0

Newest Issue


Editorial

ASME J of Nuclear Rad Sci. 2017;4(1):010202-010202-2. doi:10.1115/1.4038276.
FREE TO VIEW

The Reviewers of the Year Award is given to reviewers who have made an outstanding contribution to the journal in terms of the quantity, quality, and turnaround time of reviews completed during the past 12 months. The prize includes a Wall Plaque, 50 free downloads from the ASME Digital Collection, and a one year free subscription to the journal.

Commentary by Dr. Valentin Fuster
Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;4(1):010204-010204-2. doi:10.1115/1.4038274.
FREE TO VIEW
Commentary by Dr. Valentin Fuster

Guest Editorial

ASME J of Nuclear Rad Sci. 2017;4(1):010301-010301-2. doi:10.1115/1.4037556.
FREE TO VIEW
Commentary by Dr. Valentin Fuster

Research Papers

ASME J of Nuclear Rad Sci. 2017;4(1):011001-011001-15. doi:10.1115/1.4038162.
FREE TO VIEW

Two computational fluid dynamic (CFD) benchmarks have been performed to assess the prediction accuracy and sensitivity of CFD codes for heat transfer in different geometries. The first benchmark focused on heat transfer to water in a tube (first benchmark), while the second benchmark covered heat transfer to water in two different channel geometries (second benchmark) at supercritical pressures. In the first round with the experimental data unknown to the participants (i.e., blind calculations), CFD calculations were conducted with initial boundary conditions and simpler CFD models. After assessment against measurements, the calculations were repeated with the refined boundary conditions and material properties in the follow-up calculation phase. Overall, the CFD codes seem to be able to capture the general trend of heat transfer in the tube and the annular channel but further improvements are required in order to enhance the prediction accuracy. Finally, sensitivity analyses on the numerical mesh and the boundary conditions were performed. It was found that the prediction accuracy has not been improved with the introduction of finer meshes and the effect of mass flux on the result is the strongest among various investigated boundary conditions.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;4(1):011002-011002-8. doi:10.1115/1.4037807.
FREE TO VIEW

The thermal-hydraulics program in support of the development of the Canadian supercritical water-cooled reactor (SCWR) concept has undergone several phases. It focused on key parameters such as heat transfer, critical flow, and stability of fluids at supercritical pressures. Heat-transfer experiments were performed with tubes, annuli, and bundles in water, carbon dioxide (CO2), or refrigerant flows. Data from these experiments have led to enhancement in understanding of the phenomena, improved prediction methods, and verified analytical tools. In addition, these experiments facilitated the investigation of separate effects on heat transfer (such as geometry, diameter, spacing device, and transient). Chocking flow characteristics were studied experimentally with sharp-edged nozzles of two different sizes of opening. Experimental data have been applied in improving the critical-flow correlation in support of accident analyses. A one-dimensional (1D) analytical model for instability phenomena has been developed and assessed against the latest experimental data for quantifying the prediction capability and applicability.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;4(1):011003-011003-7. doi:10.1115/1.4037117.
FREE TO VIEW

Supercritical fluids (SCFs) become more and more important in various engineering applications. In nuclear power systems, SCFs are considered as coolant of the reactor core such as the supercritical water-cooled reactor (SCWR), superconducting magnets and blankets in the fusion reactors, or as fluid in the energy conversion systems of the next generation nuclear reactors. Accurate determination of heat transfer and the temperature of the structural material (e.g., fuel rod cladding) is of crucial importance for the system design. Thus, extensive studies on heat transfer to SCFs have been carried out in the past five decades and are still ongoing worldwide. However, no breakthrough is recognized or expected in the near future. In this paper, the status, main challenges, and future R&D needs are briefly reviewed. Three aspects are taken into consideration, i.e., experimental studies, numerical analysis, and model development for the prediction of heat transfer coefficient (HTC). Several key challenges and also the important subjects of the future R&D needs are identified. They are (a) data base for turbulence quantities, (b) multisolution of wall temperature, (c) extensive Reynolds-averaged Navier–Stokes (ERANS) method, and (d) new prediction method for HTC.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;4(1):011004-011004-14. doi:10.1115/1.4037720.
FREE TO VIEW

