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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.
Kyungdoo Kim, Won-Pyo Chang, Kun-Joong Yoo, Seon-Hwa Lee, Chong-Bae Lee
Nuclear Technology | Volume 122 | Number 2 | May 1998 | Pages 125-131
Technical Paper | RETRAN | doi.org/10.13182/NT98-A2856
Articles are hosted by Taylor and Francis Online.
A multiple failure event, a stuck-open pressurizer spray valve along with pressurizer pressure transmitter failure, which occurred on February 25, 1995, at Kori nuclear unit 2, is simulated using the best-estimate thermal-hydraulic computer code RETRAN03/MOD000. The simulation was performed to validate the predictive capabilities of RETRAN03 against plant data. The results would be useful in evaluation of the emergency operation procedures. The transient was simulated for 5000 s until the reactor coolant system pressure was stabilized and hot standby condition could be achieved. The simulation results and their corresponding plant data, especially for the evolutions of all the major thermal-hydraulic parameters, are compared and analyzed. Relatively good agreement between the plant data and the code prediction has been obtained; however, the simulation cannot duplicate the plant data for the low-flow condition that was encountered near the end of the transient.
