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Colin Judge: Testing structural materials in Idaho’s newest hot cell facility
Idaho National Laboratory’s newest facility—the Sample Preparation Laboratory (SPL)—sits across the road from the Hot Fuel Examination Facility (HFEF), which started operating in 1975. SPL will host the first new hot cells at INL’s Materials and Fuels Complex (MFC) in 50 years, giving INL researchers and partners new flexibility to test the structural properties of irradiated materials fresh from the Advanced Test Reactor (ATR) or from a partner’s facility.
Materials meant to withstand extreme conditions in fission or fusion power plants must be tested under similar conditions and pushed past their breaking points so performance and limitations can be understood and improved. Once irradiated, materials samples can be cut down to size in SPL and packaged for testing in other facilities at INL or other national laboratories, commercial labs, or universities. But they can also be subjected to extreme thermal or corrosive conditions and mechanical testing right in SPL, explains Colin Judge, who, as INL’s division director for nuclear materials performance, oversees SPL and other facilities at the MFC.
SPL won’t go “hot” until January 2026, but Judge spoke with NN staff writer Susan Gallier about its capabilities as his team was moving instruments into the new facility.
Koshi Mitachi, Takahisa Yamamoto, Ritsuo Yoshioka
Nuclear Technology | Volume 158 | Number 3 | June 2007 | Pages 348-357
Technical Paper | Fission Reactors | doi.org/10.13182/NT07-A3846
Articles are hosted by Taylor and Francis Online.
In this paper, an improved design for a small molten-salt reactor (MSR) that uses neutron flux flattening, which is referred to as FUJI-U3, is proposed. This reactor is a 200-MW(electric) power reactor, and its core contains graphite (as the moderator) and fuel salt. The fuel salt is composed of ThF4 as the fertile material, 233UF4 as the fissile material, and LiF-BeF2 as both the solvent and heat transfer medium. A basic improvement in FUJI-U3 is the introduction of the design concept of a three-region core in order to avoid the replacement of graphite, which is achieved by reducing the maximum neutron flux. Since there is a limit for irradiation growth in graphite, this reduction in the maximum neutron flux contributes to a longer lifetime of the graphite. Based on calculations using the nuclear analysis code SRAC95 and the burnup analysis code ORIGEN2, it is concluded that there is no need to replace the graphite moderator of FUJI-U3 for 30 yr. Further, the chemical-processing interval of the fuel salt is studied for 7.5, 15, and 30 yr. An increase in this time interval will also contribute to reduce maintenance and cost.