Canadian Nuclear Laboratories (CNL) has recently expanded the supercritical heat transfer (SCHT) databank with additional data provided by the Nuclear Power Institute of China (NPIC). These additional data cover flow conditions beyond the current databank, and are applicable for improving or validating existing correlations. The expanded databank comprises more than 41,000 points of heat-transfer measurements with different fluids flowing vertically upward in tubes, annuli, and bundles at supercritical (SC) pressures. It has been applied in assessing the prediction accuracy of 24 heat-transfer correlations, which were derived from experimental data obtained with water or nonaqueous fluids (such as carbon dioxide) flowing in tubes. For the correlation assessment, a sensitivity analysis has been performed by applying the measured wall temperature as an independent parameter. The assessment against the bundle data was based on cross-sectional-averaged flow conditions and the hydraulic diameter. The iterative approach (i.e., without prior knowledge of the wall temperature) overpredicted the wall temperature, which is conservative in safety analyses.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;4(1):011005-011005-7. doi:10.1115/1.4038061.
FREE TO VIEW

While supercritical water is a perfect coolant with excellent heat transfer, a temporary decrease of the system pressure to subcritical conditions, either during intended transients or by accident, can easily cause a boiling crisis with significantly higher cladding temperatures of the fuel assemblies. These conditions have been tested in an out-of-pile experiment with a bundle of four heated rods in the supercritical water multipurpose loop (SWAMUP) facility coconstructed by CGNPC and SJTU in China. Some of the transient tests have been simulated at KIT with a one-dimensional (1D) matlab code, assuming quasi-steady-state flow conditions, but time dependent temperatures in the fuel rods. Heat transfer at supercritical and at near-critical conditions was modeled with a recent look-up table of Zahlan (2015, “Derivation of a Look-Up Table for Trans-Critical Heat Transfer in Water Cooled Tubes,” Ph.D. dissertation, University of Ottawa, Ottawa, ON, Canada.), and subcritical film boiling was modeled with the look-up table of Groeneveld et al. (2003, “A Look-Up Table for Fully Developed Film Boiling Heat Transfer,” Nucl. Eng. Des., 225(1), pp. 83–97.). Moreover, a conduction controlled rewetting process was included in the analyses, which is based on an analytical solution of Schulenberg and Raqué (2014, “Transient Heat Transfer During Depressurization From Supercritical Pressure,” Int. J. Heat Mass Transfer, 79(12), pp. 233–240.). The method could well reproduce the boiling crisis during depressurization from supercritical to subcritical pressure, including rewetting of the hot zone within some minutes, but the peak temperature was somewhat under-predicted. Tests with a lower heat flux, which did not cause such phenomena, could be predicted as well. In another test with increasing pressure, however, a boiling crisis was also observed at a heat flux, which was significantly lower than the critical heat flux (CHF) predicted by the CHF look-up table of Groeneveld et al. (2007, “The 2006 CHF Look-Up Table,” Nucl. Eng. Des., 237(15–17), pp. 1909–1922.). The paper is summarizing the physical models and the numerical approach. Comparison with experimental data is used to discuss the applicability of the method for the design of supercritical water-cooled reactors (SCWR).

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;4(1):011006-011006-5. doi:10.1115/1.4038215.
FREE TO VIEW

Critical heat flux (CHF) experiment with uniform heating was performed in a tube of 8.2 mm in inner diameter and 2.4 m in heated length. The water flowed upward through the test section. The pressure covered the range from 8.6 to 20.8 MPa, mass flux 1157 to 3776 kg/m2s, inlet quality −2.79 to −0.08 (subcooling 19–337 °C), and local quality −0.97 to 0.53. For the pressure close to the near-critical point, the CHF decreased substantially with the pressure increasing. For the subcooling larger than a certain value, the CHF was related to the local condition. But for low subcooling and saturated condition, the CHF was related to the total power. The present results were in agreement with the previous experiment for the same local subcooled condition. Based on the present experimental results with subcooled and saturated conditions an empirical relation of the CHF was presented.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;4(1):011007-011007-14. doi:10.1115/1.4038161.
FREE TO VIEW

