A computational fluid dynamics (CFD) analysis was performed to investigate the hydraulic response of the flow field inside the pressurized water reactor (PWR) steam generator (SG) secondary side and the connected part of main feed water pipe to an abrupt main feed water line break (FWLB) accident. To realistically analyze the transient flow field situation, the flow field was assumed to be occupied initially by highly compressed subcooled water except that the upper part of the SG secondary side where steam occupied as in the practical case and the break was assumed to occur at the circumferential weld line between the feed water nozzle and the main feed water pipe. This would result in a subcooled water flashing flow from the SG through the short-broken pipe end to the surrounding atmosphere, which was numerically simulated in this study. Typical results of the prediction in terms of the fluid transient velocity and pressure were illustrated and discussed. To examine the physical validity of the present numerical simulation of the subcooled water flashing flow, the transient mass flow rates predicted in this study were compared with the other previous numerical predictions based on the subcooled water nonflashing (no phase change) flow or saturated water flashing flow assumptions and the prediction by a simple analysis method.

References

1.
Shier
,
W. G.
, and
Levine
,
M. M.
,
1980
, “
PWR Steamline Break Analysis Assuming Concurrent Steam Generator Tube Rupture
,”
ANS/ASME Topical Meeting on Reactor Thermal-Hydraulics
,
Saratoga, NY
,
Oct. 9
, Paper No. CONF-801002-9.
2.
Kalra
,
S.
, and
Adams
,
G.
,
1980
, “
Thermal Hydraulics of Steam Line Break Transients in Thermal Reactors—Simulation Experiments
,”
ANS International Conference
,
American Nuclear Society, Washington, DC
,
Nov. 17–21
, Paper No. CONF-801107.
3.
Wolf
,
L.
,
1982
, “
Experimental Results of Coupled Fluid-Structure Interactions During Blowdown of the HDR-Vessel and Comparisons With Pre- and Post-Test Predictions
,”
Nucl. Eng. Des.
,
70
(
3
), pp.
269
308
.
4.
Saha
,
P.
,
Ghosh
,
A.
,
Das
,
T. K.
, and
Ray
,
S.
,
1993
, “
Numerical Simulation of Pressure Wave Time History Inside a Steam Generator in the Event of Main Steam Line Break and Feedwater Line Break Transients
,” 29th National Heat Transfer Conference, Atlanta, GA, Aug. 8–11, pp.
131
140
.
5.
Joo
,
H. G.
,
Jeong
,
J.-J.
,
Cho
,
B.-O.
,
Lee
,
W. J.
, and
Zee
,
S. Q.
,
2003
, “
Analysis of the OECD Main Steam Line Break Benchmark Problem Using the Refined Core Thermal-Hydraulic Nodalization Feature of the MARS/MASTER Code
,”
Nucl. Technol.
,
142
(
2
), pp.
166
179
.
6.
Kang
,
K.-H.
,
Park
,
H.-S.
,
Cho
,
S.
,
Choi
,
N.-H.
,
Bae
,
S.-W.
,
Lee
,
S.-W.
,
Kim
,
Y.-S.
,
Choi
,
K.-Y.
,
Baek
,
W.-P.
, and
Kim
,
M.-Y.
,
2011
, “
Experimental Study on the Blowdown Load During the Steam Generator Feedwater Line Break Accident in the Evolutionary Pressurized Water Reactor
,”
Ann. Nucl. Energy
,
38
(
5
), pp.
953
963
.
7.
Gallardo
,
S.
,
Querol
,
A.
, and
Verdú
,
G.
,
2012
, “
Simulation of a Main Steam Line Break With Steam Generator Tube Rupture Using Trace
,”
PHYSOR
,
Knoxville, TN
,
Apr. 15–20
, pp.
2131
2144
.
8.
Gallardo
,
S.
,
Querol
,
A.
, and
Verdú
,
G.
,
2014
, “
Improvements in the Simulation of a Main Steam Line Break With Steam Generator Tube Rupture
,”
Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo
,
Paris, France
,
Oct. 27–31
, Paper No. 05104.
9.
Hamouda
,
O.
,
Weaver
,
D. S.
, and
Riznic
,
J.
,
2015
, “
Loading of Steam Generator Tubes During Main Steam Line Breaks
,” Canadian Nuclear Safety Commission, Ottawa, ON, Canada, Report No. RSP-0305.
10.
Hamouda
,
O.
,
Weaver
,
D. S.
, and
Riznic
,
J.
,
2016
, “
An Experimental Model Study of Steam Generator Tube Loading During a Sudden Depressurization
,”
ASME J. Pressure Vessel Technol.
,
138
(4), p.
041302
.
11.
Moody
,
F. J.
,
1965
, “
Maximum Flow Rate of a Single Component, Two-Phase Mixture
,”
ASME J. Heat Transfer
,
87
(
1
), pp.
134
142
.
12.
Henry
,
R. E.
, and
Fauske
,
H. K.
,
1971
, “
The Two-Phase Critical Flow of One-Component Mixtures in Nozzles, Orifices, and Short Tubes
,”
ASME J. Heat Transfer
,
93
(
2
), pp.
