Phenomena of vapor transport Phenomena of vapor transport in SGTR analysisin SGTR analysis

Pavel Kudinov and Nam DinhDivision of Nuclear Power Safety

Royal Institute of Technology (KTH)Stockholm, Sweden

1. SGTR induced threats 2. Risk assessment of SG tube leakage and rupture3. Some facts and statistics about SGTR4. Cracks and ruptures: variety of conditions5. Vapor bubbles formation and transport phenomena6. Summary

SGTR

EFIT

15..25 MPa, 330..500 oC

LFR

0.3 MPa, 400..500 oC

SGTR-Induced ThreatsSGTR-Induced Threats

Rupture-induced pressure shock wave

Steam Generation-Induced Sloshing

Steam Explosion

Steam Transport to the Reactor Core

• 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

Consequences

Not Acceptable

10-6

10-5

10-4

10-3

10-2

End of Spectrum

Probability

1/year

Risk = Probability * Consequence

Acceptable

SG tube Leak and Rupture SG tube Leak and Rupture Risk AssessmentRisk Assessment

SG tube Leak and Rupture are need to be evaluated against their Probability and Consequences

SGTR in PWR

SGTR in LFR, EFIT?

Tube DegradationDuring 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.

US NRC about tube degradationUS NRC about tube degradation

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/stress-corrosion 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 TypesSteam 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 Degradation TypesSteam Generator Degradation Types

N Date Plant Leak Rate

liter/day

Cause

1 Jan. 1990 St. Lucie 1 11 Foreign Object

2 Mar. 1990 TMI 1 5451 Fatigue

3 May 1990 Millstone 2 Cracked Plug

4 Aug 1990 North Anna 2 151 Cracked Plug

5 Nov. 1990 Oconee 2 492 Fatigue

6 Nov. 1990 Shearon Harris 189 Loose Part

7 Dec. 1990 Maine Yankee 5451 PWSCC

8 Apr. 1991 San Onofre 1 568 Sleeve Joint

9 Apr. 1991 Millstone 2 265 U-bend SCC

10 May 1991 Millstone 2 265 Tube Sheet Circumferential Crack

11 Jan. 1992 McGuire 1 946 Freespan Crack

12 Mar. 1992 ANO 2 1363 Tube Sheet Circumferential Crack

13 Mar. 1992 Prairie Island 1 545 Roll Transition Zone Axial Crack

14 May 1992 McGuire 1 19  

15 Sep. 1992 Prairie Island 1 329  

16 Nov. 1992 McGuire 1 946  

17 Nov. 1992 Trojan 757 Sleeve Weld Circumferential Crack

18 Mar. 1993 Palo Verde 2 908 Upper Bundle Freespan Inter Granular Stress Corrosion Cracking

19 Jun. 1993 Kewaunee 379 Leaking Plug

N Date Plant Leak Rate

liter/day

Cause

20 Aug. 1993 McGuire 1 700 Sleeve Failure

21 Sept 1993 Palo Verde 3 397 Freespan crack

22 Oct 1993 McGuire 1 700 Circ. crack in sleeved tube

23 Oct. 1993 Braidwood 1 1136 Freespan Cracks

24 Nov. 1993 San Onofre 3 189 Loose parts degradation and leaking welded plugs

25 Nov. 1993 Farley 2  

26 Jan. 1994 McGuire 1 379 Leaking Sleeve

27 Mar. 1994 Oconee 3 545 Fatigue

28 Mar. 1994 S. Texas 606 Leaking Plug

29 Mar. 1994 Zion 2 5451 Tubesheet Crevice Inter Granular Attack OD

30 Jul. 1994 Oconee 2 545 Fatigue

31 Jul. 1994 Maine Yankee 189 Circumferential Crack

32 Feb. 1996 Zion 1 Foreign object

33 Aug. 1996 Byron 2 454 Loose Part

34 May 1996 Vogtle 1 Foreign object

35 Nov. 1996 ANO 2 246 Axial Crack

36 June 1997 McGuire 2 250 ODSCC at TSP

37 Nov. 1997 Oconee 1 1514 2 Welded Plugs

38 Dec. 1998 Farley 1 341 2 Freespan Cracks

Steam generator tube leakageSteam generator tube leakageUSA NRC statistics 1990-2000USA NRC statistics 1990-2000

Known Steam Generator Tube Rupture Known Steam Generator Tube Rupture Accidents in the World Accidents in the World

1975-20021975-2002

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 leakageSteam generator tube leakageCrack Morphology and Leak RateCrack Morphology and Leak Rate

First leak at 17.2MPa

Maximum leak rate4.28 l/min at 34.5MPa

First leak at 25MPa

Maximum leak rate0.25 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 Wear degradation of steam generator tubsgenerator 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 forBurst characteristics foraxial notchesaxial 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:Cracks and Ruptures:Variety of Initial ConditionsVariety 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

…

1.0E-03

1.0E-02

1.0E-01

1.0E+00

1.0E+01

1.0E+02

1.0E+03

1.0E+04

1.0E+05

1.0E+06

0.1 1 10 100d, mm

ReEoWe

Vapor bubbles formation Vapor bubbles formation and transport phenomenaand transport phenomena

Shape and size of the bubbles:

Wecrit ~ 10 => dmax~10 mm

Eo(dmax)~10 => oblate ellipsoid

Beznosov et al, 2005

Steam Bubble Size DistributionSteam Bubble Size Distribution

Water: 22-24 MPa, 150-250 oC

14x2 mm tube

10 mm discharge

2000 mm depth

52 mm

Short wavelength due to high-pressure discharge.

2CAP g

Measured average velocity of a bubble ~0.3 m/s

Terminal speed of rising bubbles with dmax~10mm 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

Vapor bubbles formation Vapor bubbles formation and transport phenomenaand transport phenomena

Terminal velocity

0.00

0.10

0.20

0.30

0.40

0.50

0.60

0.70

0.80

0.1 1 10 100d, mm

m/s

Jamialahmadi

Mendelson

Lehrer

Mendelson:

Lehrer:

Size distributions of water droplets Size distributions of water droplets

Beznosov et al, 2005

Life time of small droplet Life time of small droplet on a hot surface on a hot surface

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

Time scale is ~10s of seconds for droplets ~1mm in diameter

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.

Vapor bubbles formation Vapor bubbles formation and transport phenomenaand transport phenomena

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

SummarySummary