Abstract
After the two most significant nuclear accidents in history – the Chernobyl Reactor Four explosion in Ukraine(1986) and the Fukushima Daiichi accident in Japan (2011) –, the Final Safety Analysis Report (FSAR) included a new chapter (19) dedicated to the Probabilistic Safety Assessment (PSA) and Severe Accident Analysis (SAA), covering accidents with core melting. FSAR is the most important document for licensing of siting, construction, commissioning and operation of a nuclear power plant. In the USA, the elaboration of the FSAR chapter 19 is according to the review and acceptance criteria described in the NUREG-0800 and U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.200. The same approach is being adopted in Brazil by National Nuclear Energy Commission (CNEN). Therefore, the FSAR elaboration requires a detailed knowledge of severe accident phenomena and an analysis of the design vulnerabilities to the severe accidents, as provided in a PSA – e.g., the identification of the initiating events involving significant Core Damage Frequency (CDF) are made in the PSA Level 1. As part of the design and certification activities of a plant of reference, the Laboratory of Risk Analysis, Evaluating and Management (LabRisco), located in the University of São Paulo (USP), Brazil, has been preparing a group of specialists to model the progression of severe accidents in Pressurized Water Reactors (PWR), to support the CNEN regulatory expectation – since Brazilian Nuclear Power Plants (NPP), i.e., Angra 1, 2 and 3, have PWR type, the efforts of the CNEN are concentrated on accidents at this type of reactor. The initial investigation objectives were on completing the detailed input data for a PWR cooling system model using the U.S. NRC MELCOR 2.2 code, and on the study of the reference plant equipment behavior – by comparing this model results and the reference plant normal operation main parameters, as modeled with RELAP5/MOD2 code.
Highlights
Introduction must clearly explain the context of the article
The inner reactor pressure vessel (RPV) is divided into five control volumes that represent the annular descending channel (CV 100) which is connected with the cold leg, the lower plenum (LP) (CV 101), the reactor core (CV 102), the core bypass (CV 103) and the upper head (CV 104) which connects with the hot leg
A steady state calculation using the model for the reference plant reactor cooling system – presented in the previous section (Section 4) – was conducted to verify the success achieved in the use of MELCOR code
Summary
Introduction must clearly explain the context of the article. State the precise objective and hypothesis to be discussed. The main objective of the work presented in this paper was modeling and analyzing the operational parameters such as pressure, temperature, and flow rate, until the “achievement of the steady state of a reference PWR NPP, using the severe accident MELCOR code version 2.2. The results of parameters under normal operating conditions (steady state conditions) are compared with those obtained with RELAP5/MOD2 code [12,13]. This model will be used to simulate different types of severe accident in this reference plant, supporting deterministic and probabilistic analysis, as required to licensing of new plants, as described in NUREG 0800 [14]. This paper presents a short description of the MELCOR code
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