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Comprehensive nuclear materials 5 11 material performance in helium cooled systems Comprehensive nuclear materials 5 11 material performance in helium cooled systems Comprehensive nuclear materials 5 11 material performance in helium cooled systems Comprehensive nuclear materials 5 11 material performance in helium cooled systems Comprehensive nuclear materials 5 11 material performance in helium cooled systems
5.11 Material Performance in Helium-Cooled Systems R Wright and J Wright Idaho National Laboratory, Idaho Falls, ID, USA C Cabet Commissariat a l’Energie Atomique, Gif-sur-Yvette, France ß 2012 Elsevier Ltd All rights reserved 5.11.1 Introduction 252 5.11.2 5.11.3 5.11.3.1 5.11.3.2 5.11.3.3 5.11.3.4 5.11.3.5 5.11.3.6 5.11.3.7 5.11.3.8 5.11.4 5.11.4.1 5.11.4.2 5.11.4.3 5.11.4.4 5.11.4.5 5.11.5 5.11.5.1 5.11.5.2 5.11.5.3 5.11.6 5.11.7 5.11.8 5.11.9 5.11.10 References Experience with VHTR Systems Comparison of IHX Concepts Shell-and-Tube Plate and Fin Etched Plate Microchannel Heat Exchangers Plate-Stamped Heat Exchanger Foam IHX Capillary IHX Ceramic IHX Heat Exchanger Alloys Regulatory Issues Alloy 617 (52Ni–22Cr–13Co–9Mo) Alloy 230 (57Ni–22Cr–14W–2Mo–La) Alloy 800H (42Fe–33Ni–21Cr) Alloy X (47Ni–22Cr–9Mo–18Fe) Welding Base Metal Preparation and Filler Metal Selection Preheating, Interpass Temperatures, and Postweld Heat Treatment Nontraditional Joining Methods Control Rod Materials Core Barrel Materials Environmental Effects of VHTR Atmospheres on Materials Aging Effects Summary 252 254 254 255 255 256 256 257 257 257 257 259 259 262 263 264 264 265 265 265 266 269 269 273 275 276 Abbreviations AGCNR AVR DLOC GE GMAW GTAW HE HTGRs HTR-10 Advanced gas-cooled nuclear reactor Arbeitsgemeinschaft Versuchsreaktor Depressurized loss of coolant General Electric Gas-metal-arc welding Gas-tungsten-arc welding Heat exchanger High-temperature gas reactors High-temperature reactor-10MWth in China HTTR IHX INL NGNP ODIN ORNL PBMR PCHE PCS High-temperature engineering test reactor in Japan Intermediate heat exchanger Idaho National Laboratory Next generation nuclear plant Online Data & Information Network Oak Ridge National Laboratory Pebble bed modular reactor Printed circuit heat exchanger, Heatrics Division Ltd Power conversion system 251 252 Material Performance in Helium-Cooled Systems PFHE PSHE RCS RCSS RSS SMAW THTR VHTR Plate and fin heat exchanger Plate stamped heat exchanger Reactivity control system Reactor control and shutdown system Reserve shutdown system Shielded metal arc welding Thorium Hochtemperatur Reaktor Very high-temperature reactor 5.11.1 Introduction Over the past decade, there has been renewed interest in very high-temperature reactor (VHTR) technology This type of reactor is of interest because of a number of unique characteristics, including passive safety, electricity production on a more modest scale compared to light water plants that might be more compatible with the electrical distribution system in developing countries, and very high outlet temperature that can be used for process heat or hydrogen production The relative value of electricity production or process heat applications varies considerably with world economic conditions Currently, it appears that steam for process heat and hydrogen production will drive development of this technology rather than electricity production There are currently two operating VHTR prototypes, the high-temperature engineering test reactor (HTTR) in Japan and the high-temperature reactor (HTR-10) in China The HTTR is a 30 MWt (megawatts thermal output) prismatic core reactor and the HTR-10 is a 10 MWt pebble bed prototype reactor Both the operating reactors are designed to investigate electricity production with the VHTR technology; however, each program has parallel activities to develop process heat and hydrogen production as well The challenges for high-temperature materials are not significantly different for either prismatic or pebble bed reactor designs Interest in specific applications for VHTR technology is evolving rapidly It appears that the most significant immediate interest is in a reactor with an outlet temperature on the order of 750 C with a steam generator for either electricity generation or process heat This technology would use a relatively mature conventional steam generator technology and is expected to present lower technical risk Higher outlet temperatures using a heat exchanger between the primary helium coolant and a secondary gas are viewed to be higher risk development projects that offer the opportunity for outlet temperatures from 850 to 950 C for hydrogen production by thermochemical processes or higher temperature process heat for industrial applications The material issues associated with reactor internals are not affected significantly by the reactor outlet temperature; however, heat exchangers operating at the higher outlet temperatures represent significantly different issues compared to steam generators The focus of this chapter is on higher outlet temperature systems because of the development challenges The next generation nuclear plant (NGNP) being developed in the United States is one particular VHTR concept that is under very active development and is typical of the development around the world This reactor is being developed to produce hydrogen as well as electricity Conceptual designs call for a gascooled reactor with an outlet temperature greater than the 850 C required to efficiently operate the hydrogen generation plant, with a maximum of 950 C While the design concepts are not yet final, it is highly probable that helium will be the primary coolant in the reactor The primary material in the core will be graphite, and the prime candidates for high-temperature metallic components are the nickel-based alloys Alloy 617 or Alloy 230 An artist’s representation of one concept for the reactor and power conversion vessel and the associated hydrogen generation plants is shown in Figure In this representation, a heat exchanger carries most of the reactor thermal output to a secondary circuit that powers a turbine for electricity generation An additional heat exchanger takes $10% of the thermal energy of the reactor and diverts it as process heat to the hydrogen production plant The most critical metallic component in the VHTR system is