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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.
Kazuyuki Takase, Yasuo Ose, Hajime Akimoto
Fusion Science and Technology | Volume 39 | Number 2 | March 2001 | Pages 1050-1055
Safety and Environment | doi.org/10.13182/FST01-A11963382
Articles are hosted by Taylor and Francis Online.
Damage of cooling tubes of plasma facing components (PFCs) results in water discharge into a vacuum vessel (W) of a fusion reactor. Flashing in vacuum, water pool boiling and impingement-jet on a surface of the PFC are the main heat transfer phenomena responsible for steam production that causes a rapid pressurization of the W. This is called an in-vessel loss-of-coolant accident (LOCA) event or ingress-of-coolant event (ICE). The ICE event is one of the most severe accidents in the fusion reactors.
The integrated ICE test facility was constructed to demonstrate the safety design approach of International Thermonuclear Experimental Reactor (ITER) and obtain validation data for the ITER safety analysis codes. Then, an experimental study was performed using the integrated ICE test facility and at the same time the code validation study with the TRAC code was carried out. The pressure rise characteristics in the current ITER machine during the ICE event were analyzed numerically using the verified TRAC-PF1 code and the effects of the relief pipe diameter and suppression tank volume regarding to the pressure rise due to the ICE events were clarified quantitatively from the present analytical results.
