skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: Simulation of reflooding on two parallel heated channel by TRACE

Abstract

In case of Loss-Of-Coolant accident (LOCA) in a Boiling Water Reactor (BWR), heat generated in the nuclear fuel is not adequately removed because of the decrease of the coolant mass flow rate in the reactor core. This fact leads to an increase of the fuel temperature that can cause damage to the core and leakage of the radioactive fission products. In order to reflood the core and to discontinue the increase of temperature, an Emergency Core Cooling System (ECCS) delivers water under this kind of conditions. This study is an investigation of how the power distribution between two channels can affect the process of reflooding when the emergency water is injected from the top of the channels. The peak cladding temperature (PCT) on LOCA transient for different axial level is determined as well. A thermal-hydraulic system code TRACE has been used. A TRACE model of the two heated channels has been developed, and three hypothetical cases with different power distributions have been studied. Later, a comparison between a simulated and experimental data has been shown as well.

Authors:
 [1]
  1. Department of Nuclear Engineering, Chalmers University of Technology, Gothenburg (Sweden)
Publication Date:
OSTI Identifier:
22608564
Resource Type:
Journal Article
Resource Relation:
Journal Name: AIP Conference Proceedings; Journal Volume: 1754; Journal Issue: 1; Conference: ICME 2015: 11. international conference on mechanical engineering, Dhaka (Bangladesh), 18-20 Dec 2015; Other Information: (c) 2016 Author(s); Country of input: International Atomic Energy Agency (IAEA)
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; BWR TYPE REACTORS; COMPARATIVE EVALUATIONS; COOLANTS; ECCS; FISSION; FISSION PRODUCTS; FLOW RATE; HEAT; LOSS OF COOLANT; LOSSES; MASS; POWER DISTRIBUTION; REACTOR CORES; SIMULATION; THERMAL HYDRAULICS; TRANSIENTS; WATER

Citation Formats

Zakir, Md. Ghulam. Simulation of reflooding on two parallel heated channel by TRACE. United States: N. p., 2016. Web. doi:10.1063/1.4958436.
Zakir, Md. Ghulam. Simulation of reflooding on two parallel heated channel by TRACE. United States. doi:10.1063/1.4958436.
Zakir, Md. Ghulam. 2016. "Simulation of reflooding on two parallel heated channel by TRACE". United States. doi:10.1063/1.4958436.
@article{osti_22608564,
title = {Simulation of reflooding on two parallel heated channel by TRACE},
author = {Zakir, Md. Ghulam},
abstractNote = {In case of Loss-Of-Coolant accident (LOCA) in a Boiling Water Reactor (BWR), heat generated in the nuclear fuel is not adequately removed because of the decrease of the coolant mass flow rate in the reactor core. This fact leads to an increase of the fuel temperature that can cause damage to the core and leakage of the radioactive fission products. In order to reflood the core and to discontinue the increase of temperature, an Emergency Core Cooling System (ECCS) delivers water under this kind of conditions. This study is an investigation of how the power distribution between two channels can affect the process of reflooding when the emergency water is injected from the top of the channels. The peak cladding temperature (PCT) on LOCA transient for different axial level is determined as well. A thermal-hydraulic system code TRACE has been used. A TRACE model of the two heated channels has been developed, and three hypothetical cases with different power distributions have been studied. Later, a comparison between a simulated and experimental data has been shown as well.},
doi = {10.1063/1.4958436},
journal = {AIP Conference Proceedings},
number = 1,
volume = 1754,
place = {United States},
year = 2016,
month = 7
}
  • To evaluate the applicability of the reflood analysis code REFLA for ordinal pressurized water reactors to the analysis of reflooding phenomena in light water high conversion reactors (LWHCRs) with tight-lattice cores, a numerical simulation of the NEPTUN LWHCR test was performed with the REFLA code. The NEPTUN LWHCR test was performed at the Swiss Federal Insitute for Reactor Research with a test section simulating the tight-lattice core of an LWHCR. The results indicate no potential problems in the use of REFLA for the simulation of reflooding behavior in tight-lattice rod bundles. To improve the code, however, it is recommended tomore » modify models of core heat transfer at a high flooding rate and core water distribution (integration of droplet flow) in the axial direction, and to investigate core pressure drop and horizontal cross flow.« less
  • Two analytical methods are described for the direct determination of beryllium in petroleum and petroleum products by heated vaporization atomic absorption (HVAA). The methods are applicable to the determination of 1 to 50 ng Be/g with a precision (relative standard deviation) of 10 percent at the 30 to 40 ng/g level. The methods were cross-checked in cooperating laboratories and the results indicate that reliable analyses can be obtained when the methods are applied in other laboratories. (auth)
  • An engineering model was developed to simulate the thermal-hydraulic phenomena in pressurized water reactor cores during bottom reflooding. The model couples the fluid thermal hydraulics and radial heat transfer in the fuel rods. The system dynamics were formulated in terms of a set of ordinary differential equations, which were integrated using the Gear integration package. A dynamic nodal scheme, which moves with the quench-front location, was utilized to predict the fuel rod temperatures. Model predictions and comparisons with full-scale experiments are provided, and show good agreement with the FLECHT-SEASET and Slab Core Test Facility data. The proposed methodology was foundmore » to be computationally fast when compared with previous approaches, and can be readily integrated with other modules to simulate the complete reactor coolant system.« less
  • A model was developed to simulate thermal-hydraulic phenomena in the downcomer and lower plenum of a pressurized water reactor during reflooding. The system dynamics were formulated in terms of a set of time-dependent ordinary differential equations that were integrated numerically. A model was developed to simulate the oscillatory flow in the downcomer-lower plenum-core system. A numerical procedure was devised for solving the governing global momentum equation. This procedure is shown to be numerically stable and computationally efficient. The developed model for downcomer and lower plenum was coupled to the core thermal-hydraulic model, and predictions were made for FLECHT-SET experimental data.more » The results compared well with the experiment.« less