the intermediate heat exchanger (IHX) This heat exchanger will operate at a reactor outlet temperature of up to 950 C In addition, the reactor system is intended to have a license period of 60 years The combination of very high-temperature operation and long duration of service restricts material choices for the heat exchanger to a small number of coarse-grained solid-solution strengthened alloys that provide stability and creep resistance and have high chromium content for environmental resistance 5.11.2 Experience with VHTR Systems Very early in the development of nuclear power for electricity generation or process heat, the concept of an inert gas-cooled, high-temperature reactor was explored Material Performance in Helium-Cooled Systems 253 Power conversion unit Intercooler Generator Turbine Low-pressure compressor Primary heat rejection High-pressure compressor Recuperator Commercial power Blower Power for electrolysis Pump Heat exchanger Pebble-bed or prismatic reactor Blower Heat exchanger Hydrogen production (electrolysis) Hydrogen Hydrogen production (thermochemical) Hydrogen Heat exchanger Figure An artist’s conception of a very high-temperature gas-cooled reactor and associated hydrogen production plants Table Design characteristics of VHTRs that have been built and operated Country of origin Thermal power MWt Net electric power MWe Maximum core outlet temp ( C) Helium pressure MPa Steam temp ( C) Reactor type Vessel material Date of operation Dragon AVR Peach bottom Ft St Vrain THTR-300 HTTR OECD/Britain Germany USA USA Germany Japan 21.5 – 750 2.0 – Sleeve Steel 1964–1975 46 13 950 1.1 505 Pebble Steel 1966 115 40 725 2.25 538 Sleeve Steel 1967 842 330 775 4.8 538 Block PCRVa 1979–1989 750 300 750 3.9 530 Pebble PCRV 1985 30 10 950 Prism Steel 1997 a Prestressed concrete reactor vessel Source: Simon, R A.; Capp, P D Operating experience with the dragon high temperature reactor experiment In Proceedings of the Conference on High Temperature Reactors, Petten, NL, Apr 22–24, 2002; pp 1–6 Burnette, R D.; Baldwin, N L Specialists Meeting on Coolant Chemistry, Plate-Out and Decontamination in Gas Cooled Reactors, Juelich, FRG, Dec 1980; International Atomic Energy Agency, 1980; pp 132–137 Shaw, E N History of the Dragon Project – Europe’s Nuclear Power Experiment; Pergamon: New York, 1983 Baăumer, B.; et al AVR Experimental High-Temperature Reactor; 21 Years of Successful Operation for a Future Energy Technology; Association of German Engineers (VDI), The Society for Energy Technologies: Duăsseldorf, Germany, 1990 Baumer, R.; Kalinowski, I Energy 1991, 16(1/2), 59–70 Brey, H L Energy 1991, 16(1/2), 47–58 Fuller, C H Design Requirements, Operation and Maintenance of Gas-Cooled Reactors, San Diego, CA, Sept 21–23, 1988; International Atomic Energy Agency, 1989; pp 55–61 The Peach Bottom reactor in the United States and the European Dragon project were among the first to seriously address the technical issues associated with high-temperature environmental interaction between the cooling gas and metallic components.1–3 Proposals for a VHTR with an outlet temperature of 1000 C or above were put forward in the late 1970s The Arbeitsgemeinschaft Versuchsreaktor (AVR) was the first experimental pebble bed reactor A commercial demonstration scale pebble bed, the Thorium Hochtemperatur Reaktor (THTR), was developed based on AVR experience A summary of important design characteristics for gas-cooled VHTRs that have been operated to date are given in Table 1.1–7 254 Material Performance in Helium-Cooled Systems The HTTR in Japan is the only one of the reactors listed in the table that is still in operation The HTR-10 is not included in the table since there is no extensive operating experience with this reactor as yet Operating experience with these reactors has shown that the primary helium coolant tends to contain H2O, H2, N2 as well as carbon-containing compounds CO, CO2, and CH4 at concentrations of a few parts per million Impurities are introduced through adsorption on the fuel, leaks into the coolant, and lubricants from components like the helium circulators In the reactors that are currently under consideration, the gas pressure is typically between and MPa The coolant is circulated at high velocity, reaching velocities over 100 m sÀ1 in some designs 5.11.3 Comparison of IHX Concepts The IHX design for the VHTR will be influenced by a number of interrelated considerations, including the required separation distance between the reactor and the hydrogen production or other process heat plant, the heat losses from the intermediate loop piping, the operating pressure, the working fluid in the secondary loop, and the target efficiency of the hydrogen or process heat plant The required separation distance will affect the intermediate loop piping size, the intermediate loop pumping requirements, and the piping heat losses to the environment The intermediate loop pressure is critical; a low pressure will produce a high pressure differential between the primary and secondary sides of the IHX and high stress on the IHX A high intermediate loop pressure will produce a high pressure differential across the intermediate loop pipe walls and within the hydrogen production or process heat equipment Pressure drops within the IHX affect the pumping power requirements, which also depend on the intermediate loop working fluid, and the fluid temperature and pressure, and will have an effect on the overall VHTR cycle efficiency The IHX may be arranged in parallel or in series with the VHTR power conversion system (PCS) In a serial arrangement, the total primary system flow (reactor outlet gas) passes through the IHX The IHX receives gas of the highest possible temperature for delivery to the hydrogen production process (with slightly cooler gas going to the PCS), and must be large enough to handle the full primary flow A parallel configuration splits the reactor outlet gas flow, with only a portion entering the IHX for the hydrogen or process heat plant, and the remainder of primary flow going to a direct cycle power generation turbine This results in the smallest possible IHX and the highest overall electrical power efficiency but lower process heat efficiency because of the cooler gas reaching that process Specific IHX designs under consideration include countercurrent tube and shell, plate and fin, involute