Table Of ContentSUPERCRITICAL WATER NUCLEAR STEAM SUPPLY SYSTEM:
INNOVATIONS in MATERIALS, NEUTRONICS & THERMAL-HYDRAULICS
DE – FG03 – 01SF22328
Nuclear Energy Research Initiative Project 2001 - 091
Final Report for 3 Year Grant (August 2001 - September 2004)
PRINCIPAL INVESTIGATORS:
M. Anderson M. L. Corradini K. Sridharan P. Wilson
Phone: 608-263-1646; Fax: 608-263-7451 [email protected]
COLLABORATING ORGANIZATIONS:
D. Cho, T.K.Kim, S.Lomperski; Argonne National Laboratory
Phone: 630-252-4595; 630-252-7981
ANNUAL REPORT EXECUTIVE SUMMARY
In the 1990’s supercritical light-water reactors were considered in conceptual designs. A nuclear
reactor cooled by supercritical water would have a much higher thermal efficiency with a once-
through direct power cycle, and could be based on standardized water reactor components (light
water or heavy water). The theoretical efficiency could be improved by more than 33% over that
of other water reactors and could be simplified with higher reliability; e.g., a boiling water reactor
without steam separators or dryers.
Research Objectives and Summary: To make such a system technologically feasible, advances
are required in high-temperature materials with improved corrosion and wear resistance (cladding
and pressure structural boundaries), in neutronics to improve fuel-cycle versatility with these
advanced materials as well as in neutronics and thermal-hydraulics to insure efficient heat
removal and passive safety and stability. Our research objectives are:
• Employ innovative plasma-based surface modification techniques to improve material
compatibility under supercritical conditions. These techniques are being applied to
cladding and structural materials with proven bulk properties, with the goal of mitigating
surface-initiated degradation phenomena of corrosion, oxidation and wear under
supercritical thermal-hydraulic conditions. Test results indicate a structured oxide layer
that seems to be similar after 100 hours to over years. Plasma surface modification has
been shown to reduce this oxide layer substantially given a base alloy type.
• Neutronics analyses would identify ranges of alternative fuel cycles, including variations
in enrichment, refueling schedules, recycling and conversion/breeding. These analyses
have indicated that a mixed-spectrum reactor with annular core design holds promise as a
reactor system that can meet required performance features and burn its actinide waste.
• Thermal-hydraulic studies focused on SC heat transfer and loop flow stability associated
with large density changes for natural and forced circulation of supercritical water.
Scaled CO simulant experiments and associated model development have identified
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regimes of natural convection loop stability and provide us with a basis for further work.
TASK I - SCWR Cladding Materials: Supercritical Water Corrosion Loop,
Materials Testing in Supercritical Water Environment, and Materials Analysis
The concept of using supercritical water-cooling for next generation of nuclear
reactors has gained considerable momentum in recent years. Among the reactor systems
recommended by the U.S. Department of Energy’s Generation IV reactor program, the
supercritical-water-cooled reactor (SCWR) system has been highly ranked because of
improved economics. The improved economics are due to high thermal efficiency and
plant simplification supported by the unchanged phase of the coolant in the reactor.1
Consortiums consisting of industry, universities, and national laboratories in the United
States, Canada, Europe, Japan, and Korea have been actively collaborating on
investigations of issues critical to the development supercritical water cooled nuclear
reactors, such as thermal hydraulics, neutronics, materials, and reactor design.2-5
Supercritical water (water at or above 374°C and 22.1 Mpa) cooling is presently used
in a number of fossil power plants, providing enhanced efficiency, improved fuel usage
and reduction in the emissions of carbon dioxide, nitrogen oxides, and sulfur oxides.6,7 To
obtain higher efficiencies, modern fossil power plants are utilizing supercritical water at
temperatures in excess of 600°C and pressures as high as 34.5 MPa.8,9 Under these
aggressive conditions materials degradation becomes an important source of concern.
