Klein, A.C.; Sze, D.K.
An assessment is made of the gamma radiation hazards likely to be found around a fusion reactor heat transfer and tritium breeding loop which employs a vanadium alloy for the blanket and first wall structure and the ferritic-steel HT9 for the remainder of the loop. The coolant/tritium breeding fluid is the molten metallic salt FliBe. Since the radiation levels near the primary loop components are found to be less than 100 mR/hr 3-5 days after shutdown after three years of continuous full power operation, limited hands-on maintenance could be allowed. The very short half-lives of the predominant corrosion products make this result possible and make such a system very attractive
Lee, J.D. (ed.)
This report contains separate articles of seven aspects of the MINIMARS programs. The areas discussed are Fusion Engineering Design Center, Halo Model and Computer Code, safety design, the University of Wisconsin blankets, activation product transport in a FLiBe-VANADIUM alloy HT-9 system, a halo scraper/direct converter system, and heat transport power conversion. The individual articles are cataloged separately. (WRF)
Pahl, R.G.; Lahm, C.E.; Hayes, S.L.
Steady-state testing of HT9 clad metallic fuel at high temperatures was initiated in EBR-II in November of 1987. At that time U-10 wt. % Zr fuel clad with the low-swelling ferritic/martensitic alloy HT9 was being considered as driver fuel options for both EBR-II and FFTF. The objective of the X447 test described here was to determine the lifetime of HT9 cladding when operated with metallic fuel at beginning of life inside wall temperatures approaching ∼660 degree C. Though stress-temperature design limits for HT9 preclude its use for high burnup applications under these conditions due to excessive thermal creep, the X447 test was carried out to obtain data on high temperature breach phenomena involving metallic fuel since little data existed in that area
Windisch, Charles F.; Henager, Charles H.; Engelhard, Mark H.; Bennett, Wendy D.
Raman microprobe spectroscopy was applied in studies of high-temperature air oxidation of a ferritic alloy (HT-9) in the absence and presence of zirconia coatings with the objective of evaluating the technique as a way to quickly screen candidate cladding materials and actinide-based mixed oxide fuel mixtures for advanced nuclear reactors. When oxidation was relatively uniform, Raman spectra collected using microscope optics with low spatial resolution were found to be similar to those collected with conventional Raman spectroscopy. These spectra could be used to identify major oxide corrosion products and follow changes in the composition of the oxides due to heating. However, when the oxidation films were comprised of multiple layers of varying composition, or with layers containing metallic phases, techniques with higher depth resolution and sensitivity to zero-valence metals were necessary. The requirements were met by combining Raman microprobe using different optical configurations and x-ray photoelectron spectroscopy.
DiMelfi, R.J.; Gruber, E.E.; Kramer, J.M.; Hughes, T.H.
The martensitic-ferritic alloy HT-9 is slated for long-term use as a fuel-cladding material in the Integral Fast Reactor. Analysis of published high-temperature mechanical property data suggests that secondary carbide precipitation would occur during service life causing substantial strengthening of the as-heat-treated material. Aspects of the kinetics of this precipitation process are extracted from calculations of the back stress necessary to produce the observed strengthening effect under various creep loading conditions. The resulting Arrhenius factor is shown to agree quantitatively with shifts to higher strength of crept material in reference to the intrinsic strength of HT-9. The results of very low constant strain-rate high-temperature tensile tests on as-heat-treated HT-9 that focus on the transition in strength with precipitation will be presented and related to rupture-life
Latimer, T.W.; Andrew, J.F.
Out-of-pile tests were conducted for 1000 h at 600 0 C and 725 0 C to determine the compatibility of helium- and sodium-bonded (U,Pu)C + 9-12 vol percent (U,Pu) 2 C 3 with austenitic alloys Type 330 stainless steel, A286, M813, PE-16, Inconel 706, and Inconel 718. The compatibility of ferritic alloy HT-9 was determined only in the sodium-bonded tests. As a reference alloy, Type 316 stainless steel was also tested. At 600 0 C, none of the alloys showed evidence of carburization in the sodium-bonded tests and only Inconel 706 and 718 showed significant compatibility effects in the helium-bonded (contact) tests. at 725 0 C, carbide precipitation was observed in the matrix and grain boundaries of the alloys to depths of 50 to 150 μm in the helium-bonded tests. In the sodium-bonded tests, no effects were observed in alloys M813, PE-16, and HT-9; there was precipitation in the matrix and grain boundaries to depths of 45 to 125 μm in the remaining alloys
Booker, M.K.; Sikka, V.K.; Booker, B.L.P.
A modified 9 Cr-1 Mo ferritic steel has been selected for development by the US Department of Energy as an alternative structural material for breeder reactor applications in which type 304 stainless steel or 2 1/4 Cr-1 Mo steel is currently being used. This developmental alloy is a modification (primarily through additions of niobium and vanadium) of a commercial 9 Cr-1 Mo steel already available in the US. A similar commercial alloy is available in the UK. Meanwhile, a 9 Cr-2 Mo alloy (EM12) is now used in France and a 12 Cr-1 Mo alloy (HT9) has been recommended for high temperature service in both Europe and the US. This paper compares the yield strength and ultimate tensile strength properties of the American modified 9 Cr-1 Mo steel with those of the American and British commercial 9 Cr-1 Mo steels. In addition, the creep rupture properties of the modified alloy are compared with those of the two commercial 9 Cr-1 Mo steels, 2 1/4 Cr-1 Mo steel, HT9, EM12, and type 304 stainless steel. The overall conclusion of the study is that the modified 9 Cr-1 Mo steel displays tensile and creep strengths matching or exceeding those of the other ferritic materials examined, and being at least comparable to those of type 304 stainless steel from room temperature to about 625 0 C. 15 references, 11 figures, 3 tables
Martensitic stainless steels have been developed for both in-core applications in advanced liquid metal fast breeder reactors (LMFBR) and for first wall and structural materials applications for commercial fusion reactors. It can now be shown that these steels can be expected to maintain properties to levels as high as 175 or 200 dpa, respectively. The 12Cr-1Mo-0.5W-0.2C alloy HT-9 has been extensively tested for LMFBR applications and shown to resist radiation damage, providing a creep and swelling resistant alternative to austenitic steels. Degradation of fracture toughness and Charpy impact properties have been observed, but properties are sufficient to provide reliable service. In comparison, alloys with lower chromium contents are found to decarburize in contact with liquid sodium and are therefore not recommended. Tungsten stabilized martensitic stainless steels have appropriate properties for fusion applications. Radioactivity levels are being less than 500 years after service, radiation damage resistance is excellent, including impact properties, and swelling is modest. This report describes the history of the development effort. (author)