heat exchangers, microchannel heat exchangers, and the printed circuit heat exchanger (PCHE) The design has a significant influence on the required material properties Tube-and-shell designs have the advantage of technological maturity, use heavy gauge materials, and are fabricated using conventional fusion welding methods For the most simple tube-and-shell configuration, it has been estimated that 13 tons of high-temperature alloy is required per megawatt of heat transfer capability; helical designs can reduce this value to about 1.2 tons MWÀ1 Compact heat exchanger designs have the potential for greater heat transfer efficiency; it is estimated that some of these designs will require only 0.2 tons of alloy per MW The compact designs are much less technologically mature and increase the demands on material performance Some compact designs have wall thicknesses of less that 1mm which places a premium on corrosion resistance and have significant stress concentrations that will lead to increased demand for creep resistance In addition, several of these design concepts require diffusion bonding of multiple sheets of material or brazing in complex geometries Neither of these joining methods has been used yet in nuclear applications, and nondestructive inspection methods have not been well developed 5.11.3.1 Shell-and-Tube A shell-and-tube heat exchanger is the most common type of heat exchanger It consists of a number of tubes (often finned) placed inside a volume (shell) One of the fluids runs through the tubes while the second fluid runs across and along the tubes to be heated In one variation of this concept, the heat transport fluid will flow on the shell side, allowing the tubes to contain the catalysts necessary for hydrogen production In the simple configurations, the tube axis is parallel to that of the shell The VHTR IHXproposed design features the tubes arranged in a helical configuration This type of arrangement increases efficiency because of increased surface area and reduces the size, providing the potential to decrease the cost of materials Tube-and-shell heat exchangers represent relatively mature technology Material Performance in Helium-Cooled Systems that has been widely commercialized in both nuclear and fossil energy systems A helical design was extensively tested for the AVR reactor program and a similar system is in use in the HTTR in Japan 5.11.3.2 Plate and Fin The plate and fin heat exchanger (PFHE) transfers heat between two fluids by directing flow through baffles so that the fluids are separated by metal plates with very large surface areas The fluids spread out over the plate, which facilitates the fastest possible transfer of heat This design has a major advantage over a conventional heat exchanger because the size of the heat exchanger is less for a comparable heat transfer capability However, the candidate heat exchanger materials have relatively low thermal conductivities and will reduce the efficiency of a finned structure Brazing is typically used to join the fins to the plate Brazed plate heat exchangers are used in many industrial applications, although usually at low or even cryogenic temperatures Although brazed products have been developed for high-temperature aerospace applications, the strength and creep properties of brazed joints in an IHX for a hightemperature reactor are of great concern The unit cell heat exchanger is a typical modular plate-fin design that is being developed by Brayton Energy An example is shown in Figure Many of these individual unit cells would be grouped into larger heat exchanger assemblies Integration of the modules within the vessel and with the interfacing piping is critical Offset fin plate heat exchangers have very large heat transfer area density and effective countercurrent flow 5.11.3.3 (a) (b) Figure Unit cell heat exchanger (a) primary side plate, (b) the unit cell showing countercurrent flow 255 Etched Plate Etched plate heat exchangers are diffusion-bonded, highly compact heat exchangers that can achieve a thermal effectiveness of over 98% in a single unit Compact heat exchangers are four to six times smaller and lighter than conventional shell-and-tube heat exchangers of the equivalent heat transfer capability (Figure 3) The small size gives the compact diffusionbonded heat exchangers significant benefits over conventional heat exchangers across a range of industries They are well established in the oil production, petrochemical, and refining industries In addition, they are suitable for a range of corrosive and high-purity Figure The diffusion-bonded heat exchanger in the foreground undertakes the same thermal duty, at the same pressure drop, as the stack of three shell-and-tube exchangers behind 256 Material Performance in Helium-Cooled Systems Hot channel Cold channel Figure Printed circuit heat exchanger configuration for the model streams and are particularly advantageous when space is limited and weight is critical The most widely commercialized etched plate heat exchanger is a PCHE developed by Heatric Division of Meggitt (UK) Ltd PCHE consists of metal plates on the surface of which millimeter-scale semicircular fluid-flow channels are photochemically milled, using a process analogous to that used for the manufacture of electronic printed circuit boards The plates are then stacked and diffusion-bonded together to fabricate a heat exchanger core shown schematically in Figure Heatric reports pressure capability in excess of 70 MPa and the ability to withstand temperatures ranging up to 900 C Note that the channels are straight in this schematic, but in reality they have a zigzag configuration Flow distributors can be integrated into plates or welded outside the core, depending on the design The channel diameter, plate thickness, channel angles, and other attributes can be varied, so each PCHE is custom-built to fit a specified task Channel dimensions are generally between and 0.2 mm and the thickness of the web of material left after milling is typically less than mm.8 The current fabrication limits are 1.5 m  0.6 m plates and 0.6 m stack height The diffusion-bonded blocks made from several hundred individual sheets are modular and multiple blocks can be welded together to form larger units 5.11.3.4 Microchannel Heat Exchangers Microchannel heat exchangers, produced, for example, by Velocys, also