Severe Accident Sequences Research Articles

Coupled AC2-CFD simulations for a high-pressure core melt accident scenario

Abstract Even though very unlikely to occur, severe accident scenarios in nuclear power plants have to be analyzed. During high-pressure core meltdown scenarios in a pressurized water reactor the primary circuit should fail first. Previous analyses found that a free convection flow within the vertical steam generator (SG) tubes with a simultaneous stratified gas counterflow in the hot legs could arise. This phenomenon leads to higher thermal loads on individual SG tubes which might then fail leading to a containment bypass and the release of radioactive material into the environment. Lumped parameter system codes used for safety analyses do not provide the models necessary to simulate phenomena like mixing in three-dimensional flows and could not consider local turbulence effects. Computational fluid dynamic (CFD) codes provide such capabilities but are much more computationally expensive. Coupling of a system code with a CFD code can therefore be used to simulate such phenomena. The advantages of both approaches can be maximized by splitting up the simulation domain between the codes, depending on the expected flow conditions. The system code AC2 coupled with the CFD code OpenFOAM was used to simulate part of the severe accident transient. Free convection in the hot leg and the U-tubes of the vertical SG was observed in case of high-pressure severe accident sequences. The thermal load of individual SG tubes has been estimated from the results. These loads can be used as inputs for structural-mechanical analyses to estimate which part of the primary circuit would fail first.

Dynamic event tree analysis of a severe accident sequence in a boiling water reactor experiencing a cyberattack scenario

Over the last decade, probabilistic risk assessment (PRA) for nuclear power plants shifted its focus toward simulation-based methods of evaluating safety and risk aspects. Classical reliability modeling based on static Boolean structures does not capture the dynamic nature of plants experiencing accident sequences. Dynamic PRA extends conventional reliability modeling, adding a higher degree of variability to events occuring during accident sequences. This paper demonstrates the capability of a new dynamic event tree (DET) analysis tool applicable to severe accident sequences. A cyberattack was assumed to penetrate the instrumentation system of a boiling water reactor during its hot shutdown phase, altering the thermal–hydraulic variables being processed. The analysis focused on scenarios that lead to the core being uncovered, in which the peak cladding temperature is reached due to the disabling of the cooling systems. A limiting surface is defined, associating the cyberattack timing to the recovery time limit that would compromise the core integrity.

Uncertainty quantification for a severe accident sequence in a SFP in the frame of the H-2020 project MUSA: First outcomes

The Management and Uncertainties of Severe Accidents (MUSA) project, funded in HORIZON 2020 and coordinated by CIEMAT (Spain), aims at consolidating a harmonized approach for the analysis of uncertainties and sensitivities associated with Severe Accidents (SAs) focusing on Source Term (ST). In this framework, the objectives of the Innovative Management of Spent Fuel Pool Accidents (IMSFP – WP6), led by IRSN (France), are to quantify and rank the uncertainties affecting accident analyses in a Spent Fuel Pool (SFP), to review existing and contemplated SA management measures and systems and to assess their possible benefits in terms of reduction of radiological consequences.To quantify the propagation of the uncertainties of the input parameters to the output uncertainties of severe accident codes (ASTEC, MELCOR, RELAP/SCDAP), a diverse set of uncertainty quantification (UQ) tools (DAKOTA, RAVEN, SUNSET, SUSA) are used. The statistical framework used by the different UQ-tools is similar e.g. pure random (Monte Carlo) and Latin hypercube sampling (LHS).Fourteen partners from three different world regions are involved in the WP6 activities. The target of this paper is to describe the achievements during the first three years of the project. In a first part, a description is given of the SFP accidental scenario, of the key target variables and radionuclides chosen as ST Figures of Merit (FoM) and of the identified uncertainty sources in models and input parameters. A key element when defining the SFP scenario has been the consideration (or not) of the reactor building, as it is expected to significantly affect analyses. In a second part, the first insights coming out from the calculation phase of the project are presented. The review of existing SA management measures is also exposed, as well as systems whose benefits will be assessed in the second phase of the project. Finally, challenges that arise from such an exercise are discussed, as well as major difficulties found when applying UQ methodologies to SFP scenarios and solutions adopted.