179
187
.
13.
Weisman
,
J.
, and
Tentner
,
A.
,
1978
, “
Models for Estimation of Critical Flow in Two-Phase Systems
,”
Prog. Nucl. Energy
,
2
(
3
), pp.
183
197
.
14.
Johnson
,
R. W.
,
1998
,
The Handbook of Fluid Dynamics
,
CRC Press
,
New York
, pp.
27
40
.
15.
Jo
,
J. C.
, and
Moody
,
F. J.
,
2015
, “
Transient Thermal-Hydraulic Responses of the Nuclear Steam Generator Secondary Side to a Main Steam Line Break
,”
ASME J. Pressure Vessel Technol.
,
137
, p.
041301
.
16.
Jo
,
J. C.
,
Min
,
B. K.
, and
Jeong
,
J. J.
,
2016
, “
Evaluation of a Numerical Analysis Model for the Transient Response of Nuclear Steam Generator Secondary Side to a Sudden Steam Line Break
,”
ASME J. Pressure Vessel Technol.
,
139
(
3
), p.
031305
.
17.
Jo
,
J. C.
, and
Moody
,
F. J.
,
2016
, “
Effects of a Venturi Type Flow Restrictor on the Thermal-Hydraulic Response of the Secondary Side of a Pressurized Water Reactor Steam Generator to a Main Steam Line Break
,”
ASME J. Pressure Vessel Technol.
,
138
, p.
041304
.
18.
Jo
,
J. C.
,
Jeong
,
J. J.
, and
Moody
,
F. J.
,
2016
, “
Transient Hydraulic Response of a Pressurized Water Reactor Steam Generator to a Feedwater Line Break Using the Nonflashing Liquid Flow Model
,”
ASME J. Pressure Vessel Technol.
,
139
(3), p.
031302
.
19.
Jo
,
J. C.
,
Jeong
,
J. J.
,
Yun
,
B. J.
, and
Moody
,
F. J.
,
2018
, “
Numerical Prediction of a Flashing Flow of Saturated Water at High Pressure
,”
Nucl. Eng. Technol.
,
50
(
7
), pp.
1173
1183
.
20.
KHNP
,
2011
, “
APR+ Standard Safety Analysis Report
,” KHNP, Seoul, South Korea, Report No. 10.4-17.
21.
Simones-Moreira
,
J. R.
,
Vieira
,
M. M.
, and
Angelo
,
E.
,
2002
, “
Highly Expanded Flashing Liquid Jets
,”
J. Thermophys. Heat Transfer
,
16
(
3
), pp.
415
424
.
22.
Lin
,
J. C.
,
Gruen
,
G. E.
, and
Quapp
,
W. J.
,
1980
, “
Critical Flow in Small Nozzles for Saturated and Subcooled Water at High Pressure
,” ASME Paper No. CONF-801102-10.
23.
Ardron
,
K. H.
, and
Furness
,
R. A.
,
1976
, “
A Study of the Critical Flow Models Used in Reactor Blowdown Analysis
,”
Nucl. Eng. Des.
,
39
(
2–3
), pp.
257
266
.
24.
Lien
,
P.
,
2016
, “
Scaling in Complex Thermal-Hydraulic Phenomena
,”
ASME
Paper No. PVP2016-63047.
25.
ANSYS
,
2012
, “
ANSYS CFX User's Guide-14
,”
ANSYS
,
New York
.
26.
Menter
,
F. R.
,
1994
, “
Two Equation Eddy-Viscosity Turbulence Models for Engineering Applications
,”
AIAA J.
,
32
(
8
), pp.
1598
1604
.
27.
Menter
,
F. R.
,
Kuntz
,
M.
, and
Langtry
,
R.
,
2003
, “
Ten Years of Industrial Experience With the SST Turbulence Model
,”
Turbulence, Heat and Mass Transfer
, K. Hanjalic, Y. Nagano, and M. Tummers, eds., Vol. 4, Begell House, New York, pp. 625–632.
28.
Moody
,
F. J.
,
1990
,
Introduction to Unsteady Thermofluid Mechanics
,
Wiley
,
New York
.
29.
Park
,
B. S.
, and
Lee
,
S. Y.
,
1994
, “
An Experimental Investigation of the Flash Atomization Mechanism
,”
Atomization Sprays
,
4
(
2
), pp.
159
179
.
30.
Witlox
,
H. W. M.
, and
Bowen
,
P. J.
,
2002
, “
Flashing Liquid Jets and Two-Phase Dispersion—A Review
,” Det Norske Veritas, Oslo, Norway, Report No. 403/2002.
31.
Polanco
,
G.
,
Holdo
,
A. E.
, and
Munday
,
G.
,
2010
, “
General Review of Flashing Jet Studies
,”
J. Hazard. Mater.
,
173
(
1–3
), pp.
2
18
.
32.
Sokolowski
,
L.
, and
Kozlowski
,
2012
, “
Assessment of Two-Phase Critical Flow Models Performance in RELAP5 and TRACE against Marviken Critical Flow Tests
,” NRC, Washington, DC, Report No. NUREG/IA-0401.
You do not currently have access to this content.