feature a compact design similar to the etched plate design; however, the manufacturing process is somewhat different They are constructed from diffusion-bonded corrugated sheets rather than 30Њ Figure Plate-stamped heat exchanger concept etched plates The layers of corrugated sheet form many small-diameter channels that result in a high surface area/volume ratio and a high heat transfer coefficient 5.11.3.5 Plate-Stamped Heat Exchanger The plate-stamped heat exchanger (PSHE) concept consists of a set of modules, each being composed of a stacking of plates stamped with corrugated channels The plates are stacked in such a way as to cross the channels of two consecutive plates and therefore to allow the different channels to communicate through the width of the plate as shown on the left in the figure A general view of a plate is shown in Figure Assembly of the plates into an IHX module is accomplished by welding only on the edges of the plates No joining is performed in the active part of the plates, which gives the module relatively good flexibility Therefore, this concept is thought to accommodate the thermal stresses better than the other concepts of plate IHXs The location of the welded joints is also favorable to inspection, even if Material Performance in Helium-Cooled Systems this remains a difficult question The joining processes which seem to be the most relevant are laser or electron beam welding due to the capability to perform narrow-gap joints and to avoid the overlapping of the welds of two consecutive plates It should also be noted that the thickness of the PSHE plates is the largest among the metallic plate types IHX (1.5 mm), which means that it is the most favorable concept with respect to corrosion life These reasons suggest that the PSHE concept may be the most promising among the plate IHXs 5.11.3.6 Foam IHX The foam IHX concept is based on stacking plates separated by metallic foam The barrier between the fluids is constituted by the separated plates and the fluids flow through the foam (see Figure 6) It is a new technology for heat exchanger application for which very high efficiency has been claimed Several concerns have been identified regarding this type of IHX concept The pressure losses induced by the foam are particularly high Loss of small fragments of the foam is hardly avoidable and the geometry of the foam leads to an increased risk of clogging by graphite dust 5.11.3.7 Capillary IHX A concept with thread tubes between two tube-plates with external shell including bellows has been investigated The diameter of the tubes is 2–3 mm This kind of heat exchangers is currently being developed on an industrial scale The small size of the tubes allows a sharp reduction in size and mass, but some difficulties arise at the same time, including the concern that the vibration risk is increased so that the supporting system needs to be very robust The number of tubes Figure Foam heat exchanger concept 257 reaches very high values, which increases the complexity of manufacturing, notably as assembly by narrow gap welding is required Demonstration of the elements necessary for successful implementation of the technology is mainly based on technological feasibility tests like demonstration of individual tube to tube-plate welds by laser techniques The results confirm the feasibility for limited thickness of the plate (a small mock-up is shown in Figure 7) 5.11.3.8 Ceramic IHX The development of IHXs made of ceramics is still at the research stage Ceramic heat exchangers under development are either tubular or plate IHXs (mostly PFHE for the ceramic plate IHXs) Tube-and-shell heat exchangers based on SiC composite tubes have been developed for fossil energy applications for example Joining the fiber-reinforced composite tubes to tube sheets and accommodating thermal expansion are the dominant technical challenges Their resistance to aggressive environment is remarkable and they can operate at very high temperatures, >1000 C Small monolithic compact designs have been developed from silicon carbide and silicon nitride through conventional ceramic forming and firing routes In addition to technical issues, the cost of ceramic tubes of sufficient size for a VHTR IHX remains problematic Table provides a summary-level comparison of the significant attributes of the different IHX concept alternatives 5.11.4 Heat Exchanger Alloys The desire for higher temperature operation resulted in the evolution of the materials under consideration, from stainless steels to iron-based high-temperature 258 Material Performance in Helium-Cooled Systems Figure Capillary heat exchanger mock-up Table Comparison of IHX concept alternatives PCHE PFHE PSHE Tubular IHX Maturity Stress behavior Sensitivity to corrosion Compactness Numerous developments in conventional industry Numerous developments in conventional industry Numerous developments in conventional industry High stress levels years lifetime seems very challenging High stress levels years lifetime seems very challenging Challenging but best stress accommodation among the plate IHXs Limit of state of the art Sensitive 26 MW mÀ3 Very sensitive 24 MW mÀ3 Sensitive 35 MW mÀ3 Better than plates but still sensitive Very sensitive (loss of fragments risk) Very sensitive 0.4 MW mÀ3 Foam IHX Industrial components in operation R&D No results Capillary IHX Industrial developments No results Ceramic IHX R&D Difficult design because of fragile behavior alloys to nickel-based alloys (see Chapter 2.08, Nickel Alloys: Properties and Characteristics) An extensive German program in the 1980s carried out exhaustive studies of the corrosion behavior of the iron-based Alloy 800H for control rods and nickel-based Alloy 617 for structural applications.9–12 The Japanese HTTR program extensively studied Alloy X and developed a variation known as XR with improved properties for some applications, while retaining Alloy 800H for the control rods.13 Compositions of these candidate alloys are given in Table 3.13–16 Based on creep resistance above 850 C, the leading candidate alloys for VHTRs are Alloy 617 and Alloy 230 A common characteristic of the alloys that have been put in service in high-temperature gas-cooled Resistant Comparable to other plate IHXs Better than classical tubular IHX Comparable to other plate IHXs reactors is that they rely primarily on the formation of a