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Pressure suppression containment has been adopted extensively in boiling water reactor design due to its remarkable pressure suppression capacity. To study the pressure response characteristics of the containment in the initial stage of the loss-of-coolant accident, it is necessary to build a scaling facility to simulate the prototype. Analysis of the steam mass and energy that are released from the break as well as of the steam condensation that occurs in the suppression pool is conducted, respectively, for the prototype pressurized water reactor. The suppression pipe system resistance is also scaled with a certain ratio. The length and diameter of the main pipe and the branch pipe as well as the orifice diameter are calculated. At last, the prototype and the experimental facility are modeled by MELCOR to analyze the steam mass and energy release. The results show that the containment pressure evolution trends of the prototype and the scaled experimental facility are consistent. The pressure difference exists at some times, with a maximum relative error of 4%. [ABSTRACT FROM AUTHOR]

The small modular PWR (SMPWR) usually adopts integral design. Under severe accident, the system responses are different from those large PWRs. It is necessary to study the severe accident behavior of the SMPWR. A MELCOR model is developed for SMPWR and its steady-state results are in good agreement with the design values. Severe accidents induced by secondary pipeline break accidents are simulated, and no pressure relief measures are taken to keep the primary loop under high pressure. The mitigation effects of passive containment air cooling system (PAS) and passive cavity injection system (PCIS) are evaluated under different cases. The results show that under high pressure conditions, PCIS can effectively cool the lower head. The earlier the PCIS operates, the more significant the mitigation effect can be. In addition, PAS can effectively reduce the peak pressure and temperature in the containment. This study can provide a reference for the formulation of severe accident management guidelines on SMPWRs.

LVR-15 is a light-water-tank-type research reactor placed in a stainless-steel vessel under a shielding cover located in the Research Centre Rez (CVR) near Prague. It is operated at a steady-state power of up to 10 MWt under atmospheric pressure and is cooled by forced circulation. In 2011, the fuel was replaced, going from high-enriched uranium (HEU) to low-enriched uranium (LEU). After 2017, the State Office for Nuclear Safety (SUJB) asked CVR to evaluate the LVR-15 under Design Extended Conditions B (DEC-B). For this reason, a new model was developed in the MELCOR code, which allows for modelling the progression of a severe accident (SA) in light-water nuclear power plants and estimating the behaviour of the reactor under SA conditions. The model was built by collecting information about the LVR-15. Since the research reactor can have different core configurations according to the location of the core components, the core configuration with the most fuel (hottest campaign K221) was selected. Then, to create the radial nodalisation, the details of the core components were obtained and grouped in five radial rings and 27 axial levels. The simulation was run with the boundary conditions collected from campaign K221, and the results were compared with the reference values of the campaign with a negligible percentage of error. For the coolant inlet and outlet temperature, the reference values were 318.18 K and 323.5 K, respectively, while for the simulation, the steady state reached 319 K for the inlet temperature and 324 K for the outlet temperature. Additionally, the cladding temperature of the hottest assembly was compared with the reference value (353.72 K) and the steady-state simulation results (362 K). In future work, different transients leading to severe accidents will be simulated. When simulating the LVR-15 reactor with MELCOR, specific attention is required for the aluminium-cladded fuel assemblies, as the model requires some assumptions to cope with the phenomenological limitations. [ABSTRACT FROM AUTHOR]

This paper presents the results of a severe accident parametric sensitivity study performed on a model of a generic integral pressurized water reactor (iPWR). The analyzed sequence is a loss of coolant accident (LOCA)-type scenario, postulating that safety systems are not available. In this work, a MELCOR 2.2 input deck of a generic iPWR was developed based on publicly available data. The iPWR used in this work is a generic iPWR with a thermal power of about 160 MWth, characterized by a compact steam generator and a submerged containment configuration. A hypothetical scenario considered in this work was an unmitigated small break LOCA leading to a severe accident with partial core degradation, due to the postulated assumptions. In the presented paper, 16 sensitivity cases were calculated and analyzed, focusing mainly on heat transfer, decay heat, and core degradation parameters. The selected parameters and their combination caused partial core degradation and showed significant differences in investigated variables such as core degradation and hydrogen generation, as well as examined CPU time consumption. This a preparatory work performed in the framework of the Horizon Euratom SASPAM-SA project, which aims to investigate the applicability and transfer of the operating large light-water reactor knowledge and know-how to the iPWR, taking into account European licensing analysis needs for the severe accident and emergency planning zone. [ABSTRACT FROM AUTHOR]

During the Fukushima Daiichi nuclear accident in 2011, hydrogen explosions occurred in all units from Unit 1 to Unit 3. Consequently, one of the lessons learned from the Fukushima Daiichi accident is the necessity of implementing hydrogen control and mitigation strategies for most Nuclear Power Plants (NPPs). This paper focuses on the incorporation of Passive Autocatalytic Recombiners (PARs) during the design phase of a small modular Pressurized Water Reactor (SMR-PWR) project. The numerical analyses are conducted using the MELCOR v. 2.2 code. Two scenarios are compared: the Total Loss of Feed Water (TLOFW) severe accident with and without PARs. Saphiro’s diagram is utilized to investigate whether the mixture's composition (hydrogen, air, steam) is flammable for both scenarios. It has been observed that the inclusion of PARs leads to a reduction in hydrogen risk (detonative or deflagrative) as the final hydrogen concentration values fall below the flammability limit. This study is preliminary, and further research is required.

