Severe Accident Analysis Code Research Articles

Parametric study of molten core concrete interaction and assessment of corium coolability in the light of MCCI experiments

Molten Corium Concrete Interaction (MCCI) is an important phenomenon observed in the late phase of reactor severe accidents. A realistic analysis of MCCI can save cost and efforts to mitigate severe accidents. The heat transfer mechanisms in MCCI, such as melt eruption and water ingression, are the primary focus of current research owing to complications in their quantification and replication in severe accident analysis codes. This study investigated MCCI using top flooding strategy with MELCOR 2.2 code for two concrete types, and three experiments that were conducted under (OECD/MCCI-1 program). The results demonstrated that the ablation process terminated in the case of Limestone Common Sand (LCS) concrete. In the simulation of siliceous (SIL) concrete, the maximum and average melt front locations confirmed the asymmetrical ablation progression. The LCS concrete showed a smaller axial ablation front in this study; however, the comparison with the concrete ablation volume (0.21 m3) showed an exact match, and the difference between the experimental and predicted ablated mass of concrete was only 0.61%. The quantitative comparison between the prediction of the present study, CORQUENCH code, and experimental results is provided. The parametric values of MCCI phenomenon are also predicated and model improvement for MELCOR 2.2 code are discussed as well.

An analytical model to decompose mass transfer and chemical process contributions to molecular iodine release from aqueous phase under severe accident conditions

Radioactive iodine is a representative fission product to be quantified for the safety assessment of nuclear facilities. In integral severe accident analysis codes, the iodine behavior is usually described by a multi-physical model of iodine chemistry in aqueous phase under radiation field and mass transfer through gas-liquid interface. The focus of studies on iodine source term evaluations using the combination approach is usually put on the chemical aspect, but each contribution to the iodine amount released to the environment has not been decomposed so far. In this study, we attempted the decomposition by revising the two-film theory of molecular-iodine mass transfer. The model involves an effective overall mass transfer coefficient to consider the iodine chemistry. The decomposition was performed by regarding the coefficient as a product of two functions of pH and the overall mass transfer coefficient for molecular iodine. The procedure was applied to the EPICUR experiment and suppression chamber in BWR.

Analysis of ex-vessel debris bed formation of multi-material and multiphase composition based on coupled system of MELCOR2 and THERMOS–JBREAK/MSPREAD

THERMOS is being developed to analyze the multidimensional process by which molten debris released from the lower head of the reactor pressure vessel forms a non-uniform debris bed. THERMOS consists of four modules, JBREAK, DPCOOL, MSPREAD and REMELT. To evaluate a wide range of ex-vessel debris bed formation, the authors have developed the coupled system composed of the integrated severe accident (SA) analysis code MELCOR2, THERMOS-JBREAK and THERMOS-MSPREAD. JBREAK simulates the molten jet release and impingement on the cavity floor using the separated flow model consisting of the molten jet and particle debris. MSPREAD simulates the melt spread as the multi-layered structure composed of the partially solidified molten debris, the upper and lower crusts growing on both sides using the shallow water equation.The conversion algorithm of the multi-material and multiphase composition of the released molten jet calculated by MELCOR2, which is based on lumped parameters, to the composition suitable for THERMOS, which is based on distributed parameters, is recognized to be critical to the success of this system and has been implemented as the Python script MELTHER. This conversion also includes the contribution of suspended particles embedded in the molten jet to the viscosity of partially solidified molten debris. As the JBREAK–MSPREAD interface, the method of overlapping the mesh systems of the two codes was implemented to solve the multidimensional flow equations in the near impinging region. Filtration of the molten debris inside the particle layer at the jet-impinging point was modeled using the particle debris deposition and molten debris filtration models.This coupled system was applied to a station blackout scenario in a typical BWR3 Mark-I plant, where approximately 14 tons of mixed molten and particle debris were released from the lower head and impinged on the pedestal floor. The calculation space was set to handle molten jet falling, impingement on the floor, and spreading from the pedestal to the outside dry well region. Complex anisotropic melt spreading was analyzed on the dry pedestal floor with two sumps covered with thin lids and a low weir at the bottom of the slit made in the cylindrical wall separating the pedestal floor and outer dry well regions.A series of sensitivity analyses were performed varying a combination of the initial suspended particle fraction and melt filtration inside the particle layer. Deceleration of the molten debris flow is significant due to the melt-particle bed frictional force and therefore the timing of molten debris running over the sump lids is faster when there is no particle layer. In this case, a large circulating flow is generated on the pedestal floor, and this flow promotes cooling and increases the viscosity. Influences of the initial suspended particle fraction on the melt viscosity become more pronounced in the peripheral region, and spreading behaviors are more susceptible to protrusions on the pedestal floor.