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.

Simulations of Core Damage Progression for TMI-2 Severe Accident Using CINEMA Computer Code

As an integrated computer code development for severe accident sequence analysis in Korea, CINEMA has been developing from an initiation event to a containment failure. The CINEMA computer code is composed of CSPACE, SACAP, and SIRIUS, which are capable of simulating core melt progression with thermal hydraulic analysis of the RCS (reactor coolant system), severe accident analysis of the containment, and fission product analysis in the vessel and the containment, respectively. The severe accident progression in TMI unit 2 has been analyzed as a part of a validation of the CINEMA computer code. This analysis has been performed to validate CINEMA models on the core melt progression, in particular, RCS thermal hydraulic behavior during core melt progression, fuel cladding oxidation with hydrogen generation, and fuel melting with relocation to the lower part of the core. The CINEMA results on main parameters, such as RCS pressure and an integrated hydrogen generation mass are compared with the TMI-2 data. The CINEMA results have shown that the RCS pressure is very similar to the TMI-2 data. The CINEMA results and measured total hydrogen production are very similar, which were approximately 465 kg and 460 kg, respectively.

CINEMA-SMART development and its application to SBLOCA analysis for Korean small integral reactor

The Korean small integral reactor of SMART (System-integrated Modular Advanced ReacTor) was developed, which contained core, reactor coolant pumps, steam generators and pressurizer within a single reactor pressure vessel. In the code development of severe accident sequence analysis for the SMART, a CINEMA (Code for INtegrated severe accidEnt Management Analysis)-SMART has been developing from an initiation event to a containment failure including fission product behavior in the vessel and the containment. The CINEMA-SMART is composed of CSPACE, SACAP (Severe Accident Containment Analysis Package), and SIRIUS (SImulation of Radioactive nuclide Interaction Under Severe accident), which are capable of core melt progression with thermal hydraulic analysis of the RCS (Reactor Coolant System), severe accident analysis of the containment, and fission product analysis, respectively. The main sequence of a SBLOCA (Small Break Loss Of Coolant Accident) for the SMART has been analyzed using CINEMA-SMART computer code. This analysis is focused on an evaluation of core melt progression and containment thermal hydraulic behavior during a severe accident. The CINEMA results have shown that the integrity of the containment can be maintained during the SBLOCA accident in the SMART.

Experimental investigation on the impact of cable fire products from flame-retardant cables on catalysts used in passive auto-catalytic recombiners

Cable fires in nuclear power plants have a significant probability to occur at any time in the course of a severe accident or may as well be the initiating event of a severe accident sequence. During the combustion process, enormous amounts of aerosols alongside specific combustion gases (e.g. CO, CO2) can be released. Both nature and amount of the cable fire products depend on the combustion conditions. The effect of cable fire products distributing inside the containment and getting in contact with the catalyst surfaces of passive auto-catalytic recombiners (PARs) is of vital interest for safety analyses, in order to assess the hydrogen mitigation efficiency under these conditions.The newly built REKO-Fire facility at Forschungszentrum Jülich combines a flow tube reactor for catalyst investigation with a steady-state tube furnace for the constant generation of cable fire products at varying combustion conditions. That way, simultaneous exposure of 5 × 5 cm2 catalyst samples to cable fire products and hydrogen/air mixtures is possible, enabling to quantify the influence of gaseous and particulate components on the catalysts’ start-up behavior. The installation has been used in the present study to investigate the effect of cable fire products obtained from flame-retardant power cables under three different fire conditions on the start-up of two types of catalysts for hydrogen recombination.For well-ventilated cable fire, neither gaseous nor particulate (mainly soot) cable fire products seem to affect the onset of the catalytic H2 conversion for both Pt and Pd-based catalysts. In under-ventilated fire conditions, the Pt-based catalyst is significantly deactivated, while the only impairment for the Pd-based catalyst is observed at very low hydrogen concentrations. For cable fire products generated from oxidative pyrolysis, the overall picture is ambiguous. On the one side it is obvious that deactivation and start-up delay occur for both catalyst types. However, no clear conclusion can be taken from the experimental data concerning the effect of exposure time. The presence of carbon monoxide in the atmosphere as well as particulate depositions from cable pyrolysis seem to be the most relevant mechanisms for catalyst deactivation and deserve further investigation.

