Phenomena of vapor transport in SGTR analysis

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Phenomena of vapor transport in SGTR analysis
Pavel Kudinov and Nam Dinh Division of Nuclear
Power Safety Royal Institute of Technology
(KTH) Stockholm, Sweden

• SGTR induced threats
• Risk assessment of SG tube leakage and rupture
• Some facts and statistics about SGTR
• Cracks and ruptures variety of conditions
• Vapor bubbles formation and transport phenomena
• Summary

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15..25 MPa, 330..500 oC
SGTR
0.3 MPa, 400..500 oC
LFR
EFIT
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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

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SG tube Leak and Rupture Risk Assessment
SG tube Leak and Rupture are need to be evaluated
against their Probability and Consequences
SGTR in PWR
SGTR in LFR, EFIT?
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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.,
non-radioactive) 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.
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Steam Generator Degradation Types
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Steam Generator Degradation Types
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Steam generator tube leakage USA NRC statistics
1990-2000
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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

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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
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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
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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
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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

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Vapor bubbles formation and transport phenomena
Shape and size of the bubbles Wecrit 10 gt
dmax10 mm Eo(dmax)10 gt oblate ellipsoid
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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
Short wavelength due to high-pressure discharge.
Measured average velocity of a bubble 0.3 m/s
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Vapor bubbles formation and transport phenomena
Terminal speed of rising bubbles with dmax10mm
is 0.2 - 0.3 m/s Effective density of vapor
bubble (with water droplet inside) dose not
affect terminal velocity Importance of
resolution of 3D structure of the coolant flow
for reliable prediction of void flux into the core
Mendelson
Lehrer
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Size distributions of water droplets
Beznosov et al, 2005
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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
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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.
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Summary

• High uncertainty still remains in SGTR
• Probability
• Conditions
• Consequences
• With no operating experience SGTR may become
  bottleneck for licensing
• More efforts needed in design for
• Prevention of SGTR occurrence and
• Mitigation of its consequences