Preliminary analysis focusing on in-vessel thermal hydraulics in loss-of-heat removal systems in a sodium-cooled fast reactor

Accident scenarios for the safety evaluation of sodium-cooled fast reactors (SFRs) are categorized into two groups: Anticipated Transient Without Scram (ATWS) and Loss-of-Heat Removal System (LOHRS). The latter is subdivided into two scenarios—Loss-of-Heat Sink (LOHS) and Loss-of-Reactor Level (LORL)—depending on the initiating event. With the development of SFRs, research on ATWS has been paid much attention to due to its rapid accident progression, which could lead to significant radioactive release. Conversely, the significance of LOHRS has increased in recent decades due to lessons learned from accidents at the Fukushima Daiichi Nuclear Power Plant of Tokyo Electric Power Company. However, compared with ATWS, finding and evaluation tools for LOHRS have been limited. Hence, code development is initiated to enhance the severe accident analysis code AZORES, which the Nuclear Regulation Authority developed for the safety evaluation of SFR accidents. This enhancement is achieved by coupling AZORES with plant dynamics code ADYTUM in order to deal with more detailed thermal hydraulics behavior in reactor cooling system. The paper describes the coupling methodology, focusing on the thermal–hydraulic model as the first step of code development and its application. The coupling has been confirmed to be well achieved through a preliminary analysis of LOHS, comparing the results of coupled and single AZORES. The comparison of the analysis results between LOHS and LORL by the AZORES coupled with ADYTUM has been conducted to identify the key phenomena in accident progression. The comparison shows that the coolant leak area and the pressure drop at the piping of the reactor cooling system affect accident progression in LOHS. Consequently, the coupled AZORES is capable of predicting in-vessel thermal–hydraulic behavior in LOHRS successfully, reflecting accident conditions and plant design appropriately.

Empirical modeling of ex-vessel debris bed formation in a pre-flooded reactor cavity

To secure the containment integrity of a nuclear power plant (NPP) in severe accident situation of ex-vessel corium release into the reactor cavity, a pre-flooding strategy of a reactor cavity has been adopted in various NPPs, including Korean PWRs. Molten corium released into the pre-flooded water pool is fragmentized and forms a particulate debris bed at the bottom of the cavity. To prevent the containment failure, it is important to properly remove the heat from the debris bed, which can be accomplished by water ingression into the bed. The water ingression characteristics are highly dependent on the debris bed shape. In this study, several main parameters affecting the debris bed shape were determined with respect to phenomena of free-fall & sedimentation, particle dispersion by steam spike and convection flow, and self-leveling. Then empirical model was developed and validated using the results of DEFCON (Debris bed Formation and COolability experimeNt) experiments. The model’s error was within ±10% when only free-fall & sedimentation were present. When steam-spike or convection was added to free-fall & sedimentation condition, the error increased to within ±20%. When all three were simultaneously present (free-fall & sedimentation, steam-spike, and convection), the model predicted most of the experimental results within ±20% error, except for several cases where convection had a significant impact. The effect of self-leveling itself was within ±10%. Considering all phenomena related to particulate debris bed formation, most cases had an error of ±20%. In addition, the debris bed shape formed by poly-disperse particles was predicted appropriately by the mode using the mass mean diameter of the particle size distribution, with an error range of ±20%. Consequently, it is expected that the proposed model of the debris bed shape can be utilized in an integrated severe accident analysis code for the evaluation on the ex-vessel debris bed coolability in the pre-flooded reactor cavity.