The potential for recriticality and sub-critical boron concentrations is analyzed during the relocation of the fuel rods in the assembly, which we call late phase of a severe accident, via coupling between MELCOR and whole-core Monte Carlo analyses by Serpent 2. The recriticality, initiated during the early phase, is found to maintain when the fuel assemblies containing intact fuel rods are submerged by the cooling water. It is also found that the effect of the negative reactivity insertion via remaining fission products in the fuel debris increases as the burnup increases. The sub-critical boron concentrations during the late phase are found to be 76∼544 ppm lower than those during the early phase. Therefore, it can be concluded that the boron concentration that prevents recriticality not only during the early phase but also during the late phase is the sub-critical boron concentration during the early phase.

LVR-15 is a light-water-tank-type research reactor placed in a stainless-steel vessel under a shielding cover located in the Research Centre Rez (CVR) near Prague. It is operated at a steady-state power of up to 10 MWt under atmospheric pressure and is cooled by forced circulation. In 2011, the fuel was replaced, going from high-enriched uranium (HEU) to low-enriched uranium (LEU). After 2017, the State Office for Nuclear Safety (SUJB) asked CVR to evaluate the LVR-15 under Design Extended Conditions B (DEC-B). For this reason, a new model was developed in the MELCOR code, which allows for modelling the progression of a severe accident (SA) in light-water nuclear power plants and estimating the behaviour of the reactor under SA conditions. The model was built by collecting information about the LVR-15. Since the research reactor can have different core configurations according to the location of the core components, the core configuration with the most fuel (hottest campaign K221) was selected. Then, to create the radial nodalisation, the details of the core components were obtained and grouped in five radial rings and 27 axial levels. The simulation was run with the boundary conditions collected from campaign K221, and the results were compared with the reference values of the campaign with a negligible percentage of error. For the coolant inlet and outlet temperature, the reference values were 318.18 K and 323.5 K, respectively, while for the simulation, the steady state reached 319 K for the inlet temperature and 324 K for the outlet temperature. Additionally, the cladding temperature of the hottest assembly was compared with the reference value (353.72 K) and the steady-state simulation results (362 K). In future work, different transients leading to severe accidents will be simulated. When simulating the LVR-15 reactor with MELCOR, specific attention is required for the aluminium-cladded fuel assemblies, as the model requires some assumptions to cope with the phenomenological limitations.

This paper presents the results of a severe accident parametric sensitivity study performed on a model of a generic integral pressurized water reactor (iPWR). The analyzed sequence is a loss of coolant accident (LOCA)-type scenario, postulating that safety systems are not available. In this work, a MELCOR 2.2 input deck of a generic iPWR was developed based on publicly available data. The iPWR used in this work is a generic iPWR with a thermal power of about 160 MWth, characterized by a compact steam generator and a submerged containment configuration. A hypothetical scenario considered in this work was an unmitigated small break LOCA leading to a severe accident with partial core degradation, due to the postulated assumptions. In the presented paper, 16 sensitivity cases were calculated and analyzed, focusing mainly on heat transfer, decay heat, and core degradation parameters. The selected parameters and their combination caused partial core degradation and showed significant differences in investigated variables such as core degradation and hydrogen generation, as well as examined CPU time consumption. This a preparatory work performed in the framework of the Horizon Euratom SASPAM-SA project, which aims to investigate the applicability and transfer of the operating large light-water reactor knowledge and know-how to the iPWR, taking into account European licensing analysis needs for the severe accident and emergency planning zone.

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.

In past years the Paul Scherrer Institute (PSI, Switzerland) and the Karlsruhe Institue of Technology (KIT, Germany)) collaborated to develop a model to account for the active role of nitrogen in the air oxidation of a Zircalloy cladding. The “PSI-KIT Nitriding Model for Zirconium based Fuel Cladding” model was implemented at PSI into PSI-MELCOR 1.8.6.In order to make a preliminary evaluation of the effect of the new model on the evolution of full-scale spent fuel pool accidents, one spent fuel pool event was analyzed using the PSI research version of PSI-MELCOR 1.8.6, which includes the nitriding model. To adapt an existing input deck for the calculations, a sensitivity study was conducted to find an optimal nodalization for the analyses. The nitriding model results were compared to those calculated with the MELCOR 1.8.6-PSI without the new nitriding model.The results demonstrate the effect of the nitriding reactions in spent fuel pool accident progression. Moreover, they confirm the impact of ZrN formation during cladding oxidation in air when the oxidation reactions lead to oxygen starvation inside the fuel assemblies. The nitriding reaction led to higher chemical heat generation during the accident and to an earlier failure of the cladding than when the effect of nitrogen reactions was not considered.It should be noted that the nitriding model, as implemented in the PSI version of MELCOR 1.8.6 has not yet been conclusively validated. Thereby the results presented in this paper should be treated as a preliminary demonstration of the capabilities of the model.

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.

