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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.
E. A. Mogahed, H. Y. Khater, J. F. Santarius
Fusion Science and Technology | Volume 39 | Number 2 | March 2001 | Pages 639-643
Fusion Materials | doi.org/10.13182/FST01-A11963310
Articles are hosted by Taylor and Francis Online.
A tritium-breeding blanket design is investigated for a D-T Field-Reversed Configuration (FRC) scoping study. The thrust of our initial effort on the blanket has been to seek solutions as close to present-day technology as possible, and we have therefore focused on steel structure with helium coolant. The simple FRC cylindrical geometry has allowed us reasonable success due to the low FRC magnetic field and relatively easy maintenance. In this design the breeder is Li2O tubes. The design is modular with 10 modules each 2.5 m long. The inner radius of the first wall is 2.0 m and the FW/blanket/shield thickness is about 2 m. The surface heat flux will be radiation dominated, fairly uniform, and relatively low, because most of the charged particles follow the magnetic flux tubes to the end walls. The neutron wall loading is 5 MW/m2. In this design the surface heat flux equals 0.19 MW/m2. The maximum Li2O tube temperature is 1003°C. The helium exit temperature from the heat exchanger is about 800°C which allows a thermal efficiency of about 52%. The local tritium breeding ratio (TBR) equals 1.1 and is sufficient because in the FRC geometry the plasma has nearly full coverage. The helium pumping power is 1 MW. The coolant routing is optimized to limit the steel maximum temperature to 635°C. The same concept would be applicable to a spherical torus and spheromak.