The APR1400 SOARCA study: Insights into the evolution of severe accidents and fission product source term analysis results

The Korean utility, KHNP (Korea Hydro and Nuclear Power Co.), carried out a three-year research study in July 2016 (named the Korean SOARCA: K-SOARCA), which is similar with the United State (US) SOARCA (State-of-the-Art Reactor Consequence Analysis) study. One of the two reference reactors designated for the study was Advanced Power Reactor of 1400 MWe (APR1400). To consider the integral response of the reactor, steam generators, and containment to severe accidents initiated by the representative accident sequences selected from the plant-specific Probabilistic Safety Assessment (PSA) results, dedicated modeling was performed for a broad spectrum of plant safety/mitigation systems and relevant phenomena influencing the evolution of the accident. The influence of dedicated accident mitigation strategies on the evolution of accident progression and fission product source terms was also investigated to assess their effectiveness in coping with relevant severe accident sequences. Especially, relating to the APR1400 SOARCA, the latest versions of two severe accident analysis codes, MELCOR 2.2 and MAAP5.04, were used to compare the analysis results of interest. This paper presents best-practice analysis results for the evolution of key accidents and plant response including the environmental release of radiological fission product Cs (cesium), as implemented through the APR1400 SOARCA study, and relevant insights.

Research on the performance enhancement of accident tolerant fuel for eliminating core melting based on ISAA

Based on the Integrated Severe Accident Analysis (ISAA) code, this paper studies the influence of thermo-physical properties of Accident Tolerant Fuel (ATF) materials on the plant responses to severe accidents, and further proposes the requirements of eliminating core melting on ATF performance and reactor components. The equation of fuel temperature rise is derived from the steady-state heat conduction equation of fuel rods, and three fuel properties, including thermal conductivity, density and specific heat capacity, which affect core melting, are obtained. By selecting ATF with IMDP(Inert Matrix Dispersion Pellet) as the core material and SiC as the cladding material and changing different physical parameters, it is found that fuel density and cladding density have the greatest influence on the fuel temperature rise, and appropriately increased fuel density can extend the core melting time. In addition, to eliminate core melting, the ATF parameters in ISAA are modified step by step to improve the thermodynamic properties of the fuel, and the core is finally made to reach thermal equilibrium and to realize the safety of nuclear power plants without human intervention. According to the calculation results, the supporting capacity of the cladding, the supporting capacity of the core supporting plate, and the melting point of the fuel are the main factors limiting the core melting, which can be eliminated by appropriately improving these properties.

Analysis of loss of cooling accident in VVER-1000/V446 spent fuel pool using RELAP5 and MELCOR codes

Following the Fukushima nuclear disaster, the simulation of accidents in the spent fuel pool has become more noticeable. Despite the low amount of decay heat power, the consequences of the accidents in a spent fuel pool (SFP) can be severe due to the high content of long-lived radionuclides and lack of protection by the pressure vessel.In this study, the loss-of-cooling accident (LOFA) for the VVER-1000/V446 spent fuel pool is simulated by employing RELAP5 and MELCOR 1.8.6 as the best estimate and severe accident analysis codes, respectively. For two cases with different total power levels, decay heat of spent fuels is calculated by ORIGEN-II code. For modeling SFP of a VVER-1000, a qualified nodalizations are considered in both codes. During LOFA in SFP, the key sequences such as heating up of the pool water, boiling and reducing the water level, uncovering the spent fuels, increasing the temperature of the spent fuels, starting oxidation process (generating Hydrogen and extra power), the onset of fuel melting, and finally releasing radionuclides are studied for both cases. The obtained results show a reasonable consistency between the RELAP5 and MELCOR codes, especially before starting the oxidation process.

Influence of active and passive equipment for advanced pressurized water reactor on thermal hydraulic and source term behavior in severe accidents

Extensive studies have been carried out on the behavior of core degradation and fission products of common pressurized water reactors (PWRs). However, few of them have investigated the relationship between thermal hydraulic and fission product behavior in advanced passive PWRs. Due to the impact of thermal hydraulic behaviors in different accident sequences on the release and transportation of fission products, an integrated severe accident analysis (ISAA) code with highly coupled thermal hydraulic and source term calculations is required to simultaneously analyze thermal hydraulic and source term behavior. For advanced passive PWRs, important safety systems that may affect the behavior of the core and fission products should be considered. It is therefore necessary to simulate the thermal hydraulic and fission product behavior of advanced passive PWRs. In this study, the ISAA code is adopted to simulate the occurrence of a hypothetical double ended cold leg LBLOCA of HPR1000 in three scenarios of equipment failure. The results show that the high-temperature fuel rods and cladding materials exhibit delayed failure at the lower position of the active core, whereas earlier failure at higher position during the reflooding. Active and passive equipment affects fuel temperature, the oxidation conditions of the fuel, the interaction of fission products and structural materials, and the state of the fuel, thereby affecting the release of fission products in the fuel. HPR1000 only relies on passive equipment to relieve the core degradation in severe accidents, realize the in-vessel retention of melt, and eliminate the ex-vessel release possibility of fission product. It is hoped that the results can provide references for HPR1000 to formulate the severe accident management guidelines (SAMG).

