Phenomena of vapor transport
in SGTR analysis
Pavel Kudinov and Nam Dinh
Division of Nuclear Power Safety
Royal Institute of Technology (KTH)
Stockholm, Sweden
1. SGTR induced threats
2. Risk assessment of SG tube leakage and rupture
3. Some facts and statistics about SGTR
4. Cracks and ruptures: variety of conditions
5. Vapor bubbles formation and transport phenomena
6. Summary
15..25 MPa,
330..500 oC
SGTR
0.3 MPa,
400..500 oC
LFR
EFIT
SGTR-Induced Threats
• Dynamic Loadings and Impact on Reactor Equipment
 Causing Secondary Failures
• Transport of Steam to the Core and Core Voiding
 Reactivity Insertion with Potential for Power
Excursion
 Rupture-induced pressure shock wave
 Steam Generation-Induced Sloshing
 Steam Explosion
 Steam Transport to the Reactor Core
SG tube Leak and Rupture
Risk Assessment
SG tube Leak and Rupture are need to be evaluated against
their Probability and Consequences
Risk = Probability * Consequence
10-2
Not
Acceptable
Probability
1/year
10-3
SGTR in PWR
10-4
SGTR in LFR, EFIT?
10-5
10-6
End of
Spectrum
Acceptable
Consequences
US NRC about tube degradation
Tube Degradation
During the early-to-mid 1970s, when all plants, except one, had mill annealed Alloy 600 steam generator tubes, thinning of
the mill annealed Alloy 600 steam generator tube walls due to the chemistry of the water flowing around them was the
dominant cause of tube degradation. However, all plants have changed their water chemistry control programs since
then, virtually eliminating the problem with tube thinning.
After tube thinning, tube denting became a primary concern in the mid to late-1970s. Denting results from the corrosion of
the carbon steel support plates and the buildup of corrosion product in the crevices between tubes and the tube support
plates. Measures have been taken to control denting, including changes in the chemistry of the secondary (i.e., nonradioactive) side of the plant. But other phenomena continue to cause tube cracking in plants with mill annealed Alloy 600
tubes.
The extensive tube degradation at pressurized-water reactors (PWRs) with mill annealed Alloy 600 steam generator
tubes has resulted in tube leaks, tube ruptures, and midcycle steam generator tube inspections. This degradation
also led to the replacement of mill annealed Alloy 600 steam generators at a number of plants and contributed to the
permanent shutdown of other plants.
As mill annealed Alloy 600 steam generator tubes began exhibiting degradation in the early 1970s, the industry pursued
improvements in the design of future steam generators to reduce the likelihood of corrosion. In the late 1970s, Alloy 600
tubes were subjected to a high temperature thermal treatment to improve the tubes’ resistance to corrosion. This
thermal treatment process was first used on tubes installed in replacement steam generators put into service in the early
1980s. Thermally treated Alloy 600 is presently used in the steam generators at 17 plants. Although no significant
degradation problems have been observed in plants with thermally treated Alloy 600 steam generator tubes, plants which
replaced their steam generators since 1989 have primarily used tubes fabricated from thermally treated Alloy 690, which is
believed to be even more corrosion resistant than thermally treated Alloy 600. Thermally treated Alloy 690 is presently used
in the steam generators at 27 plants.
Most of the newer steam generators, including all of the replacement steam generators, have features which make the
tubes less susceptible to corrosion-related damage. These include using stainless steel tube support plates to minimize the
likelihood of denting and new fabrication techniques to minimize mechanical stress on tubes.
Steam Generator Degradation Types
Types of SG tube
degradation in PWR
Definition
Denting
The physical deformation of the Inconel Alloy 600 tubes as they
pass through the support plate. Caused by a buildup of corrosive
material in the space between the tube and the plate.
Fatigue cracking
Caused by tube vibration.
Fretting
The wearing of tubes in their supports due to flow induced vibration.
Intergranular attack/stresscorrosion cracking
Caused when tube material is attacked by chemical impurities from
the secondary-loop water. It occurs primarily within tube sheet
crevices and other areas where impurities concentrate.
