Severe Accident Analysis Research Articles

Chemical interaction of CsOH vapor with UO2 and Fe-Zr melt

ABSTRACT At TEPCO’s Fukushima Daiichi Nuclear Power Station, it is estimated that considerable amounts of cesium still remain in the reactors from the analysis results using the severe accident analysis codes and the reverse analysis from contaminated water. Since cesium is known to form stable compounds with uranium and zirconium, chemisorption experiments with uranium dioxide pellets and iron-zirconium melts for cesium hydroxide vapor were carried out using the similar method as for stainless steel. As the results, formations of cesium uranate, Cs2UO4, and cesium zirconate, Cs2ZrO3, were confirmed, indicating that cesium was chemisorbed on both of the uranium dioxide pellets and the iron-zirconium melts in an Ar-H2-H2O flow and an Ar-H2 flow, respectively. Therefore, it was considered that cesium released from fuel might be trapped by chemisorption with fuels and/or iron-zirconium melts during light water reactor severe accidents.

Analysis of Double-Ended Guillotine Break Accident of Surge Line in ACP100 Based on Coupling Method: Development and Validation of the Coupling Method

The ACP100 is a small modular reactor (SMR) designed and built in China, featuring an integrated primary loop and passive safety systems. In SMRs, the interaction between the reactor coolant system, containment vessel, and other systems is highly interdependent. Therefore, it is necessary to use high-precision analysis codes, which can be achieved by coupling multiple codes. This paper is the first part of a study which investigates the LOCA in the ACP100 using a coupled platform. In this paper, a coupling platform was developed for transient analysis of SMR accidents, based on the system code NUSOL-SYS and the Integrated Severe Accident Analysis (ISAA) code. An inter-process communication module was developed adopting shared memory and event objects to exchange data between the two codes. A coupling interface defining the data to be exchanged was proposed. Both NUSOL-SYS and ISAA were modified to perform synchronous time step control. The coupling platform were validated through a hypothetical scenario and Edwards’ pipe blowdown experiment, demonstrating exact consistency and high accuracy. This coupling platform offers a new method for SMR accident analysis, providing a foundation for future work.

Numerical investigation of SUPERFACT-1 experiment by a MOX fuel pin thermo-mechanical model

Preparations are currently underway to develop ISAA-Na code, a sodium-cooled fast reactors (SFRs) specific version of an integrated severe accident analysis code. This version aims to evaluate both protected and unprotected accidents in SFRs. Given the crucial role of fuel pin thermo-mechanical behavior in SFR safety analysis, an advanced model for MOX(Mixed OXide) fuel pin thermo-mechanical performance under operation conditions has been developed and validated. This validation involved simulating the SUPERFACT-1 fast reactor irradiation experiment using ISAA-Na. The resulting calculation data were compared with experimental measurements and the outcomes of other codes dedicated to fuel pin performance analysis (TRANSURANUS, MACROS, andGERMINAL) to assess the fuel pin thermo-mechanical model. The overall code assessment reveals that the model achieves satisfactory performance in simulating fuel pin behavior, with predicted errors falling within an acceptable range.

Study on the lower head failure in severe accidents Part II: Thermal-mechanical coupling behavior analysis on CAP1400 reactor lower head