Creep, fatigue, and corrosion have been identified to be significant materials’ degradation
mechanisms.10 Corrosion in particular has been identified as a critical problem because
the temperature and the oxidative nature of supercritical water increases the kinetics of
corrosion mechanisms. For example, heavy iron oxide (magnetite) generation inside the
boiler tubes of fossil plants operating at supercritical temperature has been identified as
the dominant fouling mechanism.2
Austenitic stainless steels and ferritic/martensitic steels and nickel-based alloys have
been widely used in supercritical water-cooled fossil power plants. Austenitic steels and
nickel-based alloys such as 304H, Inconel 617, and Inconel 625 exhibit superior high
temperature creep characteristics and have been used for components such as superheater
and reheater tubes where temperatures exceed 620°C.11
Ferritic/martensitic steels containing 9-12%Cr (e.g., HCM12A, T91, T92, T122),
have been used especially in thick sections such as headers and steam pipes where higher
thermal fatigue and thermal shock resistance are required.12 Ferritic steels with tungsten
additions have been considered as candidate materials for ultra-super-critical water fossil
power plant because of their high creep rupture strength.13
Austenitic stainless steels, such as 316L have been shown to exhibit good overall
corrosion resistance in deionized water at temperatures of 300-500°C, however localized
corrosion such as pitting (at 300°C) or crevice corrosion (at 500°C) have been observed.
Generally, for all alloys corrosion resistance has been noted to improve with increasing
Cr content.14-16 Multi-phased nickel-based-superalloys such as Inconel 625 showed a
tendency for the formation of a protective film when exposed to deionized water between
450°C and 500°C, however minor pit development and grain boundary attack have been
observed.17 A review of the evolution of ferritic and austenitic steels and their
applications for power plant components is presented in references 11and 18.
The presence of irradiation and its associated effects may enhance materials
degradation processes such as corrosion, further increasing the importance of materials
design and selection for supercritical water-cooled nuclear reactors. A recent workshop
on advanced reactor materials,19 identified the most promising materials for SCWR
system to be ferritic and martensitic steels (e.g., T122, T91, NF616, 9Cr-2WVTa), oxide-
dispersion strengthened (ODS) alloys (e.g., MA957), nickel-based alloys (e.g., Inconel
625, 690, 718), and austenitic steels (e.g., SS316LN).3 Ferritic steels were considered to
be promising because of earlier studies demonstrating their superior swelling resistance
compared to austenitic stainless steels.20
This work reports the results of corrosion of four candidate alloys in supercritical
water. 316 stainless steel, 347 stainless steel and Inconel 718 are candidates for higher
temperature SCW core components, while Zircaloy-2 is a candidate for the SCWR water
rod box where inner box temperatures are expected to range from 280 to 300°C. All of
the alloys, but stainless steel 347 were exposed to supercritical water at the University of
Wisconsin while the stainless steel 347 was exposed for about 30 years at the Genoa 3
supercritical water fossil power plant.
EXPERIMENTAL
Supercritical Water Corrosion Loop
The natural circulation supercritical water corrosion loop designed and constructed at
the University of Wisconsin is rectangular, 3 meters tall and 2 meters wide with a
maximum power input of 100kW. The main loop is constructed with Inconel 625, which
allows for operation within ASME specifications at temperatures of up to 600°C at
pressures of 25 MPa. The loop has two heater sections on the bottom and on the side of
the loop. To transfer the heat to the supercritical water, molten lead is used as a heat
transfer fluid, which surrounds the inner 2” Inconel tube and encases Chromolux
immersion heaters, which can provide heat fluxes of up to 800 kW/m2 to the inner pipe.
There are two cooling sections on the top horizontal section and the left vertical section
that remove the heat from the loop. This results in a change in density of the fluid to drive
a natural circulation flow velocity up to 1 m/s.