It is now well known that heat transfer to fluid at supercritical pressure in a confined channel shows complex behaviors. This is due to the strong variations of the thermal–physical properties resulting from the changes of pressure and temperature. To improve the reliability and efficiency of the supercritical water-cooled reactors (SCWRs) to be designed, the understanding of supercritical fluid flow in the fuel assemblies is very important. The study reported here reconsiders a simplified geometry made of a trapezoid channel enclosing an inner rod to simulate the triangular arrangement of a fuel assembly. Large eddy simulation (LES) with the wall adapting local eddy viscosity (WALE) model is used to simulate the forced convection flow in the channel. Supercritical water at 25 MPa is used as the working fluid. The Reynolds number of flow based on the hydraulic diameter and the bulk velocity is 10,540, while the heat flux from the inner rod wall has been varied from 10 kW/m2 to 75 kW/m2. Large unsteady flow structures are observed to be present due to the nonuniformity of the cross section of the flow channel. The characteristics of the flow structures and their effect on the local heat transfer are analyzed using instantaneous velocities, spectrum analysis, and correlation analysis. The swinging flow structures in the wide gap are much weaker than those in the narrow gap. The behaviors of such large flow structures are influenced by the strong spatial and temporal variations of the properties. When the temperature distribution follows Tb < Tpc < Tw, the mixing parameters due to the large flow structures, including mixing coefficient and effective mixing velocity in the gap, are also significantly influenced.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;4(1):011008-011008-7. doi:10.1115/1.4037747.
FREE TO VIEW

Within the Generation-IV International Forum, Canadian Nuclear Laboratories (CNL) led the conceptual fuel bundle design effort for the Canadian supercritical water cooled reactor (SCWR). The proposed fuel rod assembly for the Canadian SCWR design comprised of 64-elements with spacing between elements maintained using the wire-wrap spacers. Experimental data and correlations are not available for the fuel-assembly concept of the Canadian SCWR. To analyze the thermalhydraulic performance of the new bundle design, CNL is using computational fluid dynamics (CFD) as well as the subchannel approach. Simulations of wire-wrapped bundles can benefit from the increased fidelity and resolution of a CFD approach due to its ability to resolve the boundary layer phenomena. Prior to the application, the CFD tool has been assessed against experimental heat transfer data obtained with bundle subassemblies to identify the appropriate turbulence model to use in the analyses. In the present paper, assessment of CFD predictions was made with the wire-wrapped bundle experiments performed at Xi'an Jiaotong University (XJTU) in China. A three-dimensional CFD study of the fluid flow and heat transfer at supercritical pressures for the rod-bundle geometries was performed with the key parameter being the fuel rod wall temperature. This investigation used Reynolds-averaged Navier–Stokes turbulence models with wall functions to investigate the behavior of flow through the wire-wrapped fuel rod bundles with water subjected to a supercritical pressure of 25 MPa. Along with the selection of turbulence models, CFD results were found to be dependent on the value of turbulent Prandtl number used in simulating the experimental test conditions for the wire-wrapped fuel rod configuration. It was found that the CFD simulation tends to overpredict the fuel wall temperature, and the predicted location of peak temperature differs from the measurement by up to 65 deg.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;4(1):011009-011009-6. doi:10.1115/1.4037119.
FREE TO VIEW

Burnable poison (BP) is used to control excess reactivity in supercritical water cooled reactor (SCWR). It helps reduce the number of control rods. Over all BP designs, the design in which rare-earth oxide mixes with fuel is widely used in SCWR. BP has influence on fuel assembly neutronics performance. After comparing four kinds of rare-earth oxide, Er2O3 is chosen as BP for the annular fuel assembly. The effect of different BP loading patterns on assembly power distribution is analyzed. The safety of annular fuel assembly is estimated with different BP contents. Core performance with and without BP is compared. The results had shown that the core radial power peaking factor decreased after introducing BP. It was also shown that the core axial power peaking factor increased, and the power peak moved toward the top of the core. The reason of this effect was studied. Two optimizations were given based on this study: decreasing the temperature of lower plenum and increasing the gradients of axial enrichments. By applying these optimizations, core axial power peaking factor and maximum cladding surface temperature decreased.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;4(1):011010-011010-7. doi:10.1115/1.4037818.
FREE TO VIEW

