Candidate Structural Materials for In-Core VHTR Application
- ORNL
- Idaho National Laboratory (INL)
- PBMR
Graphite moderated gas cooled reactors led the way into the nuclear age with the Chicago Pile-1 reactor, which provided the first sustained critical reaction in December, 1942. The first commercial nuclear plant, Calder Hall in the UK, went critical in 1956 with an outlet gas temperature of {approx}345 C. As depicted in Fig. 1, in five decades since Calder Hall, outlet temperature increased rapidly, reaching a plateau of {approx}950 C. This apparent ceiling is in large part due to limitations in the structural materials utilized within the core (e.g. control systems) and primary loop (hot duct, heat-exchangers etc.) Simply, the operating temperatures of Generation III (HTGR's) are very near performance limits of the structural alloys used, both in terms of elevated temperature and as-irradiated properties. This limitation remains today and is the reason the outlet temperature of the Generation IV Very High Temperature Reactor (VHTR) continues to be revised downward from the original optimistic goal of {approx}1200 C, to it's current target outlet temperature of {approx}950 C, a temperature consistent with the previous generation of HTGR's. An example of the challenges facing Generation IV VHTR is found by considering the control rods. For the Fort St. Vrain Reactor the control system consisted of thirty tubes each containing B4C control material. Alloy 800, originally developed by Inco in the 1950's under the trade-name Incoloy 800 and 800H had found widespread application is steam generators, turbines and was selected for this control rod application. These materials are included for Class 1 Nuclear Component by ASME section III. In addition to Ft. St Vrain, alloy 800H has found control rod application in the German HTR and Japanese HTTR reactors and is the primary choice Pebble Bed Modular Reactor (a HTGR) reactivity control system. Figure 2 gives the ASME allowed stress for Alloy 800H. Due to the loss in creep rupture strength, the allowed design stress for decreases substantially above {approx}500 C. For the operating temperature (and current maximum code temperature) of the Fort St. Vrain reactor ({approx}760 C) the control rod designed from Alloy 800H would be allowed approximately 20 MPa primary stress. For the anticipated temperature of the VHTR that stress would likely be under 10 MPa in the non-irradiated condition. Selection of VHTR in-core structural material will depend on the ability for the material to withstand extreme temperatures and irradiation. The use of Alloy 800H for control rod applications is preferred, but will require designing with a relatively low strength material which becomes brittle upon modest irradiaton. Presently, more fidelity in both design and the available Alloy 800H database are required to judge the lifetime this alloy will have in service. The only currently envisioned alternative to Alloy 800H are the CFC and SiC/SiC composites, which are currently challenged by code qualification issues. As with Alloy 800H, the composite irradiation data-base is inadequate, though it appears that the CFC service lifetime will be limited due to its anisotropic dimensional change under irradiation to doses somewhat higher than Alloy 800H (1-5 x 10{sup 25} n/m{sup 2}), while SiC/SiC composites appear to retain their mechanical properties to doses in excess of the VHTR lifetime dose. As the path to qualification of both CFC and SiC/SiC are similar, it appears that SiC/SiC is more attractive for high-dose Gen IV application unless high thermal conduction is required.
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR)
- Sponsoring Organization:
- USDOE Office of Science (SC)
- DOE Contract Number:
- DE-AC05-00OR22725
- OSTI ID:
- 1023813
- Resource Relation:
- Conference: American Nuclear Society, Anaheim, CA, USA, 20080610, 20080610
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
36 MATERIALS SCIENCE
ALLOYS
BUILDING MATERIALS
CHLOROFLUOROCARBONS
CONTROL ELEMENTS
CONTROL SYSTEMS
CREEP
GAS COOLED REACTORS
GRAPHITE
HEAT EXCHANGERS
HTTR REACTOR
INCOLOY 800
INCOLOY 800H
IRRADIATION
LIFETIME
MECHANICAL PROPERTIES
RUPTURES
STEAM GENERATORS
TARGETS
THERMAL CONDUCTION
TURBINES
VRAIN REACTOR