To ensure the applicability of accident-tolerant fuels, their behaviors under various accidental conditions must be assessed. While the dependences of the behavior of single physical parameters can be investigated in single- or separate-effect experiments, and more complex phenomena can be investigated using integral-effect tests, the behavior of an entire system as complex as a nuclear power plant core must be investigated using computer code modeling. One of the most commonly used computer codes for the assessment of severe accidents is MELCOR 2.2. In version 18019, the authors enabled the modeling of the behavior of the nuclear fuel with FeCrAl cladding (namely, alloy B136Y3) for the first time, using the GOX model. The ability of this model to reasonably accurately predict the behavior of FeCrAl cladding in accident conditions with quenching was verified in this work by modeling the QUENCH-19 experiment carried out in the Karlsruhe Institute of Technology on the QUENCH experimental device and by subsequent comparison of the MELCOR calculation results with the experiment. This article proves that the GOX model can be used to evaluate the behavior of FeCrAl cladding and that the results can be considered conservative. [ABSTRACT FROM AUTHOR]

One of the most important environmental and safety concerns in nuclear fusion plants is the confinement of radioactive substances into the reactor buildings during both normal operations and accidental conditions. For this reason, hydrogen build-up and subsequent ignition must be avoided, since the pressure and energy generated may threaten the integrity of the confinement structures, causing the dispersion of radioactive and toxic products toward the public environment. Potentially dangerous sources of hydrogen are related to the exothermal oxidation reactions between steam and plasma-facing components or hot dust, which could occur during accidents such as the in-vessel loss of coolant or a wet bypass. The research of technical solutions to avoid the risk of a hydrogen explosion in large fusion power plants is still in progress. In the safety and environment work package of the EUROfusion consortium, activities are ongoing to study solutions to mitigate the hydrogen explosion risk. The main objective is to preclude the occurrence of flammable gas mixtures. One identified solution could deal with the installation of passive autocatalytic recombiners into the atmosphere of the vacuum vessel pressure suppression system tanks. A model to control the PARs recombination capacity as a function of thermal-hydraulic parameters of suppression tanks has been modeled in MELCOR. This paper aims to test the theoretical effectiveness of the PAR intervention during an in-vessel loss of coolant accident without the intervention of the decay heat removal system for the Water-Cooled LithiumLead concept of EU-DEMO. [ABSTRACT FROM AUTHOR]

Design basis accidents are investigated continuously for the European DEMO reactor accompanying its development. One selected postulated initial event (PIE) is a loss of vacuum (LOVA) in vacuum vessel (VV) with large ingress of air induced by rupture in a VV penetration. It has been investigated for the helium cooled pebble bed (HCPB) blanket concept according to the DEMO baseline 2017. The associated primary heat transfer system (PHTS) and the related systems in the tokamak building, from the VV to the PHTS vault and galleries, are considered for the investigation. The LOVA is postulated to occur at a port seal of the electron cyclotron equatorial port plug on the side of the closure plate with (i) a small leak of 1.0 × 10 ^−3 m ^2 , or (ii) a large break size of 1.0 × 10 ^−2 m ^2 . Air ingress from one port cell into the VV leads to the VV pressurization and the fusion power termination followed by an unmitigated plasma disruption. A loss of off-site power for 32 h is assumed to coincide with the disruption. An in-vessel loss of coolant accident (LOCA) is considered as a consequence if the affected first wall (FW) reaches the defined temperature of 1000 °C. The radioactive inventories in the VV (tritium, W-dust) can mobilize towards the VVPSS, the affected systems in the building and the environment due to pressurization, venting and leak conditions. MELCOR 1.8.6 for fusion is applied for this deterministic safety analysis. The resulting releases of radioactivity to the environment are then provided for dose calculation using the computer systems UFOTRI and COSYMA. Outcomes of this LOVA analysis are critically discussed: the transient evolutions of different cases are compared; hydrogen production is detected in case of aggravating FW failure; the source terms (tritium, W-dust) are transported to the connected systems; and the dose results from the environmental releases are provided.

Since Fukushima nuclear power plant (NPP) accident in 2011, the importance of research on various severe accident phenomena has been emphasized. Particularly, detailed analysis of combustion risk is necessary following the containment damage caused by combustion in the Fukushima accident. Many studies have been conducted to evaluate the risk of local hydrogen concentration increases and flame propagation using computational code. In particular, the potential for combustion by local hydrogen concentration in specific areas within the containment has been emphasized. In this study, the process of flame propagation generated inside a reactor core to containment during a loss of coolant accident (LOCA) was analyzed using MELCOR 2.1 code. Later in the LOCA scenario, it was expected that hydrogen combustion occurred inside the reactor core owing to oxygen inflow through the cold leg break area. The main driving force of the oxygen intrusion is the elevated containment pressure due to the molten corium–concrete interaction. The thermal and mechanical loads caused by the flame threaten the integrity of the containment. Additionally, the containment spray system effectiveness in this situation was evaluated because changes in pressure gradient and concentrations of flammable gases greatly affect the overall behavior of ignition and subsequent containment integrity.

