VGB POWERTECH 7 (2021) - International Journal for Generation and Storage of Electricity and Heat

VGB PowerTech 7 l 2021
Verification of SPACE code based on an MSGTR experiment at the ATLAS-PAFS facility
Study on verification of SPACE code
based on an MSGTR experiment at
the ATLAS-PAFS facility
Kyungho Nam
Kurzfassung
Studie zur Verifizierung des SPACE-Codes
auf der Grundlage eines MSGTR-
Experiments in der ATLAS-PAFS-Anlage
Ein MSGTR-Unfall (Multiple Steam Generator
Tube Rupture) wird in Korea als ein Unfall definiert,
bei dem mehr als fünf U-Rohre eines
Dampferzeugers ausfallen. Um die thermohydraulischen
Phänomene eines MSGTR-Unfalls zu
verstehen, wurde von KAERI eine experimentelle
Studie durchgeführt. Das Experiment wurde
durchgeführt, um die transienten Phänomene
zu simulieren, die durch den Bruch von fünf U-
Rohren verursacht werden, und um die Wärmeabfuhrkapazität
des passiven Hilfsspeisewassersystems
(PAFS) während des Transienten zu
validieren.
In diesem Beitrag wurde ein MSGTR-Experiment
in der ATLAS-PAFS-Versuchsanlage mit
dem SPACE-Code simuliert, um die Vorhersagefähigkeit
dieses Codes für Mehrfachversagensunfälle
zu überprüfen, die in der Auslegungserweiterung
enthalten sind. Das mit dem SPACE-
Code ermittelte instationäre Verhalten des
Gesamtsystems zeigte ähnliche Trends wie die
experimentellen Ergebnisse in Bezug auf Faktoren
wie Systemdruck, Massendurchsatz und
kollabierter Wasserstand auf der Komponente.
Zusätzlich wurde eine Sensitivitätsanalyse unter
Verwendung der experimentellen Korrelation
für PAFS durchgeführt, die als Option im
SPACE-Code im Wandkondensationsmodell
enthalten ist. Als Ergebnis der Sensitivitätsberechnung
wird auch empfohlen, das PAFS-Modell
im SPACE-Code anzuwenden, um genauere
Vorhersageergebnisse über den PAFS-Betrieb zu
erhalten, indem eine Sicherheitsanalyse des
Kernkraftwerks APR+ durchgeführt wird. l
Author
Kyungho Nam
Associate research engineer
Korea Hydro & Nuclear Power Co., LTD.
Central Research Institute
Safety Analysis Group
Deajeon, Republic of Korea
A Multiple Steam Generator Tube Rupture
(MSGTR) accident is defined in Korea as an
accident in which more than five U-tubes of a
steam generator break down. To understand
the thermal hydraulic phenomena of a
MSGTR accident, an experimental study was
conducted by KAERI. The experiment was
conducted to simulate the transient phenomena
caused by the rupture of five U-tubes,
and to validate the heat removal capacity of
the Passive Auxiliary Feedwater System
(PAFS) during the transient.
In this paper, an MSGTR experiment at the
ATLAS-PAFS test facility was simulated using
the SPACE code to verify the prediction capability
of this code for multiple failure accident,
which is involved in the design extension
condition. The overall system transient
behavior obtained using SPACE code showed
similar trends with the experimental results
in terms of factors such as the system pressure,
mass flow rate, and collapsed water
level on the component. Additionally, a sensitivity
analysis was conducted using the experimental
correlation for PAFS which is included
in the wall condensation model as an
option in SPACE code. As a sensitivity calculation
results, it is also recommended that the
PAFS model in SPACE code be applied to obtain
more accurate prediction results about
the PAFS operation by performing a safety
analysis of the APR+ nuclear power plant.
1. Introduction
1.1 Background
Following the Fukushima nuclear disaster
in 2011 and based on the lessons learned
from the accident, there have been many
changes to the relevant safety design criteria
and/or regulations around the world.
The Nuclear Safety and Security Commission
(NSSC) in Korea has required plantspecific
accident management plans, which
extend beyond design basis accidents to
include severe accidents. The revised regulation
in Korea has determined a list of
multiple failure accidents that must be considered
for any accident management plan
[1]. Multiple failure accidents should be
considered for Design Extension Conditions,
which is defined by the International
Atomic Energy Agency (IAEA) Specific
Safety Requirement [2, 3].
The Multiple Steam Generator Tube Rupture
(MSGTR) accident is selected as one of
the multiple failure accidents by the Korean
NSSC, and it is defined as an accident in
which more than five U-tubes of a steam
generator rupture. In a MSGTR accident,
the pressurized primary coolant leaks into
the secondary system and thus exposes radioactive
material. Therefore, it is important
that the extent of the leak is limited
and that the pressure drop across the break
be kept as low as possible to reduce the radioactive
release. Compared to single tube
rupture, MSGTR causes quicker depressurization
of the reactor coolant system
(RCS) and places greater demand on the
inventory makeup process.
To elucidate the thermal hydraulic process
of the MSGTR accident, an experimental
study using the Advanced Test Loop for Accident
Simulation (ATLAS) facility was
conducted by the Korea Atomic Energy Research
Institute (KAERI) [4]. The experiment
simulated the rupture of five steam
generator tubes, and the results showed
that the Passive Auxiliary Feedwater System
(PAFS) adopted in the Advanced Power
Reactor Plus (APR+) had sufficient cooling
capacity to mitigate the accident. The
PAFS is one of the advanced safety features
of a passive cooling system that allow it to
replace a conventional active Auxiliary
Feed-water System (AFWS) [5]. In a typical
Nuclear Power Plant (NPP), a motordriven
or turbine-driven auxiliary feedwater
is supplied after the wide-range water
level of a steam generator is decreased below
the low steam generator level. However,
to confirm the cooling capability of
PAFS compared to that of AFWS, an experimental
scenario was conducted in which
the PAFS was supplied to an intact SG instead
of auxiliary feedwater. The PAFS is a
passive system capable of condensing the
steam generated in a steam generator and
feeding the condensed water to the steam
generator using gravity.
For current safety analyses of Korean nuclear
power plants, thermal-hydraulic safe-
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