The Canadian supercritical water-cooled nuclear reactor (SCWR) is a 2540 MWth channel-type SCWR concept that employs 336 fuel channels in the reactor core. Each fuel channel includes a pressure tube that is submerged in a heavy water moderator and contains a removable fuel assembly. The fuel assembly is designed so that all in-core components exposed to high radiation fields (other than the pressure tube) are part of the fuel assembly, which is removed from the reactor core as part of the assembly after three operating cycles. This design feature significantly reduces the likelihood of component failures due to radiation damage. To achieve high (>45%) power conversion efficiency, the Canadian SCWR operates at a supercritical water pressure (25 MPa) and high temperatures (350 °C at the inlet, 625 °C at the outlet). These conditions lead to fuel cladding temperatures close to 800 °C. Because of the reduced material strength at this temperature and higher fission gas production of the fuel, collapsible fuel cladding is selected over internally pressurized cladding. To increase heat transfer and to reduce cladding temperatures, turbulence-inducing wire-wraps are employed on fuel elements. Numerical models have been developed to analyze the thermal-structural behavior of Canadian SCWR fuel at normal and accident conditions. It was found that axial ridging, a possible failure mechanism with collapsed fuel cladding, can be avoided if the cladding thickness is larger than 0.4 mm. Detailed numerical analysis showed that the maximum fuel cladding temperature for the worst-case accident scenario is below the melting point by a small margin. This result was obtained with conservative assumptions, suggesting that the actual margin is greater. Hence, one of the design goals, the exclusion of the possibility of melting of the fuel, which is called the “no-core-melt” concept, seems attainable. However, this needs to be demonstrated more rigorously by removing the conservative assumptions in the analysis and performing supporting experimental work. This paper presents a description of the Canadian SCWR fuel assembly concept, its unique features, the rationale used in the concept development and the results of various numerical analyses demonstrating the performance and characteristics of the Canadian SCWR fuel channel.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;4(1):011011-011011-11. doi:10.1115/1.4037895.
FREE TO VIEW

The Canadian pressure-tube super critical water-cooled reactor (PT-SCWR) is an advanced generation IV reactor concept which is considered as an evolution of the conventional Canada Deuterium Uranium (CANDU) reactor that includes both pressure tubes and a low temperature and pressure heavy water moderator. The Canadian PT-SCWR fuel assembly utilizes a plutonium and thorium fuel mixture with supercritical light water coolant flowing through the high-efficiency re-entrance channel (HERC). In this work, the impact of fuel depletion on the evolution of lattice physics phenomena was investigated starting from fresh fuel to burnup conditions (25 MW d kg−1 [HM]) through sensitivity and uncertainty analyses using the lattice physics modules in standardized computer analysis for licensing evaluation (SCALE). Given the evolution of key phenomena such as void reactivity in traditional CANDU reactors with burnup, this study focuses on the impact of fission products, 233U breeding, and minor actinides on fuel performance. The work shows that the most significant change in fuel properties with burnup is the depletion of fission isotopes of Pu and the buildup of high-neutron cross section fission products, resulting in a decrease in cell k with burnup as expected. Other impacts such as the presence of protactinium and uranium-233 are also discussed. When the feedback coefficients are assessed in terms of reactivity, there is considerable variation as a function of fuel depletion; however, when assessed as Δk (without normalization to the reference reactivity which changes with burnup), the net changes are almost invariant with depletion.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;4(1):011012-011012-8. doi:10.1115/1.4037719.
FREE TO VIEW