Under severe accident conditions with core meltdown, large amounts of molten material relocate to the lower head at high temperatures. Significant thermal and mechanical loads may lead to failure of the reactor pressure vessel (RPV) lower head, spreading radioactive fission products. Lower head failure time and location are crucial for accident mitigation and consequence assessment. This study developed a thermodynamic failure analysis model for the lower head by investigating material failure mechanisms, coupling this into an Integrated Severe Accident Analysis code (ISAA), and conducting experimental validation and applied research. Part I mainly introduces the thermodynamic analysis module development, validation against OLHF experiments, and preliminary failure criteria analysis. Results indicate the model accurately predicts lower head failure behavior, and Larson-Miller life fraction and Hedl-Dorn parameter criteria accurately assess lower head integrity. Part II presents applied research after validation and preliminary analysis. Using the improved ISAA, melt pool behavior and lower head thermal–mechanical response were analyzed for a CAP1400 small break loss-of-coolant accident (SBLOCA), assuming multiple safety system failures. Results indicate the bottom of the lower head melt pool didn’t form a hard shell due to high heat release rate, only an oxide shell layer around the second layer. Decay heat from melt pool components cannot be dissipated, leading to lower head failure. Further External Reactor Vessel Cooling (ERVC) heat transfer model analysis and research are needed in ISAA. The two developed mechanical analysis models predict material failure at high temperatures, and the timing and location of lower head failure are consistent, at approximately 16,000 s at 25.38°±5.47° on the lower head. The predicted failure mode is similar to Part I validation, indicating both Larson-Miller life fraction and Hedl-Dorn parameter criteria provide consistent lower head integrity judgments.

Severe accident analysis induced by secondary pipeline break in a small modular PWR

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.

Validation of EPG/SAG Rev. 4 of the Kuosheng Nuclear Power Plant Using the MAAP5 Code

The Boiling Water Reactor Owner’s Group released Emergency Procedure Guidelines and Severe Accident Guidelines Revision 4 (EPG/SAG Rev. 4) in 2018. The major improvement to EPG/SAG Rev. 4 was Contingency 1 (Alternate Level/ Pressure Control). Contingency 1 coordinates the reactor pressure vessel (RPV) water level and RPV pressure control action to prolong the availability of steam-driven injections and optimize the transfer to motor-driven systems. In this study, the effectiveness of the EPG/SAG Rev. 4 Contingency 1 strategy was compared with those of EPG/SAG Revision 2 Contingency 1 and Specific Major Incident Guidelines (SMI) using the Modular Accident Analysis Program, Version 5 (MAAP5). SMI was developed by the Taiwan Power Company to mitigate a Fukushima-like accident. The surrogate plant that analyzed is the Kuosheng Nuclear Power Plant (NPP). The Kuosheng NPP is BWR-6 Mark-III containment. MAAP5 is an integral severe accident analysis program that simulates the responses of a light water reactor power plant during a severe accident. This program has been used extensively for probabilistic risk assessments and for verification and validation of mitigation actions specified in severe accident management guidelines. The simulation scenarios were extended loss of alternative-current power and loss of ultimate heat sink. The low-capacity, motor-driven portable pump was the only available system for RPV injection in the first hour of the accident. In this time period, the RPV water level and pressure were controlled by reactor core isolation cooling and safety relief valves. After this study, the strategy of EPG/SAG Rev. 4 Contingency 1 was successfully validated, and the effectiveness of minimum pre-depressurization RPV water level and the low-capacity, motor-driven portable pump were also demonstrated in this study.

New RMC Energy Deposition Treatment and its Application in Multi-Physics Simulation

Accurate estimation of energy deposition is important in core physics and severe accident analyses for design optimizations. In this study, a new energy deposition treatment is implemented in the Reactor Monte Carlo (RMC) code, offering multiple modes with varying levels of fidelity and computational requirements. The most precise mode is utilized in coupling simulations between RMC and the subchannel thermal-hydraulic analysis code SUBCHAN, incorporating an explicit moderator heating fraction in the coupling interface. The new treatment is verified against references from MCNP, Serpent, and OpenMC for three light water reactor (LWR) assembly cases, and great agreement is achieved. Energy deposition in different materials and components is emphasized in Kilowatt Reactor Using Stirling TechnologY (KRUSTY) modeling, and the results obtained using different modes are compared. The RMC-SUBCHAN coupling calculations for the three LWR assembly cases, employing the most accurate model, reveal a maximum increase of 94.6 K in the control rod centerline temperature, with a normalized energy deposition of 35.9% in the control rod regions. In the assembly case with gadolinium (Gd) burnable poison, a temperature increase of 7.3 K is observed in the Gd rod centerline, while the coolant outlet temperature decreases by 1.6 K due to the reduced explicit moderator heating fraction of 2.1%, compared to the constant 2.6% in the previous coupling scheme.