tenacious chromia scale for long-term protection from environmental interaction with the gas-cooling environment.9,10,12,17 The alloys are also primarily solid-solution strengthened with carbides on the grain boundaries to stabilize the microstructure and enhance the creep resistance Sustaining such a protective surface oxide requires sufficient oxygen partial pressure The primary coolant gas of choice for VHTRs is helium Although the helium is nominally pure and thus considered to be inert, there are inevitably impurities at the parts per million by volume (ppm) levels in the coolant in operating high-temperature reactors Although at low levels, the impurities can significantly affect the performance of materials, Material Performance in Helium-Cooled Systems Table 259 Compositions of potential high-temperature alloys for VHTR (compositions in wt%) Alloy Ni Fe Cr Co Mo Al Alloy 617 UNS N06617 Alloy 230 UNS N06230 Alloy 800H UNS N08810 Alloy X UNS N06002 44.5 20–24 10–15 8–10 0.8–1.5 Bal 20–24 1–3 0.2–0.5 30–35 39.5 19–23 Bal 17–20 20.5–23 W 8–10 0.1 C 0.6 0.05–0.15 13–15 0.15–0.6 0.5–2.5 Ti 0.2–1 Si Mn 0.05–0.15 0.25–0.75 0.3–1 0.15–0.6 0.05–0.1 0.03 0.05–0.15 45 mm with large carbide precipitates rich in W, presumably of the M6C type After aging, Alloy 230 typically exhibits M6C and M23C6 precipitates After aging for 1000 h at 850 C, very small carbide precipitates rich in Cr and M23C6 were observed along the grain boundaries No grain coarsening was observed.25 Creep strength is believed to be brought about by solid-solution strengthening, low stacking fault energy, and precipitation of M23C6 carbides on glide dislocations.26,27 However, a negative impact of M23C6 on room temperature ductility was also reported After aging at 871 C for 8000 h, the room temperature tensile elongation of Alloy 230 decreased from $50% to 35%, with a precipitation of M23C6 observed in microstructural examination, but an additional 8000 h of aging did not further decrease 263 ductility.26 Significant microstructural changes were also observed after thermal aging in air for 10 000 h at temperatures ranging from 750 to 1050 C After the 750 C aging, coarser intergranular precipitation of M23C6 and coarse and blocky intra- and intergranular precipitates of M6C were observed After the 850–1050 C aging, the M6C carbides were irregular in shape After aging at 1050 C, the secondary intragranular M23C6 appeared to have dissolved A decrease in toughness and ductility coincided with the appearance of the intragranular M23C6 and reached a minimum after the aging at 850 C The toughness and ductility recovered after the aging at 1050 C.28 There is less characterization of Alloy 230 compared to Alloy 617 The major known large-scale study was tensile and creep tests by Haynes International Creep times ranged from 15.3 to 28 391 h Like Alloy 617, Alloy 230 is not currently qualified for use in ASME Code Section III, although it is allowed in Section VIII, Division (for nonnuclear service) At present, the database for Alloy 230 is significantly smaller than that for Alloy 617 and a much larger effort is required to develop an Alloy 230 Code Case for elevated temperature application Some recent data on environmental effects of exposure to prototypical VHTR chemistries are given in the following sections and creep–fatigue properties are included in Figures and 10 5.11.4.4 Alloy 800H (42Fe–33Ni–21Cr) This alloy is the only iron-based alloy under consideration, although it has a solid-solution strengthened austenitic structure like the other three alloys Upon aging, precipitates can form and somewhat reduce the tensile and creep ductility Alloy 800H has the lowest creep rupture strength and the lowest resistance to oxidation of the four alloys There is an additional variant of this alloy, 800HT, that has a composition similar to that of 800H, but has an additional specification for coarse grain size The majority of material that is currently available in this alloy series is Alloy 800HT, which also meets the specification for Alloy 800H Among the four candidate materials, Alloy 800H is the only one that is Code qualified for use in nuclear systems, but only for temperatures up to 760 C and a maximum service time of 300 000 h Alloy 800H was the primary high-temperature alloy used in the German HTGR programs and an enormous amount of data were obtained However, only very limited data from the German HTGR programs 264 Material Performance in Helium-Cooled Systems are currently available on the mechanical properties of this alloy beyond 800 C, especially in impure helium environments 5.11.5 Welding All of the solid-solution alloys that have been mentioned are readily welded using conventional fusion welding methods Alloys 617 and 230 are described in more detail later as prototypical of these materials Alloy 617 has excellent weldability Alloy 617 filler metal is used for gas-tungsten-arc (GTAW) and gasmetal-arc welding (GMAW) The composition of the filler metal matches that of the base metal, and deposited weld metal is comparable to the wrought alloy in strength and corrosion resistance.29 Alloy 230 is also readily welded by GTAW and GMAW Shielded metal-arc welding (SMAW) and resistance welding techniques can also be used Submerged-arc welding Stress (MPa) Alloy X has the best oxidation resistance of the four alloys, although its carburization resistance is the worst Above 700 C, Alloy X can form embrittling phases that result in property degradation The creep rupture strength is not as good as Alloy 617 or 230 The limitations of this alloy will be similar to the draft code case for Alloy 617 in terms of grain size, product form, and limitations on service time A limited database exists for Alloy X for conditions typical of a VHTR, but the high-temperature scaling in Hastelloy X has been less than optimal As a result, a modified version, Alloy XR, has been developed in Japan; however, the United States has little access to Alloy XR material, either for evaluation or for ASME Code qualification Japanese are currently using Alloy XR in a heat exchanger in the HTTR at temperatures of 850–950 C The material is codified in Japan for nuclear use, which would likely accelerate code acceptance in ASME An extensive environmental database and HTGR experience exist However, the database may be limited to large grain material, similar to the