Analysis on the discharge characteristics and spreading behavior of an ex-vessel core melt in the SMART

The aim of this research is to analyze the characteristics of a core melt discharged from the reactor vessel and the spreading behavior the core melt in the reactor cavity of the SMART. First, a severe accident sequence under conservative conditions is simulated by the MELCOR code to obtain the conditions for an analysis of the spreading behavior and coolability of the ex-vessel melt. Second, the spreading behavior and coolability of the ex-vessel melt are analyzed by the MELTSPREAD code. The level, temperature, and pressure of the water in the cavity as well as the temperature, mass, composition, and discharge velocity of the melt were utilized to construct the ex-vessel analysis. The melt spread only to part of the cavity, and that the height of the corium in a static state was less than 25 cm. The characteristics of a small modular reactor on the spreading behavior and coolability of melt were analyzed. In the SMART, the amount of melt discharged into the cavity is relatively small and the area of the cavity is sufficiently large when compared to a high-power pressurized water reactor. It was found that the coolability of an ex-vessel core melt can be sufficiently secured.

Dose Calculation for Emergency Control Room HVAC Filter

NPP Krsko is introducing Emergency Control Room (ECR) as part of safety upgrades. According to 10CFR50 Appendix A, GDC 19, both main control room and emergency control room should have adequate radiation protection to permit operators to shutdown the plant and keep it in safe shutdown conditions without receiving more than 50 mSv effective whole body dose, within 30 days from accident initiation. One of the important prerequisites to achieve that is proper operation of control room HVAC. In this work we are focused to calculation of gamma doses from radioactive materials accumulated in HEPA and charcoal filters during 30 days of HVAC operation. The dose at selected points around the filter was calculated using Microshield 10.0 point kernel code. The radioactive gamma source is calculated using RADTRAD 3.03 for plant's severe accident SGTR sequence calculated with MAAP 4.0.7 code. Calculated dose rates at peak filter activity are compared against results obtained with SCALE 6.2 MAVRIC shielding sequence (Monaco Monte Carlo functional module and CADIS methodology). The reasonable agreement between point kernel and hybrid Monte Carlo results was obtained.

Validation of ASTEC2.1 using QUENCH-12 for VVER-Reactors

Russian VVER-type reactors are one of the most common used commercial reactors in the world. The validation of severe accident codes using experimental data is focused in core degradation both in-vessel and ex-vessel phenomena such as the QUENCH-12 tests performed at the Karlsruhe Institute of Technology (KIT). This step is needed before integral codes are applied to evaluate the behavior of VVER-plants under severe accident conditions. The main goal of the QUENCH-12 test was to evaluate the hydrogen generation resulting from the injection of cold quench water into an overheated and oxidized bundle of fuel rod simulators representing the VVER-fuel rods. Such phenomena are expected to occur during a severe accident sequence e.g. LOCA or as part of an accident management measure. This paper describes the investigations done to simulate the QUENCH-12 test using the ASTEC V2, which is developed by IRSN to simulate all physical and chemical phenomena during severe accident condition. A model of the QUENCH-12 test facility was developed for ASTEC first time taking into account the specific material data, geometry and boundary conditions. The evaluation of the results have shown, that ASTEC V2 is able to predict the main trends of key-parameters during the all test phases in good agreement with the measured data. The integral hydrogen generation is only over-predicted during the quench phase. As next steps, the model will be improved by taking into account the real material properties of the ZrNb-cladding typical of Russian reactors instead of the ones of Zr. Also the sensitivity studies regarding the electrical heater resistance will be carried out.

