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Thermo-fluid analysis of water cooled research reactors in natural convection; Analise termofluidodinamica de reatores nucleares de pesquisa refrigerados a agua em regime de conveccao natural

Abstract

The STHIRP-1 computer program, which fundamentals are described in this work, uses the principles of the subchannels analysis and has the capacity to simulate, under steady state and transient conditions, the thermal and hydraulic phenomena which occur inside the core of a water-refrigerated research reactor under a natural convection regime. The models and empirical correlations necessary to describe the flow phenomena which can not be described by theoretical relations were selected according to the characteristics of the reactor operation. Although the primary objective is the calculation of research reactors, the formulation used to describe the fluid flow and the thermal conduction in the heater elements is sufficiently generalized to extend the use of the program for applications in power reactors and other thermal systems with the same features represented by the program formulations. To demonstrate the analytical capacity of STHIRP-l, there were made comparisons between the results calculated and measured in the research reactor TRIGA IPR-R1 of CDTN/CNEN. The comparisons indicate that the program reproduces the experimental data with good precision. Nevertheless, in the future there must be used more consistent experimental data to corroborate the validation of the program. (author)
Publication Date:
Jul 01, 2004
Product Type:
Thesis/Dissertation
Report Number:
INIS-BR-8749
Resource Relation:
Other Information: Tese (Ph.D.)
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 97 MATHEMATICAL METHODS AND COMPUTING; COMPARATIVE EVALUATIONS; EXPERIMENTAL DATA; FLUID FLOW; NATURAL CONVECTION; POWER REACTORS; REACTOR CHANNELS; RESEARCH AND TEST REACTORS; S CODES; STEADY-STATE CONDITIONS; THEORETICAL DATA; THERMAL CONDUCTION; THERMAL HYDRAULICS; TRIGA-BRAZIL REACTOR; WATER COOLED REACTORS; COMPUTER CODES; CONVECTION; DATA; ENERGY TRANSFER; ENRICHED URANIUM REACTORS; EVALUATION; FLUID MECHANICS; HEAT TRANSFER; HOMOGENEOUS REACTORS; HYDRAULICS; HYDRIDE MODERATED REACTORS; INFORMATION; IRRADIATION REACTORS; ISOTOPE PRODUCTION REACTORS; MASS TRANSFER; MECHANICS; NUMERICAL DATA; REACTOR COMPONENTS; REACTORS; SOLID HOMOGENEOUS REACTORS; THERMAL REACTORS; TRIGA TYPE REACTORS; WATER MODERATED REACTORS
OSTI ID:
21440491
Research Organizations:
Universidade Estadual de Campinas (FEQ/UNICAMP), SP (Brazil). Fac. de Engenharia Quimica. Sistemas de Processos Quimicos e Informatica
Country of Origin:
Brazil
Language:
Portuguese
Other Identifying Numbers:
TRN: BR11V1354036260
Availability:
Available from INIS in electronic form
Submitting Site:
BRN
Size:
250 p. pages
Announcement Date:
Jun 15, 2011

Citation Formats

Veloso, Maria Auxiliadora Fortini. Thermo-fluid analysis of water cooled research reactors in natural convection; Analise termofluidodinamica de reatores nucleares de pesquisa refrigerados a agua em regime de conveccao natural. Brazil: N. p., 2004. Web.
Veloso, Maria Auxiliadora Fortini. Thermo-fluid analysis of water cooled research reactors in natural convection; Analise termofluidodinamica de reatores nucleares de pesquisa refrigerados a agua em regime de conveccao natural. Brazil.
Veloso, Maria Auxiliadora Fortini. 2004. "Thermo-fluid analysis of water cooled research reactors in natural convection; Analise termofluidodinamica de reatores nucleares de pesquisa refrigerados a agua em regime de conveccao natural." Brazil.
@misc{etde_21440491,
title = {Thermo-fluid analysis of water cooled research reactors in natural convection; Analise termofluidodinamica de reatores nucleares de pesquisa refrigerados a agua em regime de conveccao natural}
author = {Veloso, Maria Auxiliadora Fortini}
abstractNote = {The STHIRP-1 computer program, which fundamentals are described in this work, uses the principles of the subchannels analysis and has the capacity to simulate, under steady state and transient conditions, the thermal and hydraulic phenomena which occur inside the core of a water-refrigerated research reactor under a natural convection regime. The models and empirical correlations necessary to describe the flow phenomena which can not be described by theoretical relations were selected according to the characteristics of the reactor operation. Although the primary objective is the calculation of research reactors, the formulation used to describe the fluid flow and the thermal conduction in the heater elements is sufficiently generalized to extend the use of the program for applications in power reactors and other thermal systems with the same features represented by the program formulations. To demonstrate the analytical capacity of STHIRP-l, there were made comparisons between the results calculated and measured in the research reactor TRIGA IPR-R1 of CDTN/CNEN. The comparisons indicate that the program reproduces the experimental data with good precision. Nevertheless, in the future there must be used more consistent experimental data to corroborate the validation of the program. (author)}
place = {Brazil}
year = {2004}
month = {Jul}
}