Alloy 617 draft code case Also, similar to the Alloy 617 draft code case, Alloy XR may have issues with weldments that need to be addressed It is uncertain if this alloy is readily available as a commercial product Figures 11–14 compare the creep rupture strength, oxidation behavior, carburization behavior, and allowable stress for the four alloys, respectively Alloy X 60 Alloy 230 50 Alloy 617 40 30 20 10 1000 100 Time to rupture (h) 10 10 000 100 000 Figure 11 Creep rupture strength at 962 C in air Thickness (mm) Alloy X (47Ni–22Cr–9Mo–18Fe) Alloy 800H 70 20 10 −10 −20 −30 −40 −50 −60 −70 −80 −90 Alloy 800H Alloy 617 Oxide scale Alloy 230 Affected zone Alloy X Internal oxide Figure 12 Schematic representation of isothermal oxidation behavior after 800 h exposure at 950 C in helium environment 24 Normalized weight gain 5.11.4.5 80 Alloy X 19 Alloy 230 Alloy 617 14 −1 200 400 600 800 1000 1200 Exposure time (h) Figure 13 Mass change as a function of time in H2–5.5% CH4–4.5% CO2 carburizing environment at 1000 C is not recommended, as this process is characterized by high heat input to the base metal and slow cooling of the weld These factors can increase weld restraint and promote cracking The as-welded properties of these alloys are given in Table 7.30 The welds exhibit room temperature strength that matches or is slightly Material Performance in Helium-Cooled Systems 265 better than the base metal, but a considerable decrease in ductility is observed at elevated temperatures, as shown in Table alloy, HASTELLOY S alloy, or HASTELLOY W alloy welding products may all be considered, depending upon the particular case.29,30 5.11.5.1 Base Metal Preparation and Filler Metal Selection 5.11.5.2 Preheating, Interpass Temperatures, and Postweld Heat Treatment Prior to any welding operation, the welding surface and adjacent regions should be thoroughly cleaned with an appropriate solvent All greases, oils, corrosion products, and other foreign matter should be completely removed It is preferable, but not necessary, that the alloy be in the solution-annealed condition when welded.30 Alloys 617 and 230-W™ (AWS A5.14, ERNiCrWMo-1) filler wire are recommended for joining Alloy 617 and 230, respectively, by GTAW or GMAW The filler metals are not specifically designed for nuclear application For dissimilar metal joining of Alloy 230 to nickel-, cobalt-, or iron-based materials, 230-W filler wire, Alloy 556™ Preheat is not required, generally room temperature (typical shop conditions) is adequate Interpass temperature should be maintained below 93 C Auxiliary cooling methods may be used between weld passes, as needed, providing that such methods not introduce contaminants Postweld heat treatment is not generally required either Table shows the nominal welding parameters based on welding conditions used in the Haynes International laboratories and should serve as a guide for performing typical GTAW and GMAW operations on Alloy 230 All processes used 230-W filler wire.30 5.11.5.3 Maximum allowable stress (MPa) 250 Alloy X Alloy 230 200 Alloy 617 Alloy 800H 150 100 50 0 200 400 600 800 1000 Temperature ( ЊC) Figure 14 Allowable stress for heat exchanger materials for plate, sheet, and strip forms from the ASME boiler and pressure vessel code Section VIII Table Nontraditional Joining Methods As noted earlier in the description of IHX designs, several of the compact heat exchanger design concepts will require the joining of sheet product to be either diffusion bonding or brazing Diffusion bonding of these alloys is relatively well developed because of applications in aerospace systems that require this fabrication method The etch plate compact design fabricated from austenitic stainless steel has been commercialized for petrochemical applications, and limited diffusion bonding studies have been completed using Alloy 617 Characterization of diffusion-bonded stacks of sheet indicates that mechanical properties comparable to base metal can be achieved at room temperature The details of diffusion bonding parameters are considered proprietary by the IHX vendors, and it is not clear whether temperatures sufficiently high to cause carbide dissolution and/or grain growth is a matter of concern Room-temperature tensile properties of joints in as-welded condition Alloy Specimen Yield strength (0.2% offset) (MPa) Tensile strength (MPa) Elongation (%) Reduction of area (%) 61729 GMAWa GTAWb GMAWc 510 542 490 761 823 785 43.3 37.3 48.2 42.0 38.3 23030 a Alloy Filler Metal 617 Average of ten tests Alloy Filler Metal 617 Average of 17 tests Alloy 230-W filler wire b c 266 Material Performance in Helium-Cooled Systems Table Tensile properties of 230 base and weld metals 23 C GMAW deposit weld metal Cold-rolled and 1232 C solution annealed (sheet) Hot-rolled and 1232 C solution annealed (plate) Vacuum investment castings (as-cast) 538 C 871 C UTS YS EL UTS YS EL UTS YS EL 785 838 840 615 490 422 375 325 48.2 47.2 47.7 37.8 610 699 690 450 435 303 251 230 34.8 53.7 54.6 38.2 310 308 315 285 275 234 242 185 45.4 75.0 99.5 19.0 Source: HAYNESW 230W Alloy Haynes International, Inc Publication H-3000H, 2004 Table Weld parameters for Alloy 230 Welding method GMAW Configuration (mm) Thickness > 2.3 1.1 dia wire Technique Stringer bead or slight weave 100–130a 18–21 4.3–4.8 12.7–19.1 203–356 Torch, 50 Ar-25% He Current (A) Voltage (V) Feed rate (m minÀ1) Stick-out (mm) Travel speed (mm minÀ1) Gas flow (l minÀ1) Gas GTAW Auto Manual Square butt joints 1.0/1.6/3.2 thick, 1.6 electrode with 45 included shape No filler metal added 50/80/120b 8.0/8.5/9.5 V or U groove, >3.6 thick, 3.6 dia wire, 3.6 electrode with 30 included shape Stringer bead interpass T < 100 C 120 root, 140–150 fillb 11–14 10/12/12 Shield, 14.2 backing, 4.7 Argon 102–152 Shield, 14.2–16.5 backup, 4.7 Argon a DCEP, torch flow CFPH DCEN Source: HaynesW 230W Alloy, Haynes International High-Temperature Alloys b Very little information on brazing these alloys is available A general concern is that low melting point braze materials could result in poor elevated temperature properties in structures fabricated by these methods 5.11.6 Control Rod Materials The pebble bed modular reactor (PBMR) is the most complete recent design for a VHTR The reactor was designed to operate at about 400 MWt and primarily to produce electricity Recent changes in the global economic climate have