The EU-DEMO must demonstrate the possibility of generating electricity through nuclear fusion reactions. Moreover, it must denote the necessary technologies to control a powerful plasma with adequate availability and to meet the safety requirements for plant licensing. However, the extensive radioactive materials inventory, the complexity of the plant, and the presence of massive energy sources require a rigorous safety approach to fully realize fusion power's environmental advantages. The Tokamak building barrier design must address two main issues: radioactive mass transport hazards and energy-related or pressure/vacuum hazards. Safety studies are performed in the frame of the EUROfusion Safety And Environment (SAE) work package to support design improvement and evaluate the thermal-hydraulic behavior of confinement building environments during accident conditions in addition to source term mobilization. This paper focuses on developing a thermal-hydraulic model of the EU-DEMO Tokamak building. A preliminary model of the heat ventilation and air conditioning system and vent detritiation system is developed. A loss-of-coolant accident is studied by investigating the Tokamak building pressurization, source term mobilization, and release. Different nodalizations were compared, highlighting their effects on source term estimation. Results suggest that the building design should be improved to maintain the pressure below safety limits; some mitigative systems are preliminarily investigated for this purpose. [ABSTRACT FROM AUTHOR]

Severe Accidents Management (SAM) in nuclear power plants is an important safety issue. SAMs require assessing the thermal-hydraulic conditions of the nuclear reactor during the accident. This paper investigates three accident scenarios, Station Black-Out (SBO), SBO with Large Break Loss of Coolant Accident (LBLOCA), and SBO with Small Break Loss of Coolant Accident (SBLOCA) in the WWER1000/V446 reactor using the MELCOR1.8.6 code. The RELAP3.2-SCDAP code is also used to evaluate the accuracy of the reactor pressure vessel parameters. The purpose of this study was to determine the operating conditions of the reactor during the accident scenarios with emphasis on hydrogen production. The results showed that the SBO has the highest hydrogen production (2150 kg) compared to other accidents and the maximum fuel temperature occurred sooner than the other SAs (11800 seconds). The results obtained by the two codes and the Final Safety Analysis Report (FSAR) of the WWER1000/V446 reactor showed good agreement. The results of this analysis help make the appropriate decision in SAM.

One of the coming Small Modular Reactor (SMR) designs is the integrated Pressurized Water Reactor (iPWR), which merges years of knowledge and experience in light water reactors with new demands from the market, such as flexibility, construction optimization, and reliability. Generally, SMRs are considered to be inherently safe and, in many cases, safer than generation III designs. Most of the SMR designs rely on passive safety systems at different levels of passive operation. One of the ways to examine new concepts is a numerical simulation using a reliable tool. However, when new features come up, there are also new challenges for existing tools, even though they are continuously updated following the technology's evolution. To investigate if implemented new features give expected results and to increase knowledge about SMRs modeling and simulations, accident transients of iPWR were calculated using MELCOR 2.2 code and analyzed. The work aims to understand better numerical simulations of the analyzed iPWR accident scenarios and code reliability. To do so, the paper is divided into two parts; in Part 1, the authors present MELCOR 2.2 iPWR input deck description, including nodalization and steady-state conditions as well as the analysis of an iPWR LOCA-type scenario in the design basis accident domain. In addition, design basis accident progression is compared to beyond design basis scenario in which safety systems are postulated partly failing. The analyses of beyond design basis scenarios in which core degradation is observed are published separately in second part of that paper. The obtained results indicate that MELCOR2.2 can simulate the thermal–hydraulic response of an iPWR in accident scenarios. Values received in this study in steady state and accident simulations are mostly expected and coherent with general knowledge of accident progression. [ABSTRACT FROM AUTHOR]

Radioactive aerosol may be released from the molten core during a severe accident in a nuclear power plant. Pool scrubbing, a representative aerosol scrubbing process incorporated in several severe accident scenarios, has been modeled in MELCOR. Under gas injection conditions with a low Weber number, gas flow regimes above the nozzle are globule and swarm. Based on visual scrubbing experiments, we investigated the gas flow and aerosol scrubbing characteristics under low-Weber-number injection conditions. We systematically analyzed the relationship between scrubbing efficiency and gas flow behavior through experiments and the applicability of the aerosol scrubbing model in MELCOR to different pool scrubbing conditions. Also, the variation of the globule and swarm shapes with the inlet pressure was obtained to evaluate the applicability of the gas flow model in MELCOR. Additionally, the bubble size distribution and Sauter average diameter were used to characterize the swarm flow characteristics based on the image processing results. Finally, we used a modified scrubbing model to compare the experimental results and evaluate the most suitable swarm characterization parameters. [Display omitted] • The relationship between scrubbing efficiency and gas flow behavior was analyzed. • The applicability of the gas flow and aerosol scrubbing model in MELCOR was evaluated. • The modified models were used to evaluate the swarm characterization parameter. [ABSTRACT FROM AUTHOR]

One of the important issues raised from the Fukushima-Daiichi accident is the safety of multi-unit sites when simultaneous accidents occur at the site and recently a multi-unit PSA methodology is being developed worldwide. Since all operation modes of the plant should be considered in the multi-unit PSA, the accident analysis needs to be performed for shutdown operation modes, too. In this study, a station blackout during the mid-loop operation is selected as a reference scenario. The overall accident progression for the mid-loop operation is slower than that for the full-power operation because the residual heat per mass of coolant is about 6 times lower than that in the mid-loop scenario. Though the fractions of Cs released from the core to the RCS in both operation modes are almost the same, the amount of Cs delivered to the containment atmosphere is quite different due to the chemisorption in the RCS. While 45.5% of the initial inventory is chemisorbed on the RCS surfaces during the full-power operation, only 2.2% during the mid-loop operation. The containment remains intact during the mid-loop operation, though 83.9% of Cs is delivered to the containment.

