LOCA analysis of BWR-4/Mark-I nuclear power plant with TRACE
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J.-J. Huang
,
S.-W. Chen
Abstract
This study established an RCS-Containment coupled model that integrates the reactor coolant system (RCS) and the containment system by using the TRACE code. The coupled model was used in both short-term and long-term loss of coolant accident (LOCA) analyses. Besides, the RELAP5/CONTAN model that only contains the containment system was also developed for comparison. For short-term analysis, three kinds of LOCA scenarios were investigated: the recirculation line break (RCLB), the main steam line break (MSLB), and the feedwater line break (FWLB). For long-term analysis, the dry-well and suppression pool temperature responses of the RCLB were studied. The analysis results of RELAP5/CONTAN and TRACE models are benchmarked with those of FSAR and RELAP5/GOTHIC models, and it appears that the results of the above four models are consistent in general trends.
Abstract
In dieser Studie wurde ein kombiniertes TRACE-Modell erstellt, das das Reaktorkühlsystem (RCS) und das Containment-System zusammenführt. Das gekoppelte Modell wurde sowohl für kurzfristige als auch für langfristige Kühlmittelverlust-Unfallanalysen (LOCA) verwendet. Daneben wurde zum Vergleich auch ein RELAP5/CONTAN-Modell nur für das Containment entwickelt. Für die Kurzzeitanalyse wurden drei Arten von LOCA-Szenarien untersucht: der Bruch der Rezirkulationsleitung (RCLB), der Frischdampfleitungsbruch (MSLB) und der Speisewasserleitungsbruch (FWLB). Für die Langzeitanalyse wurden die Temperaturrückwirkungen des RCLB auf das Drywell und das Suppressionsbecken untersucht. Die Analyseergebnisse der Modelle RELAP5/CONTAN und TRACE werden mit denen der Modelle FSAR und RELAP5/GOTHIC verglichen, und es zeigt sich, dass die Ergebnisse der oben genannten vier Modelle in den allgemeinen Trends übereinstimmen.
Nomenclature
- ADS
Automatic Depressurization System
- BWR
Boiling Water Reactor
- CS
Core Spray
- CSNPP
Chinshan Nuclear Power Plant
- DBA
Design Basis Accident
- ECCS
Emergency Core Cooling Systems
- FSAR
Final Safety Analysis Report
- FWLB
Feedwater Line Break
- HPCI
High-Pressure Coolant Injection System
- LOCA
Loss of Coolant Accident
- LPCI
Low-Pressure Coolant Injection System
- MSIV
Main Steam Isolation Valves
- MSL
Main Steam Lines
- MSLB
Main Steam Line Break
- NPSH
Net Positive Suction Head
- OLTP
Original Licensed Thermal Power
- PCT
Peak Cladding Temperature
- P/T
Pressure and Temperature
- RCIC
Reactor Core Isolation Cooling System
- RCLB
Recirculation Line Break
- RCS
Reactor Coolant System
- RHR
Residual Heat Removal
- RPV
Reactor Pressure Vessel
- SNAP
Symbolic Nuclear Analysis Package
- SRVs
Safety/Relief Valves
- TPC
Taiwan Power Company
- TRACE
TRAC/RELAP Advanced Computational Engine
- fi
interfacial unit volume force
- fwg
unit volume force between wall and gas mixture
- fwl
unit volume force between wall and liquid
- eg
gas mixture internal energy
- el
liquid internal energy
gravity vector
vapor enthalpy of the bulk vapor if the vapor is condensing or the vapor saturation enthalpy if the liquid is vaporizing
- P
total pressure
- qdg
power deposited directly to the gas mixture
- qdl
power deposited directly to the liquid
- qwg
heat-transfer rate per unit volume between wall and gas mixture
- qwl
heat-transfer rate per unit volume between wall and liquid
- qwsat
heat-transfer rate per unit volume between wall and saturation fluid
- t
time
gas mixture velocity vector
liquid velocity vector
- α
void fraction
- ρg
gas mixture density
- ρl
liquid density
- Γ
interfacial mass-transfer rate (positive from liquid to gas)
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© 2021 Walter de Gruyter GmbH, Berlin/Boston
Artikel in diesem Heft
- Frontmatter
- Experimental investigation on pool boiling for downward facing heating with different concentrations of Al2O3 nanofluids
- 244Cm contributions to the alpha source term of CANDU reactors
- Methodology for integral analysis of ATWS for Kuosheng nuclear power station
- LOCA analysis of BWR-4/Mark-I nuclear power plant with TRACE
- Fuel integrity evaluation method for BWR-6 plant Kuosheng during ATWS events
- Uncertainty studies on hydrogen source term with MAAP5 code
- Dynamic fault tree analysis of auxiliary feedwater system in a pressurized water reactor
- Impact of control rod insertion during burnup on PWR fuel assembly isotopic composition
- Temperature adjusted cross section libraries used for criticality calculations
Artikel in diesem Heft
- Frontmatter
- Experimental investigation on pool boiling for downward facing heating with different concentrations of Al2O3 nanofluids
- 244Cm contributions to the alpha source term of CANDU reactors
- Methodology for integral analysis of ATWS for Kuosheng nuclear power station
- LOCA analysis of BWR-4/Mark-I nuclear power plant with TRACE
- Fuel integrity evaluation method for BWR-6 plant Kuosheng during ATWS events
- Uncertainty studies on hydrogen source term with MAAP5 code
- Dynamic fault tree analysis of auxiliary feedwater system in a pressurized water reactor
- Impact of control rod insertion during burnup on PWR fuel assembly isotopic composition
- Temperature adjusted cross section libraries used for criticality calculations