caused reconsideration of the design for a VHTR in South Africa; however, the analysis that went into the design and selection of materials for the control rods is illustrative of the most recent analysis of these issues A schematic of the PBMR core is shown in Figure 15 The design outlet gas temperature for the PBMR was 900 C; the core was designed to be 11m high and 3.7 m in diameter, and the annulus filled with about 452 000 60-mm-diameter fuel pebbles.31 The PBMR builds on the German experience of the AVR and THTR; however, it will use a direct cycle to produce power rather than a steam generator, and it will have an annular core configuration with a solid graphite central reflector The annular core produces several advantages: it shifts the peak power radially outward, thus enabling significantly higher output; it enhances the fuel safety margin; and, by increasing the neutron flux in the outer graphite reflector, it increases the effectiveness of the control and shutdown systems.32 The reactor control and shutdown system (RCSS) has two components: the reactivity control system (RCS) and the reserve shutdown system (RSS) The RCS consists of 12 control rods and 12 shutdown rods, located in the outer reflector.33 They are evenly spaced around the core and at a radial distance of about 70mm from the inner surface of the reflector (see Figure 16).34 During normal operation, the control rods, which Material Performance in Helium-Cooled Systems Reactor pressure Core barrel Top Side Cold gas riser Center Pebble bed Bottom Inlet Inl Hot gas Outlet Figure 15 Schematic of the pebble bed modular reactor (PBMR) annular pebble bed reactor Reproduced from Kriel, W Material selection: High-temperature metallic materials Slides, Sept 21–22, 2005 Core barrel 267 penetrate a maximum distance of 1.5m into the core,34 are used for minor reactivity adjustments to keep the reactor critical, provide reactivity compensation for xenon poisoning effects during load following effects,35 and allow for some excess reactivity so that the reactor may continue operation for some time if no fuel is being loaded.36 They are also used for hot shutdown purposes.33 The reactor power is actually adjusted by regulating the mass flow rate of the gas inside the primary circuit rather than by adjusting the control rods.32,33 During scram, the additional 12 shutdown rods are lowered to the bottom of the active core In the event of a loss of electrical power, insertion of the rods is by gravity The first set of control rods will drop, and later the shutdown rods will drop, should the need arise The RSS consists of eight storage containers of 10-mm-diameter small absorber spheres containing B4C that can be fed by gravity into eight channels in the central reflector The RSS serves as both the secondary shutdown system and the cold shutdown system It must be activated in addition to the RCSS to bring the PBMR to a cold shutdown condition (100 C) The control rod design is similar to that of previous metal control rods A schematic is shown in Figure 17 A number of annular B4C rings are encased between two tubes of Alloy 800H, to form a section about a meter long One unique feature is that Reactor pressure vessel Side reflector barrel Reactor inlet pipes Annular core Reactor outlet pipe Center reflector Gas riser channels Small absorber sphere channels Control rod Figure 16 Top view of the PBMR core, showing the location of the control rod channels, fuel pebbles, small absorber sphere channels, and other features Reproduced from PBMR Data and boundary conditions to be used in VSOP, TINTE, and MCNP PBMR 400 MW ( Th) reactor models 268 Material Performance in Helium-Cooled Systems Control rod drive mechanism RCS chain Control rod segment Control rod link Secondary shock absorber Figure 17 Schematic of the control rod assembly in the PBMR Reproduced from Broom, N.; Smit, K PBMR Design Methodology Slides, Oak Ridge, TN, 12th April 2005 the inner tube is much thinner than the outer tube These sections are mechanically linked to form an articulated control rod several meters long One difference from past designs (e.g., the AVR) is that the control rod is suspended from the drive mechanism by a chain, rather than a cable A secondary shock absorber is in place in the channel below the control rod to protect it and the core structure in the event of a chain failure Additional shock absorbers within the drive mechanism dampen the impact load on the control rod drives during scram.33 A control rod guide tube (not shown in Figure 16) connects the control rod drive mechanism to the core structure to guide the control rod into the core.37 During normal operation, the temperature of the control rods is estimated to be from about 650 to 700 C,38,39 and the temperature resulting from a depressurized loss of coolant (DLOC) event is estimated to be only 850 C.39 The end-of-life fast fluence is reported as  1022 (E > 0.1 MeV),38 and the thermal fluence is reported as  1021 n cmÀ2.39 The secondary shock absorbers have an operating temperature of 900 C; during DLOC, they can be subjected to temperatures of up to 1100 C for short periods Under these conditions, the use of Alloy 800H was justified by the PBMR program because the high-temperature strength and creep resistance are sufficiently qualified for long-term normal operation at 700 C, and limited operation above 850 C under abnormal events can be tolerated according to available data The response of Alloy 800H to neutron irradiation at the temperatures expected in the control rod sleeves is not well characterized The PBMR design notes that the irradiation response has been characterized to high levels of fast fluence at lower temperatures and Alloy 800H control rods have had extensive qualification and service in previous German VHTR programs Limited data from older VHTR programs on irradiation effects in Alloy 800H at temperatures above 600 C suggest that helium embrittlement from (n, a) reactions associated with thermal neutrons is the predominant degradation mechanism Recent work has examined property changes associated with irradiation to 1.45 dpa at temperatures of 580 and 660 C.40,41 Significant strengthening was observed along with a sharp decrease in ductility Material irradiated at 660 C and subsequently tensile tested at 700 C showed tensile elongation of less than 0.5% The mechanisms