Water chemistry in the loop is controlled by extracting water from the loop, running it
through a chemistry system and re-injecting the water. The chemistry loop contains a
filter for controlling conductivity, a permeable membrane for extracting dissolved
oxygen, and measuring instruments for monitoring chemistry. A variety of Siemens
pressure transmitters (with an error less than 0.1% of the measured value) are being used
throughout the system. Approximately 50 E-type thermocouples are being used
throughout the loop to measure water, lead, and piping temperatures. LabView software
is being used for data acquisition of all temperatures, pressures and heater states. Figure
1 shows the output from the online diagnostics for monitoring conductivity, dissolved
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oxygen content, temperature, and pressure during a 7-day run at 400 C. Following start-
up, system control is maintained within a fine tolerance band.
Materials
The nominal composition of the alloys used for this research is as follows (in weight %):
1. 316 austenitic stainless steel: 16-18% Cr, 10-14% Ni, 1.75-2.50% Mo, 0.08%C, 2%
Mn, 1.0% Si, 0.2% P, 0.1% S, and balance Fe
2. 347 austenitic stainless steel: 17-19% Cr, 9-13% Ni, 2% Mn, 1.0% Si, 0.04-0.1% C,
0.04% P, 0.03% S, 1% (Nb+Ta), and balance Fe
3. IN 718: 52.80% Ni, 18.63% Fe, 18.42%Cr, 2.90% Si, 0.56% Al, 5.12% Nb, 0.1% Si,
1.01%%Ti, 0.19% Co, 0.04% C, 0.07% Mn, 0.04% Cu, 0.02% Ta, 0.007% P, 0.004% B
4. Zircaloy-2: 1.32% Sn, 0.18% Fe, 0.10% Cr, 0.12%O, and balance Zr
Materials Characterization
Surface examination of the samples after testing in supercritical water was performed
using scanning electron microscopy (SEM). The objective of this examination was to
quantify features of corrosion typical to each type of alloy such as pitting and fissures,
particulate formation, and preferential attack of microstructural features. The Energy
Dispersive Spectroscopy (EDS) capability of the SEM was used for chemical analysis of
regions of attack or particulates. In addition, scanning electron microscopy was
performed after mounting the samples in cross-section to examine and measure the
thickness and composition of the oxide layer formed on the surface, as well as the
structure of the oxide and its adhesion to the substrate alloy.
Elemental composition as a function of depth below surface of thin oxide film
corrosion product was also performed with Auger Electron Spectroscopy (AES). The
most important information to be acquired from AES measurements was the thickness of
the oxide film as inferred from the tailing off of the oxygen content as the sputter depth
approaches the oxide-alloy interface. AES also provides information on oxide
composition and stoichiometry.
RESULTS AND DISCUSSION
Corrosion Testing in Supercritical Water Corrosion Loop
Surface examination of the samples after exposure to supercritical water for periods
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ranging from 3 to 7 days at 400 C and 500 C showed that each alloy acquired certain
characteristic corrosion features (Figures 2 and 3). For Zircaloy-2, fissures and
particulates were observed, and their size was noted to increase with duration of exposure
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to supercritical water. Typically, the fissure size after 7 days at 400 C was in the 5 to
10µm size range. Particulates were analyzed by EDS and noted to be predominantly Zr-
and Sn-oxide. Stainless steel developed sub-micron sized pits after 3 days exposure at
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400 C. Fine particulate debris observed in this case was identified to be oxides of Fe and
Cr. For Inconel 718, high magnification imaging showed strong evidence of preferential
attack at the niobium-rich intermetallic precipitates. This observation suggests that multi-
phased Fe-, Ni- based alloys may be prone to corrosion attack despite their high elevated
temperature strength. In general, the 316 austenitic stainless steel exhibited the least
corrosion damage.
