular value decomposition to the time-series results of a severe accident code to develop a reduced order modeling (ROM). Using the ROM, the probabilistic safety margin analysis for the station blackout with a total loss of feedwater capabilities at a boiling water reactor is carried out. The dominant parameters to the accident progression are assumed to be the down-time and the recovery-time of the reactor core isolation cooling system, and decay heat. To reduce the number of RELAP5/SCDAPSIM analyses while maintaining the prediction accuracy of ROM, we develop a data sampling method based on adaptive sampling, which selects the new sampling data based on the dissimilarity with the existing training data for ROM. Our ROM can rapidly reproduce the time-series results and can estimate core conditions. By reproducing multiple results by ROM, a time-dependent core damage probability distribution is calculated instead of the direct use of RELAP5/SCDAPSIM.

Analysis of loss of cooling and loss of coolant severe accident scenarios in VVER-1000/V446 spent fuel pool

The spent fuel storage system is designed to store and cool the spent fuel within several years with account for scheduled refueling and emergency unloading of the entire reactor core. The Fukushima nuclear disaster in 11 March 2011 showed that the safe storage of spent fuel assemblies in the spent fuel pool (SFP) is very vital since many radioactive fission products could be simply released into the environment if a severe accident occurred in this equipment. In this study, severe accident scenarios at the VVER-1000/V446 spent fuel pool are studied using the MELCOR 1.8.6 severe accident analysis code. Since the most likely events of the SFP are the loss of cooling and loss of coolant, scenarios leading to such accidents (i.e. six scenarios for different leakage rate and leak elevation) are modeled and investigated. The most important results of this study are the calculation of the occurrence time of the key events, the amount of generated hydrogen that has great importance in development of hydrogen mitigation strategies, and the amount of radioactive materials released into the containment for each scenario.

Performance analysis of automatic depressurization system in advanced PWR during a typical SBLOCA transient using MIDAC

The aim in the present work is to simulate accident scenarios of AP1000 during the small-break loss-of-coolant accident (SBLOCA) and investigate the performance and behavior of automatic depressurization system (ADS) during accidents by using MIDAC (The Module In-vessel Degradation severe accident Analysis Code). Four types of accidents with different hypothetical conditions were analyzed in this study. The impact on the thermal-hydraulic of the reactor coolant system (RCS), the passive core cooling system and core degradation was researched by comparing these types. The results show that the RCS depressurization becomes faster, the core makeup tanks (CMT) and accumulators (ACC) are activated earlier and the effect of gravity water injection is more obvious along with more ADS valves open. The open of the only ADS1-3 can't stop the core degradation on the basis of the first type of the accident. The open of ADS1-3 has a great impact on the injection time of ACC and CMT. The core can remain intact for a long time and the core degradation can be prevent by the open of ADS-4. The all results are significant and meaningful to understand the performance and behavior of the ADS during the typical SBLOCA.

Comparative analysis of simulations of LIVE-L10 and -L11 experiments using different lower head modules of AC2

Abstract The stabilization of molten core material in the lower head in case of a severe accident by external cooling of the reactor vessel is regarded as an effective severe accident management measure. In the experiments LIVE-L10 and -L11 the late phase melt pool behaviour of the corium is investigated under different cooling conditions – the former under sub-cooled convection, the latter under nucleate boiling conditions. In this work the experiments are calculated with the severe accident analysis code AC2 – ATHLET-CD 3.1A. Objective of the simulations conducted is the analysis and assessment of the code's capability to simulate the most relevant phenomena that occur during the tests. The simulations are performed with two different lower head modules implemented in ATHLET-CD, AIDA (Analysis of the Interaction between Core Debris and the reactor pressure vessel during severe Accidents) and LHEAD (extended Lower Head module). The simulation results, analysed in comparison with the experimental results, show the capability of both modules to reproduce the respective experiments.

Cesium chemisorbed species onto stainless steel surfaces: An atomistic scale study

Under the scope of Fukushima Daiichi Nuclear Power Station (1-F) severe accident (SA), Cs retention is of high interest, as it impacts Cs distribution, decommissioning and dismantling work of the reactor. To derive consistent and appropriate models for such process, accurate thermodynamic properties of Cs chemisorbed species are required by the SA analysis codes. In particular, for CsFeSiO4, a newly identified Cs chemisorbed species under conditions similar to 1-F SA, the thermodynamic data are unknown in literature. We propose in this work the obtention of the fundamental properties of this substance by theoretical approaches. The consistency and appropriateness of derived computational methodology have been investigated by calculating the thermodynamic properties of relatively known Cs–Si–O substances. It was found that our computational methodology provides excellent agreement with literature data lying between 1 and 4% for the formation energy, 1–5% for standard entropy and heat capacity. The thermodynamic properties of CsFeSiO4 as function of temperature have been estimated for the first time using harmonic and quasi-harmonic approximations, values being consistent with both methodologies.

