Critical heat flux experiment specifications, test section basic and detailed design, data acquisition analysis and correlations for a small Pressurized Water Reactor (PWR) (up to 50 MWt)

• Grant number: 10/11113-2
• Support Opportunities: Research Projects - Thematic Grants
• Start date: June 01, 2012
• End date: November 30, 2014
• Field of knowledge: Engineering - Nuclear Engineering
• Principal Investigator: José Roberto Castilho Piqueira
• Grantee: José Roberto Castilho Piqueira
• Host Institution: Escola Politécnica (EP). Universidade de São Paulo (USP). São Paulo , SP, Brazil
• Pesquisadores principais: José Jaime da Cruz
• Associated researchers: André da Silva Serra ; Andre Luis Ferreira Marques ; Antonio Luis de Campos Mariani ; Diego Paolo Ferruzzo Correa ; Fuad Kassab Junior ; Gherhardt Ribatski ; Itamar Iliuk ; José Osvaldo de Souza Guimarães ; Juliana Pacheco Duarte ; Luciano Pagano Júnior ; Ricardo Paulino Marques ; Ricardo Sbragio ; Rodney Aparecido Busquim e Silva

Abstract

CTMSP is building a small Pressurized Water Reactor (LABGENE) with Brazilian design and technology. To demonstrate the safety case for this reactor, the occurrence of CHF must be investigated, empirical CHF correlations must be established, and ultimately the Brazilian Nuclear Commission (CNEN) requirements must be fulfilled. The construction of this reactor is underway and a set of CHF experiments, which may take place in Brazil and in the U.S., followed by the establishment of correlations (and validation) for different conditions of operation, are needed. One of the most important requirements in the design of PWRs is to avoid the occurrence of Critical Heat Flux (CHF) (also known as Boiling Crisis or Departure from Nucleate Boiling -- DNB). The design criterion specifies that PWRs should always operate below the CHF to keep the fuel cladding temperature within safety limits. Fuel rods overheating due to the occurrence of a boiling crisis, during a power transient, can yield clad failure with radioactive products leakage to the coolant. To predict this type of phenomenon, it is a common practice to perform simulations of the reactor operational and transient conditions in thermal-hydraulic loops by using electrically heated rods. Such simulations represent an important aspect of the reactor safety analysis. Therefore, CHF is one of important thermal-hydraulic parameter limiting the available power and the plant operation. To be able to license and operate a PWR, the CHF must be extensively investigated and empirical/experimental correlations must be established. Therefore, the University of Sao Paulo, the Massachusetts Institute of Technology and the Technological Navy Center at Sao Paulo are proposing a collaboration to generate new knowledge, improve the nuclear courses and train graduate students in the area of CHF. With this collaboration, the EPUSP, MIT and CTMSP intend to provide a bridge to bring nuclear engineering knowledge in Brazil to an international level. (AU)

Scientific publications

(References retrieved automatically from Web of Science and SciELO through information on FAPESP grants and their corresponding numbers as mentioned in the publications by the authors)

ILIUK, ITAMAR; BALTHAZAR, JOSE MANOEL; TUSSET, ANGELO MARCELO; CASTILHO PIQUEIRA, JOSE ROBERTO. Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code. ADVANCES IN MECHANICAL ENGINEERING, v. 8, n. 1, JAN 2016. (10/11113-2)

LOPES, DENISE A.; ZIMMERMANN, ANGELO J. O.; SILVA, SELMA L.; PIQUEIRA, J. R. C.. Thermal cycling effect in U-10Mo/Zry-4 monolithic nuclear fuel. JOURNAL OF NUCLEAR MATERIALS, v. 473, p. 136-142, MAY 2016. (10/11113-2)