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Licensed Unlicensed Requires Authentication Published by De Gruyter March 30, 2021

LOCA analysis of BWR-4/Mark-I nuclear power plant with TRACE

LOCA-Analyse des Kernkraftwerks BWR-4/Mark-I mit TRACE
J.-J. Huang, S.-W. Chen, J.-R. Wang, C. Shih and H.-T. Lin
From the journal Kerntechnik

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)

References

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Received: 2020-09-14
Published Online: 2021-03-30
Published in Print: 2021-04-30

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