Sensitivity and uncertainty analysis of the vessel lower head failure mode and melt release conditions in Nordic BWR using MELCOR code

Nordic boiling water reactors (BWR) employ ex-vessel debris coolability as a severe accident management (SAM) strategy. Melt release mode into a deep water pool located in the lower drywell is a major source of uncertainty for success of such strategy. The main goal of this work is to identify the major contributors to the uncertainty in prediction of vessel failure mode, timing and melt release conditions in Nordic BWR using MELCOR severe accident analysis code. There is a forest of control rod guide tubes (CRGTs) and instrumentation guide tubes (IGTs) in the lower head of a BWR. Failure of the penetrations is considered in this work along with the gross creep rupture of the vessel lower head. Modelling of penetrations failure in MELCOR code is based on parametric models, allowing a user to select different assumptions, such as temperature threshold for penetrations failure, modelling of penetrations assembly and respective failure modes, mode of solid and liquid debris ejection from the vessel. In this work we perform sensitivity analysis to the MELCOR modelling options and sensitivity parameters in unmitigated station blackout scenario (SBO). Results of analysis suggest that (i) vessel breach due to penetration failure is predicted to occur before creep-rupture failure of the vessel lower head, (ii) the mode of debris ejection from the vessel has the dominant effect on the likelihood of creep-rupture failure of the vessel lower head and debris ejection rate from the vessel.

Confirmation of severe accident code modeling in light of the findings at Fukushima Daiichi NPPs

On March 11th 2011 the Great East Japan Earthquake and Tsunami hit the Fukushima Daiichi Nuclear Power Station. The Fukushima Daiichi Units 1 to 3 lost all DC and AC power supplies, which set in motion a chain of events that led to releases of radioactivity to the environment. Since then, TEPCO has made many efforts to investigate the accident progressions and the status of the reactors and containment vessels, mainly by using severe accident analysis code, such as MAAP, MELCOR and SAMPSON. However, there still exist several tens of unsolved issues to be investigated for the fully understanding of the accident. In recent years, massive amounts of information of the accident plants were obtained through the investigation activities by using robots or other devices. The investigation results suggested us the possibilities that some phenomena identifications established before the accident were incorrect. And also, there are many difficulties to reproduce the accident progressions of the Fukushima Daiichi unit 1 to 3, due to lack of knowledge real BWR plant accident phenomena. In this paper, we will introduce the special characteristics, which made difficult to reproduce the accident. And we will discuss how to use the investigation results obtained in Fukushima Daiichi NPP for the SA code validation by comparing with investigation results and calculation results.

An experimental investigation of influencing chemical factors onCs-chemisorption behavior onto stainless steel

ABSTRACT A more accurate cesium hydroxide (CsOH)-chemisorption model is required to improve the estimation of the Cs distribution in Fukushima Daiichi Nuclear Power Station by severe-accident (SA) analysis codes. The current CsOH-chemisorption model incorporated in SA analysis codes was developed with insufficient information of chemical factors such as H2/H2O ratio, concentration of chemically affecting elements, heating time, and surface condition of the stainless steel (SS). Therefore, we have conducted experimental tests for CsOH-chemisorption onto SS type-304 (SS304) in order to examine the effect and dependence of such chemical factors on the CsOH-chemisorption behavior in detail. It was found that the first-order surface-reaction rate constant was influenced by not only temperature, as already known, but also H2/H2O ratio, CsOH concentration in the gas phase, and silicon content in SS304. Such chemical factors should be considered for the construction of the improved CsOH-chemisorption model. Another important finding is that the chemisorption behavior at lower temperatures, around 873 K, could differ from those above 1073 K. Namely, Cs–Fe–O compounds would form as the main chemisorbed Cs compounds at 873 K while Cs–Si–Fe–O compounds at more than 1073 K.

Numerical simulation of zircaloy-water reaction based on the moving particle semi-implicit method and combined analysis with the MIDAC code for the nuclear-reactor core melting process

During the severe accident in PWR (Pressurized Water Reactor), local phenomena (e.g., eutectic reaction, candling effect, zircaloy-water reaction) occurring in the core have a significant impact on the core melting process. However, the traditional severe accident analysis codes cannot simulate the local phenomena accurately. To solve this problem, the Computational Fluid Dynamics (CFD) method is considered. By coupling CFD method with traditional severe accident codes, the accident sequence in the core melting process can be simulated accurately. In this study, considering the complexity of physical and chemical phenomena involved in the core melting process, the zircaloy-water reaction is selected for analysis. The zircaloy-water reaction kinetics models in the Module In-vessel Degradation Analysis Code (MIDAC) are summarized. The correlations for zircaloy temperature below 1800K are compared and the results show that the Cathcart-Pawel correlation induces a considerable overestimation of the reaction rates above 1460 K. Then based on the Moving Particle Semi-implicit (MPS) method, a numerical analysis model is developed to simulate zircaloy-water reaction. A zircaloy-water reaction kinetics correlation with higher accuracy is obtained by simulations and added to MIDAC. The TMI-2 accident is selected to evaluate the newly developed correlation by comparing with other correlations in MIDAC. The results indicate that different correlations can greatly influence the structure of corium pool in the core and the melt behaviors during reflooding and long-term cooling stage. However, it can hardly influence the relocation of debris bed to the lower plenum and behaviors of debris bed in the lower plenum.