Cross-sectional examination after mounting the samples in a conductive mount was
performed with the goal of examining the oxide film thickness. Figure 4 shows a cross-
sectional view of the three alloys after exposure to supercritical water for 7 days at
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400 C. The Zircaloy-2 exhibited a distinct oxide 1 to 1.5µm thick, whereas the Inconel
718 showed surface perturbations due to pitting. Stainless steel showed no visible
oxidation or pitting effects. For samples exposed to supercritical water for 7 days at
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500 C, cross-sectional examination showed a pronounced growth of the oxide film for
the Zircaloy-2, with the oxide film thickness being in the range of 6 to 8µm. A very thin,
nearly indistinguishable oxide film was observed on stainless steel. These observations
clearly point to the unsuitability of zirconium alloys in the supercritical water at
temperatures of 400°C or greater, at least from the standpoint of corrosion. However,
zirconium alloy still proves to be adequate for the lower temperature water rod box
application. Austenitic stainless steel on the other hand exhibited good corrosion
resistance in supercritical water environment.
For corrosion resistant alloys where the oxide thickness is typically on the order of
fractions of a micron, the SEM cross-sectional approach was deemed unsuitable because
of edge retention effects during sample preparation. For these samples surface sensitive
Auger Electron Spectroscopy (AES) was performed for the determination of the
composition versus depth profile by using the sequential sputtering and chemical analysis
capability of the AES technique. Results of AES analysis for 316 austenitic stainless
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steel samples in the untested condition and after exposure to supercritical water at 300 C
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and 500 C for 7 days are shown in Figure 5. The tailing of the oxygen peak and the
upsurge in the elemental peaks of the substrate provide a measure of the oxide thickness.
The original output of AES analysis yields composition as a function of sputter time.
However, the sputter time was correlated to the thickness of the oxide film, by
conducting profilometry of the sputter crater produced for each alloy for a predetermined
sputter time. AES analysis also shows that when 316 stainless steel is exposed to
supercritical water a stoichiometric oxide of predominantly Fe (and some Cr) forms at the
surface (figs. 5a and 5b). An oxygen diffusion profile is observed below this oxide.
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Furthermore, the oxide film grows to a thickness about twenty times greater at 500 C
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compared to 300 C, during a 7-day exposure period.
Long-term Corrosion Testing
A variety of components that were subjected to prolonged exposure (~30 years) to
supercritical water under various conditions have been procured from Genoa 3
Supercritical fossil power plant located in Genoa, WI. The availability of these
components provided a unique opportunity to examine the long-term effects of corrosion
in materials when exposed to supercritical water over time periods that could not be
simulated in laboratory tests. In addition to surface corrosion, these samples also
provided an opportunity to gain an insight into the changes in the bulk microstructure of
the alloys from thermal effects after prolonged holding times.
An AISI 347 stainless steel superheater pendant was exposed to supercritical water
since the inception of the plant in 1969 and was removed for testing in the spring of 2003.
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The component was exposed to supercritical water at a temperature of 900F (482 C), and
a pressure of 3600psi (24.82Mpa).
Fig. 6a shows cross-sectional microstructure of the surface oxide corrosion product
and a portion of the base alloy. Figures 6b and 6c show EDS line-scan analysis
performed between points 1 and 21 and selective EDS analysis from point 16 to 21 on the
micrograph shown in Fig. 6a, respectively. Cross-sectional microstructure (Fig. 6a)
clearly shows that an oxide layer in excess of 100µm developed on the surface as a result
of corrosion. The oxide layer exhibits porosity and the pore density in the oxide film
varies across the oxide film thickness. The pore density is particularly high near the
alloy-oxide interface and some delamination of the oxide is also observed at this
interface.
Figure 6b shows the oxide to be predominantly that of Fe and Cr, except at the oxide
surface, which is depleted of Cr and consists of predominantly Fe-oxide. Although, the
Si content of the alloy is quite low, a marked upsurge in Si concentration is observed at
the alloy-oxide interface. Fig. 6a also shows grain boundary coarsening and attack in the
bulk alloy near the oxide-alloy interface due to corrosion. Elemental analysis from points
16 to 21 (fig. 6c, cutting across the grain boundaries) shows an upsurge of Cr content at
the grain boundaries, possibly due to diffusion of Cr to the grain boundaries during the
prolonged thermal exposure. However, a similar profiling for an untested 347 stainless
steel will have to be performed to conclusively support this proposal. Fig. 6d shows a
lower magnification micrograph where a comparison can be made between the grain
boundary structure of the base alloy near the oxide-alloy interface with those farther in
the interior of the base alloy. Preferential corrosion attack of the grain boundaries near
the oxide-alloy interface is clearly evident.
