atw 2017-09

atw Vol. 62 (2017) | Issue 8/9 ı August/September
ENVIRONMENT AND SAFETY 538
Simulation of Total Loss of Feed Water
in ATLAS Test Facility Using SPACE Code
Minhee Kim and Seyun Kim
1 Introduction After the Fukushima accident, simulation of beyond design extension conditions (DEC) are
important for developing the RCS cool down strategy and recovery action. Additional Failure of the safety components
are also considered in terms of sufficient safety margin with application of proper emergency operating procedures.
In order to examine the knowledge of
the actual phenomena and provide
the best guideline for accident
manage­ment, OECD-ATLAS joint
­project has been launched within the
OECD/NEA since April 2014 [1].
KAERI (Korea Atomic Energy Research
Institute) started operation of ATLAS
(Advanced Thermal-hydraulic Test
Loop for Accident Simulation) in order
to investigate the multiple responses
during anticipated transients and
­postulated accidents. The reference
plant of ATLAS is the APR1400, which
has a rated thermal power of 1400
MW and a loop arrangement of 2 hot
legs and 4 cold legs for the RCS. Major
topics of the OECD-ATLAS joint projects
was prolonged station black out
(SBO), small break loss of coolant
­accident (SBLOCA) during SBO,
total loss of feedwater (TLOFW)
with ­additional failure, medium-size
­LOCA, and counterpart tests.
In this paper, the target scenario is
OECD-ATLAS A3.1 test, which is a
TLOFW with additional failures. A
­total loss of feedwater (TLOFW) accident
is used to represent an accident
involving the failure of cooling by the
secondary cooling system. In a very
short period of time, the turbine and
the reactor trip either directly due to
turbo-pump trip or indirectly due to a
low narrow-range level in steam
­generator. Previous studies have
­focused on accidents involving a
TLOFW accident to demonstrate the
use of feed and bleed operation [2-5].
Reventós et al. [6] characterized the
procedure for feed and bleed operation
in a TLOFW accident and provided
plant behavior analysis under
the partial availability of systems at
the Ascó NPP. The maximum time
­allowed to start the procedure and
some considerations for heat sink
­recovery in a TLOFW accident were
also addressed. Sherry et al. [7] identified
the most important parameters
likely to influence whether core
­damage would occur during a TLOFW
accident.
The present paper is focused on the
thermal-hydraulic analysis for TLOFW
accident with additional failures of
ATLAS facility using SPACE code,
which has been developed in recent
years by the Korea Hydro & Nuclear
Power Company (KHNP). In order to
assess the capability of SPACE code,
the calculation results are compared
with experimental data in terms of
steady-state and transient behavior.
Based on the observed thermal-­
hydraulic features, comparative
­studies can provide physical insight
to the system response and confirm
the effectiveness of RCS cool down
strategy according to the emergency
operating guideline.
2 Code descriptions
The Safety and Performance Analysis
Code for Nuclear Power Plants
(SPACE) has been developed in recent
years by the Korea Hydro & Nuclear
Power Co. through collaborative works
with other Korean nuclear industries
and research institutes, and is approved
by Korea Institute of Nuclear
Safety (KINS) in March 2017. The
SPACE is a best-estimate two-phase
three-field thermal-hydraulic analysis
code used to analyze the safety and
performance of pressurized water
­reactors. Each field equation is
­discretized based on finite volume
­approach on a structured mesh and
an unstructured mesh together with
an one-dimensional pipe meshes [7].
The semi-implicit scheme is used for
time integration method. The SPACE
code is package of input and output,
hydrodynamic model, heat structure
model, and reactor kinetics model.
The input package performs a
reading the input and restart files, a
parsing the data files, an allocating
the memory, and checking the unit
conversion. Hydrodynamic model
package is composed of hydraulic
solver, constitutive models, special
process models, and component
­models. The hydraulic solver is based
on two-fluid, three-field governing
equations, which are composed of gas,
continuous liquid, and droplet fields.
Therefore, SPACE code have the advantage
in solving a dispersed liquid
field as well as vapor and continuous
liquid fields in comparison with existing
nuclear reactor system analysis
codes, such as RELAP5 (ISL, 2001),
TRACE (NRC, 2000), CATHARE
(­Robert et al., 2003), and MARS-KS
(KAERI, 2006). Constitutive models
are composed of correlations by the
flow regime map to simulate the mass,
momentum, and energy distributions,
such as surface area and surface heat
transfer, surface-wall friction, droplet
separation and adhesion, and wall-­
fluid heat transfer. In order to ­simulate
the physical phenomena of the plant
system, special process and system
components are modeled. Major
­special process and component
­models are critical flow model,
­counter current flow limit model,
­off-take model, abrupt area change
model, 2-phase level tracking model,
pump model, safety injection tank
model, valve model, pressurizer
­model, and separator model, etc.
The package of heat structure
model calculates the heat addition
transfer and removal. The heat structure
model includes transient heat
conduction of rectangular or cylindrical
geometry, and has various
boundary conditions of convection,
radiation, user defined variables such
as temperature, heat flux, and heat
transfer coefficient.
Nuclear fission heat of a nuclear
­fuel rod is calculated by point kinetics
approximation and treated as a heat
source in the heat conduction equation.
Reactivity feedbacks are considered
in terms of moderator density,
moderator temperature, fuel temperature,
boron concentration, reactor
scram, and power defect. Decay heat
of ANS-73, -79, and -2005 models are
also implemented.
The 3.0 version of the code was
­released through various validation
and verification using the separated or
integral loop test data and the plant
operating data. The approved code
Environment and Safety
Simulation of Total Loss of Feed Water in ATLAS Test Facility Using SPACE Code ı Minhee Kim and Seyun Kim