Modeling of fission product release during severe accidents with the fuel performance code ALCYONE

This paper presents simulations of four tests performed on medium to high burnup fuel during the VERCORS and VERDON experimental programs. The tests are representative of a Severe Accident (SA) sequence with a temperature increase up to fuel-clad melting and oxidizing/reducing conditions within the furnace. The simulations are performed with the fuel performance code ALCYONE where irradiated fuel thermochemistry and fission gas release are coupled. In this paper, the impact of the radial burnup and Fission Product (FP) profiles within the fuel pellet on the FP release from the sample during the SA sequence is studied. Simulations of the fuel sample behavior during nominal irradiation in commercial reactors are first performed to assess the initial state of the fuel. The simulations of the SA sequences include a burnup dependent fission gas release model. The simulated release curves of various volatile and semi-volatile FPs (Xe, I, Te, Cs, Mo and Ba) are compared successfully to on-line measurements. The impact of the burnup and FP radial profiles on both the thermochemical equilibria within the pellet and the FP release kinetics is discussed. It is shown that the FP release from the fuel pellets is not significantly increased by the consideration of the burnup and FP radial profiles. This conclusion is due to the limited radial extension of the peaked burnup radial profile in the fuel pellet and to the uniform temperature of the fuel samples.

Analysis of an unmitigated 2-inch cold leg LOCA transient with ASTEC and MELCOR codes

The analyses of postulated severe accident sequences play a key role for the international nuclear technical scientific community for the study of the effect of possible actions to prevent significant core degradation and mitigate source term release. To simulate the complexity of phenomena involved in a severe accident, computational tools, known as severe accident codes, have been developed in the last decades. In the framework of NUGENIA TA-2 ASCOM project, the analysis of an unmitigated 2-inch cold leg LOCA transient, occurring in a generic western three-loops PWR-900 MWe, has been carried out with the aim to give some insights on the modelling capabilities of these tools and to characterize the differences in the calculations results. The ASTEC V2.2b code (study carried out with ASTEC V2, IRSN all rights reserved, [2021]), and MELCOR 2.2 code have been used in this code-to-code benchmark exercise. In the postulated transient, the unavailability of all active injection coolant systems has been considered and only the injection of accumulators has been assumed as accident mitigation strategy.

Analysis on containment thermal hydraulic behaviour under passive containment heat removal system operation condition for HPR1000

HPR1000 NPP is designed with Passive Containment Heat Removal System (PCS) following the technical route called active and passive combined concept. The installed PCS system will have effect on the thermal hydraulic (T-H) behavior in the containment, and the complex T-H environment will influence the heat removal and operating character of PCS system backwards. The main goal of this work is to evaluate the effectiveness of the PCS system of HPR1000 and analyse the complex T-H environment of HPR1000 when the PCS system is operating. The HPR1000 containment is modelled by the containment T-H code and the PCS system is modelled based on experiment results. Then, the severe accident sequence initiated by large break LOCA is calculated by an integrated computer code, which outputs mass and energy release sources from reactor coolant primary system as boundaries for the containment T-H code. The results show that HPR1000 PCS system has enough heat removal capability under selected severe accident, which could ensure the containment meet the design requirement. Also, the nonuniformity of temperature and gases in the containment will disappear when the mass and energy releases turn stable, and the operating of PCS system does not cause a huge effect on containment inhomogeneity.

Numerical assessment for entry condition of severe accident management guidelines in a Swedish nuclear power plant

The entry conditions of severe accident management guidelines (SAMGs) in pressurized water reactors (PWRs) rely on the indication of core exit temperature (CET). Yet, the setpoints for the CET may be different from plant to plant. Most Westinghouse PWR designs adopt the setpoint of CET at 650℃ as the entry condition of the SAMGs, since this setpoint is an effective indicator of core damage in a wide spectrum of accident sequences. Motivated by the interest in the verification & validation of SAMS after the Fukushima accidents, the present study is conducted numerically to verify the effectiveness of the CET setpoint for the transition from emergency operation procedures (EOPs) to SAMGs in a Swedish nuclear power plant. For this purpose, six representative severe accident sequences covering the main contributors to the core damage frequency (CDF) are analyzed using the MELCOR code. Moreover, different CET readings and alternative entry conditions are also investigated. The simulation results show that the average CET = 650 °C is the effective setpoint as the entry condition of SAMGs, i.e., given this setpoint the transition from EOPs to SAMGs will take place slightly before the occurrence of core degradation, which secures the intended mitigation of SAMGs while keeping EOPs active as long as possible. On the other hand, it is too conservative if the maximum CET = 650 °C is used as the setpoint of entry condition of SAMGs, i.e., it will result in an excessive realization of SAMGs over EOPs. The coolant temperature in the primary circuits, the water level in the RPV and the hydrogen concentration in the containment can also be applied as reference indications of core damage states in the accident management.

