Abstract
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.
Highlights
Probabilistic Safety Assessment (PSA) is a key part of a Nuclear Power Plant (NPP) licensing process [1]
The input model of the plant is a node representation of the core, reactor pressure vessel (RPV), Steam Generators (SG), the reactor coolant pumps (RCPs), and the PZR that are enclosed in a steel containment, which is unrounded by a water pool used as shielding and ultimate heat sink, comprising about 72 control volumes (CV) and 87 flow paths (FL)
The following remarks were obtained from these simulation results: a) The core level begins to drop because the steam flow rate into the SG #2 is decreasing – caused by the feedwater pumps tripped
Summary
PSA is a key part of a NPP licensing process [1]. It considers the elaboration and updating of probabilistic models that estimate the risk associated with the operation, allowing the risk monitoring from the plant design to the decommissioning, for both operational as regulatory matters. After modeling the phenomena that occur due to the sequence of events indicated in PSA Level 1, the PSA Level 2 analyzes the behavior of the containment, evaluates the release of radionuclides, providing frequency estimation and release of radioactive material (i.e., the LERF). Based on the results of the PSA Level 1 of a reference plant (in the design phase), a study was started with the aim of analyzing the most impacting events to the CDF After a brief description of the MELCOR code, this paper presents the reference plant and a summary of the model developed to simulate the evolution of severe accidents in this plant. The results of the LOFW simulation are presented, including a description of the sequence of events and a discussion of the concentration of hydrogen in the plant containment
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