Technical research on rapid source term calculation and prediction technology of nuclear accidents based on Bayesian network

ABSTRACT How to evaluate and predict potential source terms quickly and effectively is very important for emergency response. In this paper, the traditional forward analysis calculation based on deterministic method was combined with probability theory calculation, and a large number of typical accident sequences were calculated with severe accident analysis software to obtain data of source terms. According to expert knowledge base, probabilistic safety analysis, and other information, a model of nuclear power plant was established based on Bayesian network algorithm, so as to carry out researches on rapid source term calculation and prediction methods. Based on the Bayesian network model of a specific power plant and the pre-calculated source term results of typical accident scenarios, the best available knowledge and observation results of events were used to estimate the relevant typical source terms and their possibilities, which provide a strong basis for effectively grasping and predicting the accident process, quickly predicting source terms and release characteristics of radioactive materials, and quickly and accurately evaluating the radioactive consequences.

Performance evaluation of an improved pool scrubbing system for thermally-induced steam generator tube rupture accident in OPR1000

An improved mitigation system for thermally-induced steam generator tube rupture accidents was introduced to prevent direct environmental release of fission products bypassing the containment in the OPR1000. This involves injecting bypassed steam into the containment, cooling, and decontaminating it using a water coolant tank. To evaluate its performance, a severe accident analysis was performed using the MELCOR 2.2 code for OPR1000. Simulation results show that the proposed system sufficiently prevented the release of radioactive nuclides (RNs) into the environment via containment injection. The pool scrubbing system effectively decontaminated the injected RN and consequently reduced the aerosol mass in the containment atmosphere. However, the decay heat of the collected RNs causes re-vaporization. To restrict the re-vaporization, an external water source was considered, where the decontamination performance was significantly improved, and the RNs were effectively isolated. However, due to the continuous evaporation of the feed water caused by decay heat, a substantial amount of steam is released into the containment. Despite the slight pressurization inside the containment by the injected and evaporated steam, the steam decreased the hydrogen mole fraction, thereby reducing the possibility of ignition.

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.

Investigation of aerosol transportation phenomena in pool scrubbing by combining bubble plume measurements and single-bubble mass transfer analysis

The fission product filtration capability of pool scrubbing is an uncertain factor in severe accident risk analysis. Limited discussions on aerosol transportation phenomena exist due to insufficient understanding of aerosol mass transfer from single bubbles and bubble plume behavior. This study aims to investigate aerosol transportation during pool scrubbing by integrating individual bubble mass transfer into the overall bubble plume.Decontamination factor (DF) measurements were conducted for various nozzle submergences, followed by analysis of bubble diameter distributions in the corresponding bubble plume. Findings revealed that the MELCOR model overestimates bubble break-up, contrary to the expected positive impact on DF. To address this discrepancy, a computational fluid dynamics simulation was employed to analyze mass transfer coefficients considering different bubble diameters and aerosol properties. By integrating the mass transfer of single bubbles, the discrepancy in the DF tendencies was attributed to the underestimation of aerosol mass transfer coefficients in individual bubbles.

Evaluation of uncertainties associated with best estimates of parameters in the deterministic analysis of a severe accident

This paper describes an approach to evaluate the uncertainties of the results of deterministic severe accident analysis. The approach was developed in order to follow best practices of best estimate methodology and to fulfill the Russian regulatory requirements NP-001–15 in the area of atomic energy use. In accordance with these requirements, deterministic safety analyses must be accompanied with the evaluation of errors and uncertainties of the obtained results. The application of the proposed approach is demonstrated through the example of a deterministic analysis of a severe LOCA initiated by a large break at a generic nuclear power plant with VVER-1000. Within the given example a best estimate value and associated uncertainty of hydrogen mass generated at in-vessel stage is analyzed.

