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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.
Y. Iwai, Y. Misaki, T. Hayashi, T. Yamanishi, S. Konishi, M. Nishi, R. Ninomiya, S. Yanagimachi, S. Senrui, H. Yoshida
Fusion Science and Technology | Volume 41 | Number 3 | May 2002 | Pages 1126-1130
Isotope Separation | Proceedings of the Sixth International Conference on Tritium Science and Technology Tsukuba, Japan November 12-16, 2001 | doi.org/10.13182/FST02-A22759
Articles are hosted by Taylor and Francis Online.
The water detritiation system (WDS) of tritium plant for the International Thermonuclear Experimental Reactor (ITER) was designed. The concept of the Combined Electrolysis Catalytic Exchange (CECE) process was selected for the WDS. The design conditions are (a) tritium concentration of waste water: 3.7 × 1010∼3.7 × 1011 Bq/kg, (b) waste water flow rate: 20 kg/h (1100 mol/h), a net working rate: 300 days, annual capacity: 150 tons/year (c) tritium concentration in the H2 discharged to environment: 6.5 x 101 Bq/m3, (d) tritium concentration in the H2O vapor discharged to environment: 3.7 x 103 Bq/m3, (e) tritium concentration in the electrolyzer: ∼ 1.85 × 1013 Bq/kg. Tritium concentration in the electrolyzer is determined considering the lifetime of the electrolyzer which depends on tritium concentration. Design value of height of a unit (30cm) of water-hydrogen isotopic exchange column and the correlation between the column internal flow rates and the column diameter were determined based on similar system for Japanese advanced thermal reactor (Fugen) moderated with heavy water.