Uncertainty and sensitivity analysis of the Phébus FPT3 test

Since the accident in the Fukushima Daichii Nuclear Power plant specific attention has been paid to the analyses and possible mitigation of severe accidents, as well as to design extension conditions. There is also a growing interest in applying not only the so-called best estimate codes to evaluate the postulated severe accident sequences, but to complement these studies with uncertainty analyses. The Phébus FP (Fission Product) research program, conducted by IRSN and CEA, aimed to study the physical phenomena that govern core degradation as well as, the release, transfer, and deposition of fission products and their behaviour in the containment. In this paper the degradation phase of the Phébus FPT3 experiment is modelled and simulated with the help of the best estimate system code ATHLET-CD 3.2, the core degradation module of AC2. The study is complemented by an uncertainty analysis as well, applying the so-called GRS method. The paper considers uncertain model (e.g. parameters concerning Zr oxidation) as well as experimental (e.g. material properties) parameters and determines the 95/95 tolerance limits for the main figures of merit. Furthermore, a sensitivity analysis is performed, and the Spearman’s rank correlation coefficient is determined for each uncertain input parameter.

Best-practice severe accident analysis for the OPR1000 short-term SBO sequence using MELCOR2.2 and MAAP5

A short-term station blackout (STSBO) accident has been considered as one of risk dominant sequences for nuclear power plants (NPPs). Unless appropriate mitigation actions are not taken, a rapid damage of the reactor core and thereby a severe accident may result in. This paper presents best practice analysis results on the evolution of the severe accident sequences and fission product (FP) source terms expected during the STSBO accident of a reference pressurized water reactor. In addition, this study aims to analyze the impact of dedicated severe accident mitigation/management actions on the results of interest through the established modeling approach. As the modeling and simulation tools for the analysis, MELCOR2.2 and MAAP5.05, were used to compare the analysis results of interest such as the integrities of the reactor coolant system (RCS) pressure boundary and the reactor/containment building, the generation of flammable gases (such as hydrogen and carbon monoxide), and the environmental release of the FP cesium. As a key finding, this study demonstrated that both codes similarly predicted the environmental release of the FP cesium as well as the general trend of the evolution of the severe accident sequences and plant responses. Of particular interest was the substantial difference in predicting the generation of the flammable gases by the two codes. Relevant insights are summarized in terms of particular points of interest.

Mitigation of severe accidents for SFR and associated event sequence assessment

This article deals with the assessment of severe accident event sequence and mitigation strategy elaboration performed in the frame of ASTRID SFR Gen IV reactor project (Le Coz et al., 2013), including the Japanese/French collaboration involving JAEA, MFBR, Framatome and CEA. In the first part of the paper, the mitigation strategy and the approach defined for the design of mitigation features, which are transverse tubes called DCS-M-TT are presented. By considering 21 DCS-M-TTs into the core as defined at the end of the conceptual design, which enable to extract the core materials and to significantly reduce the core reactivity within the order of 10 s in some typical accidental conditions, the accidental power excursions in ASTRID should not lead to a significant mechanical energy release, thus ensuring the integrity of the main vessel. Nevertheless, some loading assumptions independent from event evolutions have been defined in order to design the main vessel and the core catcher in order to cover extreme accident consequences and to get rid of event sequence dependency when designing the reactor. Then, SIMMER-IV (3D) the main calculation results of ULOF simulations whose goal was to verify the DCS-M-TT efficiency and to support a PIRT are highlighted in the paper. They confirmed that fuel discharge through discharge tubes has a large mitigating impact on power excursions. The last part of this paper deals with an approach dedicated to the description of severe accident event sequences through generic event trees (GETs), which logically summarize the representative courses of accident progression showing relationship between critical causal events and major outcomes of the accidents. A set of trees has been elaborated for the description of each event sequence able to cause severe accidents (ULOF: unprotected loss of flow, UTOP: unprotected transient overpower and USAF: unprotected sub-assembly fault). The event-tree bifurcation points towards one branch or another of the possible event evolution could be examined alternatively with a help of phenomenological event chart (PECs). Afterwards, in association with the GET analysis, a PIRT (Phenomenon identification ranking table) on ULOF has been developed conjointly by Japanese and French sides in order to prioritize R&D needs. About 100 phenomena have been retained and classified depending on the various time periods of the accident event sequences; the primary phase, the short-term relocation phase, the expansion phase and the long-term relocation and cooling phase.