Main outcomes of the Phebus FPT1 uncertainty and sensitivity analysis in the EU-MUSA project

The Management and Uncertainties of Severe Accidents (MUSA) project was funded in HORIZON 2020 and is coordinated by CIEMAT (Spain). The project aims at consolidating a harmonized approach for the analysis of uncertainties and sensitivities associated with Severe Accidents (SAs) analysis, focusing on source term figures of merit. The Application of Uncertainty Quantification (UQ) Methods against Integral Experiments (AUQMIE – Work Package 4 (WP4)), led by ENEA (Italy), was devoted to apply and test UQ methodologies adopting the internationally recognized PHEBUS FPT1 test. FPT1 was chosen to test UQ methodologies because, even though it is a simplified SA scenario, it was representative of the in-vessel phase of a severe accident initiated by a break in the cold leg of a PWR primary circuit.WP4 served as a platform to identify and discuss the issues encountered in the application of UQ methodologies to SA analyses (e.g. discuss the UQ methodology, perform the coupling between the SA codes and the UQ tools, define the results post-processing methods, etc.). The purpose of this paper is to describe the MUSA PHEBUS FPT1 uncertainty application exercise with the related specifications and the methodologies used by the partners to perform the UQ exercise. The main outcomes and lessons learned of the analysis are: scripting was in general needed for the SA code and uncertainty tool coupling and to have more flexibility; particular attention should be devoted to the proper choice of the input uncertain parameters; outlier values of figures of merit should be carefully analyzed; the computational time is a key element to perform UQ in SA; the large number of uncertain input parameters may complicate the interpretation of correlation or sensitivity analysis; there is the need for a statistically solid handling of failed calculations.

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.

Research and Application of Steam Condensation Heat Transfer Model Containing Noncondensable Gas on a Wall Surface

Steam condensation plays an important role in various engineering processes due to its excellent heat transfer performance. However, condensation in the presence of noncondensable gas has attracted great attention in recent years since noncondensable gas will have a negative effect on condensation heat transfer. The present study proposes a comprehensive model coupled with convective heat transfer, liquid film heat transfer and steam condensation for the heat transfer of condensation with noncondensable gas and uses it in the Program Integrated for Severe Accident Analysis (PISAA) for a nuclear power plant. The condensation heat transfer model has good universality, the calculation process is stable with less iteration and a fast convergence and it is verified and validated by comparing the simulation results of the PISAA and those from traditional containment analysis codes, as well the experiments from the Wisconsin condensation tests; then, a sensitivity analysis for the parameters of the heat transfer coefficient is performed. The validation results show that the average error of the condensation heat transfer coefficient is approximately 10%, and the maximum error does not exceed 30%. The deviation from the experimental data is limited in the acceptable range, which could fulfill the requirement for the analysis of containment accidents in nuclear power plants.

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.

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

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

KIT reactor safety research for LWRs: Research lines, numerical tools, and prospects

This paper describes the main research lines of the Karlsruhe Institute of Technology (KIT) relevant to Light Water Reactors (LWR) that permit both the safety evaluation of different reactor designs and the prediction of the radiological source term including its impact in the case of hypothetical severe accidents. The numerical tools for core analysis, nuclear power plant analysis of design basis and severe accidents, and the estimation of the radioactive dispersions of fission products after a core meltdown accident are described and selected results are discussed. The analytical investigations at KIT are complemented by experimental investigations at two facilities, namely the COSMOS-H/-L and the QUENCH facilities devoted to safety-relevant Thermal-Hydraulic (TH) phenomena and in-vessel severe accident early-phase for different kinds of fuel rods including Accident Tolerant Fuel (ATF). These data are crucial for the validation of the numerical tools used for safety demonstration. The paper is complemented by a brief description of the in-house tools under development for improved core/plant analysis based on the multi-physics/-scale coupling methodologies, where the in-house tools are combined with external system TH, Monte Carlo, and thermo-mechanic code to increase the prediction accuracy of the core/plant behaviour under stationary and accidental conditions. Selected results are presented and discussed. It can be stated that the long-lasting safety research activities at KIT devoted to LWR led to the development of a computation route combining external with in-house codes that is very much appropriate for safety evaluations and risk assessment; it includes unique multi-physics/-scale coupled tools for improved core and plant analysis of different reactor designs, e.g. PWR, VVER, BWR, SMR, and research reactors.

