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Title: Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

Abstract

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unitmore » 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.« less

Authors:
;  [1]
  1. Nuclear and Biophysics Research Group, Dept. of Physics, Bandung Institute of Technology, Jl.Ganesha 10, Bandung, 40132 (Indonesia)
Publication Date:
OSTI Identifier:
22004173
Resource Type:
Journal Article
Resource Relation:
Journal Name: AIP Conference Proceedings; Journal Volume: 1448; Journal Issue: 1; Conference: ICANSE 2011: 3. international conference on advances in nuclear science and engineering, Bali (Indonesia), 14-17 Nov 2011; Other Information: (c) 2012 American Institute of Physics; Country of input: International Atomic Energy Agency (IAEA)
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; 22 GENERAL STUDIES OF NUCLEAR REACTORS; COMPUTERIZED SIMULATION; EARTHQUAKES; FUKUSHIMA-1 REACTOR; FUKUSHIMA-2 REACTOR; FUKUSHIMA-3 REACTOR; FUKUSHIMA-4 REACTOR; HIGH PRESSURE COOLANT INJECTION; ISOLATION CONDENSERS; LOSS OF COOLANT; MELTDOWN; NUCLEAR POWER PLANTS; RCIC SYSTEMS; REACTOR CORES; REACTOR FUELING; RELIEF VALVES; SHUTDOWN; STEADY-STATE CONDITIONS; TEMPERATURE DISTRIBUTION

Citation Formats

Su'ud, Zaki, and Anshari, Rio. Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident. United States: N. p., 2012. Web. doi:10.1063/1.4725470.
Su'ud, Zaki, & Anshari, Rio. Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident. United States. doi:10.1063/1.4725470.
Su'ud, Zaki, and Anshari, Rio. Wed . "Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident". United States. doi:10.1063/1.4725470.
@article{osti_22004173,
title = {Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident},
author = {Su'ud, Zaki and Anshari, Rio},
abstractNote = {Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.},
doi = {10.1063/1.4725470},
journal = {AIP Conference Proceedings},
number = 1,
volume = 1448,
place = {United States},
year = {Wed Jun 06 00:00:00 EDT 2012},
month = {Wed Jun 06 00:00:00 EDT 2012}
}
  • A loss-of-coolant accident in a pressurized water, nuclear power plant is one which permits coolant to escape from the primary system. If such an accident were allowed to proceed uninhibiied by corrective measures, the core may lose sufficient coolant such as to permit core heatup. In order to design a system to maintain the core cool, it is necessary to evaluate the coolant blowdown process which occurs after rupture and thereby establish the pressure- time and volume-time relationships of the primary coolant after rupture. The coolant blowdown process after rupture is complex because the two-phase expansion of water and steammore » obtains after saturation pressure is attained. The analysis of this process utilizes heat, mass and volume balances of the reactor coolant to establish the thermodynantic state of the reactor coolant at any time after fupture within conservative limits. (auth)« less
  • As part of the AP600 design certification program, a series of component separate effects tests and two integral systems tests of the nuclear steam supply system were performed. These tests were designed to provide data necessary to validate Westinghouse safety analysis codes for AP600 applications. In addition, the tests have provided the opportunity to investigate the thermal-hydraulic phenomena that are expected to be important in AP600 transients. One series of integral systems tests was undertaken on the SPES-2 facility in Italy, a full-height, full-pressure, 1/395th-power and -volume scale simulation of the AP600 nuclear steam supply system and passive safety features.more » A series of thirteen design-basis events were simulated at SPES-2 to obtain data for verification and validation of the computer models used for the safety analysis of the AP600. The modeled initiating events included a series of small-break loss-of-coolant accidents (SBLOCAs), single steam generator tube ruptures, and a main steamline break. The results of the analyses of the SPES-2 test data, performed to investigate the performance of the safety-related systems are reported. These analyses were also designed to demonstrate, through mass and energy inventory calculations, mass and energy balances, and event timing analyses, the applicability of the SPES-2 tests for computer model verification and validation. It is concluded that the key thermal-hydraulic phenomena that characterize the SBLOCA and non-LOCA transients have been successfully simulated in the SPES-2 facility, and the test results can be used to validate the AP600 safety analysis computer codes. The SPES-2 tests demonstrate that the AP600 passive safety-related systems successfully combine to provide a continuous removal of core decay heat. The SPES-2 tests also showed no adverse interactions between the passive safety-related system components or with the nonsafety-related systems.« less
  • A loss-of-coolant-accident (LOCA) analysis has been completed for the Ignalina nuclear power plant (INPP) located in northeastern Lithuania near the borders of Latvia and Belorus. The INPP site has two RBMK-1500 reactors; the RBMK-1500 is a boiling water, graphite-moderated, pressure tube reactor with the capability of producing up to 4800 MW(thermal). Currently, the power level of INPP is limited to 4200 MW(thermal); thus, the analysis results presented in this paper have been obtained for an initial power level of 4200 MW(thermal).
  • A digital computer analysis of the loss-of-coolant accident in the primary system of a multi-circuit core nuclear power plant in the event of a complete severance of a pressure or jumper tube is presented. The time-dependent mass, momentum, and energy balance differential equations are expressed in finite difference form and solved numerically on an IBM-7090 digital computer together with the equations of state, system boundary conditions, and constraints. The system mass flow rate, pressure, and enthalpy distribution are calculated together with the other important system properties as functions of time during the transient operation following the break. The application ofmore » the analysis to the Carolinas-Virginia Tube Reactor indicates that the loss-of-coolant accident could lead to flow starvation in the reactor core and steam formation in the primary pump with subsequent core damage if no corrective action were taken. The flow starvation and steam formation problems are solved by the operation of a high pressure, high capacity emergency injection pump with fast starting characteristics. (auth)« less