of embrittlement are not yet completely clear and irradiation experiments to higher fluence at higher temperature will likely be required to examine this issue The PBMR design includes specialized equipment to remove and replace the control rods, as well as storage for used control rods The RCSS will be inspected every years during the scheduled maintenance outage, and repaired as necessary These outages are planned to last 30–50 days, depending on the other maintenance scheduled, with the exception of a 180-day shutdown after 24 years to replace the core reflector.34,42 Eventually, it is hoped that a VHTR similar in design to the PBMR reactor can run with an outlet temperature of 1000 C or even higher In this case, carbon fiber-reinforced carbon composites (Cf/C) or silicon carbide fiber-reinforced silicon carbide composites (SiCf/SiC) must be considered for the more challenging temperatures of the control rods.39 Experience with irradiation of SiCf/SiC composites for nuclear fusion applications suggests that these materials have superior resistance to property degradation from neutron irradiation as well as resistance to higher temperatures and could potentially have Material Performance in Helium-Cooled Systems lifetimes comparable to the life of the plant As with ceramic IHXs, application of these advanced materials in a nuclear system would require considerable further development and cooperation with appropriate standards and regulatory organizations 5.11.7 Core Barrel Materials Another evidently important metallic internal structure shown in Figures 15 and 16 is the core barrel The function of the core barrel is to mechanically contain the shape of the graphite blocks making up the core and to channel the flow of the primary coolant Although schematics in Figures 15 and 16 are specific to a pebble bed design, the core barrel is essentially identical in design and function for a prismatic design as well The temperatures, neutron fluence, and mechanical loads on the core barrel are moderate and the PBMR design, for example, proposed the use of Type 316 stainless steel for this application Alloy 800H is also a leading candidate for the core barrel material Although the demands on the material are modest in terms of mechanical loading and neutron irradiation, it is a very large structure ($8 m high by 3.5 m diameter and 50 mm in thickness) that will need to be fabricated on site by welding in most cases Both Type 316 stainless steel and Alloy 800H are available in the required size and are readily fabricated 5.11.8 Environmental Effects of VHTR Atmospheres on Materials All the high-temperature reactor systems operated to date had extensive gas cleanup systems associated with the helium coolant flow These systems are intended to keep the total impurity levels in the helium below typically 10 ppm Particularly in the early reactors, where the fuel was either not intended to contain the fission products or was ineffective in this function, the cleanup systems were also intended to capture radionuclides.1,3 Capture of tritium that is produced (at least in part) by transmutation of lithium impurities in the graphite remains an important function of the cleanup system In the AVR and THTR reactors, active control was maintained on the H2O and CO concentrations to reduce oxidation of the graphite reflectors, and the other impurities were routinely found to reach acceptable steady-state levels without active control.4–7,43,44 It has been noted that the 269 cleanup systems may play a secondary role in maintaining gas chemistry, with the massive amount of graphite at high temperature present in all of the reactor designs playing a dominant role.1 Molecular sieves are effective in capturing most of the gaseous impurities; however, they have difficulty capturing H2 and CO To resolve this problem, the gas stream is passed over a bed of CuO that oxidizes the H2 to H2O and CO to CO2 upstream of the molecular sieve where these gases are effectively removed The Peach Bottom plant attempted the use of heated Ti getters for hydrogen and tritium; however, these were not effective and oxidation of the H2 prior to removal is now accepted practice.2 In a typical plant, up to about 20% of the gas stream is diverted to the cleanup system each hour Table 10 shows the impurity levels reported for steady-state operation for several of the VHTRs.1,2,43,44 As shown in the table, at steady state, all of the reactors for which operating data are available had similar levels of impurities Some caution should be exercised when comparing the data for different plants, since, in some cases, there are varying values reported in different publications for the same plant This may be associated with conversion from partial pressure of impurities (the preferred units for corrosion studies) to ppm by volume (the typical units used for comparison of one plant to another) Several plants have undergone extensive postmortem analysis of the core internals and heat exchangers.1,2 There are reports of some oxidation and at least one report of massive deposition of carbon on the internals, as discussed in more detail in the following paragraphs; however, there have been no problems with failure of components on the primary side associated with environmental effects There have been a large number of experimental studies and modeling of the effect of VHTR helium on the high-temperature alloys listed in Table 3.10–12,16,17,45–54 Depending on the specific proposed application, different model chemistries have been developed, and the testing has focused on these Several of the model impurity chemistries are shown in Table 11.45–52,54 Comparison of the values in Table 11 with actual operating experience suggests that the model chemistries tend to have higher impurity levels of some species than those found in operating reactors This is notable for H2 in particular It is not clear why these particular values were chosen; however, it can be noted that several of the proposed applications were for process heat for coal gasification, and there was concern that hydrogen would diffuse from the process plant into the primary 270 Material Performance in Helium-Cooled Systems Table 10 Impurities reported in the helium coolant during steady-state operation of VHTRs (in ppm) Dragon Peach bottom Fort St Vrain AVR THTR H2O H2 CO CO2 CH4 O2 N2 0.1 0.5 0.15