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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.
Nicholas T. Saltos,* Richard N. Christensen, Tunc Aldemir
Nuclear Technology | Volume 83 | Number 1 | October 1988 | Pages 93-109
Technical Paper | Heat Transfer and Fluid Flow | doi.org/10.13182/NT88-A34178
Articles are hosted by Taylor and Francis Online.
A methodology is presented to determine the transient temperature distributions in fuel bundles under loss-of-coolant-accident (LOCA) conditions using a recently developed variational technique for the solution of radial-azimuthal heat conduction in the fuel rods and the modified view factor concept proposed by Uchida and Nakamura to model the radiative heat transfer between the rods. The variational technique is based on the Lebon-Labermont restricted variational principle and represents the temperature distribution in the rods at a given time during the LOCA via parabolic and circular trial functions in the radial and azimuthal directions, respectively. The methodology is implemented to a 4 × 4 boiling water reactor fuel bundle under typical LOCA conditions to investigate the effects of changes in rod heat transfer characteristics and simplifying modeling assumptions on predicted rod temperature distributions. The results show that these effects depend on the rod location in the assembly and LOCA phase under consideration and indicate that same degree of modeling detail may not be necessary for all the rods in the bundle at all times during the LOCA. An important advantage of the methodology is that it selects the optimum number of trial functions for each rod and for each time step in the simulation to reduce the computation time without compromising solution accuracy. Other advantages are that (a) the variational technique is faster than finite difference techniques for comparable accuracy and uses the same algorithm for one-dimensional radial and two-dimensional solutions, and (b) formulation of the heat conduction problem in the rods is compatible with the modular accident analysis codes already in use in the nuclear industry.