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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.
Dae-Hyun Hwang, Kyong-Won Seo, Chung-Chan Lee
Nuclear Technology | Volume 158 | Number 2 | May 2007 | Pages 219-228
Technical Paper | Nuclear Reactor Thermal Hydraulics | doi.org/10.13182/NT07-A3837
Articles are hosted by Taylor and Francis Online.
Critical heat flux (CHF) in rod bundles is a parameter of great importance for the thermal-hydraulic design and safety analysis of advanced light water reactors. An experimental investigation has been conducted for the 19-rod hexagonal test bundles with a tightly spaced nonsquare arrangement of heater rods. The parametric effects on the CHF were examined for the heated length, the unheated rod, and the nonuniform axial power shape. As a result, a pertinent CHF correlation has been developed on the basis of the bundle cross-sectional averaged conditions. The available CHF database for rod bundles with square and nonsquare rod pitches was employed for the assessment of representative CHF correlations that were applicable to the round tubes and rod bundles. The database covered a wide range of operating conditions and test bundle geometries that are applicable to advanced light water reactors. The prediction accuracy of the CHF correlations was evaluated on the basis of the local thermal-hydraulic conditions calculated by a subchannel analysis code.