As the amount of fission product released from ISLOCA was overestimated because of conservative assumptions in the past, several studies have been recently conducted to evaluate the actual release amount. Among several pathways for the ISLOCA, most studies were focused on the pathway with the highest possibility. However, different ISLOCA pathways may have different fission product release characteristics. In this study, fission product behavior was analyzed for various pathways at the Westinghouse two-loop plant using MELCOR. Four pathways are considered: the pipes from a cold leg, from a downcomer, from a hot leg to the outlet of RHR heat exchanger, and the pipe from the hot leg to the inlet of RHR pump (Pathway 1–4). According to the analysis results, cladding fails at around 2.5 h in Pathways 1 and 2, and on the other hand, about 3.3 h in Pathways 3 and 4 because the ISLOCA pathways affect the safety injection flow path. While the release amount of cesium and iodine ranges between 20 and 26% in Pathways 1 to 3, Pathway 4 allows only 5% to the environment because the break location is submerged. Also, as more than 90% of cesium released to the environment passes through the personnel door, reinforcing the pressure capacity of the doors would be a significant factor in the accident management of the ISLOCA.

It is necessary to evaluate the released gases distribution (especially hydrogen) to avoid hydrogen accumulation and maintain containment integrity against the pressure loads due to its deflagration or detonation. For this purpose, this work uses coupling of the FLUENT and MELCOR to evaluate the produced gases distribution in the WWER1000/V446 during in-vessel and ex-vessel phases of the Station Black Out (SBO) along with Large Break Loss of Coolant Accident (LBLOCA). The results indicate that 30 cm mesh size is sufficiently fine to evaluate the gases behavior in the containment. Also, driving force of the hydrogen/carbon monoxide (CO) and steam are mainly in the form of the plume due to buoyancy effects and jet flow caused by momentum, respectively. Furthermore, hydrogen accumulates in the upper compartments more than in other areas and its average volume fraction reaches to 12.7% at the end of the calculations. In addition, it can be concluded that steam condensation on the walls increases the hydrogen volume fraction up to 2%. [ABSTRACT FROM AUTHOR]

IFMIF-DONES will be an experimental facility designed to irradiate material samples under conditions similar to those expected in future fusion power reactors. As a radiological facility, it requires the precise design of specific safety subsystems to ensure the protection of the public, the plant personnel and the environment. This is the case of the Argon Purification Subsystem (ArPS), which is responsible for ensuring the inertization of those rooms where liquid Li may be present, involving a risk of fire and hazardous releases. In this context, it is important to set Ar purity requirements as design input for the ArPS to ensure this inertization while being technically sound due to the large amounts of volumes involved. To support this design process, a MELCOR simulation model of the IFMIF-DONES ArPS has been developed. Using this model, we perform a parametric study to analyse the performance of this subsystem under different configurations. The goal is to keep the impurity concentrations of the inert rooms stable during a year of operation in order to minimize the risk of Li reactions, while ensuring dynamic confinement and reducing Ar consumption. In addition, the transition between maintenance and operation modes is considered, analysing the effect of different transition periods and injected Ar inventory. • The MELCOR modelling of the IFMIF-DONES argon purification system is described. • Different parameters are analysed from simulations of inert rooms of the building. • We study how to keep impurities stable in inert rooms over a year of operation. • We analyse how different parameters affect the maintenance-operation transition. [ABSTRACT FROM AUTHOR]

• Material model for FeCrAl implemented in MELCOR-2.2 with increased detail. • Recovered PWR LOCA scenario analyzed for high burnup ATF and reference zircaloy case. • For scenario analyzed, additional coping time from FeCrAl is important. • Increased burnup increases fission product inventory. • Lower fission product release with FeCrAl compared to zircaloy maintained at high burnup. It has been established that incremental improvements in beyond design basis accident performance can be achieved through accident tolerant fuel (ATF). However, they have the potential to recover margin with respect to conventional fuel and therefore enhance plant economics through uprate or cycle length increase. To realize this potential, it is necessary to quantify the reduction in source term due to use of ATF, and correspondingly how this is affected by increasing the cycle length and/or burnup. This requires development of a risk informed analysis methodology for ATF under high burnup conditions, which is being developed within LWRS. To this end, a MELCOR simulation of FeCrAl ATF in a recovered Large Break LOCA (LBLOCA) scenario has been developed, using a model based on the Zion Pressurized Water Reactor (PWR). The new user defined material capability, along with the inclusion of detailed neutronics- and depletion-derived parameters such as core power distribution, decay heat behavior, and fission product inventories, allows a more detailed simulation of FeCrAl clad material properties and behavior than has previously been possible using MELCOR. These detailed FeCrAl results were compared with a zircaloy clad model to investigate the differences between the two clad types and quantify the benefits of FeCrAl ATF with an 18-month cycle. Next, the fuel cycle with ATF was extended to 24 months, to determine whether any additional safety margin provided by FeCrAl ATF could be leveraged to implement high burnup FeCrAl-clad fuel while retaining the same operating and safety limits as current zircaloy-clad fuel. For the particular scenario analyzed, the delay to fuel failure from using ATF was of the same order as the LOCA recovery time, and hence significant in reducing fission product release, with the 24-month FeCrAl core performing better than the 18-month zircaloy core. It is noted that for other transients, the reduction in release due to using FeCrAl may be less significant. The material model developed here can be used in such further studies in support of determining the overall source term. [ABSTRACT FROM AUTHOR]