Uncertainty Quantification in Support of Severe Accident Analysis Code User Confidence Using MELCOR-DAKOTA

Abstract Uncertainty and Sensitivity analysis methods are often used in severe accident analysis for validating the complex physical models employed in the system codes that simulate such scenarios. This is necessitated by the large uncertainties associated with the physical models and boundary conditions employed to simulate severe accident scenarios. Uncertainty analysis involves primarily the estimation of the distribution of a key figure-of-merit (FOM) following the propagation of uncertainty in selected input parameters through a system code. The input parameters are sampled within defined ranges based on assigned probability distribution functions (PDFs) for the required number of code runs/realizations using stochastic sampling techniques. Input parameter selection is based on their importance to the key FOM, which is determined by the parameter identification and ranking table (PIRT). Sensitivity analysis investigates the contribution of each uncertain input parameter to the uncertainty of the selected FOM. In this study, the integrated severe accident analysis code MELCOR was coupled with a design analysis kit for optimization and terascale applications (DAKOTA), an optimization and uncertainty quantification tool in order to investigate the effect of input parameter uncertainty on hydrogen generation. The methodology developed was applied to the Fukushima Daiichi unit 1 nuclear power plant (NPP) accident scenario, which was modeled in another study. The results show that there is approximately 22.46% uncertainty in the amount of hydrogen generated as estimated by a single MELCOR run given the uncertainty in selected input parameters. The sensitivity analysis results also reveal that MELCOR input parameters: COR_SC 1141(melt flow rate per unit width at breakthrough candling), COR_ZP (porosity of fuel debris beds), and COR_EDR (characteristic debris size in the core region) contributed most significantly to the uncertainty in hydrogen generation.

Study on the effect of flow blockage due to rod deformation in QUENCH experiment

During a loss-of-coolant accident (LOCA) in the pressurized water reactor (PWR), there is a possibility that high temperature and internal pressure of the fuel rods lead to ballooning of the cladding, which causes a partial blockage of flow area in a subchannel. Such flow blockage would influence the core coolant flow, thus affecting the core heat transfer during a reflooding phase and subsequent severe accident. However, most of the system analysis codes simulate the accident process based on the assumed channel blockage ratio, resulting in the fact that the simulation results are not consistent with the actual situation. This paper integrates the developed core Fuel Rod Thermal-Mechanical Behavior analysis (FRTMB) module into the self-developed severe accident analysis code ISAA. At the same time, the existing flow blockage model is improved to make it possible to simulate the change of flow distribution due to fuel rod deformation. Finally, the ISAA-FRTMB is used to simulate the QUENCH-LOCA-0 experiment to verify the correctness and effectiveness of the improved flow blockage model, and then the effect of clad ballooning on core heat transfer and subsequent parts of core degradation is analyzed.

A study of RCS depressurization strategy of CPR1000 SAMG

Abstract CPR1000 severe accident management guideline (SAMG) is a product of the Westinghouse Owners Group (WOG) SAMG and the French SAMG (called GIAG) combination. It adopts the structure of WOG SAMG. However, it follows the approach of GIAG performing reactor coolant system (RCS) depressurization immediately after entering SAMG. This is a special feature of CPR1000 SAMG. The purpose of this paper is to evaluate this RCS depressurization strategy. The simulations of a station blackout (SBO) accident are performed for CPR1000 with RCS depressurization strategy adopted at both core exit temperature (CET) of 650 and 1100 °C together with opening different number of safety relief valves (SRV) using an integral severe accident analysis code. The optimized RCS depressurization strategy is provided to delay the reactor pressure vessel (RPV) failure time. And some comments are provided for future improvement.

Performance evaluation of Accident Tolerant Fuel under station blackout accident in PWR nuclear power plant by improved ISAA code

The Accident Tolerant Fuel (ATF) is a new concept of fuel, which can not only withstand the consequences of the accident for a longer time, but also maintain or improve the performance under operating conditions. ISAA is a self-developed severe accident analysis code, which uses modular structures to simulate the development processes of severe accidents in nuclear plants. The basic version of ISAA is developed based on UO2–Zr fuel. To study the potential safety gain of ATF cladding, an improved version of ISAA, referred to as ISAA-ATF, is introduced to analyze the station blackout accident of PWR using ATF cladding. The results show that ATF cladding enable the core to maintain a longer time compared to zirconium alloy cladding, thereby enhancing the accident mitigation capability. Meanwhile, the generation of hydrogen is significantly reduced and delayed, which proves that ATF can improve the safety characteristics of the nuclear reactor.