Pitting
The result of local breakdown in the protective film on the tube.
Active corrosion occurs at the site of breakdown.
Stress-corrosion cracking
(inside diameter)
Cracking of steam generator tubes occurring at the tangent point
and apex of U-bend tubes, at the tube sheet roll transition, and in
tube dents. It occurs when Inconel Alloy 600 tubing is exposed to
primary-loop water.
Tube wear
A thinning of tubes caused by contact with support structures either
as the tubes vibrate or as feedwater entering the vessel impinges on
the tube bundle at that location.
Wastage
A general corrosion caused by chemical attack from acid phosphate
residues in areas of low water flow.
Steam Generator Degradation Types
Types of SG tube degradation in PWR
Types of SG tube degradation in LFR
Denting
Fatigue cracking
Fretting
Intergranular attack/stress-corrosion
cracking
Pitting
Stress-corrosion cracking (inside
diameter)
Tube wear
Wastage
?
Steam generator tube leakage
USA NRC statistics 1990-2000
N
Date
Plant
1
Jan. 1990
St. Lucie 1
2
Mar. 1990
TMI 1
3
May 1990
Millstone 2
4
Aug 1990
North Anna 2
5
Nov. 1990
6
Leak
Rate
liter/day
11
Cause
N
Date
Plant
Leak
Rate
liter/day
Cause
Foreign Object
20
Aug. 1993
McGuire 1
700
Sleeve Failure
Fatigue
21
Sept 1993
Palo Verde 3
397
Freespan crack
Cracked Plug
22
Oct 1993
McGuire 1
700
Circ. crack in sleeved tube
151
Cracked Plug
23
Oct. 1993
Braidwood 1
1136
Oconee 2
492
Fatigue
24
Nov. 1993
San Onofre 3
189
Nov. 1990
Shearon Harris
189
Loose Part
Loose parts degradation and leaking
welded plugs
7
Dec. 1990
Maine Yankee
5451
25
Nov. 1993
Farley 2
8
Apr. 1991
San Onofre 1
568
Sleeve Joint
26
Jan. 1994
McGuire 1
379
Leaking Sleeve
9
Apr. 1991
Millstone 2
265
U-bend SCC
27
Mar. 1994
Oconee 3
545
Fatigue
10
May 1991
Millstone 2
265
Tube Sheet Circumferential Crack
28
Mar. 1994
S. Texas
606
Leaking Plug
11
Jan. 1992
McGuire 1
946
Freespan Crack
29
Mar. 1994
Zion 2
12
Mar. 1992
ANO 2
13
Mar. 1992
Prairie Island 1
14
May 1992
McGuire 1
15
Sep. 1992
16
5451
PWSCC
5451
Freespan Cracks
Tubesheet Crevice Inter Granular
Attack OD
1363
Tube Sheet Circumferential Crack
30
Jul. 1994
Oconee 2
545
Fatigue
545
Roll Transition Zone Axial Crack
31
Jul. 1994
Maine Yankee
189
Circumferential Crack
19
32
Feb. 1996
Zion 1
Prairie Island 1
329
33
Aug. 1996
Byron 2
Nov. 1992
McGuire 1
946
34
May 1996
Vogtle 1
17
Nov. 1992
Trojan
757
Sleeve Weld Circumferential
Crack
35
Nov. 1996
ANO 2
246
Axial Crack
18
Mar. 1993
Palo Verde 2
908
Upper Bundle Freespan Inter
Granular Stress Corrosion
Cracking
36
June 1997
McGuire 2
250
ODSCC at TSP
37
Nov. 1997
Oconee 1
1514
2 Welded Plugs
38
Dec. 1998
Farley 1
341
19
Jun. 1993
Kewaunee
379
Leaking Plug
Foreign object
454
Loose Part
Foreign object
2 Freespan Cracks
Known Steam Generator Tube Rupture
Accidents in the World
1975-2002
Single SGTR is a rare event
Multiple SGTR (MSGTR)
has never occurred
The reasons for reduction of
SGTR frequency during
past years are:
 enhancement of SG
production technology
 chemistry control
during operation
 regular inspections
and better regulation
Steam generator tube leakage
Crack Morphology and Leak Rate
First leak at 17.2MPa
Maximum leak rate
4.28 l/min at 34.5MPa
First leak at 25MPa
Maximum leak rate
0.25 l/min at 31.7MPa