Figure 6e shows a different type of oxide structure observed in other regions of the
same 347 stainless steel sample. Here a densification of oxide appears to have occurred
in the vicinity of the alloy-oxide interface and a high concentration of pores is observed
further away from the interface. As with the image shown in figure 6a, the Si
concentration was noted to be elevated in the region of increased porosity. The origin of
the distribution of porosity in the oxide film is presently being investigated.
CONCLUSIONS
A natural circulation supercritical water corrosion loop with a maximum power input
of 100kW has been designed and built at the University of Wisconsin. Surface analysis
has been performed of 316 austenitic stainless steel, Inconel 718, and Zircaloy-2 after
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corrosion testing in this supercritical water loop at temperatures of 300 C to 500 C for
exposure periods of up to 7 days. For Zircaloy-2 surface fissures, a few microns in size,
developed in the initial stages of testing and at higher temperatures an oxide film several
microns thick developed on the surface. For Inconel 718, only a sub-micron thick oxide
film formed, however substantial pitting and grain boundary attack was observed
particularly in the niobium-rich regions. The 316 austenitic stainless steel exhibited the
best corrosion resistance with very marginal pitting and only a sub-micron thick oxide
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film forming on the surface even when exposed to 500 C for 7 days. Cross-sectional
examination of AISI 347 stainless steel exposed to supercritical water for period of about
30 years showed an oxide layer over a 100µm in thickness. The oxide exhibited
structural and compositional variations across its thickness. Such long-term exposure
may have also resulted in Cr diffusion to the grain boundaries in the bulk alloy. The
near-surface microstructure of the bulk alloy clearly showed preferential corrosion attack
at the grain boundaries.
TASK I REFERENCES
1. A Technology Roadmap for Generation IV Nuclear Energy Systems, U.S. DOE
Nuclear Energy Research Advisory Committee and the Generation IV International
Forum, December (2002).
2. K.A. BURRILL, “Water Chemistries and Corrosion Product Transport in Supercritical
Water in Reactor Heat Transport Systems”, Proc. Water Chemistry of Nuclear Reactor
Systems 8, British Nuclear Energy Society, Bournemouth, UK, Oct. 22-26, (2000).
3. B. CORWIN, L. Mansur, R. Nanstad, A. Rowcliffe, B. Swindeman, P. Tortorelli, D.
Wilson, I. Wright, “Materials Issues,” Supercritical Water Reactor Review Meeting,
Madison, Wisconsin, April 30, (2003).
4. J. MCKINLEY, S. Teysseyre, G.S. Was, D.B. Mitton, H. Kim, J.K. Kim, R.M.
Latanision, “Corrosion and Stress Corrosion Cracking of Austenitic Alloys in
Supercritical Water,” GENES4/ANP2003, Kyoto, Japan, Paper 1027, September (2003).
5. Y. TSUCHIYA, F. Kano, N. Saito, A. Shioiri, S. Kasahara, K. Moriya, H. Takahashi,
“SCC and Irradiation Properties of Metals under Supercritical-Water Cooled Power
Reactor Conditions,” GENES4/ANP2003, Kyoto, Japan, Paper 1096, Sep. 15-19 (2003).
6. A.F. ARMOR, G.T. Preston, “The Impact of Fossil Generation Advances on the
Emissions of CO in the United States,” Energy Conversion and Management, 37, 6-8, p.
2
671 (1996).
7. J.P. LONGWELL, E.S. Rubin, J. Wilson, “Coal: Energy for the Future,” Progress in
Energy and Combustion Science, 21,4, p. 269 (1995).
