Rodolfo Vaghetto

RoverD: Use of test data in GSI-191 risk assessment

Abstract An approach is described that would use test data to evaluate the risk associated with the concerns raised in Generic Safety Issue 191 (GSI-191). The relationship to the elements of quantitative risk-informed regulation for addressing the concerns raised in GSI-191 in pressurized water reactor (PWR) plant licensing is described. Use of experimental data from a deterministic sump performance test to establish scenario success for tested debris loads is summarized and compared to the licensing requirements in the regulations. Generation and transport of debris to the emergency core cooling system sump from a loss-of-coolant accident is described, and data are shown for a particular PWR. Application of the analysis results to a license amendment for an operating PWR is summarized.

Sensitivity Analysis of a PWR Response During a Loss of Coolant Accident Under a Hypothetical Core Blockage Scenario Using RELAP5-3D

During a Loss of Coolant Accident (LOCA) a substantial amount of debris may be generated in containment during the blowdown phase. This debris can become a major safety concern since it can potentially impact the Emergency Core Cooling System (ECCS). Debris, produced by the LOCA break flow and transported to the sump, could pass through the filtering systems (debris bed and sump strainer) in the long term cooling phase. If the debris were to sufficiently accumulate at the core inlet region, the core flow could theoretically decrease, affecting the core coolability. Under such conditions, the removal of decay heat would only be possible by coolant flow reaching the core through alternative flow paths, such as the core bypass (baffle). There are certain plant specific features that can play a major role in core cooling from this bypass flow. One of these of key interest is the pressure relief holes. A typical 4-loop Pressurized Water Reactor (PWR) was modeled using RELAP5-3D to simulate the reactor system response during the phases of a large break LOCA and the effectiveness of core cooling under full core blockage was analyzed. The simulation results showed that the presence of alternative flow paths may significantly increase core coolability and prevent cladding temperatures from reaching safety limits, while the lack of LOCA holes may lead to a conservative over-prediction of the cladding temperature.Copyright

Experimental Investigation of a Scaled Water-Cooled Reactor Cavity Cooling System

Abstract The Very High Temperature Gas-Cooled Reactor (VHTR) is one of the next-generation nuclear reactors designed to achieve high temperatures to support industrial applications and power generation. Because of the high temperature reached during normal operation, new safety features were added to its design. The reactor cavity cooling system (RCCS) is a passive safety system that will be incorporated in the VTHR. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady state) and accident scenarios. A small-scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the thermal-hydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates. A steady-state experimental run was conducted to study the behavior of the coolant under this condition. The experimental results obtained confirmed the capabilities of the system in removing the heat from the cavity and helped in identifying phenomena that may occur in this type of passive system.

Experimental Study of the Effect of Graphite Dispersion on the Heat Transfer Phenomena in a Reactor Cavity Cooling System

An experimental activity was performed to observe and study the effects of graphite dispersion and deposition on thermal-hydraulic phenomena in a reactor cavity cooling system (RCCS). The small-scale RCCS experimental facility (16.5 × 16.5 × 30.4 cm) used for this activity represents half of the reactor cavity with an electrically heated vessel. Water flowing through five vertical pipes removes the heat produced in the vessel and releases it into the environment by mixing with cold water in a large tank. The particle image velocimetry technique was used to study the velocity field of the air inside the cavity. A set of 52 thermocouples was installed in the facility to monitor the temperature profiles of the vessel, pipe walls, and air. Ten grams of a fine graphite powder (average particle size 2 μm) was injected into the cavity through a spraying nozzle placed at the bottom of the vessel. The temperatures and air velocity field were recorded and compared with the measurements obtained before the graphite dispersion, showing a decrease of the temperature surfaces that was related to an increase in their emissivity. The results contribute to the understanding of RCCS capability in an accident scenario.

Impact of Pressure Relief Holes on Core Coolability for a PWR during a Large-Break Loss-of-Coolant Accident with Core Blockage Using RELAP5-3D

Abstract During a loss-of-coolant accident (LOCA), fibrous debris and other particles generated by the jet impingement may be transported to the sump, accumulate, or even penetrate through the strainers, reaching the reactor core. Pressure relief holes and other plant-specific features may provide alternative paths to the coolant under debris-generated core blockage scenarios and can play a major role in core coolability. A typical four-loop pressurized water reactor was modeled using RELAP5-3D to simulate the reactor system response during large-break LOCA scenarios under hypothetical full core blockage conditions. Pressure relief holes were included in the input model to study the effects of these alternative flow paths on the core coolability. The comparison of the simulation results obtained with two different models (with and without pressure relief holes) proved the effectiveness of these alternative flow paths in providing sufficient flow to the core to remove the decay heat during the long-term cooling phase, maintaining the cladding temperature sufficiently below the safety limits at any time after the core blockage occurred. The results presented in this paper not only confirmed the importance of including specific geometric features of the reactor system (generally neglected) when simulating core blockage scenarios but also provided evidence that even under certain extreme core blockage conditions, core coolability may still be guaranteed.

Sensitivity Analysis of the Response of PWR Containment During a Loss of Coolant Accident Using RELAP5-3D and MELCOR

The prediction of specific parameters for the reactor containment, such as pressure and sump pool temperature, is of paramount importance when studying the thermal-hydraulic phenomena involved in the debris generation, transport, and accumulation during Loss of Coolant Accidents (LOCA). The response of the reactor containment during these events may significantly vary depending of several factors such as break size and location, and other plant-specific features. When modeling the reactor and containment response using systems codes, the predictions may also depend on the selection of physical models, correlations and their coefficients. A sensitivity analysis of the response of a typical Pressurized Water Reactor (PWR) 4-loop reactor system and associated containment during a large break LOCA was conducted using RELAP5-3D and MELCOR to investigate the influence of geometrical parameters (break location), physical models (chiked flow models), and related coefficients (discharge coefficient at the break), on the containment response. The simulation results showed how the containment response changed by varying the selected parameters and confirmed the importance of identifying and studying the factors triggering the containment engineered features (containment sprays) when simulation the containment response.© 2014 ASME

Sensitivity Study of the South Texas Project Power Plant Steady-State Simulations Using RELAP5-3D Coupled With Dakota

A RELAP5-3D input deck of the South Texas Project (STP) power plant was created in order to study the thermal-hydraulic behavior of the plant during normal operation (steady-state) and during a Loss of Coolant Accident (LOCA). It is important to study the sensitivity of selected output parameters such as the total coolant mass flow rate, the peak clad temperature, the secondary pressure, as a function of specific input parameters (reactor nominal power, vessel inlet temperature, steam generators primary side heat transfer coefficient, primary pressure etc.) in order to identify the variables that play a role in the uncertainty of the thermal-hydraulic calculations. RELAP5-3D, one of the most used best estimate thermal-hydraulic system codes, was coupled with DAKOTA, developed by Sandia National Laboratory for Uncertainty Quantification and Sensitivity Analysis in order to simplify the simulation process and the analysis of the results. In the present paper, the results of the sensitivity study for selected output parameters of the steady-state simulations are presented. The coupled software was validated by repeating one set of simulations using the RELAP5-3D standalone version and by analyzing the simulation results with respect of the physical expectations and behavior of the power plant. The thermal-hydraulic parameters of interest for future uncertainty quantification calculations were identified.Copyright