Long-term temporal analysis of fatal and severe occupational accidents in Central European forests of the Slovak Republic

Introduction: Forestry is one of the most dangerous industries worldwide. Workers handle heavy loads and are exposed to numerous adverse factors of the work environment. This study was focused on analyzing severe and fatal occupational accidents over a long period. Methods: Data were gathered from occupational safety and health databases of the State Forests of the Slovak Republic, which manages 879,113 ha of forests in Slovakia. Between 2007 and 2021, 210 severe and64 fatal accidents were recorded. The data analyses included contingency tables, Pearson χ2 test, and Cramer’s V coefficient to analyze the relationships between studied variables. Results: The mean annual number of fatalities was 14, occurring to workers aged 42 on average. Almost 80% of all fatal accidents were suffered by contractors of harvesting operations, who were 5.23 times more likely to suffer a fatality than company employees. On average, 1.28 severe and0.75 fatal accidents occurred per 1 million m3 of harvested timber. The outcomes of the correlation analysis showed a decreasing trend in incidence in the case of severe accidents. Workers ≤ 30 years or between 51 and 60 years of age were the most vulnerable in the case of fatal accidents. In the case of severe accidents, workers between 51 and 60 were the most vulnerable group. Most fatal accidents were suffered by tree fellers (72.2%) during felling (69.8%), who were in contact with falling trees and tree parts (93.7% share for this material agent). Conclusion and Practical Applications: The presented study identified the most vulnerable worker groups as well as the effects of factors on the occurrence of severe and fatal occupational accidents in forestry. The results will serve in development of occupational safety and health strategies in forest enterprises.

The Effect of Fuel Aerosol Particle Size on Aerosol Release Potential in Energetic Core Disruptive Accident Studies

In nuclear reactor accident safety studies, the radiological source term is a metric that quantifies the release of radiological material from the reactor to the environment. The present work evaluates heat transfer between high-temperature vapor bubbles and the surrounding coolant and the effect these interactions have on the source term for postulated core disruptive accident scenarios associated with an oxide-fueled, liquid metal–cooled fast reactor class. It is shown that aerosol particle size can influence heat transfer, and it is suggested that the extent of the influence depends on the fineness of the particles in the aerosol. The results are consistent with legacy experiments conducted in the Fuel Aerosol Simulant Test (FAST) facility at Oak Ridge National Laboratory and offer a more comprehensive assessment of vapor condensation by treating the bubble constituents, in the context of radiation heat transfer, as participating media. The model, which couples classical scattering theory to the equation of radiative transfer and the energy equation, provides a means for estimating size-affected radiative cooling times. Solutions are obtained via the P-1 method of spherical harmonics with improved, higher-order boundary conditions. Outcomes include the development of an “extinction-time ratio” criterion for assessing whether ejection of aerosol from the bubble to the cover region is likely. Aerosol release from the coolant pool is evaluated using this criterion with the potential to extend this work to reactor-scale accidents. A baseline evaluation is provided that shows that omission of participatory effects could lead, in a relative sense, to cooling time offsets in excess of 14%. In addition to enhancing previous evaluations of FAST results, these modeling outcomes contribute to knowledge management efforts aimed at developing a more mechanistic assessment of the source term while suggesting potential enhancements to severe accident safety analysis through the use of more comprehensive radiative heat transfer models.