One of the most important environmental and safety concerns in nuclear fusion plants is the confinement of radioactive substances into the reactor buildings during both normal operations and accidental conditions. For this reason, hydrogen build-up and subsequent ignition must be avoided, since the pressure and energy generated may threaten the integrity of the confinement structures, causing the dispersion of radioactive and toxic products toward the public environment. Potentially dangerous sources of hydrogen are related to the exothermal oxidation reactions between steam and plasma-facing components or hot dust, which could occur during accidents such as the in-vessel loss of coolant or a wet bypass. The research of technical solutions to avoid the risk of a hydrogen explosion in large fusion power plants is still in progress. In the safety and environment work package of the EUROfusion consortium, activities are ongoing to study solutions to mitigate the hydrogen explosion risk. The main objective is to preclude the occurrence of flammable gas mixtures. One identified solution could deal with the installation of passive autocatalytic recombiners into the atmosphere of the vacuum vessel pressure suppression system tanks. A model to control the PARs recombination capacity as a function of thermal-hydraulic parameters of suppression tanks has been modeled in MELCOR. This paper aims to test the theoretical effectiveness of the PAR intervention during an in-vessel loss of coolant accident without the intervention of the decay heat removal system for the Water-Cooled LithiumLead concept of EU-DEMO.

To ensure the applicability of accident-tolerant fuels, their behaviors under various accidental conditions must be assessed. While the dependences of the behavior of single physical parameters can be investigated in single- or separate-effect experiments, and more complex phenomena can be investigated using integral-effect tests, the behavior of an entire system as complex as a nuclear power plant core must be investigated using computer code modeling. One of the most commonly used computer codes for the assessment of severe accidents is MELCOR 2.2. In version 18019, the authors enabled the modeling of the behavior of the nuclear fuel with FeCrAl cladding (namely, alloy B136Y3) for the first time, using the GOX model. The ability of this model to reasonably accurately predict the behavior of FeCrAl cladding in accident conditions with quenching was verified in this work by modeling the QUENCH-19 experiment carried out in the Karlsruhe Institute of Technology on the QUENCH experimental device and by subsequent comparison of the MELCOR calculation results with the experiment. This article proves that the GOX model can be used to evaluate the behavior of FeCrAl cladding and that the results can be considered conservative.

The EU-DEMO must demonstrate the possibility of generating electricity through nuclear fusion reactions. Moreover, it must denote the necessary technologies to control a powerful plasma with adequate availability and to meet the safety requirements for plant licensing. However, the extensive radioactive materials inventory, the complexity of the plant, and the presence of massive energy sources require a rigorous safety approach to fully realize fusion power’s environmental advantages. The Tokamak building barrier design must address two main issues: radioactive mass transport hazards and energy-related or pressure/vacuum hazards. Safety studies are performed in the frame of the EUROfusion Safety And Environment (SAE) work package to support design improvement and evaluate the thermal-hydraulic behavior of confinement building environments during accident conditions in addition to source term mobilization. This paper focuses on developing a thermal-hydraulic model of the EU-DEMO Tokamak building. A preliminary model of the heat ventilation and air conditioning system and vent detritiation system is developed. A loss-of-coolant accident is studied by investigating the Tokamak building pressurization, source term mobilization, and release. Different nodalizations were compared, highlighting their effects on source term estimation. Results suggest that the building design should be improved to maintain the pressure below safety limits; some mitigative systems are preliminarily investigated for this purpose.

• The probabilistic method to propagate input uncertainty can be used to develop uncertainty analyses. • One uncertainty tool available is DAKOTA also provided as a SNAP plug-in. • MELCOR-DAKOTA coupling within SNAP is described. Uncertainty estimation to assess figures of merit characterizing evolution of a severe accident transient is a topic of current investigation in development of best-estimate plus uncertainty methodology. The probabilistic method to propagate input uncertainty is one of the methodologies used to develop Uncertainty Analyses (UAs). Using this methodology, UAs are performed by sampling probability distributions that describe the range of possible values that computer simulation model inputs can have. For each sample (or realization) of a set of uncertain input parameters, a computer simulation is performed. From the range of code simulation results obtained for each input realization, a distribution of code results is obtained. In this process, the distribution of input uncertainties is propagated to obtain a distribution of possible code results (i.e., the code output uncertainty). This probabilistic methodology is facilitated using Uncertainty Tools (UTs), which can be coupled with the accident analysis computer code to perform an UA. One of the UTs currently available is DAKOTA, developed by Sandia National Laboratories. DAKOTA is also provided as a SNAP plug-in. SNAP is a graphical user interface designed to support the use of USNRC codes (MELCOR, TRACE, etc). This paper is entirely derived from the NUREG/IA-532 issued by USNRC in April 2023 (Mascari et al., 2023) and has as a major target to summarize the main needs of UA in severe accident, the main element of the probabilistic method to propagate input uncertainty, and the workflow within SNAP to assist other interested analysts with their applications given they are members of the USNRC Cooperative Severe Accident Research Program (CSARP). [ABSTRACT FROM AUTHOR]

