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Fuel integrity evaluation method for BWR-6 plant Kuosheng during ATWS events

Methode zur Bewertung der Brennstoffintegrität der SWR-6-Anlage Kuosheng bei ATWS-Ereignissen
K.-H. Hsu and H.-C. Chien
From the journal Kerntechnik

Abstract

Anticipated Transients without Scram (ATWS) are evaluated to demonstrate that during ATWS the fuel’s integrity prevents fission product release, and that the safety features the BWR-6 plant is equipped with mitigate the ATWS event. For analyzing this scenario, a RETRAN-3D system model and hot channel model of Kuosheng NPP have been developed. From all possible events those are analyzed where the main steam isolation valves close (MSIVC) when the reactor is operating at rated power conditions. For cases with low core flows this results in higher peak cladding temperatures and thicker oxidation thicknesses. All results confirm the ATWS acceptance criteria. For sensitivity cases, increasing the boron concentration of standby liquid control system (SLCS) or injecting boron of SLCS earlier were investigated. The results show, that these measures cannot decrease the PCT value immediately, but can shorten the period of peak cladding temperature over 850 °C which results in smaller oxidation thickness.

Abstract

Anticipated Transients without Scram (ATWS) werden ausgewertet, um zu zeigen, dass während ATWS die Integrität des Brennstoffs erhalten bleibt und so die Freisetzung von Spaltprodukten verhindert und dass die Sicherheitsmerkmale, mit denen die SWR-6-Anlage ausgestattet ist, das ATWS-Ereignis abmildern. Für die Analyse dieses Szenarios wurden ein RETRAN-3D-Systemmodell und ein Heißkanalmodell des KKW Kuosheng entwickelt. Von allen möglichen Ereignissen werden diejenigen analysiert, bei denen die Frischdampfabsperrventile schließen, wenn der Reaktor mit Nennleistung betrieben wird. Für die Fälle mit geringen Kernströmungen ergeben sich höhere Spitzenhüllentemperaturen und dickere Oxidationsdicken. Alle Ergebnisse bestätigen die ATWS-Akzeptanzkriterien. Als Sensitivitätsfälle wurden die Erhöhung der Borkonzentration des Standby Liquid Control Systems oder die frühere Borinjektion des SLCS untersucht. Die Ergebnisse zeigen, dass diese Maßnahmen den PCT-Wert nicht sofort senken können, aber den Zeitraum der maximalen Plattentemperatur über 850 °C verkürzen können, was zu einer geringeren Oxidationsdicke führt.

Nomenclature

1-D

one-dimensional

AOO

anticipated operational occurrence

ARI

alternate rod insertion

ATWS

anticipated transient without scram

BE

best estimate

BWR

boiling water reactor

EPRI

Electric Power Research Institute

HPCS

high pressure core spray system

HTC

heat transfer coefficient

MSIVC

main steam isolation valve closure

MSIV

main steam isolation valves

NPP

nuclear power plant

OLTP

original licensed thermal power

PCT

peak cladding temperature

PWR

pressurized water reactor

ROC-AEC

Republic of China-Atomic Energy Council

RPT

recirculation pumps trip

RPV

reactor pressure vessel

RCIC

reactor core isolation cooling system

RRCS

redundant reactivity control system

RTP

rated thermal power

SLCS

standby liquid control system

SPU

stretch power uprate

SRV

safety relief valve

TSV

turbine stop valve

TCV

turbine control valve

FW

feedwater

References

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Received: 2019-10-17
Published Online: 2021-03-30
Published in Print: 2021-04-30

© 2021 Walter de Gruyter GmbH, Berlin/Boston

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