8. K. NATESAN, A. Purohit, D.L. Rink, “Fireside Corrosion of Alloys for Combustion
Power Plants,” Power Plant Chemistry, 4, 9 (2002).
9. F. MASUYAMA, “History of Power Plants and Progress in Heat Resistant Steels”,
ISIJ International, 41, 6 p. 621 (2001).
10. M. BETHMONT, “Damage and Lifetime of Fossil Power Plant Components,”
Materials at High Temperatures, 15, 3/4 p. 231 (1998).
11. R. VISWANATHAN, W.T. Bakker, “Materials for Boilers in Ultra Supercritical
Power Plants,” Proc. Intl. Joint Power Generation Conference (IJPGC2000-15049),
Miami Beach, Florida, July 23-26, (2000).
12. Z. KLENOWICZ, K. Darowicki, “Waste Incinerators: Corrosion Problems and
Construction Materials – Review,” Corrosion Reviews, 19, 5-6, p. 467 (2001).
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High Strength 9 to 12% Chromium Containing Creep Resistant Steels, Creep and
Fracture of Engineering Materials and Structures,” Key Engineering Materials, 171, 1 p.
427 (2000).
14. D.B. MITTON, N. Eliaz, J.A. Cline, R.M. Latanision, “Assessing Degradation
Mechanisms in Supercritical Water Oxidation Systems,” Corrosion, paper 01352, NACE
Intl., Houston, TX (2001).
15. D.B. MITTON, J.H. Yoon, H.S. Kim, N. Eliaz, and R.M. Latanison “Corrosion
Behavior of Nickel-Based Alloys in Supercritical Water Oxidation Systems,” Ind. Eng.
Chem. Res., 39 p. 4689 (2000).
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Understanding of Corrosion in Supercritical Water Oxidation Systems for the Destruction
of Hazardous Waste Products,” Mater. Tech., 16, p. 44 (2001).
17. N. ELIAZ, D.B. Mitton, R.M. Latanision, “Review of Materials Issues in
Supercritical Water Oxidation Systems and the Need for Corrosion Control,” Trans.
Indian Inst. Metals, 56, 3, p. 305 (2003).
18. R. VISWANATHAN, W.T. Bakker, “Materials for Ultra Supercritical Fossil Power
Plants,” Report TR-114750, EPRI, Palo Alto, January (2000).
19. T.R. ALLEN, Workshop on Higher Temperature Materials for Advanced Nuclear
Energy Systems, DOE Office of Nuclear Energy, Science and Technology, La Jolla, CA,
March 18 (2002).
20. D.R. HARRIES, Proc. Topical Conf. on Ferritic Steels for Use in Nuclear Energy
Technologies, eds. J.W. Davis and D.J. Michel (The Metallurgical Society of AIME,
Warrendale, PA, p. 141 (1984).
Fig. 1. Output from online diagnostics in the University of Wisconsin Supercritical Water
corrosion loop for monitoring conductivity, dissolved oxygen, pressure, and temperature
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of supercritical water (taken during a 7-day run at 400 C).
(a) 3000 X (b) 2000 X (c) 2000 X
Fig. 2. Nucleation of corrosion features on the surface of (a) Zircaloy-2, (b) 316 stainless
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steel, and (c) Inconel 718 alloys, after exposure to supercritical water at 400 C for 3 days.
(a) 400 X (b) 400 X (c) 400 X
Fig. 3. Scanning electron microscopy of the surface of (a) Zircaloy-2, (b) 316 stainless
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steel, and (c) Inconel 718 alloys after exposure to supercritical water at 400 C for 7 days.
(a) 4000 X (b) 4000 X (c) 2000 X
Figure 4. Cross-sectional SEM view of the alloys after exposure for 7 days in supercritical water
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at 400 C showing, (a) a well-developed oxide film for the zirconium alloy, (b) no visible oxide
for stainless steel, and (c) surface perturbations due to pitting for Inconel 718.