• Do version control of hand-written input decks. • Wrapper script to check output and monitor logs. • Substitution allows automated parameter sweeps. • Using batch system features gives better throughput. MELCOR is a software tool for accident analysis with a long history in the fission sector. A model has been developed for use in MELCOR, using proposed layouts of EU DEMO and assumptions based on current design choices, to facilitate safety design and development for fusion power plants. There seems however to be a misconception that MELCOR is difficult to run and that the tasks must be executed sequentially. Accident analysis also requires reruns of multiple slightly different scripts. We have streamlined the running of MELCOR tasks to allow simultaneous execution on a computing cluster. The input decks are version controlled within UKAEA's internal GitLab repository, and shell scripts wrap the separate MELCOR invocations, embed specific job parameters, and keep separate the inputs and outputs of multiple separate instances. We illustrate the streamlined running using two examples: a matrix representing potential responses to an incident by different valves, and a parameter sweep across a range of response times for a single valve. The new technique is shown to be a much faster method of carrying out sensitivity studies and allowing further investigations into specific areas of concern in more detail. [ABSTRACT FROM AUTHOR]

• A scenario of SGTR + SLB was simulated in a three-loop Western 1000 MWe PWR. • A comparison is carried out between MELCOR 2.2 and CATHARE 2 codes. • The analysis is focused on thermal–hydraulic and FP releases. • The results are compared, discussed and the differences highlighted. The European project denominated Reduction of Radiological Consequences of design basis and design extension Accidents (R2CA) was launched in September 2019, with a very broad participation: 11 countries, 17 participating organisations, including international organisations, utilities, regulators, technical support organisations, researchers and developers and was coordinated by IRSN. The main goal of the project was to assess the conservatisms in the radiological releases calculations in nuclear power plant (NPP) studies. The work here presented is focused on developing new calculation methodologies and updating computer code models to carry out more detailed assessments of source terms resulting from a steam generator tube rupture (SGTR) accident in the Design Extension conditions A (DEC-A) domain. For this purpose, participants in Work Package 2 (WP2) of R2CA developed and implemented simplified models in the Severe Accident (SA) and Thermal-Hydraulic (TH) codes that take into account iodine spiking and iodine transport phenomena within primary and secondary circuit. These methodologies will not only provide a better estimation of the radiological consequences, but should also serve to improve accident management procedures, innovative instrumentation development and early detection tools. The new approaches are tested in a SGTR + Steam Line Break Outside Containment scenario without significant fuel degradation in a three-loop Western 1000 MW e PWR. During the transient, operators are also assumed to implement emergency operating procedures (EOPs) in line with Westinghouse EOPs to limit the release of radioactivity and control the evolution of plant parameters. The calculations are performed by Tractebel and CIEMAT with the American SA tool MELCOR and Bel-V with the French TH code CATHARE. The results highlighted some limitations of implemented models in predicting iodine behaviour as well as small discrepancies in the TH evolution of the transient, both of which were analysed and discussed in detail. [ABSTRACT FROM AUTHOR]

• The safe storage of Spent Nuclear Fuel (SNF) within a dry cask system must be guaranteed for any operating condition. • The influence of a partial to complete blockage of air inlet openings on the PCT is assessed for the HI-STORM 100S system. • The cooling and the safe storage of the SNF seem to be ensured also in out-of-the ordinary situations. • The results from the adiabatic (most extreme) case indicate the importance of a tempestive accident management. • MELCOR showed good adaptability and flexibility, even in simulating scenarios out of its natural domain of application. The safe storage of Spent Nuclear Fuel (SNF) within a dry cask system must be guaranteed for any operating condition. In particular, the Peak Cladding Temperature (PCT) is currently employed as a measure of the fuel integrity in normal conditions. The same fuel integrity has to be ensured also in off-normal and accidental conditions by keeping the PCT below the regulatory limits. In this regard, the work proposed in this paper aims at evaluating the performance of a concrete-based dry cask in off-normal and accidental conditions. More specifically, the influence of a partial to complete blockage of the air inlet openings on the PCT is assessed by means of a MELCOR model of the HI-STORM 100S cask. Attention is also paid to the temperature of the external concrete cask in order to check the integrity of the concrete itself. [ABSTRACT FROM AUTHOR]

MELCOR is a state-of-the-art code used to model severe accidents in conventional nuclear reactors Humphries et al., 2021. The purpose of this work was to enhance the material modelling in MELCOR via one-way coupling to the equilibrium thermodynamics code Thermochimica. One-way coupling from MELCOR to Thermochimica was implemented to allow Thermochimica to predict the phase fractions and compositions within each simulation cell. Two test case scenarios were simulated in MELCOR using a generic boiling water reactor design that included either conventional Zircaloy-2 or accident-tolerant FeCrAl fuel cladding/canister material. FeCrAl cladding/canister material was examined in the second case because the improved oxidation kinetics make FeCrAl a potential choice for future accident-tolerate cladding material. The results from these two cases were analysed with Thermochimica to demonstrate potential areas for improvement of MELCOR's material modelling capabilities. • MELCOR simulations comparing Zry-2 to FeCrAl cladding under severe accidents. • Coupled Thermochimica to MELCOR for enhanced materials chemistry predictions. • OECD/NEA TAF-ID database used for thermodynamic calculations. [ABSTRACT FROM AUTHOR]