MAAP5 simulation of AP1000 severe accident induced by pressurizer safety valve stuck-open

The Modular Accident Analysis Program version 5(MAAP5) is a computer code that can simulate the response of light water reactor power plants during severe accident sequences. The present work aims to simulate the severe accident of AP1000 with the Modular Accident Analysis Program version 5(MAAP5). The pressurizer (PZR) safety valve stuck-open is essentially a small break loss-of-coolant accident which becomes one of the major concerns on severe accident initiating event of pressurized water reactor (PWR). For such improbable accidents, four cases of PZR safety valve stuck-open accidents are simulated for comparison. This analysis is performed at 100% core power with different assumptions. For all cases, the primary system pressure can long term be maintained at a low level, and the ERVC system has enough capability to remove the heat from the molten pool. The total mass of hydrogen production in the core of the cases without auto depressurization system (ADS) is more significant than those cases with ADS. The results are meaningful and significant for comprehending the detail process of AP1000 severe accident, which is the basic standard for establishing the severe accident management guidelines.

Preliminary validation of ASTEC V2.2.b with the QUENCH-20 BWR bundle experiment

ASTEC is an integral severe accident code which aims to simulate entire severe accident transients of the nuclear reactors. New BWR-related models are implemented in the new versions of European ASTEC integral code by IRSN for BWR-typical core structures such as absorber blades, canister, etc. during the In-vessel core degradation after a severe accident in a BWR. The goal is to improve the simulation of the whole severe accident sequence in a BWR, such as the Fukushima accidents. An important step is the validation of the new models (oxidation, molten material formation and melt relocation, etc.) before the code is used to simulate severe accident scenarios in BWRs. The QUENCH-20 BWR test was specifically performed to investigate the oxidation, hydrogen generation, melt formation and oxidation during the reflooding of the test section representing a quarter of a SVEA-96 Optima-2 fuel assembly. The data measured during the QUENCH-20 BWR test e.g. temperature of the fuel rod simulators, of the blade and of the canister as well as the hydrogen production, quench front movement is used for the validation of the code’s prediction capability. In this work, the code ASTEC V.2.2.b is used to simulate the behavior of the QUENCH-20 test section. For this purpose, a detailed model of the QUENCH facility and test section was developed. This model was used to analyze the severe accident phenomena taking into account the same boundary conditions i.e. electrical power, mass flow rates, pressure, etc. as in the test. It was found out that the evolution of the temperature distribution in the bundle is reasonably evaluated by ASTEC for the entire duration of the transient. Besides, oxidation processes are also well reproduced. ASTEC predicts a total hydrogen generation of about 53 g while the measured value is around 57 g. In addition, ASTEC rather well predicted that about 15% of this hydrogen amount comes from the B4C-oxidation. Such good overall agreement between ASTEC and experimental measurements leads to further validation possibilities on BWR plant applications.

Extension of a Level 2 PSA Event Tree Based on Results of a Probabilistic Dynamic Safety Analysis of Induced Steam Generator Tube Rupture

This paper presents the approach of extending a classical generic event tree (ET) of a Level 2 Probabilistic Safety Analysis to the results of a probabilistic dynamic safety analysis. The example of creep-induced steam generator tube rupture has been chosen. The results of an Analysis of Thermal Hydraulics of Leaks and Transients with Core Degradation (ATHLET-CD)/Monte Carlo Dynamic Event Tree (MCDET) simulation analyzing the failure of reactor coolant system components by creeping in a scenario of a high-pressure core melt accident in a generic pressurized water reactor (PWR) have been implemented in the ET. From the results of these simulations, a set of parameters has been extracted and integrated into the ET along with their probability distributions. The effect of these parameters on both the progression of the severe accident sequence under consideration and the release categories of a generic German PWR plant, respectively, is discussed in this paper.