The authors look for an attractive light water reactor (LWR) concept, which achieves high breeding performance with respect to the compound system doubling time (CSDT). In the preceding study, a high breeding fast reactor concept, cooled by supercritical pressure light water (Super FBR), was developed using tightly packed fuel assembly (TPFA) concept, in which fuel rods were arranged in a hexagonal lattice and packed by contacting each other. However, the designed concept had characteristics, which had to be improved, such as low power density (7.4 kW/m), large core pressure loss (1.02 MPa), low discharge burnup (core average: 8 GWd/t), and low coolant temperature rise in the core (38 °C). The aim of this study is to clarify the main issues associated with improvement of the Super FBR with respect to these design parameters and to show the improved design. The core design is carried out by fully coupled three-dimensional neutronics and single-channel thermal-hydraulic core calculations. The design criteria are negative void reactivity, maximum linear heat generation rate (MLHGR) of 39 kW/m, and maximum cladding surface temperature (MCST) of 650 °C for advanced stainless steel. The results show that significant improvement is possible with respect to the core thermal-hydraulic characteristics with minimal deterioration of CSDT by replacing TPFA with the commonly acknowledged hexagonal tight lattice fuel assembly (TLFA). Further design studies are necessary to improve the core enthalpy rise by reducing the radial power swing and power peaking.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;4(1):011013-011013-6. doi:10.1115/1.4037669.
FREE TO VIEW

An optimization design of China supercritical water-cooled reactor (SCWR) with the rated electric power of 1000 MWe (CSR1000) conceptual core is proposed. Steady-state performance of the proposed core is then studied with the SCWR core steady-state analysis code system SNTA. These key parameters such as burnup performance, reactivity control capability, power distribution, maximum fuel cladding temperature, and maximum linear power density are analyzed. The relative coolant flow rate of the second flow path, which is suited with assembly power, is also presented. The study shows that the refueling cycle of CSR1000 core can be extended effectively under the optimization design.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;4(1):011014-011014-6. doi:10.1115/1.4038060.
FREE TO VIEW

Because of the high temperature and high pressure characteristics of supercritical water-cooled reactor (SCWR), the thermal hydraulic performance of SCWR is greatly different from pressurized water reactor (PWR), which makes the current PWR fuel rod performance analysis codes are no longer applicable to SCWR. In this research, the irradiation swelling, irradiation densification, thermal expansion, thermal creep, plastic deformation, irradiation creep and irradiation hardening of UO2 pellet, and stainless steel cladding were considered; the gas conductance and radiant conductance of gap heat transfer were considered, the forced convective heat transfer on the outer surface of cladding was considered. Meanwhile, the irradiation effects and the thermal effects on the materials parameters such as thermal conductivity, specific heat, and young’s modulus were also considered in this research. With the help of abaqus software, the related user-defined subroutines were developed, and the irradiation effects and thermal effects of SCWR fuel were introduced into the numerical simulation, and then completed the analysis of SCWR fuel rods’ performance under steady power conditions. Some reference suggestions for the design and development of SCWR fuel could be provided by the establishment of this numerical simulation method.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;4(1):011015-011015-7. doi:10.1115/1.4037668.
FREE TO VIEW

Supercritical water-cooled reactor (SCWR) was chosen as Generation IV reactor concept in Canada to utilize Canada's expertise and technical capabilities from past research and designs. The conceptual design of Canadian SCWR has a core outlet temperature of 650 °C at 25 MPa and a peak cladding temperature as high as 800 °C. Corrosion/oxidation resistance is an important factor in material selections and also coating considerations. Most of the reported supercritical water (SCW) test data have been obtained at temperatures up to 700 °C as no autoclave exists that can operate above 700 °C at supercritical pressures and temperatures. Superheated steam (SHS) is used as a surrogate fluid to SCW in this study to evaluate two coating materials, FeCrAlY and NiCrAl, at 800 °C. The results showed that the FeCrAlY became discolored rapidly while NiCrAl still maintained some metallic sheen after 600 h. The weight change results suggest that more oxide formation took place on FeCrAlY than NiCrAl. In particular, grain boundary oxide (Al2O3) formed on FeCrAlY surface upon exposure to steam after 300 h. Further exposure caused more intragranular Al2O3 to form, in addition to magnetite formation on the grain boundary regions. For NiCrAl samples, NiO formed after steam exposure for 300 h. Spinel and (Cr,Al)2O3 were also found after 300 h along with very limited amount of Al2O3. After 600 h, Al2O3 became well developed on NiCrAl and the coverage of spinel and Cr2O3 on the surface reduced.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;4(1):011016-011016-7. doi:10.1115/1.4037897.
FREE TO VIEW