The Probabilistic Safety Assessment (PSA) is part of a Nuclear Power Plant (NPP) licensing process. It considers the elaboration and updating of probabilistic models that estimate the risk associated to the operation, allowing the risk monitoring from the design to the plant decommissioning, for both operational as regulatory activities. The PSA identifies those components or plant systems whose unavailability contributes significantly to the Core Damage Frequency (CDF) and to the Large Early Release Frequency (LERF) of radioactive material. Based on the PSA Level 1 results for a reference plant under design, the Analysis, Evaluating and Risk Management Laboratory (LabRisco), located in the University of São Paulo (USP), Brazil, started the analytical investigation of severe accident phenomena using the US Nuclear Regulatory Commission (NRC) MELCOR2.2 code – focusing on the qualification of a group of specialists who will subsidize a PSA Level 2 for the same plant. This PSA Level 1 shows that the accident with large CDF contribution is the Loss of Feed Water Accident (LOFW). Therefore, the initial objective of the investigation was to model the progression of severe accidents during a LOFW for the reference Pressurized Water Reactor (PWR) and to analyze the response of the plant under these accident scenarios. During the course of the hypothetical LOFW in the reference plant, hydrogen was generated – by a reaction between the high temperature steam water and the fuel-cladding inside the reactor pressure vessel (RPV) but not representing a serious threat to the RPV integrity.

Pressure suppression containment has been adopted extensively in boiling water reactor design due to its remarkable pressure suppression capacity. To study the pressure response characteristics of the containment in the initial stage of the loss-of-coolant accident, it is necessary to build a scaling facility to simulate the prototype. Analysis of the steam mass and energy that are released from the break as well as of the steam condensation that occurs in the suppression pool is conducted, respectively, for the prototype pressurized water reactor. The suppression pipe system resistance is also scaled with a certain ratio. The length and diameter of the main pipe and the branch pipe as well as the orifice diameter are calculated. At last, the prototype and the experimental facility are modeled by MELCOR to analyze the steam mass and energy release. The results show that the containment pressure evolution trends of the prototype and the scaled experimental facility are consistent. The pressure difference exists at some times, with a maximum relative error of 4%.

Nuclear safety relies to a good extent on thoroughly validated codes. However, code predictions are affected by uncertainties that need to be quantified for a more accurate evaluation of safety margins. In this regard, the present paper proposes a preliminary uncertainty and sensitivity analysis of the thermal behavior of a concrete-based dry cask for spent nuclear fuel storage, employing the MELCOR code and a series of MATLAB scripts. As thermal behavior is of utmost importance for the fulfillment of United States Nuclear Regulatory Commission (USNRC) safety requirements, the Peak Cladding Temperature (PCT) has been addressed as the key Figure of Merit (FOM). Variables related to the main heat transfer mechanisms have been selected as input parameters for the uncertainty quantification, whereas heat source and heat sink, namely decay power and external air temperature, have been dealt with in a separate sensitivity analysis. The results show that the selected parameters have a weak influence on the PCT, whereas it is strongly related to the decay power and external air temperature values. In any case, PCT stays below the regulatory threshold even under the considered off-normal conditions.

Since the accident at Fukushima, one major goal of reactor safety research has been the development of accident tolerant technologies that can mitigate or delay fuel degradation during a beyond-design-basis accident. One major effort has been focused on the development of coatings for light water reactor fuel cladding. Chromium-coated zirconium-alloy clad is one of the leading options. In this work, the MELCOR systems code (version 1.8.6 User-Defined Generalized Coating) is used to evaluate the performance of Cr-coated Zr-alloy clad as compared to Zr-alloy clad and APMT FeCrAl-coated Zr-alloy clad for a pressurized water reactor (i.e., Surry) for a station blackout (SBO) accident scenario. Our focus is primarily on the accident progression behavior depending on oxidation kinetics and the assumed failure criterion for the coated cladding material. Our simulation and comparison indicate that the presence of the coating material can significantly reduce the initial rate of hydrogen generation and delay the time when hydrogen generation becomes significant. This decrease in the rate of oxidation and delay in timing can provide additional coping time for compensatory operator actions. We also note that the effect of extended auxiliary feedwater system operation (long-term SBO) can increase this additional coping time in combination with Cr-coated Zr-alloy, but it is limited by other primary system failures (e.g., hot-leg creep rupture) that will occur driven by core decay heat and independent of coated cladding effects. Finally, we observe that while the initial suppression of hydrogen generation for Cr-coated Zr-alloy clad compared to Zr-alloy is notable, the overall amount of hydrogen produced is similar since hydrogen can also be produced through competing oxidation of stainless steel components during the accident progression. Our future work is focused on the uncertainty analysis of the oxidation rate data, coating failure criteria, and severe accident modeling limitations in order to better quantify accident tolerant fuel clad benefits. [ABSTRACT FROM AUTHOR]

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