Preliminary Analysis of Core Disruptive Accident in Sodium-Cooled Fast Reactor Using Eutectic Reaction Model of Boron-Carbide Control-Rod Material

It is necessary to simulate a eutectic melting reaction and relocation behavior of boron carbide (B4C) as a control rod material and stainless steel (SS) during a core disruptive accident (CDA) in an advanced large-scale sodium-cooled fast reactor (SFR) designed in Japan. A physical model simulating the eutectic reaction and relocation of the eutectic melt was developed to incorporate into the fast reactor severe accident analysis code SIMMER-IV for the CDA numerical analysis of SFRs. This study applied the SIMMER-IV code with the newly developed model to the CDA analysis of the SFR. This analysis indicated that the SIMMER-IV code using the eutectic reaction model has successfully simulated the eutectic reaction and the upward motion of the eutectic melt in the molten core pool as well as the reactivity transient behavior caused by the molten core material relocation.

Interfacing between MAAP and MACCS to perform radiological consequence analysis

Interfacing the output of severe accident analysis with the input of radiological consequence analysis is an important and mandatory procedure at the beginning of Level 3 PSA. Such interfacing between the severe accident analysis code MELCOR and MACCS, one of the most commonly used consequence analysis codes, is relatively tractable since they share the same chemical groups, and the related interfacing software, MelMACCS, has already been developed. However, the linking between MAAP, another frequently used code for severe accident analyses, and MACCS has difficulties because MAAP employs a different chemical grouping method than MACCS historically did. More specifically, MAAP groups by chemical compound, while MACCS groups by chemical element. An appropriate interfacing method between MAAP and MACCS has therefore long been requested by users. This study suggests a way of extracting relevant information from MAAP results and providing proper source term information to MACCS by an appropriate treatment. Various parameters are covered in terms of magnitude and manner of release in this study, and special treatment is made for a bypass scenario. It is expected that the suggested approach will provide an important contribution as a guide to interface MAAP and MACCS when performing radiological consequence analyses.

Development of mechanistic cladding rupture model for integrated severe accident code ISAA. Part II: DVI line small-break LOCA in CAP1400

Cladding ballooning and rupture are the important phenomena at the early stage of a severe accident (SA). For example, the SA caused by the loss of coolant accident (LOCA), the decrease in primary loop pressure and the increase in core temperature will lead to the cladding ballooning and rupture. The cladding deformation affects the flow distribution and causes local flow blockage. Meanwhile, water vapor will enter the fuel gap from the rupture, thereby increasing the surface area of cladding that is oxidized by the steam. At present, the widely used integrated severe accident analysis codes cannot analyze fuel rods thermal-mechanical behavior at the early SA stage, and the judgment of cladding rupture is only based on simple parameter model. This paper integrates the developed FRTMB (core Fuel Rod Thermal-Mechanical Behavior analysis module) module into the integrated severe accident analysis code ISAA, so that the coupled system ISAA-FRTMB can analyze fuel rods thermal–mechanical behavior and judge cladding rupture. Part I introduces the need of developing the FRTMB module, verification of the module, and how the module interactively works in the severe accident analysis code. And evaluates the steady-state fuel rods thermal–mechanical behavior at different enrichment areas of the CAP1400 reactor. Part II focuses on analyzing fuel rods behavior during a hypothetical DVI Line (Direct Vessel Injection line) small break-out accident (break diameter d = 4 in.) of the CAP1400, and predicts the cladding rupture time and the corresponding failure temperature. Due to failure of components of the emergency core cooling system and assumed unavailability of several preventive and mitigative accident management measures (AMM), the accident develops into a severe accident scenario with core melt and reactor pressure vessel failure. The mechanical analysis results show the top of the fuel rod is the first to rupture, rather than the peak node. The primary factor affecting the fuel pellet strain is the deformation caused by thermal expansion, while the densification and swelling change little. These results demonstrate the applicability and reliability of ISAA-FRTMB in analyzing fuel rods thermal–mechanical behavior and judging cladding rupture during transient accidents.

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.