Source term evaluation is an important element in the assessment of efficiency of a Severe Accident Management (SAM) strategy. The identification of phenomena and parameters that present major contributions to the uncertainty in the magnitude and timing of the releases and quantify the uncertainty is vital for comprehensive risk analysis. In this work source term evaluation was performed using MELCOR for two accident scenarios, large break LOCA and station blackout, that leads to containment failure due to ex-vessel phenomena such as formation of non-coolable debris bed and steam explosion. Cases without containment failure were also analyzed for the effect of melt debris release characteristics on fission product release in both accident scenarios. It was observed that initial vessel failure mode was due to failure of the penetrations. When the debris being ejected was limited to only molten mass, instead of solid and molten mass, it also led to creep-rupture of the vessel lowerhead wall. This mode of ejection also shows a remarkable increase in the cesium and iodine source term release to the environment. When the containment does not fail, it is observed that there is lesser accumulation of fission products inside the containment and the retention inside the pressure suppression pool is enhanced in the scenarios with only molten debris ejection.

Nordic Boiling Water Reactors employ filtered containment venting and ex-vessel debris coolability in the deep pool located under the reactor pressure vessel as cornerstones of their severe accident management strategy. This paper focuses on the uncertainty analysis of the source term in accident sequences that result in filtered containment venting to the environment using the MELCOR code. The impact of uncertain MELCOR modeling parameters and modeling options on the timing and magnitude of the source term released to the environment has been evaluated in accident sequences initiated by a large break LOCA and SBO. The performed simulations illustrate the effect of MELCOR modeling parameters and options on the code's predictions of severe accident progression, event timing, and the magnitude of the source term released to the environment in different accident scenarios. Furthermore, the results highlight the importance of various retention mechanisms that limit the release of fission products into the environment.

Boiling Water Reactor (BWR) employs the Pressure Suppression Pool (PSP) as a heat sink to prevent overpressure of the reactor vessel and containment. Steam can be injected into the PSP through spargers in normal and accident conditions and through blowdown pipes in case of a loss of coolant accident (LOCA). There is a safety limit on the maximum PSP temperature at which such steam injection might cause dynamic loads on the containment structures. The performance of the pool can be affected if thermal stratification is developed when temperature of the hot layer grows rapidly while cold layer remains inactive. Simulation of pool behavior during realistic accident scenarios requires validated models that can sufficiently address the interaction between phenomena, safety systems and operational procedures. Direct modeling of steam injection into a water pool in long-term transients is computationally expensive due to the need to resolve simultaneously the smallest space and time scales of individual steam bubbles and the scales of the whole PSP. To enable PSP analysis for practical purposes, Effective Heat source and Effective Momentum source (EHS/EMS) models have been proposed that avoid the need to resolve steam-water interface. This paper aims to implement mechanistic approaches previously developed by authors for the simulation of transient thermal stratification and mixing phenomena induced by steam injection through spargers in a Nordic BWR PSP. The latest version of the EHS/EMS models using the 'Unit cell' approach has been validated against integral effect pool tests and applied to plant simulations. Several scenarios with boundary conditions corresponding to postulated accident sequences were simulated to investigate the possibility of stratification development and the effects of activation of different systems (e.g., blowdown pipes, high momentum nozzle) on the pool behavior.

Severe Accident Management (SAM) strategy aims to minimize exposure of public to potentially harmful radiation and the release of fission products into the environment. For assessment of SAM efficiency it is crucial to identify phenomena and parameters that present major contributions to the uncertainty in the magnitude and timing of the releases, assess their sensitivity, and quantify the uncertainty. In this work source term evaluations are performed, considering 50 uncertain input and modeling parameters that can affect the accident progression, release paths and magnitude of the release. MELCOR was used to perform evaluations for two accident scenarios initiated by: (1) large-break LOCA, and (2) station blackout, that lead to containment failure due to containment bypass or ex-vessel phenomena at RPV melt-through. Preliminary screening was performed using best-estimate and bounding assessments, where parameters were varied one-factor-at-a-time. The fraction of cesium release to the environment was found to be within ~2-35% of core inventory in case of LOCA, while it was within ~2-7% in case of SBO. This difference could be explained by the greater effect of suppression pool scrubbing in case of SBO. Releases obtained with modeling of solid core debris ejection from the vessel were marginally lower than without, for both LOCA and SBO. Dakota was used to perform Morris sensitivity analysis for the most influential parameters in the bounding assessment and the results are presented here.

Steam discharging through spargers and blowdown pipes into the Pressure Suppression Pool (PSP) is employed in Boiling Water Reactor (BWR) to prevent overpressure of the reactor vessel and containment. The capability of suppression can be reduced during the operation when the thermal stratification is developed. Direct modeling of steam injection into a water pool with long-term transient is computationally expensive due to the large-scale difference in space and time. To enable such prediction, Effective Heat source and Effective Momentum source (EHS/EMS) models are proposed. In previous work, we demonstrated the implantation of EHS/EMS models in the Computational Fluid Dynamics (CFD) tool and its application to plant simulation. In this work, we use the developed model to further investigate the thermal stratification and mixing in the PSP of a Nordic BWR. The event to be analyzed is initiated by spurious activation of one valve in the safety injection system. The focus of the simulations is to investigate the possibility of stratification development and understand the effects of the activation of different systems on pool behavior. Pool transient is simulated by CFD code (ANSYS Fluent) with EHS/EMS models and the injection conditions of the steam are derived from the simulation results performed by the system-level codes (GOTHIC).

Passive safety systems in a nuclear reactor allow to simplify the overall plant design, beside improving economics and reliability, which are considered to be among the salient goals of advanced Generation IV reactors. This work focuses on investigating the application of a self-actuated, gravity-driven shutdown system in a small lead-cooled fast reactor and its dynamic response to an initiating event. The reactor thermal-hydraulics and neutronics assessment were performed in advance. According to a first-order approximation approach, the passive insertion of shutdown assembly was assumed to be influenced primarily by three forces: gravitational, buoyancy and fluid drag. A system of kinematic equations were formulated a priori and a MATLAB program was developed to determine the dynamics of the assembly. Identifying the delicate nature of the balance of forces, sensitivity analysis for coolant channel velocities and assembly foot densities yielded an optimal system model that resulted in successful passive shutdown. Transient safety studies, using the multi-point dynamics code BELLA, showed that the gravity-driven system acts remarkably well, even when accounting for a brief delay in self-actuation. Ultimately the reactor is brought to a sub-critical state while respecting technological constraints.