The presented work consists of a test setup study of a new pneumatic material testing device based on double bellows (DBs) loading device and with miniature autoclaves enabling applications at temperature and pressure up to 650 °C and 35 MPa, respectively. It has been demonstrated that it is technically feasible to carry out well defined and controlled material testing in the supercritical water (SCW) environment using this testing system. By using this type of system, it makes possible to investigate the intrinsic role of the applied stress on the deformation behavior of material in light water reactor (LWR) conditions and also in other harsh environments like SCW conditions. In addition, the compactness and versatility of the setup makes this system particularly attractive for deployment in a hot-cell for testing of irradiated materials.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;4(1):011017-011017-5. doi:10.1115/1.4037324.
FREE TO VIEW

Concerns with greenhouse gas emissions and the uncertainty of long-term supply of fossil fuels have resulted in renewed interest in nuclear energy as an essential part of the energy mix for the future. Many countries worldwide including Canada, China, and EU are currently undertaking the design of generation IV supercritical water-cooled reactor (SCWR) with higher thermodynamic efficiency and considerable plant simplification. The identification of appropriate materials for in-core and out-of-core components to contain the supercritical water (SCW) coolant is one of the major challenges for the design of SCWR. This study is carried out to evaluate the oxidation/corrosion behaviors of bare alloy 214 and NiCrAlY coated 214 under SCW at a temperature of 700 °C/25 MPa for 1000 h. The results show that chromium and nickel based oxide forms on the bare surface after exposure in SCW for 1000 h. A dense and adhered oxide layer, consisting of Cr2O3 with spinel (Ni(Cr, Al)2O4), was observed on NiCrAlY surface after 1000 h in SCW.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;4(1):011018-011018-5. doi:10.1115/1.4037896.
FREE TO VIEW

The supercritical carbon dioxide (S-CO2) Brayton gas turbine cycle has been studied as an efficient and cost-effective option for advanced power systems. One major safety issue for any power cycle is a pipe break and the associated discharge of the working fluid and subsequent decrease in system pressure. In this paper, an S-CO2 critical flow in the nozzle tube is analyzed numerically with fluent 15.0. The Redlich–Kwong real gas equation is selected to calculate carbon dioxide density and the standard k-epsilon turbulence model is selected. Experimental data are used as a benchmark to examine the capability of the current approach. Compared with experimental data, the simulation results overestimate the critical mass flux; the error range is between 15% and 25%. The simulation results show that as L/D increases, critical mass flow decreases. As stagnation temperature increases, critical mass flow decreases. The complex thermal hydraulic behavior in the nozzle tubes is analyzed. Three flow patterns in the nozzle tube during transient critical flow are obtained and discussed. From inlet to outlet of the tube, CO2 may undergo the following phases in turn: (1) supercritical phase; (2) supercritical phase—gas phase; (3) supercritical phase—gas phase—liquid phase. The simulation results are also helpful for further experimental and theoretical research.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;4(1):011019-011019-5. doi:10.1115/1.4037718.
FREE TO VIEW

In this study, an Al-containing alloy 214 was evaluated in superheated steam at 800 °C for a duration of 600 h. The purpose of using superheated steam was to simulate the supercritical water (SCW) condition at higher temperatures where no commercial SCW rig is currently capable of reaching (800 °C and beyond). After exposure to superheated steam, the weight change and surface oxidation were analyzed. Alloy 214 experienced greater weight gain than IN 625 and Ni20Cr5Al, due to its low Cr content. Formation of both Cr2O3 and Al2O3 was observed on the surface after 300 and 600 h of exposure. However, as exposure progressed, more Ni and Mn-containing spinel started to form, signaling Cr and Al depletion on the metal substrate surface.

Commentary by Dr. Valentin Fuster

Sorry! You do not have access to this content. For assistance or to subscribe, please contact us:

  • TELEPHONE: 1-800-843-2763 (Toll-free in the USA)
  • EMAIL: asmedigitalcollection@asme.org
Sign In