Seong Sik Hwang, Hong Pyo Kim, Joung Soo Kim, Kenneth E. Kasza, Jangyul Park and William J. Shack
”Leak behavior of SCC degraded steam generator tubings of nuclear power plant”
Nuclear Engineering and Design, Volume 235, Issue 23, December 2005, Pages 2477-2484
Wear degradation of steam
generator tubs
Seong Sik Hwang, Chan Namgung, Man Kyo Jung, Hong Pyo Kim and Joung Soo Kim
”Rupture pressure of wear degraded alloy 600 steam generator tubings”
Journal of Nuclear Materials, In Press, Corrected Proof, Available online 16 May 2007
Burst characteristics for
axial notches
Seong Sik Hwang, Hong Pyo Kim and Joung Soo Kim
“Evaluation of the burst characteristics for axial notches on SG tubings”
Nuclear Engineering and Design, Volume 232, Issue 2, August 2004, P.139-143
Cracks and Ruptures:
Variety of Initial Conditions
 Flow rate depends on type, area and geometry of the opening
 Sizes and Geometry of opening depends on initial degradation
type and sizes
 “Leakage before rupture” concept: based on fact that small
leakage was often detected before (~several hours) the rupture
had occurred. Although sudden ruptures also took place in the
past
 How to detect small leakage in lead cooled systems?
 Leakage can produce small bubbles transportable to the core
 …
Vapor bubbles formation
and transport phenomena
1.0E+06
Re
Eo
We
1.0E+05
1.0E+04
1.0E+03
1.0E+02
1.0E+01
1.0E+00
1.0E-01
1.0E-02
1.0E-03
0.1
1
10
d, mm
Shape and size of the bubbles:
Wecrit ~ 10
=> dmax~10 mm
Eo(dmax)~10 => oblate ellipsoid
100
Steam Bubble Size Distribution
Water: 22-24 MPa, 150-250 oC
Beznosov et al, 2005
14x2 mm tube
10 mm discharge
2000 mm depth
52 mm
 CAP

2
g
Short wavelength due to high-pressure discharge.
Measured average velocity of a bubble ~0.3 m/s
Vapor bubbles formation
and transport phenomena
Terminal velocity
0.80
Terminal speed of rising
bubbles with dmax~10mm
is ~0.2 - 0.3 m/s
Jamialahmadi
Mendelson
Lehrer
0.70
0.60
m/s
0.50
0.40
Effective density of vapor
bubble (with water droplet
inside) dose not affect
terminal velocity
0.30
0.20
0.10
0.00
0.1
1
10
100
d, mm
Mendelson:
Lehrer:
Importance of resolution
of 3D structure of the
coolant flow for reliable
prediction of void flux into
the core
Size distributions of water droplets
Beznosov et al, 2005
Life time of small droplet
on a hot surface
Time scale is ~10s of seconds
for droplets ~1mm in diameter
Guido Bleiker and Eckehard Specht Film evaporation of drops of
different shape above a horizontal plate International Journal of
Thermal Sciences, Volume 46, Issue 9, September 2007, Pages
835-841
Vapor bubbles formation
and transport phenomena
Evaporation of water droplet in a bubble will lead to
growth of bubble diameter.
Due to evaporation initial volume of void will
increase ~2 times during ~10 seconds.
Unfortunately, big (fast rising) bubbles most likely
will not be stable due to high We number and high
turbulence level.
As a result we will have larger number of middle
size bubbles up to 10 mm in diameter.
Summary
1. High uncertainty still remains in SGTR
 Probability
 Conditions
 Consequences
2. With no operating experience SGTR may become
bottleneck for licensing
3. More efforts needed in design for
 Prevention of SGTR occurrence and
 Mitigation of its consequences